Each construction permit for a utilization facility is subject to the following terms and conditions and the applicable requirements of Sec. 50.55a; each construction permit for a production facility is subject to the following terms and conditions with the exception of paragraph (i); each early site permit is subject to the terms and conditions in paragraph (f) of this section; each manufacturing license is subject to the terms and conditions in paragraphs (e), (f), and (i) of this section and the applicable requirements of Sec. 50.55a; and each combined license is subject to the terms and conditions in paragraphs (e), (f), and (i) of this section and the applicable requirements of Sec. 50.55a until the date that the Commission makes the finding under Sec. 52.103(g) of this chapter:
(a) The construction permit shall state the earliest and latest dates for completion of the construction or modification.
(b) If the proposed construction or modification of the facility is not completed by the latest completion date, the construction permit shall expire and all rights are forfeited. However, upon good cause shown, the Commission will extend the completion date for a reasonable period of time. The Commission will recognize, among other things, developmental problems attributable to the experimental nature of the facility or fire, flood, explosion, strike, sabotage, domestic violence, enemy action, an act of the elements, and other acts beyond the control of the permit holder, as a basis for extending the completion date.
(c) Except as modified by this section and Sec. 50.55a, the construction permit shall be subject to the same conditions to which a license is subject.
(d) At or about the time of completion of the construction or modification of the facility, the applicant will file any additional information needed to bring the original application for license up to date, and will file an application for an operating license or an amendment to an application for a license to construct and operate the facility for the issuance of an operating license, as appropriate, as specified in Sec. 50.30(d) of this part.
(e)(1) Definitions. For purposes of this paragraph, the definitions in Sec. 21.3 of this chapter apply.
(1) Definitions. For purposes of this paragraph, the definitions in Sec. 21.3 of this chapter apply.
(2) Posting requirements. (i) Each individual, partnership, corporation, dedicating entity, or other entity subject to the regulations in this part shall post current copies of the regulations in this part; Section 206 of the Energy Reorganization Act of 1974 (ERA); and procedures adopted under the regulations in this part. These documents must be posted in a conspicuous position on any premises within the United States where the activities subject to this part are conducted.
(i) Each individual, partnership, corporation, dedicating entity, or other entity subject to the regulations in this part shall post current copies of the regulations in this part; Section 206 of the Energy Reorganization Act of 1974 (ERA); and procedures adopted under the regulations in this part. These documents must be posted in a conspicuous position on any premises within the United States where the activities subject to this part are conducted.
(ii) If posting of the regulations in this part or the procedures adopted under the regulations in this part is not practicable, the licensee or firm subject to the regulations in this part may, in addition to posting Section 206 of the ERA, post a notice which describes the regulations/procedures, including the name of the individual to whom reports may be made, and states where the regulation, procedures, and reports may be examined.
(3) Procedures. Each individual, corporation, partnership, or other entity holding a facility construction permit subject to this part, combined license (until the Commission makes the finding under 10 CFR 52.103(g)), and manufacturing license under 10 CFR part 52 must adopt appropriate procedures to--
(i) Evaluate deviations and failures to comply to identify defects and failures to comply associated with substantial safety hazards as soon as practicable, and, except as provided in paragraph (e)(3)(ii) of this section, in all cases within 60 days of discovery, to identify a reportable defect or failure to comply that could create a substantial safety hazard, were it to remain uncorrected.
(ii) Ensure that if an evaluation of an identified deviation or failure to comply potentially associated with a substantial safety hazard cannot be completed within 60 days from discovery of the deviation or failure to comply, an interim report is prepared and submitted to the Commission through a director or responsible officer or designated person as discussed in paragraph (e)(4)(v) of this section. The interim report should describe the deviation or failure to comply that is being evaluated and should also state when the evaluation will be completed. This interim report must be submitted in writing within 60 days of discovery of the deviation or failure to comply.
(iii) Ensure that a director or responsible officer of the holder of a facility construction permit subject to this part, combined license (until the Commission makes the finding under 10 CFR 52.103(g)), and manufacturing license under 10 CFR part 52 is informed as soon as practicable, and, in all cases, within the 5 working days after completion of the evaluation described in paragraph (e)(3)(i) or (e)(3)(ii) of this section, if the construction or manufacture of a facility or activity, or a basic component supplied for such facility or activity--
(A) Fails to comply with the AEA, as amended, or any applicable regulation, order, or license of the Commission, relating to a substantial safety hazard;
(B) Contains a defect; or
(C) Undergoes any significant breakdown in any portion of the quality assurance program conducted under the requirements of appendix B to 10 CFR part 50 which could have produced a defect in a basic component. These breakdowns in the quality assurance program are reportable whether or not the breakdown actually resulted in a defect in a design approved and released for construction, installation, or manufacture.
(4) Notification. (i) The holder of a facility construction permit subject to this part, combined license (until the Commission makes the finding under 10 CFR 52.103(g)), and manufacturing license who obtains information reasonably indicating that the facility fails to comply with the AEA, as amended, or any applicable regulation, order, or license of the Commission relating to a substantial safety hazard must notify the Commission of the failure to comply through a director or responsible officer or designated person as discussed in paragraph (e)(4)(v) of this section.
(i) The holder of a facility construction permit subject to this part, combined license (until the Commission makes the finding under 10 CFR 52.103(g)), and manufacturing license who obtains information reasonably indicating that the facility fails to comply with the AEA, as amended, or any applicable regulation, order, or license of the Commission relating to a substantial safety hazard must notify the Commission of the failure to comply through a director or responsible officer or designated person as discussed in paragraph (e)(4)(v) of this section.
(ii) The holder of a facility construction permit subject to this part, combined license, or manufacturing license, who obtains information reasonably indicating the existence of any defect found in the construction or manufacture, or any defect found in the final design of a facility as approved and released for construction or manufacture, must notify the Commission of the defect through a director or responsible officer or designated person as discussed in paragraph (e)(4)(v) of this section.
(iii) The holder of a facility construction permit subject to this part, combined license, or manufacturing license, who obtains information reasonably indicating that the quality assurance program has undergone any significant breakdown discussed in paragraph (e)(3)(iii)(C) of this section must notify the Commission of the breakdown in the quality assurance program through a director or responsible officer or designated person as discussed in paragraph (e)(4)(v) of this section.
(iv) A dedicating entity is responsible for identifying and evaluating deviations and reporting defects and failures to comply associated with substantial safety hazards for dedicated items; and maintaining auditable records for the dedication process.
(v) The notification requirements of this paragraph apply to all defects and failures to comply associated with a substantial safety hazard regardless of whether extensive evaluation, redesign, or repair is required to conform to the criteria and bases stated in the safety analysis report, construction permit, combined license, or manufacturing license. Evaluation of potential defects and failures to comply and reporting of defects and failures to comply under this section satisfies the construction permit holder's, combined license holder's, and manufacturing license holder's evaluation and notification obligations under part 21 of this chapter, and satisfies the responsibility of individual directors or responsible officers of holders of construction permits issued under Sec. 50.23, holders of combined licenses (until the Commission makes the finding under Sec. 52.103 of this chapter), and holders of manufacturing licenses to report defects, and failures to comply associated with substantial safety hazards under Section 206 of the ERA. The director or responsible officer may authorize an individual to provide the notification required by this section, provided that this must not relieve the director or responsible officer of his or her responsibility under this section.
(5) Notification--timing and where sent. The notification required by paragraph (e)(4) of this section must consist of--
(i) Initial notification by facsimile, which is the preferred method of notification, to the NRC Operations Center at (301) 816-5151 or by telephone at (301) 816-5100 within 2 days following receipt of information by the director or responsible corporate officer under paragraph (e)(3)(iii) of this section, on the identification of a defect or a failure to comply. Verification that the facsimile has been received should be made by calling the NRC Operations Center. This paragraph does not apply to interim reports described in paragraph (e)(3)(ii) of this section.
(ii) Written notification submitted to the Document Control Desk, U.S. Nuclear Regulatory Commission, by an appropriate method listed in Sec. 50.4, with a copy to the appropriate Regional Administrator at the address specified in appendix D to part 20 of this chapter and a copy to the appropriate NRC resident inspector within 30 days following receipt of information by the director or responsible corporate officer under paragraph (e)(3)(iii) of this section, on the identification of a defect or failure to comply.
(6) Content of notification. The written notification required by paragraph (e)(5)(ii) of this section must clearly indicate that the written notification is being submitted under Sec. 50.55(e) and include the following information, to the extent known.
(i) Name and address of the individual or individuals informing the Commission.
(ii) Identification of the facility, the activity, or the basic component supplied for the facility or the activity within the United States which contains a defect or fails to comply.
(iii) Identification of the firm constructing or manufacturing the facility or supplying the basic component which fails to comply or contains a defect.
(iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by the defect or failure to comply.
(v) The date on which the information of a defect or failure to comply was obtained.
(vi) In the case of a basic component which contains a defect or fails to comply, the number and location of all the basic components in use at the facility subject to the regulations in this part.
(vii) In the case of a completed reactor manufactured under part 52 of this chapter, the entities to which the reactor was supplied.
(viii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.
(ix) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to other entities.
(7) Procurement documents. Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part shall ensure that each procurement document for a facility, or a basic component specifies or is issued by the entity subject to the regulations, when applicable, that the provisions of 10 CFR part 21 or 10 CFR 50.55(e) applies, as applicable.
(8) Coordination with 10 CFR part 21. The requirements of Sec. 50.55(e) are satisfied when the defect or failure to comply associated with a substantial safety hazard has been previously reported under part 21 of this chapter, under Sec. 73.71 of this chapter, or under Sec. Sec. 50.55(e) or 50.73. For holders of construction permits issued before October 29, 1991, evaluation, reporting and recordkeeping requirements of Sec. 50.55(e) may be met by complying with the comparable requirements of part 21 of this chapter.
(9) Records retention. The holder of a construction permit, combined license, and manufacturing license must prepare and maintain records necessary to accomplish the purposes of this section, specifically--
(i) Retain procurement documents, which define the requirements that facilities or basic components must meet in order to be considered acceptable, for the lifetime of the facility or basic component.
(ii) Retain records of evaluations of all deviations and failures to comply under paragraph (e)(3)(i) of this section for the longest of:
(A) Ten (10) years from the date of the evaluation;
(B) Five (5) years from the date that an early site permit is referenced in an application for a combined license; or
(C) Five (5) years from the date of delivery of a manufactured reactor.
(iii) Retain records of all interim reports to the Commission made under paragraph (e)(3)(ii) of this section, or notifications to the Commission made under paragraph (e)(4) of this section for the minimum time periods stated in paragraph (e)(9)(ii) of this section;
(iv) Suppliers of basic components must retain records of:
(A) All notifications sent to affected licensees or purchasers under paragraph (e)(4)(iv) of this section for a minimum of ten (10) years following the date of the notification;
(B) The facilities or other purchasers to whom basic components or associated services were supplied for a minimum of fifteen (15) years from the delivery of the basic component or associated services.
(v) Maintaining records in accordance with this section satisfies the recordkeeping obligations under part 21 of this chapter of the entities, including directors or responsible officers thereof, subject to this section.
(f)(1) Each nuclear power plant or fuel reprocessing plant construction permit holder subject to the quality assurance criteria in appendix B of this part shall implement, pursuant to Sec. 50.34(a)(7) of this part, the quality assurance program described or referenced in the Safety Analysis Report, including changes to that report.
(1) Each nuclear power plant or fuel reprocessing plant construction permit holder subject to the quality assurance criteria in appendix B of this part shall implement, pursuant to Sec. 50.34(a)(7) of this part, the quality assurance program described or referenced in the Safety Analysis Report, including changes to that report.
(2) Each construction permit holder described in paragraph (f)(1) of this section shall, by June 10, 1983, submit to the appropriate NRC Regional Office shown in appendix D of part 20 of this chapter the current description of the quality assurance program it is implementing for inclusion in the Safety Analysis Report, unless there are no changes to the description previously accepted by NRC. This submittal must identify changes made to the quality assurance program description since the description was submitted to NRC. (Should a permit holder need additional time beyond June 10, 1983 to submit its current quality assurance program description to NRC, it shall notify the appropriate NRC Regional Office in writing, explain why additional time is needed, and provide a schedule for NRC approval showing when its current quality assurance program description will be submitted.)
(3) After March 11, 1983, each construction permit holder described in paragraph (f)(1) of this section may make a change to a previously accepted quality assurance program description included or referenced in the Safety Analysis Report, provided the change does not reduce the commitments in the program description previously accepted by the NRC. Changes to the quality assurance program description that do not reduce the commitments must be submitted to NRC within 90 days. Changes to the quality assurance program description that do reduce the commitments must be submitted to NRC and receive NRC approval before implementation, as follows:
(i) Changes to the Safety Analysis Report must be submitted for review as specified in Sec. 50.4. Changes made to NRC-accepted quality assurance topical report descriptions must be submitted as specified in Sec. 50.4.
(ii) The submittal of a change to the Safety Analysis Report quality assurance program description must include all pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the criteria of appendix B of this part and the Safety Analysis Report quality assurance program description commitments previously accepted by the NRC (the letter need not provide the basis for changes that correct spelling, punctuation, or editorial items).
(iii) A copy of the forwarding letter identifying the changes must be maintained as a facility record for three years.
(iv) Changes to the quality assurance program description included or referenced in the Safety Analysis Report shall be regarded as accepted by the Commission upon receipt of a letter to this effect from the appropriate reviewing office of the Commission or 60 days after submittal to the Commission, whichever occurs first.
(4) Each holder of an early site permit or a manufacturing license under part 52 of this chapter shall implement the quality assurance program described or referenced in the safety analysis report, including changes to that report. Each holder of a combined license shall implement the quality assurance program for design and construction described or referenced in the safety analysis report, including changes to that report, provided, however, that the holder of a combined license is not subject to the terms and conditions in this paragraph after the Commission makes the finding under Sec. 52.103(g) of this chapter.
(i) Each holder described in paragraph (f)(4) of this section may make a change to a previously accepted quality assurance program description included or referenced in the safety analysis report, if the change does not reduce the commitments in the program description previously accepted by the NRC. Changes to the quality assurance program description that do not reduce the commitments must be submitted to NRC within 90 days. Changes to the quality assurance program description that reduce the commitments must be submitted to NRC and receive NRC approval before implementation, as follows:
(A) Changes to the safety analysis report must be submitted for review as specified in Sec. 50.4. Changes made to NRC-accepted quality assurance topical report descriptions must be submitted as specified in Sec. 50.4.
(B) The submittal of a change to the safety analysis report quality assurance program description must include all pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the criteria of appendix B of this part and the safety analysis report quality assurance program description commitments previously accepted by the NRC (the letter need not provide the basis for changes that correct spelling, punctuation, or editorial items).
(C) A copy of the forwarding letter identifying the changes must be maintained as a facility record for three (3) years.
(D) Changes to the quality assurance program description included or referenced in the safety analysis report shall be regarded as accepted by the Commission upon receipt of a letter to this effect from the appropriate reviewing office of the Commission or 60 days after submittal to the Commission, whichever occurs first.
(ii) [Reserved]
(g) [Reserved]
(h) [Reserved]
(i) Structures, systems, and components subject to the codes and standards in 10 CFR 50.55a must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed. [21 FR 355, Jan. 19, 1956, as amended at 32 FR 4055, Mar. 15, 1967; 35 FR 11461, July 17, 1970; 35 FR 19661, Dec. 29, 1970; 36 FR 11424, June 12, 1971; 37 FR 6460, Mar. 30, 1972; 38 FR 1272, Jan. 11, 1973; 41 FR 16446, Apr. 19, 1976; 42 FR 43385, Aug. 29, 1977; 48 FR 1029, Jan. 10, 1983; 51 FR 40309, Nov. 6, 1986; 56 FR 36091, July 31, 1991; 59 FR 14087, Mar. 25, 1994; 68 FR 58809, Oct. 10, 2003; 72 FR 49497, Aug. 28, 2007; 78 FR 34248, June 7, 2013; 79 FR 65798, Nov. 5, 2014] Sec. 50.55a Codes and standards.
(a) Documents approved for incorporation by reference. The standards listed in this paragraph have been approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51. The standards are available for inspection at the NRC Technical Library, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-6239; or at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.
(1) American Society of Mechanical Engineers (ASME), Three Park Avenue, New York, NY 10016; telephone: 1-800-843-2763; http://www.asme.org/Codes/.
(i) ASME Boiler and Pressure Vessel Code, Section III. The editions and addenda for Section III of the ASME Boiler and Pressure Vessel Code are listed below, but limited to those provisions identified in paragraph (b)(1) of this section.
(A) ``Rules for Construction of Nuclear Vessels:''
(1) 1963 Edition,
(2) Summer 1964 Addenda,
(3) Winter 1964 Addenda,
(4) 1965 Edition,
(5) 1965 Summer Addenda,
(6) 1965 Winter Addenda,
(7) 1966 Summer Addenda,
(8) 1966 Winter Addenda,
(9) 1967 Summer Addenda,
(10) 1967 Winter Addenda,
(11) 1968 Edition,
(12) 1968 Summer Addenda,
(13)1968 Winter Addenda,
(14) 1969 Summer Addenda,
(15) 1969 Winter Addenda,
(16) 1970 Summer Addenda, and
(17) 1970 Winter Addenda.
(B) ``Rules for Construction of Nuclear Power Plant Components:''
(1) 1971 Edition,
(2) 1971 Summer Addenda,
(3) 1971 Winter Addenda,
(4) 1972 Summer Addenda,
(5) 1972 Winter Addenda,
(6) 1973 Summer Addenda, and
(7) 1973 Winter Addenda.
(C) ``Division 1 Rules for Construction of Nuclear Power Plant Components:''
(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda,
(4) 1975 Summer Addenda,
(5) 1975 Winter Addenda,
(6) 1976 Summer Addenda, and
(7) 1976 Winter Addenda;
(D) ``Rules for Construction of Nuclear Power Plant Components--Division 1'';
(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Summer Addenda,
(10) 1980 Winter Addenda,
(11) 1981 Summer Addenda,
(12) 1981 Winter Addenda,
(13) 1982 Summer Addenda,
(14) 1982 Winter Addenda,
(15) 1983 Edition,
(16) 1983 Summer Addenda,
(17) 1983 Winter Addenda,
(18) 1984 Summer Addenda,
(19) 1984 Winter Addenda,
(20) 1985 Summer Addenda,
(21) 1985 Winter Addenda,
(22) 1986 Edition,
(23) 1986 Addenda,
(24) 1987 Addenda,
(25) 1988 Addenda,
(26) 1989 Edition,
(27) 1989 Addenda,
(28) 1990 Addenda,
(29) 1991 Addenda,
(30) 1992 Edition,
(31) 1992 Addenda,
(32) 1993 Addenda,
(33) 1994 Addenda,
(34) 1995 Edition,
(35) 1995 Addenda,
(36) 1996 Addenda, and
(37) 1997 Addenda.
(E) ``Rules for Construction of Nuclear Facility Components--Division 1:''
(1) 1998 Edition,
(2) 1998 Addenda,
(3) 1999 Addenda,
(4) 2000 Addenda,
(5) 2001 Edition,
(6) 2001 Addenda,
(7) 2002 Addenda,
(8) 2003 Addenda,
(9) 2004 Edition,
(10) 2005 Addenda,
(11) 2006 Addenda,
(12) 2007 Edition, and
(13) 2008 Addenda.
(ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions and addenda for Section XI of the ASME Boiler and Pressure Vessel Code are listed below, but limited to those provisions identified in paragraph (b)(2) of this section.
(A) ``Rules for Inservice Inspection of Nuclear Reactor Coolant Systems:''
(1) 1970 Edition,
(2) 1971 Edition,
(3) 1971 Summer Addenda,
(4) 1971 Winter Addenda,
(5) 1972 Summer Addenda,
(6) 1972 Winter Addenda,
(7) 1973 Summer Addenda, and
(8) 1973 Winter Addenda.
(B) ``Rules for Inservice Inspection of Nuclear Power Plant Components:''
(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda, and
(4) 1975 Summer Addenda.
(5) 1975 Winter Addenda,
(6) 1976 Summer Addenda, and
(7) 1976 Winter Addenda.
(C) ``Rules for Inservice Inspection of Nuclear Power Plant Components--Division 1:''
(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Winter Addenda,
(10) 1981 Summer Addenda,
(11) 1981 Winter Addenda,
(12) 1982 Summer Addenda,
(13) 1982 Winter Addenda,
(14) 1983 Edition,
(15) 1983 Summer Addenda,
(16) 1983 Winter Addenda,
(17) 1984 Summer Addenda,
(18) 1984 Winter Addenda,
(19) 1985 Summer Addenda,
(20) 1985 Winter Addenda,
(21) 1986 Edition,
(22) 1986 Addenda,
(23) 1987 Addenda,
(24) 1988 Addenda,
(25) 1989 Edition,
(26) 1989 Addenda,
(27) 1990 Addenda,
(28) 1991 Addenda,
(29) 1992 Edition,
(30) 1992 Addenda,
(31) 1993 Addenda,
(32) 1994 Addenda,
(33) 1995 Edition,
(34) 1995 Addenda,
(35) 1996 Addenda,
(36) 1997 Addenda,
(37) 1998 Edition,
(38) 1998 Addenda,
(39) 1999 Addenda,
(40) 2000 Addenda,
(41) 2001 Edition,
(42) 2001 Addenda,
(43) 2002 Addenda,
(44) 2003 Addenda,
(45) 2004 Edition,
(46) 2005 Addenda,
(47) 2006 Addenda,
(48) 2007 Edition, and
(49) 2008 Addenda.
(iii) ASME Code Cases: Nuclear Components--(A) ASME Code Case N-722-1. ASME Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' (Approval Date: January 26, 2009), with the conditions in paragraph (g)(6)(ii)(E) of this section.
(A) ASME Code Case N-722-1. ASME Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' (Approval Date: January 26, 2009), with the conditions in paragraph (g)(6)(ii)(E) of this section.
(B) ASME Code Case N-729-1. ASME Code Case N-729-1, ``Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1'' (Approval Date: March 28, 2006), with the conditions in paragraph (g)(6)(ii)(D) of this section.
(C) ASME Code Case N-770-1. ASME Code Case N-770-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' (Approval Date: December 25, 2009), with the conditions in paragraph (g)(6)(ii)(F) of this section.
(iv) ASME Operation and Maintenance Code. The editions and addenda for the ASME Code for Operation and Maintenance of Nuclear Power Plants are listed below, but limited to those provisions identified in paragraph (b)(3) of this section.
(A) ``Code for Operation and Maintenance of Nuclear Power Plants:''
(1) 1995 Edition,
(2) 1996 Addenda,
(3) 1997 Addenda,
(4) 1998 Edition,
(5) 1999 Addenda,
(6) 2000 Addenda,
(7) 2001 Edition,
(8) 2002 Addenda,
(9) 2003 Addenda,
(10) 2004 Edition,
(11) 2005 Addenda, and
(12) 2006 Addenda.
(B) [Reserved]
(2) Institute of Electrical and Electronics Engineers (IEEE) Service Center, 445 Hoes Lane, Piscataway, NJ 08855; telephone: 1-800-678-4333; http://ieeexplore.ieee.org.
(i) IEEE standard 279-1971. (IEEE Std 279-1971), ``Criteria for Protection Systems for Nuclear Power Generating Stations'' (Approval Date: June 3, 1971), referenced in paragraph (h)(2) of this section.
(ii) IEEE Standard 603-1991. (IEEE Std 603-1991), ``Standard Criteria for Safety Systems for Nuclear Power Generating Stations'' (Approval Date: June 27, 1991), referenced in paragraphs (h)(2) and (3) of this section. All other standards that are referenced in IEEE Std 603-1991 are not approved for incorporation by reference.
(iii) IEEE standard 603-1991, correction sheet. (IEEE Std 603-1991 correction sheet), ``Standard Criteria for Safety Systems for Nuclear Power Generating Stations, Correction Sheet, Issued January 30, 1995, '' referenced in paragraphs (h)(2) and (3) of this section. (Copies of this correction sheet may be purchased from Thomson Reuters, 3916 Ranchero Dr., Ann Arbor, MI 48108; http://www.techstreet.com.)
(3) U.S. Nuclear Regulatory Commission (NRC) Public Document Room, 11555 Rockville Pike, Rockville, Maryland 20852; telephone: 1-800-397-4209; email: pdr.resource@nrc.gov; http://www.nrc.gov/reading-rm/doc-collections/reg-guides/.
(i) NRC Regulatory Guide 1.84, Revision 36. NRC Regulatory Guide 1.84, Revision 36, ``Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,'' dated August 2014, with the requirements in paragraph (b)(4) of this section.
(ii) NRC Regulatory Guide 1.147, Revision 17. NRC Regulatory Guide 1.147, Revision 17, ``Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,'' dated August 2014, which lists ASME Code Cases that the NRC has approved in accordance with the requirements in paragraph (b)(5) of this section.
(iii) NRC Regulatory Guide 1.192, Revision 1. NRC Regulatory Guide 1.192, Revision 1, ``Operation and Maintenance Code Case Acceptability, ASME OM Code,'' dated August 2014, which lists ASME Code Cases that the NRC has approved in accordance with the requirements in paragraph (b)(6) of this section.
(b) Use and conditions on the use of standards. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code (BPV Code) and the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) as specified in this paragraph. Each combined license for a utilization facility is subject to the following conditions.
(1) Conditions on ASME BPV Code Section III. Each manufacturing license, standard design approval, and design certification under part 52 of this chapter is subject to the following conditions. As used in this section, references to Section III refer to Section III of the ASME Boiler and Pressure Vessel Code and include the 1963 Edition through 1973 Winter Addenda and the 1974 Edition (Division 1) through the 2008 Addenda (Division 1), subject to the following conditions:
(i) Section III condition: Section III materials. When applying the 1992 Edition of Section III, applicants or licensees must apply the 1992 Edition with the 1992 Addenda of Section II of the ASME Boiler and Pressure Vessel Code.
(ii) Section III condition: Weld leg dimensions. When applying the 1989 Addenda through the latest edition and addenda, applicants or licensees may not apply subparagraphs NB-3683.4(c)(1) and NB-3683.4(c)(2) or Footnote 11 from the 1989 Addenda through the 2003 Addenda, or Footnote 13 from the 2004 Edition through the 2008 Addenda to Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 for welds with leg size less than 1.09 tn.
(iii) Section III condition: Seismic design of piping. Applicants or licensees may use Subarticles NB-3200, NB-3600, NC-3600, and ND-3600 for seismic design of piping, up to and including the 1993 Addenda, subject to the condition specified in paragraph (b)(1)(ii) of this section. Applicants or licensees may not use these subarticles for seismic design of piping in the 1994 Addenda through the 2005 Addenda incorporated by reference in paragraph (a)(1) of this section, except that Subarticle NB-3200 in the 2004 Edition through the 2008 Addenda may be used by applicants and licensees, subject to the condition in paragraph (b)(1)(iii)(A) of this section. Applicants or licensees may use Subarticles NB-3600, NC-3600, and ND-3600 for the seismic design of piping in the 2006 Addenda through the 2008 Addenda, subject to the conditions of this paragraph corresponding to those subarticles.
(A) Seismic design of piping: First provision. When applying Note (1) of Figure NB-3222-1 for Level B service limits, the calculation of Pb stresses must include reversing dynamic loads (including inertia earthquake effects) if evaluation of these loads is required by NB-3223(b).
(B) Seismic design of piping: Second provision. For Class 1 piping, the material and Do/t requirements of NB-3656(b) must be met for all Service Limits when the Service Limits include reversing dynamic loads, and the alternative rules for reversing dynamic loads are used.
(iv) Section III condition: Quality assurance. When applying editions and addenda later than the 1989 Edition of Section III, the requirements of NQA-1, ``Quality Assurance Requirements for Nuclear Facilities,'' 1986 Edition through the 1994 Edition, are acceptable for use, provided that the edition and addenda of NQA-1 specified in NCA-4000 is used in conjunction with the administrative, quality, and technical provisions contained in the edition and addenda of Section III being used.
(v) Section III condition: Independence of inspection. Applicants or licensees may not apply NCA-4134.10(a) of Section III, 1995 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1) of this section.
(vi) Section III condition: Subsection NH. The provisions in Subsection NH, ``Class 1 Components in Elevated Temperature Service,'' 1995 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1) of this section, may only be used for the design and construction of Type 316 stainless steel pressurizer heater sleeves where service conditions do not cause the components to reach temperatures exceeding 900[emsp14] [deg]F.
(vii) Section III condition: Capacity certification and demonstration of function of incompressible-fluid pressure-relief valves. When applying the 2006 Addenda through the 2007 Edition up to and including the 2008 Addenda, applicants and licensees may use paragraph NB-7742, except that paragraph NB-7742(a)(2) may not be used. For a valve design of a single size to be certified over a range of set pressures, the demonstration of function tests under paragraph NB-7742 must be conducted as prescribed in NB-7732.2 on two valves covering the minimum set pressure for the design and the maximum set pressure that can be accommodated at the demonstration facility selected for the test.
(2) Conditions on ASME BPV Code Section XI. As used in this section, references to Section XI refer to Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition through the 1976 Winter Addenda and the 1977 Edition through the 2007 Edition with the 2008 Addenda, subject to the following conditions:
(i) [Reserved]
(ii) Section XI condition: Pressure-retaining welds in ASME Code Class 1 piping (applies to Table IWB-2500 and IWB-2500-1 and Category B-J). If the facility's application for a construction permit was docketed prior to July 1, 1978, the extent of examination for Code Class 1 pipe welds may be determined by the requirements of Table IWB-2500 and Table IWB-2600 Category B-J of Section XI of the ASME BPV Code in the 1974 Edition and Addenda through the Summer 1975 Addenda or other requirements the NRC may adopt.
(iii) [Reserved]
(iv) [Reserved]
(v) [Reserved]
(vi) Section XI condition: Effective edition and addenda of Subsection IWE and Subsection IWL. Applicants or licensees may use either the 1992 Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection IWE and Subsection IWL, as conditioned by the requirements in paragraphs (b)(2)(viii) and (ix) of this section, when implementing the initial 120-month inspection interval for the containment inservice inspection requirements of this section. Successive 120-month interval updates must be implemented in accordance with paragraph (g)(4)(ii) of this section.
(vii) Section XI condition: Section XI references to OM Part 4, OM Part 6, and OM Part 10 (Table IWA-1600-1). When using Table IWA-1600-1, ``Referenced Standards and Specifications,'' in the Section XI, Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition, the specified ``Revision Date or Indicator'' for ASME/ANSI OM part 4, ASME/ANSI part 6, and ASME/ANSI part 10 must be the OMa-1988 Addenda to the OM-1987 Edition. These requirements have been incorporated into the OM Code, which is incorporated by reference in paragraph (a)(1)(iv) of this section.
(viii) Section XI condition: Concrete containment examinations. Applicants or licensees applying Subsection IWL, 1992 Edition with the 1992 Addenda, must apply paragraphs (b)(2)(viii)(A) through (E) of this section. Applicants or licensees applying Subsection IWL, 1995 Edition with the 1996 Addenda, must apply paragraphs (b)(2)(viii)(A), (b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of this section. Applicants or licensees applying Subsection IWL, 1998 Edition through the 2000 Addenda, must apply paragraphs (b)(2)(viii)(E) and (F) of this section. Applicants or licensees applying Subsection IWL, 2001 Edition through the 2004 Edition, up to and including the 2006 Addenda, must apply paragraphs (b)(2)(viii)(E) through (G) of this section. Applicants or licensees applying Subsection IWL, 2007 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, must apply paragraph (b)(2)(viii)(E) of this section.
(A) Concrete containment examinations: First provision. Grease caps that are accessible must be visually examined to detect grease leakage or grease cap deformations. Grease caps must be removed for this examination when there is evidence of grease cap deformation that indicates deterioration of anchorage hardware.
(B) Concrete containment examinations: Second provision. When evaluation of consecutive surveillances of pre-stressing forces for the same tendon or tendons in a group indicates a trend of pre-stress loss such that the tendon force(s) would be less than the minimum design pre-stress requirements before the next inspection interval, an evaluation must be performed and reported in the Engineering Evaluation Report as prescribed in IWL-3300.
(C) Concrete containment examinations: Third provision. When the elongation corresponding to a specific load (adjusted for effective wires or strands) during re-tensioning of tendons differs by more than 10 percent from that recorded during the last measurement, an evaluation must be performed to determine whether the difference is related to wire failures or slip of wires in anchorage. A difference of more than 10 percent must be identified in the ISI Summary Report required by IWA-6000.
(D) Concrete containment examinations: Fourth provision. The applicant or licensee must report the following conditions, if they occur, in the ISI Summary Report required by IWA-6000:
(1) The sampled sheathing filler grease contains chemically combined water exceeding 10 percent by weight or the presence of free water;
(2) The absolute difference between the amount removed and the amount replaced exceeds 10 percent of the tendon net duct volume; and
(3) Grease leakage is detected during general visual examination of the containment surface.
(E) Concrete containment examinations: Fifth provision. For Class CC applications, the applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or the result in degradation to such inaccessible areas. For each inaccessible area identified, the applicant or licensee must provide the following in the ISI Summary Report required by IWA-6000:
(1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation;
(2) An evaluation of each area, and the result of the evaluation; and
(3) A description of necessary corrective actions.
(F) Concrete containment examinations: Sixth provision. Personnel that examine containment concrete surfaces and tendon hardware, wires, or strands must meet the qualification provisions in IWA-2300. The ``owner-defined'' personnel qualification provisions in IWL-2310(d) are not approved for use.
(G) Concrete containment examinations: Seventh provision. Corrosion protection material must be restored following concrete containment post-tensioning system repair and replacement activities in accordance with the quality assurance program requirements specified in IWA-1400.
(ix) Section XI condition: Metal containment examinations. Applicants or licensees applying Subsection IWE, 1992 Edition with the 1992 Addenda, or the 1995 Edition with the 1996 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A) through (E) of this section. Applicants or licensees applying Subsection IWE, 1998 Edition through the 2001 Edition with the 2003 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and (b)(2)(ix)(F) through (I) of this section. Applicants or licensees applying Subsection IWE, 2004 Edition, up to and including the 2005 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and (b)(2)(ix)(F) through (H) of this section. Applicants or licensees applying Subsection IWE, 2004 Edition with the 2006 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and (b)(2)(ix)(B) of this section. Applicants or licensees applying Subsection IWE, 2007 Edition through the latest addenda incorporated by reference in paragraph (a)(1)(ii) of this section, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J) of this section.
(A) Metal containment examinations: First provision. For Class MC applications, the following apply to inaccessible areas.
(1) The applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or could result in degradation to such inaccessible areas.
(2) For each inaccessible area identified for evaluation, the applicant or licensee must provide the following in the ISI Summary Report as required by IWA-6000:
(i) A description of the type and estimated extent of degradation, and the conditions that led to the degradation;
(ii) An evaluation of each area, and the result of the evaluation; and
(iii) A description of necessary corrective actions.
(B) Metal containment examinations: Second provision. When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.
(C) Metal containment examinations: Third provision. The examinations specified in Examination Category E-B, Pressure Retaining Welds, and Examination Category E-F, Pressure Retaining Dissimilar Metal Welds, are optional.
(D) Metal containment examinations: Fourth provision. This paragraph (b)(2)(ix)(D) may be used as an alternative to the requirements of IWE-2430.
(1) If the examinations reveal flaws or areas of degradation exceeding the acceptance standards of Table IWE-3410-1, an evaluation must be performed to determine whether additional component examinations are required. For each flaw or area of degradation identified that exceeds acceptance standards, the applicant or licensee must provide the following in the ISI Summary Report required by IWA-6000:
(i) A description of each flaw or area, including the extent of degradation, and the conditions that led to the degradation;
(ii) The acceptability of each flaw or area and the need for additional examinations to verify that similar degradation does not exist in similar components; and
(iii) A description of necessary corrective actions.
(2) The number and type of additional examinations to ensure detection of similar degradation in similar components.
(E) Metal containment examinations: Fifth provision. A general visual examination as required by Subsection IWE must be performed once each period.
(F) Metal containment examinations: Sixth provision. VT-1 and VT-3 examinations must be conducted in accordance with IWA-2200. Personnel conducting examinations in accordance with the VT-1 or VT-3 examination method must be qualified in accordance with IWA-2300. The ``owner-defined'' personnel qualification provisions in IWE-2330(a) for personnel that conduct VT-1 and VT-3 examinations are not approved for use.
(G) Metal containment examinations: Seventh provision. The VT-3 examination method must be used to conduct the examinations in Items E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method must be used to conduct the examination in Item E4.11 of Table IWE-2500-1. An examination of the pressure-retaining bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must be conducted once each interval. The ``owner-defined'' visual examination provisions in IWE-2310(a) are not approved for use for VT-1 and VT-3 examinations.
(H) Metal containment examinations: Eighth provision. Containment bolted connections that are disassembled during the scheduled performance of the examinations in Item E1.11 of Table IWE-2500-1 must be examined using the VT-3 examination method. Flaws or degradation identified during the performance of a VT-3 examination must be examined in accordance with the VT-1 examination method. The criteria in the material specification or IWB-3517.1 must be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item E1.11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason.
(I) Metal containment examinations: Ninth provision. The ultrasonic examination acceptance standard specified in IWE-3511.3 for Class MC pressure-retaining components must also be applied to metallic liners of Class CC pressure-retaining components.
(J) Metal containment examinations: Tenth provision. In general, a repair/replacement activity such as replacing a large containment penetration, cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, pressurizers, or other major equipment; or other similar modification is considered a major containment modification. When applying IWE-5000 to Class MC pressure-retaining components, any major containment modification or repair/replacement must be followed by a Type A test to provide assurance of both containment structural integrity and leak-tight integrity prior to returning to service, in accordance with 10 CFR part 50, Appendix J, Option A or Option B on which the applicant's or licensee's Containment Leak-Rate Testing Program is based. When applying IWE-5000, if a Type A, B, or C Test is performed, the test pressure and acceptance standard for the test must be in accordance with 10 CFR part 50, Appendix J.
(x) Section XI condition: Quality assurance. When applying Section XI editions and addenda later than the 1989 Edition, the requirements of NQA-1, ``Quality Assurance Requirements for Nuclear Facilities,'' 1979 Addenda through the 1989 Edition, are acceptable as permitted by IWA-1400 of Section XI, if the licensee uses its 10 CFR part 50, Appendix B, quality assurance program, in conjunction with Section XI requirements. Commitments contained in the licensee's quality assurance program description that are more stringent than those contained in NQA-1 must govern Section XI activities. Further, where NQA-1 and Section XI do not address the commitments contained in the licensee's Appendix B quality assurance program description, the commitments must be applied to Section XI activities.
(xi) [Reserved]
(xii) Section XI condition: Underwater welding. The provisions in IWA-4660, ``Underwater Welding,'' of Section XI, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, are not approved for use on irradiated material.
(xiii) [Reserved]
(xiv) Section XI condition: Appendix VIII personnel qualification. All personnel qualified for performing ultrasonic examinations in accordance with Appendix VIII must receive 8 hours of annual hands-on training on specimens that contain cracks. Licensees applying the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section may use the annual practice requirements in VII-4240 of Appendix VII of Section XI in place of the 8 hours of annual hands-on training provided that the supplemental practice is performed on material or welds that contain cracks, or by analyzing prerecorded data from material or welds that contain cracks. In either case, training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility.
(xv) Section XI condition: Appendix VIII specimen set and qualification requirements. Licensees using Appendix VIII in the 1995 Edition through the 2001 Edition of the ASME Boiler and Pressure Vessel Code may elect to comply with all of the provisions in paragraphs (b)(2)(xv)(A) through (M) of this section, except for paragraph (b)(2)(xv)(F) of this section, which may be used at the licensee's option. Licensees using editions and addenda after 2001 Edition through the 2006 Addenda must use the 2001 Edition of Appendix VIII and may elect to comply with all of the provisions in paragraphs (b)(2)(xv)(A) through (M) of this section, except for paragraph (b)(2)(xv)(F) of this section, which may be used at the licensee's option.
(A) Specimen set and qualification: First provision. When applying Supplements 2, 3, and 10 to Appendix VIII, the following examination coverage criteria requirements must be used:
(1) Piping must be examined in two axial directions, and when examination in the circumferential direction is required, the circumferential examination must be performed in two directions, provided access is available. Dissimilar metal welds must be examined axially and circumferentially.
(2) Where examination from both sides is not possible, full coverage credit may be claimed from a single side for ferritic welds. Where examination from both sides is not possible on austenitic welds or dissimilar metal welds, full coverage credit from a single side may be claimed only after completing a successful single-sided Appendix VIII demonstration using flaws on the opposite side of the weld. Dissimilar metal weld qualifications must be demonstrated from the austenitic side of the weld, and the qualification may be expanded for austenitic welds with no austenitic sides using a separate add-on performance demonstration. Dissimilar metal welds may be examined from either side of the weld.
(B) Specimen set and qualification: Second provision. The following conditions must be used in addition to the requirements of Supplement 4 to Appendix VIII:
(1) Paragraph 3.1, Detection acceptance criteria--Personnel are qualified for detection if the results of the performance demonstration satisfy the detection requirements of ASME Section XI, Appendix VIII, Table VIII-S4-1, and no flaw greater than 0.25 inch through-wall dimension is missed.
(2) Paragraph 1.1(c), Detection test matrix--Flaws smaller than the 50 percent of allowable flaw size, as defined in IWB-3500, need not be included as detection flaws. For procedures applied from the inside surface, use the minimum thickness specified in the scope of the procedure to calculate a/t. For procedures applied from the outside surface, the actual thickness of the test specimen is to be used to calculate a/t.
(C) Specimen set and qualification: Third provision. When applying Supplement 4 to Appendix VIII, the following conditions must be used:
(1) A depth sizing requirement of 0.15 inch RMS must be used in lieu of the requirements in Subparagraphs 3.2(a) and 3.2(c), and a length sizing requirement of 0.75 inch RMS must be used in lieu of the requirement in Subparagraph 3.2(b).
(2) In lieu of the location acceptance criteria requirements of Subparagraph 2.1(b), a flaw will be considered detected when reported within 1.0 inch or 10 percent of the metal path to the flaw, whichever is greater, of its true location in the X and Y directions.
(3) In lieu of the flaw type requirements of Subparagraph 1.1(e)(1), a minimum of 70 percent of the flaws in the detection and sizing tests must be cracks. Notches, if used, must be limited by the following:
(i) Notches must be limited to the case where examinations are performed from the clad surface.
(ii) Notches must be semielliptical with a tip width of less than or equal to 0.010 inches.
(iii) Notches must be perpendicular to the surface within 2 degrees.
(4) In lieu of the detection test matrix requirements in paragraphs 1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain a representative distribution of flaw orientations, sizes, and locations.
(D) Specimen set and qualification: Fourth provision. The following conditions must be used in addition to the requirements of Supplement 6 to Appendix VIII:
(1) Paragraph 3.1, Detection Acceptance Criteria--Personnel are qualified for detection if:
(i) No surface connected flaw greater than 0.25 inch through-wall has been missed.
(ii) No embedded flaw greater than 0.50 inch through-wall has been missed.
(2) Paragraph 3.1, Detection Acceptance Criteria--For procedure qualification, all flaws within the scope of the procedure are detected.
(3) Paragraph 1.1(b) for detection and sizing test flaws and locations--Flaws smaller than the 50 percent of allowable flaw size, as defined in IWB-3500, need not be included as detection flaws. Flaws that are less than the allowable flaw size, as defined in IWB-3500, may be used as detection and sizing flaws.
(4) Notches are not permitted.
(E) Specimen set and qualification: Fifth provision. When applying Supplement 6 to Appendix VIII, the following conditions must be used:
(1) A depth sizing requirement of 0.25 inch RMS must be used in lieu of the requirements of subparagraphs 3.2(a), 3.2(c)(2), and 3.2(c)(3).
(2) In lieu of the location acceptance criteria requirements in Subparagraph 2.1(b), a flaw will be considered detected when reported within 1.0 inch or 10 percent of the metal path to the flaw, whichever is greater, of its true location in the X and Y directions.
(3) In lieu of the length sizing criteria requirements of Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch RMS must be used.
(4) In lieu of the detection specimen requirements in Subparagraph 1.1(e)(1), a minimum of 55 percent of the flaws must be cracks. The remaining flaws may be cracks or fabrication type flaws, such as slag and lack of fusion. The use of notches is not allowed.
(5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test matrix, personnel demonstration test sets must contain a representative distribution of flaw orientations, sizes, and locations.
(F) Specimen set and qualification: Sixth provision. The following conditions may be used for personnel qualification for combined Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII qualification. Licensees choosing to apply this combined qualification must apply all of the provisions of Supplements 4 and 6 including the following conditions:
(1) For detection and sizing, the total number of flaws must be at least 10. A minimum of 5 flaws must be from Supplement 4, and a minimum of 50 percent of the flaws must be from Supplement 6. At least 50 percent of the flaws in any sizing must be cracks. Notches are not acceptable for Supplement 6.
(2) Examination personnel are qualified for detection and length sizing when the results of any combined performance demonstration satisfy the acceptance criteria of Supplement 4 to Appendix VIII.
(3) Examination personnel are qualified for depth sizing when Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws are sized within the respective acceptance criteria of those supplements.
(G) Specimen set and qualification: Seventh provision. When applying Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or combined Supplement 4 and Supplement 6 qualification, the following additional conditions must be used, and examination coverage must include:
(1) The clad-to-base-metal-interface, including a minimum of 15 percent T (measured from the clad-to-base-metal-interface), must be examined from four orthogonal directions using procedures and personnel qualified in accordance with Supplement 4 to Appendix VIII.
(2) If the clad-to-base-metal-interface procedure demonstrates detectability of flaws with a tilt angle relative to the weld centerline of at least 45 degrees, the remainder of the examination volume is considered fully examined if coverage is obtained in one parallel and one perpendicular direction. This must be accomplished using a procedure and personnel qualified for single-side examination in accordance with Supplement 6. Subsequent examinations of this volume may be performed using examination techniques qualified for a tilt angle of at least 10 degrees.
(3) The examination volume not addressed by paragraph (b)(2)(xv)(G)(1) of this section is considered fully examined if coverage is obtained in one parallel and one perpendicular direction, using a procedure and personnel qualified for single sided examination when the conditions in paragraph (b)(2)(xv)(G)(2) are met.
(H) Specimen set and qualification: Eighth provision. When applying Supplement 5 to Appendix VIII, at least 50 percent of the flaws in the demonstration test set must be cracks and the maximum misorientation must be demonstrated with cracks. Flaws in nozzles with bore diameters equal to or less than 4 inches may be notches.
(I) Specimen set and qualification: Ninth provision. When applying Supplement 5, Paragraph (a), to Appendix VIII, the number of false calls allowed must be D/10, with a maximum of 3, where D is the diameter of the nozzle.
(J) [Reserved]
(K) Specimen set and qualification: Eleventh provision. When performing nozzle-to-vessel weld examinations, the following conditions must be used when the requirements contained in Supplement 7 to Appendix VIII are applied for nozzle-to-vessel welds in conjunction with Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or combined Supplement 4 and Supplement 6 qualification.
(1) For examination of nozzle-to-vessel welds conducted from the bore, the following conditions are required to qualify the procedures, equipment, and personnel:
(i) For detection, a minimum of four flaws in one or more full-scale nozzle mock-ups must be added to the test set. The specimens must comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for flaw locations specified in Table VIII S6-1. Flaws may be notches, fabrication flaws, or cracks. Seventy-five (75) percent of the flaws must be cracks or fabrication flaws. Flaw locations and orientations must be selected from the choices shown in paragraph (b)(2)(xv)(K)(4) of this section, Table VIII-S7-1--Modified, with the exception that flaws in the outer eighty-five (85) percent of the weld need not be perpendicular to the weld. There may be no more than two flaws from each category, and at least one subsurface flaw must be included.
(ii) For length sizing, a minimum of four flaws as in paragraph (b)(2)(xv)(K)(1)(i) of this section must be included in the test set. The length sizing results must be added to the results of combined Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The combined results must meet the acceptance standards contained in paragraph (b)(2)(xv)(E)(3) of this section.
(iii) For depth sizing, a minimum of four flaws as in paragraph (b)(2)(xv)(K)(1)(i) of this section must be included in the test set. Their depths must be distributed over the ranges of Supplement 4, Paragraph 1.1, to Appendix VIII, for the inner 15 percent of the wall thickness and Supplement 6, Paragraph 1.1, to Appendix VIII, for the remainder of the wall thickness. The depth sizing results must be combined with the sizing results from Supplement 4 to Appendix VIII for the inner 15 percent and to Supplement 6 to Appendix VIII for the remainder of the wall thickness. The combined results must meet the depth sizing acceptance criteria contained in paragraphs (b)(2)(xv)(C)(1), (b)(2)(xv)(E)(1), and (b)(2)(xv)(F)(3) of this section.
(2) For examination of reactor pressure vessel nozzle-to-vessel welds conducted from the inside of the vessel, the following conditions are required:
(i) The clad-to-base-metal-interface and the adjacent examination volume to a minimum depth of 15 percent T (measured from the clad-to-base-metal-interface) must be examined from four orthogonal directions using a procedure and personnel qualified in accordance with Supplement 4 to Appendix VIII as conditioned by paragraphs (b)(2)(xv)(B) and (C) of this section.
(ii) When the examination volume defined in paragraph (b)(2)(xv)(K)(2)(i) of this section cannot be effectively examined in all four directions, the examination must be augmented by examination from the nozzle bore using a procedure and personnel qualified in accordance with paragraph (b)(2)(xv)(K)(1) of this section.
(iii) The remainder of the examination volume not covered by paragraph (b)(2)(xv)(K)(2)(ii) of this section or a combination of paragraphs (b)(2)(xv)(K)(2)(i) and (ii) of this section, must be examined from the nozzle bore using a procedure and personnel qualified in accordance with paragraph (b)(2)(xv)(K)(1) of this section, or from the vessel shell using a procedure and personnel qualified for single sided examination in accordance with Supplement 6 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (G) of this section.
(3) For examination of reactor pressure vessel nozzle-to-shell welds conducted from the outside of the vessel, the following conditions are required:
(i) The clad-to-base-metal-interface and the adjacent metal to a depth of 15 percent T (measured from the clad-to-base-metal-interface) must be examined from one radial and two opposing circumferential directions using a procedure and personnel qualified in accordance with Supplement 4 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(B) and (C) of this section, for examinations performed in the radial direction, and Supplement 5 to Appendix VIII, as conditioned by paragraph (b)(2)(xv)(J) of this section, for examinations performed in the circumferential direction.
(ii) The examination volume not addressed by paragraph (b)(2)(xv)(K)(3)(i) of this section must be examined in a minimum of one radial direction using a procedure and personnel qualified for single sided examination in accordance with Supplement 6 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (G) of this section.
(4) Table VIII-S7-1, ``Flaw Locations and Orientations,'' Supplement 7 to Appendix VIII, is conditioned as follows:
Table VIII--S7-1--Modified
[Flaw locations and orientations]----------------------------------------------------------------------------------------------------------------
Parallel to weld Perpendicular to weld----------------------------------------------------------------------------------------------------------------Inner 15 percent............................................ X XOutside Diameter Surface.................................... XSubsurface.................................................. X ........................----------------------------------------------------------------------------------------------------------------
(L) Specimen set and qualification: Twelfth provision. As a condition to the requirements of Supplement 8, Subparagraph 1.1(c), to Appendix VIII, notches may be located within one diameter of each end of the bolt or stud.
(M) Specimen set and qualification: Thirteenth provision. When implementing Supplement 12 to Appendix VIII, only the provisions related to the coordinated implementation of Supplement 3 to Supplement 2 performance demonstrations are to be applied.
(xvi) Section XI condition: Appendix VIII single side ferritic vessel and piping and stainless steel piping examinations. When applying editions and addenda prior to the 2007 Edition of Section XI, the following conditions apply.
(A) Ferritic and stainless steel piping examinations: First provision. Examinations performed from one side of a ferritic vessel weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations. To demonstrate equivalency to two sided examinations, the demonstration must be performed to the requirements of Appendix VIII, as conditioned by this paragraph and paragraphs (b)(2)(xv)(B) through (G) of this section, on specimens containing flaws with non-optimum sound energy reflecting characteristics or flaws similar to those in the vessel being examined.
(B) Ferritic and stainless steel piping examinations: Second provision. Examinations performed from one side of a ferritic or stainless steel pipe weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations. To demonstrate equivalency to two sided examinations, the demonstration must be performed to the requirements of Appendix VIII, as conditioned by this paragraph and paragraph (b)(2)(xv)(A) of this section.
(xvii) Section XI condition: Reconciliation of quality requirements. When purchasing replacement items, in addition to the reconciliation provisions of IWA-4200, 1995 Addenda through 1998 Edition, the replacement items must be purchased, to the extent necessary, in accordance with the licensee's quality assurance program description required by 10 CFR 50.34(b)(6)(ii).
(xviii) Section XI condition: NDE personnel certification. (A) NDE personnel certification: First provision. Level I and II nondestructive examination personnel must be recertified on a 3-year interval in lieu of the 5-year interval specified in the 1997 Addenda and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section.
(A) NDE personnel certification: First provision. Level I and II nondestructive examination personnel must be recertified on a 3-year interval in lieu of the 5-year interval specified in the 1997 Addenda and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section.
(B) NDE personnel certification: Second provision. When applying editions and addenda prior to the 2007 Edition of Section XI, paragraph IWA-2316 may only be used to qualify personnel that observe leakage during system leakage and hydrostatic tests conducted in accordance with IWA 5211(a) and (b).
(C) NDE personnel certification: Third provision. When applying editions and addenda prior to the 2005 Addenda of Section XI, licensee's qualifying visual examination personnel for VT-3 visual examination under paragraph IWA-2317 of Section XI must demonstrate the proficiency of the training by administering an initial qualification examination and administering subsequent examinations on a 3-year interval.
(xix) Section XI condition: Substitution of alternative methods. The provisions for substituting alternative examination methods, a combination of methods, or newly developed techniques in the 1997 Addenda of IWA-2240 must be applied when using the 1998 Edition through the 2004 Edition of Section XI of the ASME BPV Code. The provisions in IWA-4520(c), 1997 Addenda through the 2004 Edition, allowing the substitution of alternative methods, a combination of methods, or newly developed techniques for the methods specified in the Construction Code, are not approved for use. The provisions in IWA-4520(b)(2) and IWA-4521 of the 2008 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, allowing the substitution of ultrasonic examination for radiographic examination specified in the Construction Code, are not approved for use.
(xx) Section XI condition: System leakage tests--(A) System leakage tests: First provision. When performing system leakage tests in accordance with IWA-5213(a), 1997 through 2002 Addenda, the licensee must maintain a 10-minute hold time after test pressure has been reached for Class 2 and Class 3 components that are not in use during normal operating conditions. No hold time is required for the remaining Class 2 and Class 3 components provided that the system has been in operation for at least 4 hours for insulated components or 10 minutes for uninsulated components.
(A) System leakage tests: First provision. When performing system leakage tests in accordance with IWA-5213(a), 1997 through 2002 Addenda, the licensee must maintain a 10-minute hold time after test pressure has been reached for Class 2 and Class 3 components that are not in use during normal operating conditions. No hold time is required for the remaining Class 2 and Class 3 components provided that the system has been in operation for at least 4 hours for insulated components or 10 minutes for uninsulated components.
(B) System leakage tests: Second provision. The NDE provision in IWA-4540(a)(2) of the 2002 Addenda of Section XI must be applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section.
(xxi) Section XI condition: Table IWB-2500-1 examination requirements. (A) Table IWB-2500-1 examination requirements: First provision. The provisions of Table IWB 2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 (Inspection Program A) and Items B3.120 and B3.140 (Inspection Program B) of the 1998 Edition must be applied when using the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section. A visual examination with magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, with a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed instead of an ultrasonic examination.
(A) Table IWB-2500-1 examination requirements: First provision. The provisions of Table IWB 2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 (Inspection Program A) and Items B3.120 and B3.140 (Inspection Program B) of the 1998 Edition must be applied when using the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section. A visual examination with magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, with a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed instead of an ultrasonic examination.
(B) [Reserved]
(xxii) Section XI condition: Surface examination. The use of the provision in IWA-2220, ``Surface Examination,'' of Section XI, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, that allows use of an ultrasonic examination method is prohibited.
(xxiii) Section XI condition: Evaluation of thermally cut surfaces. The use of the provisions for eliminating mechanical processing of thermally cut surfaces in IWA-4461.4.2 of Section XI, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, is prohibited.
(xxiv) Section XI condition: Incorporation of the performance demonstration initiative and addition of ultrasonic examination criteria. The use of Appendix VIII and the supplements to Appendix VIII and Article I-3000 of Section XI of the ASME BPV Code, 2002 Addenda through the 2006 Addenda, is prohibited.
(xxv) Section XI condition: Mitigation of defects by modification. The use of the provisions in IWA-4340, ``Mitigation of Defects by Modification,'' Section XI, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section are prohibited.
(xxvi) Section XI condition: Pressure testing Class 1, 2 and 3 mechanical joints. The repair and replacement activity provisions in IWA-4540(c) of the 1998 Edition of Section XI for pressure testing Class 1, 2, and 3 mechanical joints must be applied when using the 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section.
(xxvii) Section XI condition: Removal of insulation. When performing visual examination in accordance with IWA-5242 of Section XI of the ASME BPV Code, 2003 Addenda through the 2006 Addenda, or IWA-5241 of the 2007 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, insulation must be removed from 17-4 PH or 410 stainless steel studs or bolts aged at a temperature below 1100[emsp14] [deg]F or having a Rockwell Method C hardness value above 30, and from A-286 stainless steel studs or bolts preloaded to 100,000 pounds per square inch or higher.
(xxviii) Section XI condition: Analysis of flaws. Licensees using ASME BPV Code, Section XI, Appendix A, must use the following conditions when implementing Equation (2) in A-4300(b)(1):
For R < 0, [Delta]KI depends on the crack depth (a), and the flow stress ([sigma]f). The flow stress is defined by [sigma]f = 1/2([sigma]ys + [sigma]ult), where [sigma]ys is the yield strength and [sigma]ult is the ultimate tensilestrength in units ksi (MPa) and (a) is in units in. (mm). For -2 <= R <= 0 and Kmax- Kmin <= 0.8 x 1.12 [sigma]f[radic]([pi]a), S = 1 and [Delta]KI = Kmax. For R < -2 and Kmax- Kmin <= 0.8x 1.12 [sigma]f[radic]([pi]a), S = 1 and [Delta]KI = (1 - R) Kmax/3. For R < 0 and Kmax - Kmin
0.8 x 1.12[sigma]f[radic]([pi]a), S = 1 and [Delta]KI = Kmax-Kmin.
(xxix) Section XI condition: Nonmandatory Appendix R. Nonmandatory Appendix R, ``Risk-Informed Inspection Requirements for Piping,'' of Section XI, 2005 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, may not be implemented without prior NRC authorization of the proposed alternative in accordance with paragraph (z) of this section.
(3) Conditions on ASME OM Code. As used in this section, references to the OM Code refer to the ASME Code for Operation and Maintenance of Nuclear Power Plants, Subsections ISTA, ISTB, ISTC, ISTD, Mandatory Appendices I and II, and Nonmandatory Appendices A through H and J, including the 1995 Edition through the 2006 Addenda, subject to the following conditions:
(i) OM condition: Quality assurance. When applying editions and addenda of the OM Code, the requirements of NQA-1, ``Quality Assurance Requirements for Nuclear Facilities,'' 1979 Addenda, are acceptable as permitted by ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA-1500 of the 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section, provided the licensee uses its 10 CFR part 50, Appendix B, quality assurance program in conjunction with the OM Code requirements. Commitments contained in the licensee's quality assurance program description that are more stringent than those contained in NQA-1 govern OM Code activities. If NQA-1 and the OM Code do not address the commitments contained in the licensee's Appendix B quality assurance program description, the commitments must be applied to OM Code activities.
(ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees must comply with the provisions for MOV testing in OM Code ISTC 4.2, 1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section, and must establish a program to ensure that motor-operated valves continue to be capable of performing their design basis safety functions.
(iii) [Reserved]
(iv) OM condition: Check valves (Appendix II). Licensees applying Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM Code, 1995 Edition with the 1996 and 1997 Addenda, must satisfy the requirements of (b)(3)(iv)(A) through (C) of this section. Licensees applying Appendix II, 1998 Edition through the 2002 Addenda, must satisfy the requirements of (b)(3)(iv)(A), (B), and (D) of this section.
(A) Check valves: First provision. Valve opening and closing functions must be demonstrated when flow testing or examination methods (nonintrusive, or disassembly and inspection) are used;
(B) Check valves: Second provision. The initial interval for tests and associated examinations may not exceed two fuel cycles or 3 years, whichever is longer; any extension of this interval may not exceed one fuel cycle per extension with the maximum interval not to exceed 10 years. Trending and evaluation of existing data must be used to reduce or extend the time interval between tests.
(C) Check valves: Third provision. If the Appendix II condition monitoring program is discontinued, then the requirements of ISTC 4.5.1 through 4.5.4 must be implemented.
(D) Check valves: Fourth provision. The applicable provisions of subsection ISTC must be implemented if the Appendix II condition monitoring program is discontinued.
(v) OM condition: Snubbers ISTD. Article IWF-5000, ``Inservice Inspection Requirements for Snubbers,'' of the ASME BPV Code, Section XI, must be used when performing inservice inspection examinations and tests of snubbers at nuclear power plants, except as conditioned in paragraphs (b)(3)(v)(A) and (B) of this section.
(A) Snubbers: First provision. Licensees may use Subsection ISTD, ``Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Power Plants,'' ASME OM Code, 1995 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section, in place of the requirements for snubbers in the editions and addenda up to the 2005 Addenda of the ASME BPV Code, Section XI, IWF-5200(a) and (b) and IWF-5300(a) and (b), by making appropriate changes to their technical specifications or licensee-controlled documents. Preservice and inservice examinations must be performed using the VT-3 visual examination method described in IWA-2213.
(B) Snubbers: Second provision. Licensees must comply with the provisions for examining and testing snubbers in Subsection ISTD of the ASME OM Code and make appropriate changes to their technical specifications or licensee-controlled documents when using the 2006 Addenda and later editions and addenda of Section XI of the ASME BPV Code.
(vi) OM condition: Exercise interval for manual valves. Manual valves must be exercised on a 2-year interval rather than the 5-year interval specified in paragraph ISTC-3540 of the 1999 through the 2005 Addenda of the ASME OM Code, provided that adverse conditions do not require more frequent testing.
(4) Conditions on Design, Fabrication, and Materials Code Cases. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter is subject to the following conditions. Licensees may apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.84, Revision 36, without prior NRC approval, subject to the following conditions:
(i) Design, Fabrication, and Materials Code Case condition: Applying Code Cases. When an applicant or licensee initially applies a listed Code Case, the applicant or licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section.
(ii) Design, Fabrication, and Materials Code Case condition: Applying different revisions of Code Cases. If an applicant or licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the applicant or licensee may continue to apply the previous version of the Code Case as authorized or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use, until it updates its Code of Record for the component being constructed.
(iii) Design, Fabrication, and Materials Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless an applicant or licensee applied the listed Code Case prior to it being listed as annulled in Regulatory Guide 1.84. If an applicant or licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.84, the applicant or licensee may continue to apply the Code Case until it updates its Code of Record for the component being constructed.
(5) Conditions on inservice inspection Code Cases. Licensees may apply the ASME BPV Code Cases listed in Regulatory Guide 1.147, Revision 17, without prior NRC approval, subject to the following:
(i) ISI Code Case condition: Applying Code Cases. When a licensee initially applies a listed Code Case, the licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section.
(ii) ISI Code Case condition: Applying different revisions of Code Cases. If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case, as authorized, or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use. Licensees who choose to continue use of the Code Case during subsequent 120-month ISI program intervals will be required to implement the latest version incorporated by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of Regulatory Guide 1.147, Revision 17.
(iii) ISI Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in Regulatory Guide 1.147. If a licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.147, the licensee may continue to apply the Code Case to the end of the current 120-month interval.
(6) Conditions on Operation and Maintenance of Nuclear Power Plants Code Cases. Licensees may apply the ASME Operation and Maintenance Code Cases listed in Regulatory Guide 1.192, Revision 1, without prior NRC approval, subject to the following:
(i) OM Code Case condition: Applying Code Cases. When a licensee initially applies a listed Code Case, the licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section.
(ii) OM Code Case condition: Applying different revisions of Code Cases. If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case, as authorized, or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use. Licensees who choose to continue use of the Code Case during subsequent 120-month ISI program intervals will be required to implement the latest version incorporated by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of Regulatory Guide 1.192, Revision 1.
(iii) OM Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in Regulatory Guide 1.192. If a licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.192, the licensee may continue to apply the Code Case to the end of the current 120-month interval.
(c) Reactor coolant pressure boundary. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter and each combined license for a utilization facility is subject to the following conditions:
(1) Standards requirement for reactor coolant pressure boundary components. Components that are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section III \1,4\ of the ASME BPV Code, except as provided in paragraphs (c)(2) through (4) of this section.
(2) Exceptions to reactor coolant pressure boundary standards requirement. Components that are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in Sec. 50.2 need not meet the requirements of paragraph (c)(1) of this section, provided that:
(i) Exceptions: Shutdown and cooling capability. In the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system; or
(ii) Exceptions: Isolation capability. The component is or can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only.
(3) Applicable Code and Code Cases and conditions on their use. The Code edition, addenda, and optional ASME Code Cases to be applied to components of the reactor coolant pressure boundary must be determined by the provisions of paragraph NCA-1140, Subsection NCA of Section III of the ASME BPV Code, subject to the following conditions:
(i) Reactor coolant pressure boundary condition: Code edition and addenda. The edition and addenda applied to a component must be those that are incorporated by reference in paragraph (a)(1)(i) of this section;
(ii) Reactor coolant pressure boundary condition: Earliest edition and addenda for pressure vessel. The ASME Code provisions applied to the pressure vessel may be dated no earlier than the summer 1972 Addenda of the 1971 Edition;
(iii) Reactor coolant pressure boundary condition: Earliest edition and addenda for piping, pumps, and valves. The ASME Code provisions applied to piping, pumps, and valves may be dated no earlier than the Winter 1972 Addenda of the 1971 Edition; and
(iv) Reactor coolant pressure boundary condition: Use of Code Cases. The optional Code Cases applied to a component must be those listed in NRC Regulatory Guide 1.84 that is incorporated by reference in paragraph (a)(3)(i) of this section.
(4) Standards requirement for components in older plants. For a nuclear power plant whose construction permit was issued prior to May 14, 1984, the applicable Code edition and addenda for a component of the reactor coolant pressure boundary continue to be that Code edition and addenda that were required by Commission regulations for such a component at the time of issuance of the construction permit.
(d) Quality Group B components. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter, and each combined license for a utilization facility is subject to the following conditions:
(1) Standards requirement for Quality Group B components. For a nuclear power plant whose application for a construction permit under this part, or a combined license or manufacturing license under part 52 of this chapter, docketed after May 14, 1984, or for an application for a standard design approval or a standard design certification docketed after May 14, 1984, components classified Quality Group B \7\ must meet the requirements for Class 2 Components in Section III of the ASME BPV Code.
(2) Quality Group B: Applicable Code and Code Cases and conditions on their use. The Code edition, addenda, and optional ASME Code Cases to be applied to the systems and components identified in paragraph (d)(1) of this section must be determined by the rules of paragraph NCA-1140, Subsection NCA of Section III of the ASME BPV Code, subject to the following conditions:
(i) Quality Group B condition: Code edition and addenda. The edition and addenda must be those that are incorporated by reference in paragraph (a)(1)(i) of this section;
(ii) Quality Group B condition: Earliest edition and addenda for components. The ASME Code provisions applied to the systems and components may be dated no earlier than the 1980 Edition; and
(iii) Quality Group B condition: Use of Code Cases. The optional Code Cases must be those listed in NRC Regulatory Guide 1.84 that is incorporated by reference in paragraph (a)(3)(i) of this section.
(e) Quality Group C components. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter and each combined license for a utilization facility is subject to the following conditions.
(1) Standards requirement for Quality Group C components. For a nuclear power plant whose application for a construction permit under this part, or a combined license or manufacturing license under part 52 of this chapter, docketed after May 14, 1984, or for an application for a standard design approval or a standard design certification docketed after May 14, 1984, components classified Quality Group C \7\ must meet the requirements for Class 3 components in Section III of the ASME BPV Code.
(2) Quality Group C applicable Code and Code Cases and conditions on their use. The Code edition, addenda, and optional ASME Code Cases to be applied to the systems and components identified in paragraph (e)(1) of this section must be determined by the rules of paragraph NCA-1140, subsection NCA of Section III of the ASME BPV Code, subject to the following conditions:
(i) Quality Group C condition: Code edition and addenda. The edition and addenda must be those incorporated by reference in paragraph (a)(1)(i) of this section;
(ii) Quality Group C condition: Earliest edition and addenda for components. The ASME Code provisions applied to the systems and components may be dated no earlier than the 1980 Edition; and
(iii) Quality Group C condition: Use of Code Cases. The optional Code Cases must be those listed in NRC Regulatory Guide 1.84 that is incorporated by reference in paragraph (a)(3)(i) of this section.
(f) Inservice testing requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions. Each combined license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (f)(4) through (6) of this section must be met only after the Commission makes the finding under Sec. 52.103(g) of this chapter. Requirements for inservice inspection of Class 1, Class 2, Class 3, Class MC, and Class CC components (including their supports) are located in Sec. 50.55a(g).
(1) Inservice testing requirements for older plants (pre-1971 CPs). For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, pumps and valves must meet the test requirements of paragraphs (f)(4) and (5) of this section to the extent practical. Pumps and valves that are part of the reactor coolant pressure boundary must meet the requirements applicable to components that are classified as ASME Code Class 1. Other pumps and valves that perform a function to shut down the reactor or maintain the reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for safety-related systems (in meeting the requirements of the 1986 Edition, or later, of the BPV or OM Code) must meet the test requirements applicable to components that are classified as ASME Code Class 2 or Class 3.
(2) Design and accessibility requirements for performing inservice testing in plants with CPs issued between 1971 and 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, pumps and valves that are classified as ASME Code Class 1 and Class 2 must be designed and provided with access to enable the performance of inservice tests for operational readiness set forth in editions and addenda of Section XI of the ASME BPV incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively) in effect 6 months before the date of issuance of the construction permit. The pumps and valves may meet the inservice test requirements set forth in subsequent editions of this Code and addenda that are incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17; or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the applicable conditions listed therein.
(3) Design and accessibility requirements for performing inservice testing in plants with CPs issued after 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit under this part or design approval, design certification, combined license, or manufacturing license under part 52 of this chapter was issued on or after July 1, 1974:
(i)-(ii) [Reserved]
(iii) IST design and accessibility requirements: Class 1 pumps and valves. (A) Class 1 pumps and valves: First provision. In facilities whose construction permit was issued before November 22, 1999, pumps and valves that are classified as ASME Code Class 1 must be designed and provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively) applied to the construction of the particular pump or valve or the summer 1973 Addenda, whichever is later.
(A) Class 1 pumps and valves: First provision. In facilities whose construction permit was issued before November 22, 1999, pumps and valves that are classified as ASME Code Class 1 must be designed and provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively) applied to the construction of the particular pump or valve or the summer 1973 Addenda, whichever is later.
(B) Class 1 pumps and valves: Second provision. In facilities whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, issued on or after November 22, 1999, pumps and valves that are classified as ASME Code Class 1 must be designed and provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in editions and addenda of the ASME OM Code (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraph (a)(3)(iii) of this section), incorporated by reference in paragraph (a)(1)(iv) of this section at the time the construction permit, combined license, manufacturing license, design certification, or design approval is issued.
(iv) IST design and accessibility requirements: Class 2 and 3 pumps and valves. (A) Class 2 and 3 pumps and valves: First provision. In facilities whose construction permit was issued before November 22, 1999, pumps and valves that are classified as ASME Code Class 2 and Class 3 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later.
(A) Class 2 and 3 pumps and valves: First provision. In facilities whose construction permit was issued before November 22, 1999, pumps and valves that are classified as ASME Code Class 2 and Class 3 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later.
(B) Class 2 and 3 pumps and valves: Second provision. In facilities whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, issued on or after November 22, 1999, pumps and valves that are classified as ASME Code Class 2 and 3 must be designed and provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in editions and addenda of the ASME OM Code (or the optional ASME OM Code Cases listed in NRC Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraph (a)(3)(iii) of this section), incorporated by reference in paragraph (a)(1)(iv) of this section at the time the construction permit, combined license, or design certification is issued.
(v) IST design and accessibility requirements: Meeting later IST requirements. All pumps and valves may meet the test requirements set forth in subsequent editions of codes and addenda or portions thereof that are incorporated by reference in paragraph (a) of this section, subject to the conditions listed in paragraph (b) of this section.
(4) Inservice testing standards requirement for operating plants. Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and (3) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
(i) Applicable IST Code: Initial 120-month interval. Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during the initial 120-month interval must comply with the requirements in the latest edition and addenda of the OM Code incorporated by reference in paragraph (a)(1)(iv) of this section on the date 12 months before the date of issuance of the operating license under this part, or 12 months before the date scheduled for initial loading of fuel under a combined license under part 52 of this chapter (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.192, Revision 1, that is incorporated by reference in paragraph (a)(3)(iii) of this section, subject to the conditions listed in paragraph (b) of this section).
(ii) Applicable IST Code: Successive 120-month intervals. Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 12 months before the start of the 120-month interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the conditions listed in paragraph (b) of this section.
(iii) [Reserved]
(iv) Applicable IST Code: Use of later Code editions and addenda. Inservice tests of pumps and valves may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (a)(1)(iv) of this section, subject to the conditions listed in paragraph (b) of this section, and subject to NRC approval. Portions of editions or addenda may be used, provided that all related requirements of the respective editions or addenda are met.
(5) Requirements for updating IST programs--(i) IST program update: Applicable IST Code editions and addenda. The inservice test program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (f)(4) of this section.
(i) IST program update: Applicable IST Code editions and addenda. The inservice test program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (f)(4) of this section.
(ii) IST program update: Conflicting IST Code requirements with technical specifications. If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program. The licensee must submit this application, as specified in Sec. 50.4, at least 6 months before the start of the period during which the provisions become applicable, as determined by paragraph (f)(4) of this section.
(iii) IST program update: Notification of impractical IST Code requirements. If the licensee has determined that conformance with certain Code requirements is impractical for its facility, the licensee must notify the Commission and submit, as specified in Sec. 50.4, information to support the determination.
(iv) IST program update: Schedule for completing impracticality determinations. Where a pump or valve test requirement by the Code or addenda is determined to be impractical by the licensee and is not included in the revised inservice test program (as permitted by paragraph (f)(4) of this section), the basis for this determination must be submitted for NRC review and approval not later than 12 months after the expiration of the initial 120-month interval of operation from the start of facility commercial operation and each subsequent 120-month interval of operation during which the test is determined to be impractical.
(6) Actions by the Commission for evaluating impractical and augmented IST Code requirements--(i) Impractical IST requirements: Granting of relief. The Commission will evaluate determinations under paragraph (f)(5) of this section that code requirements are impractical. The Commission may grant relief and may impose such alternative requirements as it determines are authorized by law, will not endanger life or property or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
(i) Impractical IST requirements: Granting of relief. The Commission will evaluate determinations under paragraph (f)(5) of this section that code requirements are impractical. The Commission may grant relief and may impose such alternative requirements as it determines are authorized by law, will not endanger life or property or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
(ii) Augmented IST requirements. The Commission may require the licensee to follow an augmented inservice test program for pumps and valves for which the Commission deems that added assurance of operational readiness is necessary.
(g) Inservice inspection requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions. Each combined license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (g)(4) through (6) of this section must be met only after the Commission makes the finding under Sec. 52.103(g) of this chapter. Requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in Sec. 50.55a(f).
(1) Inservice inspection requirements for older plants (pre-1971 CPs). For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued before January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (g)(5) of this section to the extent practical. Components that are part of the reactor coolant pressure boundary and their supports must meet the requirements applicable to components that are classified as ASME Code Class 1. Other safety-related pressure vessels, piping, pumps and valves, and their supports must meet the requirements applicable to components that are classified as ASME Code Class 2 or Class 3.
(2) Design and accessibility requirements for performing inservice inspection in plants with CPs issued between 1971 and 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, components (including supports) that are classified as ASME Code Class 1 and Class 2 must be designed and be provided with access to enable the performance of inservice examination of such components (including supports) and must meet the preservice examination requirements set forth in editions and addenda of Section III or Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) in effect 6 months before the date of issuance of the construction permit. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in paragraph (a) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section), subject to the applicable limitations and modifications.
(3) Design and accessibility requirements for performing inservice inspection in plants with CPs issued after 1974. For a boiling or pressurized water-cooled nuclear power facility, whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, was issued on or after July 1, 1974, the following are required:
(i) ISI design and accessibility requirements: Class 1 components and supports. Components (including supports) that are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice examination of these components and must meet the preservice examination requirements set forth in the editions and addenda of Section III or Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular component.
(ii) ISI design and accessibility requirements: Class 2 and 3 components and supports. Components that are classified as ASME Code Class 2 and Class 3 and supports for components that are classified as ASME Code Class 1, Class 2, and Class 3 must be designed and provided with access to enable the performance of inservice examination of these components and must meet the preservice examination requirements set forth in the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular component.
(iii)-(iv) [Reserved]
(v) ISI design and accessibility requirements: Meeting later ISI requirements. All components (including supports) may meet the requirements set forth in subsequent editions of codes and addenda or portions thereof that are incorporated by reference in paragraph (a) of this section, subject to the conditions listed therein.
(4) Inservice inspection standards requirement for operating plants. Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME BPV Code (or ASME OM Code for snubber examination and testing) that become effective subsequent to editions specified in paragraphs (g)(2) and (3) of this section and that are incorporated by reference in paragraph (a)(1)(ii) or (iv) for snubber examination and testing of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components. Components that are classified as Class MC pressure retaining components and their integral attachments, and components that are classified as Class CC pressure retaining components and their integral attachments, must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of the ASME BPV Code and addenda that are incorporated by reference in paragraph (a)(1)(ii) of this section, subject to the condition listed in paragraph (b)(2)(vi) of this section and the conditions listed in paragraphs (b)(2)(viii) and (ix) of this section, to the extent practical within the limitation of design, geometry, and materials of construction of the components.
(i) Applicable ISI Code: Initial 120-month interval. Inservice examination of components and system pressure tests conducted during the initial 120-month inspection interval must comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section on the date 12 months before the date of issuance of the operating license under this part, or 12 months before the date scheduled for initial loading of fuel under a combined license under part 52 of this chapter (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using Section XI, or Regulatory Guide 1.192, Revision 1, when using the OM Code, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the conditions listed in paragraph (b) of this section.
(ii) Applicable ISI Code: Successive 120-month intervals. Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using Section XI, or Regulatory Guide 1.192, Revision 1, when using the OM Code, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section), subject to the conditions listed in paragraph (b) of this section. However, a licensee whose inservice inspection interval commences during the 12 through 18-month period after July 21, 2011, may delay the update of their Appendix VIII program by up to 18 months after July 21, 2011.
(iii) Applicable ISI Code: Optional surface examination requirement. When applying editions and addenda prior to the 2003 Addenda of Section XI of the ASME BPV Code, licensees may, but are not required to, perform the surface examinations of high-pressure safety injection systems specified in Table IWB-2500-1, Examination Category B-J, Item Numbers B9.20, B9.21, and B9.22.
(iv) Applicable ISI Code: Use of subsequent Code editions and addenda. Inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (a) of this section, subject to the conditions listed in paragraph (b) of this section, and subject to Commission approval. Portions of editions or addenda may be used, provided that all related requirements of the respective editions or addenda are met.
(v) Applicable ISI Code: Metal and concrete containments. For a boiling or pressurized water-cooled nuclear power facility whose construction permit under this part or combined license under part 52 of this chapter was issued after January 1, 1956, the following are required:
(A) Metal and concrete containments: First provision. Metal containment pressure retaining components and their integral attachments must meet the inservice inspection, repair, and replacement requirements applicable to components that are classified as ASME Code Class MC;
(B) Metal and concrete containments: Second provision. Metallic shell and penetration liners that are pressure retaining components and their integral attachments in concrete containments must meet the inservice inspection, repair, and replacement requirements applicable to components that are classified as ASME Code Class MC; and
(C) Metal and concrete containments: Third provision. Concrete containment pressure retaining components and their integral attachments, and the post-tensioning systems of concrete containments, must meet the inservice inspections, repair, and replacement requirements applicable to components that are classified as ASME Code Class CC.
(5) Requirements for updating ISI programs--(i) ISI program update: Applicable ISI Code editions and addenda. The inservice inspection program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section.
(i) ISI program update: Applicable ISI Code editions and addenda. The inservice inspection program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section.
(ii) ISI program update: Conflicting ISI Code requirements with technical specifications. If a revised inservice inspection program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program. The licensee must submit this application, as specified in Sec. 50.4, at least six months before the start of the period during which the provisions become applicable, as determined by paragraph (g)(4) of this section.
(iii) ISI program update: Notification of impractical ISI Code requirements. If the licensee has determined that conformance with a Code requirement is impractical for its facility the licensee must notify the NRC and submit, as specified in Sec. 50.4, information to support the determinations. Determinations of impracticality in accordance with this section must be based on the demonstrated limitations experienced when attempting to comply with the Code requirements during the inservice inspection interval for which the request is being submitted. Requests for relief made in accordance with this section must be submitted to the NRC no later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought.
(iv) ISI program update: Schedule for completing impracticality determinations. Where the licensee determines that an examination required by Code edition or addenda is impractical, the basis for this determination must be submitted for NRC review and approval not later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought.
(6) Actions by the Commission for evaluating impractical and augmented ISI Code requirements--(i) Impractical ISI requirements: Granting of relief. The Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines are authorized by law, will not endanger life or property or the common defense and security, and are otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
(i) Impractical ISI requirements: Granting of relief. The Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines are authorized by law, will not endanger life or property or the common defense and security, and are otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
(ii) Augmented ISI program. The Commission may require the licensee to follow an augmented inservice inspection program for systems and components for which the Commission deems that added assurance of structural reliability is necessary.
(A) [Reserved]
(B) Augmented ISI requirements: Submitting containment ISI programs. Licensees do not have to submit to the NRC for approval of their containment inservice inspection programs that were developed to satisfy the requirements of Subsection IWE and Subsection IWL with specified conditions. The program elements and the required documentation must be maintained on site for audit.
(C) Augmented ISI requirements: Implementation of Appendix VIII to Section XI. (1) Appendix VIII and the supplements to Appendix VIII to Section XI, Division 1, 1995 Edition with the 1996 Addenda of the ASME BPV Code must be implemented in accordance with the following schedule: Appendix VIII and Supplements 1, 2, 3, and 8--May 22, 2000; Supplements 4 and 6--November 22, 2000; Supplement 11--November 22, 2001; and Supplements 5, 7, and 10--November 22, 2002.
(2) Licensees implementing the 1989 Edition and earlier editions and addenda of IWA-2232 of Section XI, Division 1, of the ASME BPV Code must implement the 1995 Edition with the 1996 Addenda of Appendix VIII and the supplements to Appendix VIII of Section XI, Division 1, of the ASME BPV Code.
(D) Augmented ISI requirements: Reactor vessel head inspections--(1) All licensees of pressurized water reactors must augment their inservice inspection program with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section. Licensees of existing operating reactors as of September 10, 2008, must implement their augmented inservice inspection program by December 31, 2008. Once a licensee implements this requirement, the First Revised NRC Order EA-03-009 no longer applies to that licensee and shall be deemed to be withdrawn.
(2) Note 9 of ASME Code Case N-729-1 must not be implemented.
(3) Instead of the specified ``examination method'' requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee must perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1. A demonstrated volumetric or surface leak path assessment through all J-groove welds must be performed. If a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface examination must be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.
(4) By September 1, 2009, ultrasonic examinations must be performed using personnel, procedures, and equipment that have been qualified by blind demonstration on representative mockups using a methodology that meets the conditions specified in paragraphs (g)(6)(ii)(D)(4)(i) through (iv), instead of the qualification requirements of Paragraph -2500 of ASME Code Case N-729-1. References herein to Section XI, Appendix VIII, must be to the 2004 Edition with no addenda of the ASME BPV Code.
(i) The specimen set must have an applicable thickness qualification range of +25 percent to -40 percent for nominal depth through-wall thickness. The specimen set must include geometric and material conditions that normally require discrimination from primary water stress corrosion cracking (PWSCC) flaws.
(ii) The specimen set must have a minimum of ten (10) flaws that provide an acoustic response similar to PWSCC indications. All flaws must be greater than 10 percent of the nominal pipe wall thickness. A minimum of 20 percent of the total flaws must initiate from the inside surface and 20 percent from the outside surface. At least 20 percent of the flaws must be in the depth ranges of 10-30 percent through-wall thickness and at least 20 percent within a depth range of 31-50 percent through-wall thickness. At least 20 percent and no more than 60 percent of the flaws must be oriented axially.
(iii) Procedures must identify the equipment and essential variables and settings used for the qualification, in accordance with Subarticle VIII-2100 of Section XI, Appendix VIII. The procedure must be requalified when an essential variable is changed outside the demonstration range as defined by Subarticle VIII-3130 of Section XI, Appendix VIII, and as allowed by Articles VIII-4100, VIII-4200, and VIII-4300 of Section XI, Appendix VIII. Procedure qualification must include the equivalent of at least three personnel performance demonstration test sets. Procedure qualification requires at least one successful personnel performance demonstration.
(iv) Personnel performance demonstration test acceptance criteria must meet the personnel performance demonstration detection test acceptance criteria of Table VIII--S10-1 of Section XI, Appendix VIII, Supplement 10. Examination procedures, equipment, and personnel are qualified for depth sizing and length sizing when the RMS error, as defined by Subarticle VIII-3120 of Section XI, Appendix VIII, of the flaw depth measurements, as compared to the true flaw depths, do not exceed \1/8\ inch (3 mm) and the root mean square (RMS) error of the flaw length measurements, as compared to the true flaw lengths, do not exceed \3/8\ inch (10 mm), respectively.
(5) If flaws attributed to PWSCC have been identified, whether acceptable or not for continued service under Paragraphs -3130 or -3140 of ASME Code Case N-729-1, the re-inspection interval must be each refueling outage instead of the re-inspection intervals required by Table 1, Note (8), of ASME Code Case N-729-1.
(6) Appendix I of ASME Code Case N-729-1 must not be implemented without prior NRC approval.
(E) Augmented ISI requirements: Reactor coolant pressure boundary visual inspections \10\--(1) All licensees of pressurized water reactors must augment their inservice inspection program by implementing ASME Code Case N-722-1, subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) through (4) of this section. The inspection requirements of ASME Code Case N-722-1 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay or stress improvement.
(2) If a visual examination determines that leakage is occurring from a specific item listed in Table 1 of ASME Code Case N-722-1 that is not exempted by the ASME Code, Section XI, IWB-1220(b)(1), additional actions must be performed to characterize the location, orientation, and length of a crack or cracks in Alloy 600 nozzle wrought material and location, orientation, and length of a crack or cracks in Alloy 82/182 butt welds. Alternatively, licensees may replace the Alloy 600/82/182 materials in all the components under the item number of the leaking component.
(3) If the actions in paragraph (g)(6)(ii)(E)(2) of this section determine that a flaw is circumferentially oriented and potentially a result of primary water stress corrosion cracking, licensees must perform non-visual NDE inspections of components that fall under that ASME Code Case N-722-1 item number. The number of components inspected must equal or exceed the number of components found to be leaking under that item number. If circumferential cracking is identified in the sample, non-visual NDE must be performed in the remaining components under that item number.
(4) If ultrasonic examinations of butt welds are used to meet the NDE requirements in paragraphs (g)(6)(ii)(E)(2) or (3) of this section, they must be performed using the appropriate supplement of Section XI, Appendix VIII, of the ASME BPV Code.
(F) Augmented ISI requirements: Examination requirements for Class 1 piping and nozzle dissimilar-metal butt welds--(1) Licensees of existing, operating pressurized-water reactors as of July 21, 2011, must implement the requirements of ASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (10) of this section, by the first refueling outage after August 22, 2011.
(2) Full structural weld overlays authorized by the NRC staff may be categorized as Inspection Items C or F, as appropriate. Welds that have been mitigated by the Mechanical Stress Improvement Process (MSIP\TM\) may be categorized as Inspection Items D or E, as appropriate, provided the criteria in Appendix I of the Code Case have been met. For ISI frequencies, all other butt welds that rely on Alloy 82/182 for structural integrity must be categorized as Inspection Items A-1, A-2 or B until the NRC staff has reviewed the mitigation and authorized an alternative Code Case Inspection Item for the mitigated weld, or until an alternative Code Case Inspection Item is used based on conformance with an ASME mitigation Code Case endorsed in Regulatory Guide 1.147 with conditions, if applicable, and incorporated by reference in this section.
(3) Baseline examinations for welds in Table 1, Inspection Items A-1, A-2, and B, must be completed by the end of the next refueling outage after January 20, 2012. Previous examinations of these welds can be credited for baseline examinations if they were performed within the re-inspection period for the weld item in Table 1 using Section XI, Appendix VIII, requirements and met the Code required examination volume of essentially 100 percent. Other previous examinations that do not meet these requirements can be used to meet the baseline examination requirement, provided NRC approval of alternative inspection requirements in accordance with paragraphs (z)(1) or (2) of this section is granted prior to the end of the next refueling outage after January 20, 2012.
(4) The axial examination coverage requirements of Paragraph--2500(c) may not be considered to be satisfied unless essentially 100 percent coverage is achieved.
(5) All hot-leg operating temperature welds in Inspection Items G, H, J, and K must be inspected each inspection interval. A 25 percent sample of Inspection Items G, H, J, and K cold-leg operating temperature welds must be inspected whenever the core barrel is removed (unless it has already been inspected within the past 10 years) or 20 years, whichever is less.
(6) For any mitigated weld whose volumetric examination detects growth of existing flaws in the required examination volume that exceed the previous IWB-3600 flaw evaluations or new flaws, a report summarizing the evaluation, along with inputs, methodologies, assumptions, and causes of the new flaw or flaw growth is to be provided to the NRC prior to the weld being placed in service other than modes 5 or 6.
(7) For Inspection Items G, H, J, and K, when applying the acceptance standards of ASME BPV Code, Section XI, IWB-3514, for planar flaws contained within the inlay or onlay, the thickness ``t'' in IWB-3514 is the thickness of the inlay or onlay. For planar flaws in the balance of the dissimilar metal weld examination volume, the thickness ``t'' in IWB-3514 is the combined thickness of the inlay or onlay and the dissimilar metal weld.
(8) Welds mitigated by optimized weld overlays in Inspection Items D and E are not permitted to be placed into a population to be examined on a sample basis and must be examined once each inspection interval.
(9) Replace the first two sentences of Extent and Frequency of Examination for Inspection Item D in Table 1 of Code Case N-770-1 with, ``Examine all welds no sooner than the third refueling outage and no later than 10 years following stress improvement application.'' Replace the first two sentences of Note (11)(b)(2) in Code Case N-770-1 with, ``The first examination following weld inlay, onlay, weld overlay, or stress improvement for Inspection Items D through K must be performed as specified.''
(10) General Note (b) to Figure 5(a) of Code Case N-770-1 pertaining to alternative examination volume for optimized weld overlays may not be applied unless NRC approval is authorized under paragraphs (z)(1) or (2) of this section.
(h) Protection and safety systems. Protection systems of nuclear power reactors of all types must meet the requirements specified in this paragraph. Each combined license for a utilization facility is subject to the following conditions.
(1) [Reserved]
(2) Protection systems. For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements stated in either IEEE Std. 279, ``Criteria for Protection Systems for Nuclear Power Generating Stations,'' or in IEEE Std. 603-1991, ``Criteria for Safety Systems for Nuclear Power Generating Stations,'' and the correction sheet dated January 30, 1995. For nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.
(3) Safety systems. Applications filed on or after May 13, 1999, for construction permits and operating licenses under this part, and for design approvals, design certifications, and combined licenses under part 52 of this chapter, must meet the requirements for safety systems in IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.
(i)-(y) [Reserved]
(z) Alternatives to codes and standards requirements. Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:
(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or
(2) Hardship without a compensating increase in quality and safety. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Footnotes to Sec. 50.55a:
\1\ USAS and ASME Code addenda issued prior to the winter 1977 Addenda are considered to be ``in effect'' or ``effective'' 6 months after their date of issuance and after they are incorporated by reference in paragraph (a) of this section. Addenda to the ASME Code issued after the summer 1977 Addenda are considered to be ``in effect'' or ``effective'' after the date of publication of the addenda and after they are incorporated by reference in paragraph (a) of this section.
2-3 [Reserved].
\4\ For ASME Code editions and addenda issued prior to the winter 1977 Addenda, the Code edition and addenda applicable to the component is governed by the order or contract date for the component, not the contract date for the nuclear energy system. For the winter 1977 Addenda and subsequent editions and addenda the method for determining the applicable Code editions and addenda is contained in Paragraph NCA 1140 of Section III of the ASME Code.
5-6 [Reserved].
\7\ Guidance for quality group classifications of components that are to be included in the safety analysis reports pursuant to Sec. 50.34(a) and Sec. 50.34(b) may be found in Regulatory Guide 1.26, ``Quality Group Classifications and Standards for Water-, Steam-, and Radiological-Waste-Containing Components of Nuclear Power Plants,'' and in Section 3.2.2 of NUREG-0800, ``Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants.''
8-9 [Reserved].
\10\ For inspections to be conducted once per interval, the inspections must be performed in accordance with the schedule in Section XI, paragraph IWB-2400, except for plants with inservice inspection programs based on a Section XI edition or addenda prior to the 1994 Addenda. For plants with inservice inspection programs based on a Section XI edition or addenda prior to the 1994 Addenda, the inspection must be performed in accordance with the schedule in Section XI, paragraph IWB-2400, of the 1994 Addenda. [79 FR 65798, Nov. 5, 2014, as amended at 79 FR 66603, Nov. 10, 2014; 79 FR 73462, Dec. 11, 2014]