Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases

Federal RegisterJul 18, 2017

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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[NRC-2011-0088]

RIN 3150-AI97

Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Final rule.

SUMMARY:

The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to incorporate by reference recent editions and addenda to the American Society of Mechanical Engineers (ASME) Codes for nuclear power plants and a standard for quality assurance. The NRC is also incorporating by reference six ASME Code Cases. This action is in accordance with the NRC's policy to periodically update the regulations to incorporate by reference new editions and addenda of the ASME Codes and is intended to maintain the safety of nuclear power plants and to make NRC activities more effective and efficient.

DATES:

This final rule is effective on August 17, 2017. The incorporation by reference of certain publications listed in the regulation is approved by the Director of the Federal Register as of August 17, 2017.

ADDRESSES:

Please refer to Docket ID NRC-2011-0088 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

•

Federal Rulemaking Web site:

Go to

http://www.regulations.gov

and search for Docket ID NRC-2011-0088. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email:

Carol.Gallagher@nrc.gov.

For technical questions, contact the individuals listed in the

FOR FURTHER INFORMATION CONTACT

section of this document.

•

NRC's Agencywide Documents Access and Management System (ADAMS):

You may obtain publicly-available documents online in the ADAMS Public Documents collection at

http://www.nrc.gov/reading-rm/adams.html.

To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to

pdr.resource@nrc.gov.

For the convenience of the reader, instructions about obtaining materials referenced in this document are provided in the “Availability of Documents” section.

•

NRC's PDR:

You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

FOR FURTHER INFORMATION CONTACT:

Daniel I. Doyle, Office of Nuclear Reactor Regulation, telephone: 301-415-3748, email:

Daniel.Doyle@nrc.gov;

or Keith Hoffman, Office of Nuclear Reactor Regulation, telephone: 301-415-1294, email:

Keith.Hoffman@nrc.gov.

Both are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

The NRC is amending its regulations to incorporate by reference recent editions and addenda to the ASME Codes for nuclear power plants and an ASME standard for quality assurance. The NRC is also incorporating by reference six ASME Code Cases.

This final rule is the latest in a series of rulemakings to amend the NRC's regulations to incorporate by reference revised and updated ASME Codes for nuclear power plants. The ASME is a voluntary consensus standards body, and the ASME Codes are voluntary consensus standards. The ASME periodically revises and updates its codes for nuclear power plants by issuing new editions and addenda. The NRC's use of the ASME Codes is consistent with applicable requirements of the National Technology Transfer and Advancement Act (NTTAA). This rulemaking is in accordance with the NRC's policy to update the regulations to incorporate by reference those new editions and addenda. The incorporation by reference of the new editions and addenda will maintain the safety of nuclear power plants, make NRC activities more effective and efficient, and allow nuclear power plant licensees and applicants to take advantage of the latest ASME Codes. Additional discussion of voluntary consensus standards and the NRC's compliance with the NTTAA is set forth in Section XIV of this document, “Voluntary Consensus Standards.”

B. Major Provisions

Major provisions of this final rule include:

• Incorporation by reference of ASME Codes into the NRC's regulations and delineation of the NRC's requirements for the use of these codes, including conditions.

• Incorporation by reference of various versions of quality assurance standard NQA-1 into NRC regulations and approval for their use.

• Incorporation by reference of six ASME Code Cases.

C. Costs and Benefits

The NRC prepared a regulatory analysis (ADAMS Accession No. ML16130A522) to identify the costs and benefits associated with this final rule. The regulatory analysis prepared for this rulemaking was used to determine if the rule is cost-effective, overall, and to help the NRC evaluate potentially costly conditions placed on specific provisions of the ASME Codes and Code Cases which are the subject of this rulemaking. Therefore, the regulatory analysis focuses on the marginal difference in benefits and costs for each provision of this final rule relative to the “no action” baseline alternative. The regulatory analysis identified costs and benefits in a quantitative fashion as well as in a qualitative fashion. An uncertainty analysis was performed to evaluate the effects of uncertainties in the quantitative estimation of both costs and benefits, and this analysis showed the rule alternative is cost effective with over 99 percent certainty. The standard deviation of the cost estimate net benefit is $4.1 million.

Table 1—Cost-Benefit Summary

Objective

Alternative 2—

the rule

alternative

net benefits

(costs)

(million) (Net present value, 7% discount rate)

Industry

$11.5

NRC

3.28

Net benefit

14.7

Table 1 summarizes the costs and benefits for the alternative of proceeding with the final rule (Alternative 2) and shows that the final rule is quantitatively cost-beneficial with a net benefit of $14.7 million to both the industry and the NRC when compared to the regulatory baseline (Alternative 1). The regulatory analysis shows that implementing the final rule is quantitatively cost-effective and an efficient use of NRC and Industry resources. Uncertainty analysis shows a standard deviation of $4.08 million, resulting in a net benefit range of $8.19 million to $21.6 million. Because the

rulemaking alternative is cost-effective, the rulemaking approach is recommended.

There are several benefits associated with this final rule. The new motor-operated valve (MOV) provisions in this final rule result in over $25 million in averted costs (7-percent net present value) due to the removal of quarterly testing requirements and replacing those requirements with less frequent diagnostic and biannual testing requirements. Additionally, the provisions in this final rule will result in averted costs to the NRC and the industry from relief requests for the code cases in this final rule, in particular the ASME OMN-20 Code Case Time Period Extension provision, in excess of $5.1 million (7-percent net present value).

Qualitative factors which were considered include regulatory stability and predictability, regulatory efficiency, and consistency with the NTTAA. Table 50 in the regulatory analysis includes a discussion of the costs and benefits that were considered qualitatively. Considering non-quantified costs and benefits, the regulatory analysis shows that the rulemaking is justified because the number and significance of the non-quantified benefits outweigh the non-quantified costs. Certainly, if the qualitative benefits (including the safety benefit, regulatory efficiency, and other nonquantified benefits) are considered together with the quantified benefits, then the benefits would outweigh the identified quantitative and qualitative impacts. Therefore, integrating both quantified and non-quantified costs and benefits, the benefits of the final rule outweigh the identified quantitative and qualitative impacts attributable to the final rule.

Table of Contents

I. Background

II. Discussion

A. ASME BPV Code, Section III

B. ASME BPV Code, Section XI

C. OM Code

D. ASME Code Cases

III. Opportunities for Public Participation

IV. NRC Responses to Public Comments

V. Section-by-Section Analysis

VI. Generic Aging Lessons Learned Report

VII. Regulatory Flexibility Certification

VIII. Regulatory Analysis

IX. Backfitting and Issue Finality

X. Plain Writing

XI. Finding of No Significant Impact: Environmental Assessment

XII. Paperwork Reduction Act Statement

XIII. Congressional Review Act

XIV. Voluntary Consensus Standards

XV. Incorporation by Reference—Reasonable Availability to Interested Parties

XVI. Availability of Guidance

XVII. Availability of Documents

I. Background

The ASME develops and publishes the ASME

Boiler and Pressure Vessel Code

(BPV Code), which contains requirements for the design, construction, and inservice inspection (ISI) of nuclear power plant components; and the OM Code,

which contains requirements for inservice testing (IST) of nuclear power plant components. Until 2012, the ASME issued new editions of the ASME BPV Code every 3 years and addenda to the editions annually, except in years when a new edition was issued. Similarly, the ASME periodically published new editions and addenda of the OM Code. Starting in 2012, the ASME decided to issue editions of its BPV and OM Codes (no addenda) every 2 years. The new editions and addenda typically revise provisions of the ASME BPV and OM Codes (ASME Codes) to broaden their applicability, add specific elements to current provisions, delete specific provisions, and/or clarify them to narrow the applicability of the provision. The revisions to the editions and addenda of the ASME Codes do not significantly change philosophy or approach.

The editions and addenda of the ASME

Code for Operation and Maintenance of Nuclear Power Plants

have had different titles from 2005 to 2012 and are referred to collectively in this rule as the “OM Code.”

It has been the NRC's practice to establish requirements for the design, construction, operation, ISI examination, and IST of nuclear power plants by approving the use of editions and addenda of the ASME Codes in § 50.55a of title 10 of the

Code of Federal Regulations

(10 CFR), “Codes and standards.” The NRC approves and/or mandates the use of certain parts of editions and addenda of these ASME Codes in § 50.55a through the rulemaking process of “incorporation by reference.” Upon incorporation by reference of the ASME Codes into § 50.55a, the provisions of the ASME Codes are legally-binding NRC requirements as delineated in § 50.55a and subject to the conditions on certain specific ASME Code provisions that are set forth in § 50.55a. The editions and addenda of the ASME BPV and OM Codes were last incorporated by reference into the regulations in a final rule dated June 21, 2011 (76 FR 36232), subject to the NRC's conditions.

The ASME Codes are consensus standards developed by participants with broad and varied interests, including the NRC and licensees of nuclear power plants. The ASME's adoption of new editions of, and addenda to, the ASME Codes does not mean that there is unanimity on every provision in the ASME Codes. There may be disagreement among the technical experts, including NRC representatives, on the ASME Code committees and subcommittees, regarding the acceptability or desirability of a particular Code provision included in an ASME-approved Code edition or addenda. If the NRC believes that there is a significant technical or regulatory concern with a provision in an ASME-approved Code edition or addenda being considered for incorporation by reference, then the NRC will condition the use of that provision when it incorporates by reference that ASME Code edition or addenda. In some cases, the condition increases the level of safety afforded by the ASME Code provision or addresses a regulatory issue not considered by the ASME. In other instances, where research data or experience has shown that certain Code provisions are unnecessarily conservative, the condition may provide that the Code provision need not be complied with in some or all respects. The NRC's conditions are included in § 50.55a, typically in paragraph (b) of that regulation. In a Staff Requirements Memorandum (SRM) dated September 10, 1999, the Commission indicated that NRC rulemakings adopting (incorporating by reference) a voluntary consensus standard must identify and justify each part of the standard that is not adopted. For this rulemaking, the provisions of the 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section III, Division 1; and the 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section XI, Division 1, of the ASME BPV Code; and the 2009 Edition, 2011 Addenda, and 2012 Edition of the OM Code that the NRC is not adopting, or partially adopting, are identified in the Discussion, Regulatory Analysis, and Backfitting and Issue Finality sections of this document. The provisions of those specific editions and addenda and Code Cases that are the subject of this rulemaking that the NRC finds to be conditionally acceptable, together with the applicable conditions, are also identified in the Discussion, Regulatory Analysis, and Backfitting and Issue Finality sections of this document.

The ASME Codes are voluntary consensus standards, and the NRC's incorporation by reference of these Codes is consistent with applicable requirements of the NTTAA. Additional discussion on NRC's compliance with the NTTAA is set forth in Section XIV

of this document, “Voluntary Consensus Standards.”

This final rule reflects the NRC's redesignation of paragraphs within § 50.55a set forth in a final rule dated November 5, 2014 (79 FR 65776), as corrected on December 11, 2014 (79 FR 73461). The re-designation of paragraphs was needed to address the Office of the Federal Register's requirements in 1 CFR part 51 for incorporation by reference. For additional information on the November 2014 final rule, please consult the statement of considerations (preamble) for that final rule.

II. Discussion

The NRC regulations incorporate by reference ASME Codes for nuclear power plants. The ASME periodically revises and updates its codes for nuclear power plants. This final rule is the latest in a series of rulemakings to amend the NRC's regulations to incorporate by reference revised and updated ASME Codes for nuclear power plants. The proposed rule which led to this final rule was published on September 18, 2015 (80 FR 56820). This rulemaking is intended to maintain the safety of nuclear power plants and make NRC activities more effective and efficient.

The NRC follows a three-step process to determine acceptability of new provisions in new editions and addenda to the Codes and the need for conditions on the uses of these Codes. This process was employed in the review of the Codes that are the subject of this rule. First, the NRC staff actively participates with other ASME committee members with full involvement in discussions and technical debates in the development of new and revised Codes. This includes a technical justification of each new or revised Code. Second, the NRC committee representatives discuss the Codes and technical justifications with other cognizant NRC staff to ensure an adequate technical review. Third, the NRC position on each Code is reviewed and approved by NRC management as part of the rule amending § 50.55a to incorporate by reference new editions and addenda of the ASME Codes and conditions on their use. This regulatory process, when considered together with the ASME's own process for developing and approving the ASME Codes, provides reasonable assurance that the NRC approves for use only those new and revised Code edition and addenda, with conditions as necessary, that provide reasonable assurance of adequate protection to public health and safety, and that do not have significant adverse impacts on the environment.

The NRC is amending its regulations to incorporate by reference:

• The 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code, Section III, Division 1 and Section XI, Division 1, with conditions on their use.

• The 2009 Edition, the 2011 Addenda, and the 2012 Edition of Division 1 of the OM Code, with conditions on their use.

• ASME Standard NQA-1, “Quality Assurance Requirements for Nuclear Facility Applications,” including several editions and addenda to NQA-1 from previous years with slightly varying titles as identified in § 50.55a(a)(1)(v). More specifically, the NRC is incorporating by reference the 1983 Edition through the 1994 Edition, the 2008 Edition, and the 2009-1a Addenda to the 2008 Edition of ASME NQA-1, with conditions on their use.

• ASME BPV Code Case N-513-3, “Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,” Mandatory Appendix I, “Relations for

, and

for Through-Wall Flaws,” Approval Date: January 26, 2009. This Code Case has already been approved for use by the NRC in Regulatory Guide (RG) 1.147 (75 FR 61321; October 5, 2010), but is now being incorporated by reference in order to adopt a condition on Nonmandatory Appendix U, which requires the use of this Code Case appendix.

• ASME BPV Code Case N-729-4, “Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Section XI, Division 1,” ASME approval date: June 22, 2012, with conditions on its use.

• ASME BPV Code Case N-770-2, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1,” ASME approval date: June 9, 2011, with conditions on its use.

• ASME BPV Code Case N-824, “Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,” ASME approval date: October 16, 2012.

• ASME BPV Code Case N-852, “Application of the ASME NPT Stamp, Section III, Division 1; Section III, Division 2; Section III, Division 3; Section III, Division 5,” Approval Date: February 9, 2015.

• OM Code Case OMN-20, “Inservice Test Frequency.”

The current regulations in § 50.55a(a)(1)(ii) incorporate by reference ASME BPV Code, Section XI, 1970 Edition through the 1976 Winter Addenda; and the 1977 Edition (Division 1) through the 2008 Addenda (Division 1), subject to the existing conditions in § 50.55a(b)(2)(i) through (xxix). This amendment revises § 50.55a(a)(1)(ii) to incorporate by reference the 2009 Addenda (Division 1) through the 2013 Edition (Division 1) of the ASME BPV Code, Section XI. It also clarifies the wording and adds, removes, or revises some of the conditions as explained in this document.

The NRC is revising § 50.55a(a)(1)(iv) to incorporate by reference the 2009 Edition, 2011 Addenda, and 2012 Edition of Division 1 of the OM Code. Based on this revision, the NRC regulations will incorporate by reference in § 50.55a the 1995 Edition through the 2012 Edition of the OM Code.

The NRC reviewed changes to the Codes in the editions and addenda of the Codes identified in this rulemaking, and published a proposed rule in the

Federal Register

setting forth the NRC's proposal to incorporate by reference the ASME Codes, together with proposed conditions on their use (80 FR 56820; September 18, 2015). After consideration of the public comments received on the proposed rule (public comments are discussed in Section IV of this document, “NRC Responses to Public Comments”), the NRC concludes, in accordance with the process for review of changes to the Codes, that each of the editions and addenda of the Codes, and the 2008 Edition and the 2009-1a Addenda of NQA-1, are technically adequate, consistent with current NRC regulations, and approved for use with specified conditions set forth in this final rule. Each of the NRC conditions and the reasons for each condition are discussed in the following sections. The discussions are organized under the applicable ASME Code and Section.

There is not a separate heading for ASME quality assurance standard NQA-1 because there are three separate discussions of NQA-1—one under the heading for ASME BPV Code, Section III, one under the heading for ASME BPV Code, Section XI, and one under the heading for OM Code—because there are three conditions related to NQA-1, one in each of those areas (§ 50.55a(b)(1)(iv) for Section III, § 50.55a(b)(2)(x) for Section XI, and § 50.55a(b)(3)(i) for the OM Code). In addition, administrative and editorial changes to various paragraphs of § 50.55a are being adopted for accuracy, clarity, consistency, and general

administrative convenience. These editorial changes are not further discussed in this heading, but are described in Section V of this document, “Section-by-Section Analysis.”

Four of the six ASME Code Cases being incorporated by reference in this rulemaking (N-729-4, N-770-2, N-824, and OMN-20) are discussed in Section II.D of this document, “ASME Code Cases.” A fifth ASME Code Case, N-852, is discussed in Section II.A, “ASME BPV Code, Section III,” because the NRC's approval of that Code Case relates to a provision of Section III, which is addressed in § 50.55a(b)(1)(ix). The sixth ASME Code Case, N-513-3, is discussed in Section II.B, “ASME BPV Code, Section XI,” because the NRC's approval of that Code Case relates to a provision of Section XI, which is addressed in § 50.55a(b)(2)(xxxiv).

A. ASME BPV Code, Section III

10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section III

The NRC is clarifying that Section III Nonmandatory Appendices are not incorporated by reference. This language was originally added in a final rule published on June 21, 2011 (76 FR 36232); however, it was omitted from the final rule published on November 5, 2014 (79 FR 65776). The NRC is correcting the omission by inserting the parenthetical clause “(excluding Nonmandatory Appendices)” in § 50.55a(a)(1)(i).

10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions

The NRC is identifying prohibited subparagraphs and notes for each ASME BPV Code edition and addenda in tabular form as opposed to the narrative form of the existing regulation. No substantive change to the requirements is intended by this revision. The NRC believes that presenting the information in tabular form will increase the clarity and understandability of the regulation.

The existing condition in § 50.55a(b)(1)(ii) prohibits, for welds with leg sizes less than 1.09 t

, the use of certain Code provisions in ASME BPV Code, Section III, Division 1. The Code provisions provide stress indices for welded joints used in the design of Class 2 and Class 3 piping. The use of these indices is prohibited for welds with leg sizes less than 1.09 t

, where t

is the nominal pipe thickness because this would result in a weld that would be weaker than the pipe to which it is adjoined under these dimensions. The location of the prohibited provisions vary in the Code editions and addenda from the 1989 Addenda through the 2013 Edition, so in this final rule the NRC clearly identifies the prohibited code provisions in the editions and addenda in a tabular format.

As an editorial matter, this final rule identifies the prohibited ASME BPV Code provisions as “notes,” which is the term used by the ASME, rather than “footnotes.” The NRC is using the terminology used by the ASME for clarity.

10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance

The NRC is approving for use the version of NQA-1 referenced in the 2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code, Section III, Subsection NCA, Article 7000, which this rule is also incorporating by reference. This allows applicants and licensees to use the 2008 Edition and the 2009-1a Addenda of NQA-1 when using the 2010 and later editions and addenda of Section III.

In the 2010 Edition of ASME BPV Code, Section III, Subsection NCA, Article NCA-4000, “Quality Assurance,” was updated to require N-Type Certificate Holders to comply with the requirements of Part 1 of the 2008 Edition and the 2009-1a Addenda of ASME Standard NQA-1, “Quality Assurance Requirements for Nuclear Facility Applications,” as modified and supplemented in NCA-4120(b) and NCA-4134. In addition, NCA-4110(b) was revised to remove the reference to a specific edition and addenda of ASME NQA-1, and Table NCA-7100-2, “Standards and Specifications Referenced in Division 1,” was updated to require the 2008 Edition and 2009-1a Addenda of NQA-1 when using the 2010 Edition of Section III. In light of these changes, the NRC reviewed the 2008 Edition and the 2009-1a Addenda of NQA-1 and compared it to previously approved versions of NQA-1 and found that there were no significant differences. In addition, the NRC reviewed the changes to Subsection NCA that reference the 2008 Edition and 2009-1a Addenda of NQA-1, compared them to previously approved versions of Subsection NCA, and found that there were no significant differences. Therefore, the NRC has concluded that these editions and addenda of NQA-1 are acceptable for use.

The NRC is revising § 50.55a(b)(1)(iv) to clarify that an applicant's or licensee's commitments addressing those areas where NQA-1 either does not address a requirement in appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” or is less stringent than the comparable appendix B requirement govern the applicant's or licensee's Section III activities. The clarification is consistent with § 50.55a(b)(2)(x) and (b)(3)(i). The NQA-1 provides the ASME's method for establishing and implementing a quality assurance (QA) program for the design and construction of nuclear power plants and fuel reprocessing plants. However, NQA-1, as modified and supplemented in NCA-4120(b) and NCA-4134, does not address some of the requirements of appendix B to 10 CFR part 50. In some cases, the provisions of NQA-1 are less stringent than the comparable appendix B requirements. Therefore, in order to meet the requirements of appendix B, an applicant's or licensee's QA program description must contain commitments addressing those provisions of appendix B which are not covered by NQA-1, as well as provisions that supplement or replace the NQA-1 provisions where the appendix B requirement is more stringent.

Finally, the NRC is removing the reference in § 50.55a(b)(1)(iv) to versions of NQA-1 older than the 1994 Edition because the NRC did not receive any adverse comments from any applicant or licensee about removing versions of NQA-1 older than the 1994 Edition from the regulation. The NRC received only one comment regarding NQA-1. The comment expressed support for incorporation by reference of NQA-1 and did not respond to the NRC's request for comment regarding the removal of references to older versions of NQA-1.

10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification and Demonstration of Function of Incompressible-Fluid Pressure-Relief Valves

The NRC is revising § 50.55a(b)(1)(vii) so that the existing condition prohibiting the use of paragraph NB-7742(a)(2) of the 2006 Addenda through the 2007 Edition, up to and including the 2008 Addenda, is extended to include the editions and addenda up to the 2013 Edition, which are the subject of this rulemaking.

10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME Certification Marks

The NRC is adding § 50.55a(b)(1)(viii) to allow licensees to use either the ASME BPV Code Symbol Stamps of editions and addenda earlier than the 2011 Addenda to the 2010 Edition of the ASME BPV Code or the ASME Certification Marks with the appropriate

certification designators and class designators as specified in the 2013 Edition through the latest edition and addenda incorporated by reference in § 50.55a.

The ASME BPV Code requires, in certain instances, that components be stamped. The stamp signifies that the component has been designed, fabricated, examined and tested, as specified in the ASME BPV Code. The stamp also signifies that the required ASME BPV Code data report forms have been completed, and the authorized inspector has inspected the item and authorized the application of the ASME BPV Code Symbol Stamp.

The ASME has instituted changes in the BPV Code to consolidate the different ASME BPV Code Symbol Stamps into a common ASME Certification Mark. This action was implemented in the 2011 Addenda to the 2010 Edition of the ASME BPV Code. As of the end of 2012, ASME no longer utilizes the ASME BPV Code Symbol Stamp. Licensees, however, may not have updated to the edition or addenda that identifies the use of the ASME Certification Mark. Nevertheless, licensees are legally required to implement the ASME BPV Code Edition and Addenda identified as their current code of record. As ASME components are procured, these components may be received with the ASME Certification Mark, while the licensee's current code of record may require the component to have the ASME BPV Code Symbol Stamp. Installation of a component under such circumstances would not be in compliance with the regulations that the licensees are required to meet.

Both the ASME Certification Mark and the ASME BPV Code Symbol Stamp are official ASME methods of certifying compliance with the Code. Although these ASME Certification Marks differ slightly in appearance, they serve the same purpose of certifying code compliance by the ASME Certificate Holder and continue to provide for the same level of quality assurance for the application of the ASME Certification Mark as was required for the application of the ASME BPV Code Symbol Stamp. The new ASME Certification Mark represents a small, non-safety significant modification of ASME's trademark. As such, it does not change the technical requirements of the Code. The ASME has confirmed that the Certification Mark with designator is equivalent to the corresponding BPV Code Symbol Stamp. Based on statements made by ASME in a letter dated August 17, 2012, the NRC has concluded that the ASME BPV Code Symbol Stamps and ASME Certification Mark with code-specific designators are equivalent with respect to their certification of compliance with the BPV Code. The NRC discussed this issue in Regulatory Issue Summary 2013-07, “NRC Staff Position on the Use of American Society of Mechanical Engineers Certification Mark,” dated May 28, 2013.

10 CFR 50.55a(b)(1)(ix) Section III Condition: NPT Code Symbol Stamps

The NRC is adding § 50.55a(b)(1)(ix) to allow licensees to use the NPT Code Symbol Stamp with the letters arranged horizontally as specified in ASME BPV Code Case N-852 for the service life of a component that had the NPT Code Symbol Stamp applied during the time period from January 1, 2005, through December 31, 2015.

Public comments on the use of the NPT Code Symbol requested that the NRC accept the NPT Code Symbol Stamp having the NPT letters arranged horizontally as an acceptable NPT Stamp to certify Code compliance for fabricated items that have already been stamped prior to receiving a replacement NPT Code Symbol Stamp from the ASME. The comments requested that the NRC include acceptance of Code Case N-852 in this final rule for this purpose. Within the context of its Code rules, ASME asserts that the NPT Code Symbol Stamp having the NPT letters arranged horizontally, although differing slightly in appearance from the NPT Code Symbol Stamp as illustrated in Section III, Table NCA-8100-1 of the ASME BPV Code, 2010 Edition and earlier editions and addenda, serves the same purpose of certifying Code compliance by the ASME NPT Certificate Holder with confirmation by the Authorized Nuclear Inspector and provides the same level of quality assurance. In addition, ASME indicated that on or after January 1, 2016, the ASME will no longer authorize use of the NPT Code Symbol Stamp having the NPT letters arranged horizontally. Accordingly, on or after January 1, 2016, fabricated items will only be stamped with the NPT Code Symbol Stamp as illustrated in Section III, Table NCA-8100-1 of the ASME BPV Code, 2010 Edition and earlier editions and addenda.

The NRC agrees in general with this comment, in which the ASME asserts that the ASME NPT Code Symbol Stamp with the letters arranged horizontally to be equivalent to the “N over PT” ASME NPT Code Symbol Stamp. Therefore, using either Code Symbol Stamp serves the same purpose of certifying code compliance by the ASME Certificate Holder with confirmation by the Authorized Nuclear Inspector and provides the same level of quality assurance. The NRC also notes that the same administrative and technical requirements in the ASME Code still apply whether an ASME NPT Code Symbol Stamp with the letters arranged horizontally or an “N over PT” ASME NPT Code Symbol Stamp is applied. However, since this NPT Code Symbol Stamp having the NPT letters arranged horizontally will only be applied onto fabricated components from the time period of January 1, 2005, through December 31, 2015, the time period for when this NPT Code Symbol Stamp was applied to the component should be limited to these dates to prevent inadvertent fraudulent material. Therefore, the NRC agrees that the ASME BPV Code Case N-852 is acceptable for the service life of the component that had the NPT Code Symbol stamp applied from the time period of January 1, 2005, through December 31, 2015. In response to this comment, the NRC added § 50.55a(b)(1)(ix) to include a statement that licensees may use the NPT Code Symbol Stamp with the letters arranged horizontally as specified in ASME BPV Code Case N-852 for the service life of a component that had the NPT Code Symbol Stamp applied during the time period from January 1, 2005, through December 31, 2015. The NRC is incorporating by reference ASME BPV Code Case N-852 in § 50.55a(a)(1)(iii)(F) because it is referenced in § 50.55a(b)(1)(ix).

Although the proposed rule did not include this Code Case, the NRC has determined that the incorporation by reference of this Code Case at the final rule stage is a logical outgrowth of the proposed rule. The NRC's intent to ensure that § 50.55a identify all ASME-approved methods for labelling Code components is apparent from the statement of considerations for the proposed rule. See 80 FR 56820 (September 18, 2015) at 56823-56824. The NRC did not entirely achieve that purpose, and this resulted in public comments seeking approval of this Code Case, which supports the proposition that the public had a reasonable opportunity to either propose the correction, with conditions as the commenter believes are necessary or desirable, or to indicate why the (anticipated) correction should not be made. Therefore, the NRC concludes that it may incorporate by reference ASME BPV Code Case N-852.

B. ASME BPV Code, Section XI

10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section XI

In the proposed rule, the NRC proposed a revision to § 50.55a(a)(1)(ii) that would have clarified that Section XI Nonmandatory Appendix U of the 2013 Edition of ASME BPV Code, Section XI was not incorporated by reference and therefore not approved for use. After considering public comments, the NRC has determined that it will not exclude Appendix U from the incorporation by reference because it is the integration of ASME BPV Code Cases N-513-3, “Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,” and N-705, “Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks Section XI, Division 1,” into Section XI. The NRC has approved Code Cases N-513-3 and N-705 in RG 1.147. However, as described in the discussion for § 50.55a(b)(2)(xxxiv) in Section II.B, “ASME BPV Code Section XI,” the NRC has found it necessary to adopt two new conditions to the use of Nonmandatory Appendix U.

The NRC is adopting two conditions in the language of § 50.55a(a)(1)(ii)(C)(

) and (

) to address two inconsistencies that were identified between the NRC's position in a proposed rule regarding the acceptability of ASME Code Cases (81 FR 10780; March 2, 2016) (2016 Code Case proposed rule) and the proposed rule for this rulemaking (80 FR 56820; September 18, 2015). The first inconsistency is that the NRC's proposed conditions on ASME BPV Code Case N-799, “Dissimilar Metal Welds Joining Vessel Nozzles to Components,” in the 2016 Code Case proposed rule were not reflected in the 2015 proposed rule for this rulemaking, even though the technical content of ASME BPV Code Case N-799 has been incorporated into the 2011 Addenda and 2013 Edition of ASME BPV Code, Section XI. The second inconsistency is that the NRC's proposed disapproval of ASME BPV Code Case N-813, “Alternative Requirements for Preservice Volumetric and Surface Examination,” in the 2016 Code Case proposed rule was not reflected in the 2015 proposed rule for this rulemaking, even though the technical content of ASME BPV Code Case N-813 has been incorporated into the 2013 Edition of the ASME BPV Code, Section XI as IWB-3112(a)(3) and IWC-3112(a)(3). To address these two inconsistencies, the NRC is excluding these ASME BPV Code, Section XI items from incorporation by reference, as reflected in § 50.55a(a)(1)(ii)(C)(

) and (

) of the final rule. The NRC plans to complete the development of the regulatory approaches for examination of component-to-component welds for new construction plants and the acceptance of preservice flaws by analytical evaluation for operating plants and include them in a future rulemaking.

10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and Addenda of Subsection IWE and Subsection IWL

The NRC is revising § 50.55a(b)(2)(vi) to expressly state that licensees that implemented the expedited examination of containment during the 5-year period from September 9, 1996, to September 9, 2001, 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), when implementing the initial 120-month inspection interval for the containment ISI requirements of this section.

The expedited examination involved the completion of the first set of examinations of the first or initial 120-month containment inspection interval. It is noted that all of the operating reactors in the previously stated class would have gone past their initial 120-month inspection interval by 2011. The change removes the possibility of misinterpretation of the provision as requiring plants that do not fall in the previously stated class, such as reactors licensed after September 9, 2001, to use the 1992 Edition with 1992 Addenda or the 1995 Edition with 1996 Addenda of Subsection IWE and Subsection IWL, Section XI for implementing the initial 120-month inspection interval of the containment ISI program. Applicants and licensees that do not fall in the previously stated class must use Code editions and addenda in accordance with § 50.55a(g)(4)(i) and (ii), respectively, for the initial and successive 120-month containment ISI intervals.

10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment Examinations

The NRC is revising § 50.55a(b)(2)(viii) by removing the condition for using the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL requiring compliance with § 50.55a(b)(2)(viii)(E). To support the removal of the condition, the NRC is adding new requirements governing the performance and documentation of concrete containment examinations in § 50.55a(b)(2)(viii)(H) and (I), which are discussed separately in the next two headings.

Section 50.55a(b)(2)(viii)(E) is one of several conditions that apply to the inservice examination of concrete containments using Subsection IWL of various editions and addenda of the ASME BPV Code, Section XI, incorporated by reference in § 50.55a(a)(1)(ii). The NRC is removing the condition in § 50.55a(b)(2)(viii)(E) when applying the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL because its intent has been incorporated into the Code in the new provision IWL-2512, “Inaccessible Areas.”

10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth Provision

The NRC is adding § 50.55a(b)(2)(viii)(H) to specify the information that must be provided in the ISI Summary Report required by IWA-6000, when inaccessible concrete surfaces are evaluated under the new Code provision IWL-2512. This new condition replaces the existing condition in § 50.55a(b)(2)(viii)(E), when using the 2007 Edition with the 2009 Addenda through the 2013 Edition of Subsection IWL.

The existing condition in § 50.55a(b)(2)(viii)(E) of the current rule requires that, for Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas, and provide the evaluation information required by § 50.55a(b)(2)(viii)(E)(

), (

), and (

) in the IWA-6000 ISI Summary Report.

In the 2009 Addenda Subsection IWL, the ASME revised existing provisions IWL-1220 and IWL-2510 and added the new provision IWL-2512 intended to incorporate the condition in § 50.55a(b)(2)(viii)(E) into Subsection IWL. The IWL-2510, “Surface Examination,” was restructured into new paragraphs in IWL-2511, “Accessible Areas,” with almost the same provisions as the previous IWL-2510 and IWL-2512, “Inaccessible Areas,” to be specific to examinations required for accessible areas, and differentiate between those and the new requirements for inaccessible areas. The inaccessible areas addressed by the new IWL-2512 are: (1) Concrete surfaces obstructed by adjacent structures, parts or appurtenances (

e.g.,

generally above-grade inaccessible areas); and (2)

concrete surfaces made inaccessible by foundation material or backfill (

e.g.,

below-grade inaccessible areas).

The revised IWL-2511(a) has a new requirement that states that, “If the Responsible Engineer determines that observed suspect conditions indicate the presence of, or could result in, degradation of inaccessible areas, the requirements of IWL-2512(a) shall be met.” The new IWL-2512(a) requires the “Responsible Engineer” to evaluate suspect conditions and specify the type and extent of examinations, if any, required to be performed on inaccessible surface areas described in the previous paragraph. The acceptability of the evaluated inaccessible area would be determined either based on the evaluation or based on the additional examinations, if determined to be required. The new IWL-2512(b) further requires a periodic technical evaluation of below-grade inaccessible areas of concrete to be performed to determine and manage its susceptibility to degradation regardless of whether suspect conditions exist in accessible areas that would warrant an evaluation of inaccessible areas based on the condition in § 50.55a(b)(2)(viii)(E). Therefore, the revised IWL-2511(a) and new IWL-2512 code provisions address the evaluation and acceptability of inaccessible areas consistent with the existing condition in § 50.55a(b)(2)(viii)(E), with one exception. The exception is that the new IWL-2512 provision does not explicitly require the information specified in § 50.55a(b)(2)(viii)(E)(

), (

), and (

) of the existing condition to be provided in the IWA-6000 ISI Summary Report.

For these reasons, the NRC is identifying the information that must be provided in the ISI Summary Report required by IWA-6000 when inaccessible concrete surfaces are evaluated under the new code provision IWL-2512. This new condition replaces the existing condition in § 50.55a(b)(2)(viii)(E) when using the 2007 Edition with the 2009 Addenda through the 2013 Edition of Subsection IWL. The information required by the new condition must be provided when inaccessible concrete areas are evaluated per IWL-2512(a) for degradation based on suspect conditions found in accessible areas, as well as when periodic technical evaluations of inaccessible below-grade concrete areas required by IWL-2512(b) are performed.

10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth Provision

The NRC is adding § 50.55a(b)(2)(viii)(I) to place a condition on the periodic technical evaluation requirements in the new IWL-2512(b), for consistency with NUREG-1801, Revision 2, “Generic Aging Lessons Learned (GALL) Report,” with regard to aging management of below-grade containment concrete surfaces. The new IWL-2512(b) provision is applicable to inaccessible below-grade concrete surfaces exposed to foundation soil, backfill, or groundwater. This condition would apply only during the period of extended operation of a renewed license under 10 CFR part 54, when using IWL-2512(b) of the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL.

In the 2009 Addenda of Subsection IWL, the ASME added new Code provisions, IWL-2512(b) and (c) as well as a new line item L1.13 in Table IWL-2500-1, intended to specifically address aging management concerns with potentially unidentified degradation of inaccessible below-grade containment concrete areas and to be responsive to actions outlined in the GALL Report related to aging management of inaccessible below-grade concrete surfaces. It is noted that these new Code provisions are an enhancement to the requirement of the existing condition in § 50.55a(b)(2)(viii)(E) to specifically address aging management of inaccessible below-grade containment concrete areas and is generally acceptable to the NRC.

The new IWL-2512(b) provides requirements for systematically performing a periodic technical evaluation of concrete surfaces exposed to foundation soil, backfill, or groundwater to determine susceptibility of the concrete to deterioration that could affect its ability to perform its intended design function under conditions anticipated through the service life of the structure. It requires the technical evaluation to be performed and documented at periodic intervals not to exceed 10 years regardless of whether conditions exist in accessible areas that would warrant an evaluation of inaccessible areas by the existing condition in § 50.55a(b)(2)(viii)(E), which the NRC finds reasonable for the initial 40-year operating license period. The new IWL-2512(b) further provides the specific elements, including aging mechanisms considered, that the technical evaluation should include, as well as the definition of an aggressive below-grade environment. The new IWL-2512(c) requires that the evaluation results of IWL-2512(b) be used to define and document the condition monitoring program, if determined to be required, including required examinations and frequencies, to be implemented for the management of degradation and aging effects of the below-grade concrete surface areas. If it is determined that additional examinations are required, these examinations of inaccessible below-grade areas will be implemented in accordance with new line item L1.13 in Table IWL-2500-1 under Examination Category L-A, Concrete, with acceptance criteria based on IWL-3210. It should be noted that a technical evaluation approach, such as in IWL-2512(b), could be used, and is generally used, to determine acceptability of a below-grade inaccessible area to satisfy the condition in § 50.55a(b)(2)(viii)(E).

The technical evaluation requirements in IWL-2512(b) assist in determining the susceptibility to degradation and manage aging effects of inaccessible below-grade concrete surfaces, before the loss of intended function. The requirements are based on, and are generally consistent with, the guidance in the GALL Report, with the following two exceptions. The first exception is that IWL-2512(b) requires the technical evaluation to determine the susceptibility of the concrete to degradation and the ability to perform the intended design function through its service life at periodic intervals not to exceed 10 years. The aging management programs (AMPs) for safety-related structures (

e.g.,

Structures Monitoring) in the GALL Report require such evaluation to be performed at intervals not to exceed 5 years, which is also consistent with applicant commitments during review of license renewal applications. The second exception is that IWL-2512(b) requires that examination of representative samples of below-grade concrete be performed if excavated for any reason when an aggressive below-grade environment is present. However, the NRC notes that the AMPs (X1.S6 Structures Monitoring and X1.S7 Water Control Structures) in the GALL Report require the same examination even for a non-aggressive below-grade environment.

Based on these reasons, the NRC is adding § 50.55a(b)(2)(viii)(I) to place a condition on the periodic technical evaluation requirements in IWL-2512(b) for consistency with the GALL Report, when addressing the two exceptions previously described with respect to aging management of inaccessible below-grade concrete components of the

containment. The new condition requires that, during the period of extended operation of a renewed license, the technical evaluation under IWL-2512(b) of inaccessible below-grade concrete surfaces exposed to foundation soil, backfill, or groundwater be performed at periodic intervals not to exceed 5 years, as opposed to the 10-year interval in IWL-2512. In addition, the condition requires the examination of representative samples of the exposed portions of the below-grade concrete be performed when excavated for any reason as opposed to IWL-2512, which limits the examination to excavations in aggressive, below-grade environments. Since the GALL Report is the technical basis document for license renewal, this new condition applies only during the period of extended operation of a renewed license under 10 CFR part 54, when using IWL-2512(b) of the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL, Section XI.

10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment Examinations

The NRC is extending the applicability of the existing conditions in § 50.55a(b)(2)(ix)(A)(

) and (b)(2)(ix)(B) and (J), governing examinations of metal containments and the liners of concrete containments under Subsection IWE, to the ASME BPV Code editions and addenda which are the subject of this rulemaking (

i.e.,

the 2007 Edition with 2009 Addenda through the 2013 Edition). The last sentence of § 50.55a(b)(2)(ix) prior to this final rule stated that the referenced conditions were applicable only to addenda, but not to editions, approved by the NRC after the 2007 Edition of the ASME BPV Code. To rectify this, the NRC is revising the last sentence of § 50.55a(b)(2)(ix) to refer to the latest “edition and” addenda after the 2007 Edition which are incorporated by reference into § 50.55a.

The NRC reviewed the Code changes in Subsection IWE of the 2009 Addenda through the 2013 Edition of ASME BPV Code, Section XI, and noted that all of the changes were editorial or administrative with the intent to improve the clarity of the existing requirements or correct errors by errata. There were no changes to Subsection IWE in the Code editions and addenda that are the subject of this rulemaking that the NRC believes would require new regulatory conditions to ensure safety, nor do the changes to Subsection IWE address the NRC's reasons for adopting the conditions on the use of Subsection IWE. Accordingly, the NRC is extending the applicability of the existing conditions (by adding the words “edition and” to § 50.55a(b)(2)(ix) as discussed) without any change to the provisions of the conditions.

10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance

The NRC is approving for use the version of NQA-1 referenced in the 2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition of the ASME BPV Code, Section XI, Table IWA 1600-1, “Referenced Standards and Specifications,” which this rule is also incorporating by reference. This allows, but does not require, licensees to use the 1994 Edition or the 2008 Edition and the 2009-1a Addenda of NQA-1 when using the 2009 Addenda and later editions and addenda of Section XI.

In the 2013 Edition of ASME BPV Code, Section XI, Table IWA 1600-1 was updated to allow licensees to use the 1994 Edition or the 2008 Edition with the 2009-1a Addenda of NQA-1 when using the 2013 Edition of Section XI. In the 2010 Edition of ASME BPV Code, Section XI, IWA-1400, “Owner's Responsibilities,” Subparagraph (n)(2) was updated to reference the NQA-1 Part I, Basic Requirements and Supplementary Requirements for Nuclear Facilities. In the 2009 Addenda of the 2007 Edition of ASME BPV Code, Section XI, Table IWA-1600-1, “Referenced Standards and Specifications,” was updated to allow licensees to use the 1994 Edition of NQA-1. The NRC reviewed the 2008 Edition and the 2009-1a Addenda of NQA-1 and compared it to previously approved versions of NQA-1 and found that there were no significant differences. Therefore, the NRC has concluded that these editions and addenda of NQA-1 are acceptable for use.

The NRC is amending § 50.55a(b)(2)(x) to clarify that a licensee's commitments addressing those areas where NQA-1 either does not address a requirements in appendix B to 10 CFR part 50 or is less stringent than the comparable appendix B requirement govern the licensee's Section XI activities. The clarification is consistent with § 50.55a(b)(1)(iv) and (b)(3)(i). The ASME's method for establishing and implementing a QA program for the design and construction of nuclear power plants and fuel reprocessing plants is described in NQA-1. However, NQA-1 does not address some of the requirements of appendix B to 10 CFR part 50. In some cases, the provisions of NQA-1 are less stringent than the comparable appendix B requirements. Therefore, in order to meet the requirements of appendix B, a licensee's QA program description must contain commitments addressing those provisions of appendix B which are not covered by NQA-1, as well as provisions that supplement or replace the NQA-1 provisions where the appendix B requirement is more stringent.

Finally, the NRC is removing the reference in § 50.55a(b)(2)(x) to versions of NQA-1 older than the 1994 Edition because the NRC did not receive any adverse comments from any applicant or licensee regarding concerns about removing versions of NQA-1 older than the 1994 Edition from the regulation. The NRC received only one comment regarding NQA-1. The comment expressed support for incorporation by reference of NQA-1 and did not respond to the NRC's request for comment regarding the removal of references to older versions of NQA-1.

10 CFR 50.55a(b)(2)(xii) Section XI Condition: Underwater Welding

The NRC is revising § 50.55a(b)(2)(xii) to allow underwater welding on irradiated materials in accordance with IWA-4660, “Underwater Welding,” of Section XI, 1997 Addenda through the latest edition and addenda incorporated by reference in § 50.55a(a)(1)(ii). The conditions for which underwater welding would be permitted without prior NRC approval are based on technical factors, such as neutron fluence and, for certain material classes, helium concentration.

The existing condition in § 50.55a(b)(2)(xii) does not allow underwater welding on irradiated materials by prohibiting the use of IWA-4660, “Underwater Welding,” of Section XI, 1997 Addenda through the latest edition and addenda incorporated by reference in § 50.55a(a)(1)(ii) on materials that are irradiated; however, there are two problems with the restriction in § 50.55a(b)(2)(xii). First, the neutron fluence threshold above which a material is considered to be irradiated is not defined in § 50.55a(b)(2)(xii). Second, studies such as those documented in Boiling Water Reactor Vessel and Internals Project (BWRVIP) Report 1003020 (BWRVIP-97) have shown that reactor internals can tolerate some neutron irradiation without suffering damage to weldability, as long as the helium concentration in the material does not exceed a certain threshold. The NRC completed its Safety Evaluation of BWRVIP-97 in May 2008 and concluded that implementation of the guidelines in the BWRVIP-97 report, with some modifications as documented in the

NRC Safety Evaluation dated June 30, 2008, will provide an acceptable technical basis for the design of weld repairs based on the helium content of irradiated reactor vessel internals. The current version of § 50.55a(b)(2)(xii) does not define a threshold of helium concentration below which the material is considered to be weldable.

The most recent editions of the ASME BPV Code state in Article IWA-4660 that underwater welding may not be performed on irradiated materials other than P-No. 8 materials containing less than 0.1 atomic parts per million (appm) measured or calculated helium content generated through irradiation. Some editions and addenda of the ASME BPV Code prior to 2010 state in Article IWA-4660 that underwater welding may only be performed in applications not predicted to exceed a thermal neutron fluence of 1 × 10

n/cm

. Other editions and addenda of the ASME BPV Code prior to 2010 do not restrict the underwater welding of irradiated materials. Therefore, there is inconsistent treatment among the various editions and addenda of the ASME BPV Code on the underwater welding of irradiated materials.

Current ASME BPV Code and Code Case requirements for welding on irradiated materials, other than the underwater welding requirements specified in IWA-4660, are inconsistent. Thresholds for weldability may be stated in terms of fast neutron fluence, thermal neutron fluence, or helium concentration. In some cases, thresholds are not defined and the Code or Code Case simply states that consideration must be given to irradiation effects when welding. The NRC believes that thresholds for welding on irradiated materials should be based on the current understanding of irradiation damage, as supported by technical studies (such as BWRVIP-97) which have been evaluated by the NRC. In addition, the NRC believes that these thresholds should be consistently applied for all Code and Code Case applications.

During the public comment period for this rulemaking, a representative of ASME recommended that § 50.55a(b)(2)(xii) be revised such that it applies only to those editions and addenda earlier than the 2010 Edition. The effect of such a revision would be to allow welding on P-No. 8 materials containing less than 0.1 appm measured or calculated helium content generated through irradiation. However, this proposed revision would not be consistent with other ASME BPV Code or Code Case requirements for welding on irradiated materials, and this proposed revision does not address standards for welding on material classes other than P-No. 8. Instead the NRC is adopting conditions that would apply to all materials and which can be consistently applied for all Code and Code Case applications. The first condition, § 50.55a(b)(2)(xii)(A), is based on fast neutron fluence and applies to ferritic materials. The second condition, § 50.55a(b)(2)(xii)(B), is based on helium content and/or thermal fluence and applies to austenitic materials. For P-No. 8 austenitic materials, the evaluation of BWRVIP-97 supports a weldability threshold based on helium content and thermal fluence. For austenitic materials other than P-No. 8, there are insufficient data to support a weldability threshold based on helium content, and, therefore, the NRC is adopting a weldability threshold based on thermal fluence only.

The conditions for which underwater welding are permitted, as stated in the revision of § 50.55a(b)(2)(xii), were determined, in part, based on technical discussions in a Category 2 public meeting with industry representatives held on January 19, 2016. The NRC later presented the new conditions at a public meeting held on March 2, 2016. There were no comments on this change from the attendees at the March 2, 2016, public meeting. Summaries of the January 19 and March 2, 2016, public meetings are available in ADAMS under Accession Nos. ML16050A383 and ML16069A408, respectively.

10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth Provision

The NRC is adding § 50.55a(b)(2)(xviii)(D) to prohibit applicants and licensees from using the ultrasonic examination nondestructive examination (NDE) personnel certification requirements in Section XI, Appendix VII and Subarticle VIII-2200 of the 2011 Addenda and 2013 Edition of the ASME BPV Code. Paragraph (b)(2)(xviii) currently includes conditions on the certification of NDE personnel. In addition, the new paragraph will require applicants and licensees to use the 2010 Edition, Table VII-4110-1 training hour requirements for Levels I, II, and III ultrasonic examination personnel, and the 2010 Edition, Subarticle VIII-2200 of Appendix VIII prerequisites for personnel requirements. In the 2011 Addenda and 2013 Edition, the ASME BPV Code added an accelerated Appendix VII training process for certification of ultrasonic examination personnel based on training and prior experience, and separated the Appendix VII training requirements from the Appendix VIII qualification requirements. These new ASME BPV Code provisions will provide personnel in training with less experience and exposure to representative flaws in representative materials and configurations common to operating nuclear power plants, and they would permit personnel with prior non-nuclear ultrasonic examination experience to qualify for examinations in nuclear power plants without exposure to the variety of defects, examination conditions, components, and regulations common to operating nuclear power plants.

The impact of reduced training and nuclear power plant familiarization is unknown. The ASME BPV Code supplants training hours and field experience without a technical basis, minimum defined training criteria, process details, or standardization. For these reasons, the NRC is prohibiting the use of Appendix VII and Subarticle VIII-2200 of the 2011 Addenda and 2013 Edition. The NRC is requiring applicants and licensees using the 2011 Addenda and 2013 Edition to use the prerequisites for ultrasonic examination personnel certifications in Table VII-4110-1 and Subarticle VIII-2200, Appendix VIII in the 2010 Edition.

10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements: First Provision

The NRC is revising § 50.55a(b)(2)(xxi)(A) to modify the standard for visual magnification resolution sensitivity and contrast for visual examinations performed on Examination Category B-D components instead of ultrasonic examinations, making the rule conform with ASME BPV Code, Section XI requirements for VT-1 examinations. The character recognition rules are used in ASME BPV Code, Section XI, Table IWA-2211-1 for VT-1 tests, and are the standard tests used for resolution and contrast checks of the VT-1 equipment. This revision essentially removed a requirement that was an addition to ASME BPV Code that required 1-mil wires to be used in licensees' Sensitivity, Resolution, and Contrast Standard targets. In 2004, the NRC published NUREG/CR-6860, “An Assessment of Visual Testing,” showing that a linear target, such as a wire, is not an effective method for testing the resolution of a video camera system. In addition, Boiling Water Reactor Vessel and Internals Project Report 105696 (BWRVIP-03) was changed to eliminate a

1/2

mil wire from the Sensitivity, Resolution, and Contrast Standards due to similar concerns.

Simple line detection can be a poor performance standard, allowing detection of a highly blurred image. This does not emulate sharpness quality recognition for evaluation of weld discontinuities. The 750 μm (30 mil) and the even smaller 25 μm (1 mil) widths should not be used as performance standards because they do not determine image sharpness. This technique only measures the “visible minimum” for long linear indications, and does not measure a system's resolution or recognition limits. If the wire, or printed line, has a strong enough contrast against the background, then a linear feature well below the resolution of a system can be detected.

10 CFR 50.55a(b)(2)(xxiii) Section XI Condition: Evaluation of Thermally Cut Surfaces

The NRC is revising § 50.55a(b)(2)(xxiii) to clarify that this condition, prohibiting the ASME BPV Code provisions allowing elimination of mechanical processing of thermally cut surfaces under certain circumstances, only applies to the 2001 Edition through the 2009 Addenda.

10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator Preservice Examinations

In the proposed rule, the NRC proposed adding § 50.55a(b)(2)(xxx), with a condition regarding steam generator preservice examinations. The NRC received requests for clarification of the proposed condition, including elaboration on the kind of preservice examination that should be performed. The NRC agrees with the need for this clarification; however, during the development of the final rule, the NRC determined that additional time was needed to evaluate this proposed condition. Therefore, to ensure that this rulemaking is concluded as timely as possible, the NRC is not including this condition in this final rule and will address the need for a condition in a future rulemaking. The NRC has concluded that omitting this condition does not present a health or safety concern because licensees are currently performing appropriate steam generator preservice inspections under existing programs.

10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping Devices

The NRC is adding § 50.55a(b)(2)(xxxi) to require the use of Nonmandatory Appendix W when using a mechanical clamping device on an ASME BPV Code Class piping system. This condition, in part, clearly prohibits the use of mechanical clamping devices on small item Class 1 piping and portions of piping systems that form the containment boundary. This condition also maintains the previously required design and testing requirements for the implementation of mechanical clamping devices on ASME BPV Code Class piping systems.

In the 2010 Edition of the ASME BPV Code, a change was made to include mechanical clamping devices under the small items exclusion rules of IWA-4131. Currently in the 2007 Edition/2008 Addenda of Section XI under IWA-4133, “Mechanical Clamping Devices Used as Piping Pressure Boundary,” mechanical clamping devices may be used only if they meet the requirements of Mandatory Appendix IX of Section XI of the ASME BPV Code. Article IX-1000 (c) of Appendix IX prohibits the use of mechanical clamping devices on (1) Class 1 piping and (2) portions of a piping system that form the containment boundary.

In the 2010 Edition, IWA-4133 was modified to allow use of IWA-4131.1(c) for the installation of mechanical clamping devices. This change allowed the use of small items exclusion rules in the installation of mechanical clamping devices. Subparagraph IWA-4131.1(c) was added such that mechanical clamping devices installed on items classified as “small items” under IWA-4131, including Class 1 piping and portions of a piping system that form the containment boundary, would be allowed without a repair/replacement plan, pressure testing, services of an Authorized Inspection Agency, and completion of the NIS-2 form. The NRC, in accordance with the previously approved IWA-4133 of the 2007 Edition/2008 Addenda of the ASME BPV Code, does not believe that the ASME has provided a sufficient technical basis to support the use of mechanical clamping devices on Class 1 piping or portions of a piping system that form the containment boundary as a permanent repair. Furthermore, the NRC finds that the ASME has not provided any basis for the small item exemption allowing the installation of mechanical clamps on these components. In the 2011 Addenda of the ASME BPV Code, IWA-4131.1(c) was relocated to IWA-4131.1(d). To add clarity to the condition, the NRC has included statements such that the implementation of these paragraphs is now prohibited.

In the 2013 Edition, Mandatory Appendix IX of Section XI of the ASME BPV Code was changed to Nonmandatory Appendix W of Section XI of the ASME BPV Code. The NRC found insufficient basis to make this change, removing the mandatory requirements for the use of mechanical clamping devices on ASME BPV Code Class piping systems. By taking this action, the ASME BPV Code now allows mechanical clamping devices to be installed in various methods through interpretations of the ASME BPV Code that do not maintain the requirements for design and testing of the formerly mandatory Appendix IX. Therefore, to clarify the requirement for the implementation of mechanical clamps in ASME BPV Code Class systems, the NRC requires the use of Appendix W of Section XI when using mechanical clamping devices, and prohibits the use of mechanical clamping devices on small item Class 1 piping and portions of a piping system that form the containment boundary, as would otherwise be permitted under IWA-4131.1(c) in the 2010 Edition and IWA-4131.1(d) in the 2011 Addenda through 2013 Edition.

10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report Submittal

The NRC is adding § 50.55a(b)(2)(xxxii) to require licensees using the 2010 Edition and later editions and addenda of Section XI to continue to submit Summary Reports as required in IWA-6240 of the 2009 Addenda.

Prior to the 2010 Edition, Section XI required the preservice summary report to be submitted prior to the date of placement of the unit into commercial service, and the inservice summary report to be submitted within 90 calendar days of the completion of each refueling outage. In the 2010 Edition, IWA-6240 was revised to state, “Summary reports shall be submitted to the enforcement and regulatory authorities having jurisdiction at the plant site, if required by these authorities.” This change in the 2010 Edition could lead to confusion as to whether or not the summary reports need to be submitted to the NRC, as well as the time for submitting the reports, if they were required. The NRC concludes that summary reports must continue to be submitted to the NRC in a timely manner because they provide valuable information regarding examinations performed, conditions noted, corrective actions taken, and the implementation status of preservice inspection and ISI programs. Therefore, the NRC is adding § 50.55a(b)(2)(xxxii) to ensure that preservice and inservice summary reports will continue to be submitted within the timeframes currently

established in Section XI editions and addenda prior to the 2010 Edition.

10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed Allowable Pressure

The NRC is adding § 50.55a(b)(2)(xxxiii) to prohibit the use of Appendix G, Paragraph G-2216, in the 2011 Addenda and later editions and addenda of the ASME BPV Code, Section XI. The 2011 Addenda of the ASME BPV Code included, for the first time, a risk-informed methodology to compute allowable pressure as a function of inlet temperature for reactor heat-up and cool-down at rates not to exceed 100 degrees F/hr (56 degrees C/hr). This methodology was developed based upon probabilistic fracture mechanics (PFM) evaluations that investigated the likelihood of reactor pressure vessel (RPV) failure based on specific heat-up and cool-down scenarios.

During the ASME's consideration of this change, the NRC staff noted that additional requirements would need to be placed on the use of this alternative. For example, the NRC staff indicated that it would be important for a licensee who wishes to utilize such a risk-informed methodology for determining plant-specific pressure-temperature limits to ensure that the material condition of its facility is consistent with assumptions made in the PFM evaluations that supported the development of the methodology. One aspect of this would be evaluating plant-specific ISI data to determine whether the facility's RPV flaw distribution was consistent with the flaw distribution assumed in the supporting PFM evaluations. This consideration is consistent with a similar requirement established by the NRC in § 50.61a, “Alternative Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events.” The PFM methodology that supports § 50.61a is very similar to that which was used to support ASME BPV Code, Section XI, Appendix G, Paragraph G-2216. These concerns with the Paragraph G-2216 methodology for computing allowable pressure as a function of inlet temperature for reactor heat-up and cooldown were not addressed by the ASME. Accordingly, the NRC is prohibiting the use of Paragraph G-2216 in Appendix G of the 2010 Edition. The continued use of the deterministic methodology of Section XI, Appendix G to generate Pressure-Temperature (P-T) limits remains acceptable.

10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Nonmandatory Appendix U

The NRC is adding § 50.55a(b)(2)(xxxiv) to require that two conditions, (A) and (B), be satisfied when using Nonmandatory Appendix U of the 2013 Edition of the ASME BPV Code, Section XI. In the proposed rule, the NRC had proposed to exclude Nonmandatory Appendix U from the incorporation by reference and therefore not approve it for use. After considering public comments, the NRC has incorporated by reference Appendix U in this final rule because it integrates ASME BPV Code Cases N-513-3, “Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,” and N-705, “Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks Section XI, Division 1,” into Section XI. The NRC has approved the use of ASME BPV Code Cases N-513-3 and N-705 in RG 1.147, which allows licensees to use these code cases without prior permission from the NRC.

The first condition on the use of Appendix U is set forth in § 50.55a(b)(2)(xxxiv)(A) of this final rule and requires that an ASME BPV Code repair or replacement activity temporarily deferred under the provisions of Nonmandatory Appendix U to the 2013 Edition of the ASME BPV Code, Section XI, must be performed during the next scheduled outage. This condition is consistent with the NRC's condition on the use of ASME BPV Code Case N-513-3 in RG 1.147, Revision 17. Appendix U defines that the evaluation period is the operational time for which the temporary acceptance criteria are satisfied but not exceeding 26 months from the initial discovery of the condition. Original versions of ASME BPV Code Case N-513 stated, in part, that certain flaws may be acceptable without performing a repair/replacement activity for a limited time, not to exceed the time to the next scheduled outage. The NRC staff found that the acceptance of ASME BPV Code Case N-513 was based on allowing continued plant operation with a monitored and evaluated low safety significant degraded condition for a limited time until plant shutdown. By allowing use of this Appendix, this option is allowed rather than requiring an unnecessary plant shutdown to repair the degradation. However, the NRC believes once the plant is shut down, the degraded piping must be repaired.

The second condition on the use of Appendix U is set forth in § 50.55a(b)(2)(xxxiv)(B) of this final rule. This paragraph requires the use of the mandatory appendix in ASME BPV Code Case N-513-3 in lieu of the appendix referenced in paragraph U-S1-4.2.1(c) of Appendix U (which was inadvertently omitted from Appendix U). The NRC is incorporating by reference the mandatory appendix in ASME BPV Code Case N-513-3 in § 50.55a(a)(1)(iii)(A) because it is referenced in § 50.55a(b)(2)(xxxiv)(B).

A proposed condition on Disposition of Flaws in Class 3 Components, which was located in § 50.55a(b)(2)(xxxiv) of the proposed rule, is not included in this final rule based on public comments that the error has been corrected by ASME in published erratum.

10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RT

in the K

and K

Equations

The NRC is adding § 50.55a(b)(2)(xxxv) to specify that when licensees use the 2013 Edition of the ASME BPV Code, Section XI, Appendix A, Paragraph A-4200, if T

is available, then RT

may be used in place of RT

NDT

for applications using the K

equation and the associated K

curve, but not for applications using the K

equation and the associated K

curve.

Nonmandatory Appendix A provides a procedure based on linear elastic fracture mechanics (LEFM) for determining the acceptability of flaws that have been detected during inservice inspections that exceed the allowable flaw indication standards of IWB-3500. Sub-article A-4200 provides a procedure for determining fracture toughness of the material used in the LEFM analysis. The NRC staff's concern is related to the proposed insertion regarding an alternative based on the use of the Master Curve methodology to determine the nil-ductility transition reference temperature RT

NDT

, which is an important parameter in determining the fracture toughness of the material. Specifically, the insertion proposed to use the Master Curve reference temperature RT

, which is defined as RT

= T

+ 35 °F, where T

is a material-specific temperature value determined in accordance with ASTM E1921, “Standard Test Method for Determination of Reference Temperature, T

, for Ferritic Steels in the Transition Range,” to index (shift) the fracture toughness K

curve, based on the lower bound of static initiation critical stress intensity factor, as well as the K

curve, based on the lower bound

of crack arrest critical stress intensity factor.

While use of RT

to index the K

curve is acceptable, using RT

to index the K

curve is questionable. This concern is based on the data analysis in “A Physics-Based Model for the Crack Arrest Toughness of Ferritic Steels,” written by NRC staff member Mark Kirk and published in “Fatigue and Fracture Mechanics, 33rd Volume, ASTM STP 1417” which indicated that the crack arrest data does not support using RT

as RT

NDT

to index the K

curve. This is also confirmed by industry data disclosed in a presentation, “Final Results from the CARINA Project on Crack Initiation and Arrest of Irradiated German RPV Steels for Neutron Fluences in the Upper Bound,” by AREVA at the 26th Symposium on Effects of Radiation on Nuclear Materials (June 12-13, 2013, Indianapolis, Indiana, USA). The NRC staff recognized that the proposed insertion is consistent with ASME BPV Code Case N-629, “Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials,” which was accepted by the NRC without conditions. In addition to the current NRC effort, the appropriate ASME BPV Code committee is in the process of correcting this issue in a future revision of Appendix A of Section XI.

With this condition, users of Appendix A can avoid using an erroneous fracture toughness K

value in their LEFM analysis for determining the acceptability of a detected flaw in applicable components. Therefore, the NRC is adding a condition which permits the use of RT

in place of RT

NDT

in applications using the K

equation and the associated K

curve, but does not permit the use of RT

in place of RT

NDT

in applications using the K

equation and the associated K

curve.

10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of Irradiated Materials

The NRC is adding § 50.55a(b)(2)(xxxvi) to require licensees using ASME BPV Code, Section XI, 2013 Edition, Appendix A, Paragraph A-4400, to obtain NRC approval under § 50.55a(z) before using irradiated T

and the associated RT

in establishing fracture toughness of irradiated materials.

Sub-article A-4400 provides guidance for considering irradiation effects on materials. The NRC staff's concern is related to use of RT

based on measured T

of the irradiated materials. Specifically, the NRC staff has concerns over this sentence in the proposed insertion: “Measurement of RT

of unirradiated or irradiated materials as defined in A-4200(b) is permitted, including use of the procedures given in ASTM E1921 to obtain direct measurement of irradiated T

.”

Permission of measurement of RT

of irradiated materials, without providing guidelines regarding how to use the measured parameter in determining the fracture toughness of the irradiated materials, may mislead the users of Appendix A into adopting methodology that has not been accepted by the NRC. With this condition, users of Appendix A can avoid inappropriately using a fracture toughness K

value based on the irradiated T

and the associated RT

in their LEFM analysis for determining the acceptability of a detected flaw in applicable components.

10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements

The NRC is adding new paragraphs (g)(2)(i), (ii), and (iii) and revising current paragraphs (g) introductory text, (g)(2), (g)(3) introductory text, and (g)(3)(i), (ii), and (v) to distinguish the requirements for accessibility and preservice examination from those for inservice inspection in § 50.55a(g). In addition, consistent with other paragraphs of this section, headings are added to the subordinate paragraphs of (g) in order to enhance readability of the regulation. No substantive change to the requirements are intended by these revisions.

C. OM Code

10 CFR 50.55a(b)(3) Conditions on ASME OM Code

The NRC is revising § 50.55a(b)(3) to clarify that Subsections ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, III, and V; and Nonmandatory Appendices A through H and J through M of the OM Code are each incorporated by reference into § 50.55a. The NRC is also clarifying that the OM Code Nonmandatory Appendices incorporated by reference into § 50.55a are approved for use, but are not mandated. The Nonmandatory Appendices may be used by applicants and licensees of nuclear power plants, subject to the conditions in § 50.55a(b)(3).

10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance

The NRC is revising § 50.55a(b)(3)(i) to allow use of the 1994 Edition, 2008 Edition, and the 2009-1a Addenda of NQA-1, “Quality Assurance Requirements for Nuclear Facility Applications.” The NRC reviewed these editions and addenda, compared them to the previously approved versions of NQA-1, and found that there were no significant differences.

The NRC is removing the reference in § 50.55a(b)(3)(i) to versions of NQA-1 older than the 1994 Edition, inasmuch as these versions do not appear to be in use at any nuclear power plant. The NRC did not receive any adverse comments from any applicant or licensee regarding concerns about removing versions of NQA-1 older than the 1994 Edition from the regulation. The NRC received one comment regarding NQA-1, supporting incorporation by reference of NQA-1 but not responding to the NRC's request for comment regarding the removal of references to older versions of NQA-1. Accordingly, the NRC concludes that removal of NQA-1 versions older than the 1994 Edition will not have any adverse effect on licensees, and the final rule removes these older versions from § 50.55a(b)(3)(i).

10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) Testing

The NRC is revising § 50.55a(b)(3)(ii) to reflect the new Appendix III, “Preservice and Inservice Testing of Active Electric Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,” of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition. Appendix III of the OM Code establishes provisions for periodic verification of the design-basis capability of MOVs within the scope of the IST program. Appendix III of the OM Code reflects the incorporation of OM Code Cases OMN-1, “Alternative Rules for Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants,” and OMN-11, “Risk-Informed Testing for Motor-Operated Valves.” The NRC is adding four new conditions on the use of Mandatory Appendix III in new § 50.55a(b)(3)(ii)(A), (B), (C), and (D) to address periodic verification of MOV design-basis capability. These new conditions are discussed in the next four sections.

10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval (First Condition on Use of Mandatory Appendix III)

In the proposed rule, the NRC specified in § 50.55a(b)(3)(ii)(A) that licensees evaluate the adequacy of the diagnostic test interval for each MOV and adjust the interval as necessary, but not later than 5 years or three refueling outages (whichever is longer) from initial implementation of OM Code,

Appendix III. Paragraph III-3310(b) in Appendix III includes a provision stating that if insufficient data exist to determine the IST interval, then MOV inservice testing shall be conducted every two refueling outages or 3 years (whichever is longer) until sufficient data exist, from an applicable MOV or MOV group, to justify a longer IST interval. As discussed in a final rule published September 22, 1999 (64 FR 51386), with respect to the use of OM Code Case OMN-1, the NRC considers it appropriate to include a modification requiring licensees to evaluate the information obtained for each MOV, during the first 5 years or three refueling outages (whichever is longer) of the use of Appendix III to validate assumptions made in justifying a longer test interval.

In response to public comments, the NRC revised § 50.55a(b)(3)(ii)(A) to clarify its intent for licensees to evaluate the test interval within 5 years or three refueling outages (whichever is longer) following implementation of Appendix III to the OM Code, rather than implying that every MOV must be tested within 5 years or three refueling outages of the initial implementation of Appendix III. For example, the condition allows grouping of MOVs to share test information in the evaluation of the MOV periodic verification intervals within 5 years or three refueling outages (whichever is longer) of the implementation of OM Code, Appendix III. Therefore, § 50.55a(b)(3)(ii)(A) of this final rule states that licensees shall evaluate the adequacy of the diagnostic test intervals established for MOVs within the scope of OM Code, Mandatory Appendix III, not later than 5 years or three refueling outages (whichever is longer) from initial implementation of OM Code, Appendix III.

10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk (Second Condition on Use of Mandatory Appendix III)

The NRC is adding § 50.55a(b)(3)(ii)(B) to require that when using Mandatory Appendix III, licensees ensure that the potential increase in core damage frequency (CDF) and large early release frequency (LERF) associated with the extension is acceptably small when extending exercise test intervals for high risk MOVs beyond a quarterly frequency. As discussed in a final rule published September 22, 1999 (64 FR 51386), with respect to the use of OM Code Case OMN-1, the NRC considers it important for licensees to have sufficient information from the specific MOV, or similar MOVs, to demonstrate that exercising on a refueling outage frequency does not significantly affect component performance. The information may be obtained by grouping similar MOVs and establishing periodic exercising intervals of MOVs in the group over the refueling interval.

Section 50.55a(b)(3)(ii)(B) requires that the increase in the overall plant CDF and LERF resulting from the extension be acceptably small. As presented in RG 1.174, “An Approach for Using Probabilistic Risk Assessment [PRA] in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” the NRC considers acceptably small changes to be relative and to depend on the current plant CDF and LERF. For plants with total baseline CDF of 10

−

per year or less, acceptably small means CDF increases of up to 10

−

per year; and for plants with total baseline CDF greater than 10

−

per year, acceptably small means CDF increases of up to 10

−

per year. For plants with total baseline LERF of 10

−

per year or less, acceptably small LERF increases are considered to be up to 10

−

per year; and for plants with total baseline LERF greater than 10

−

per year, acceptably small LERF increases are considered to be up to 10

−

per year.

10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization (Third Condition on Use of Mandatory Appendix III)

The NRC is adding § 50.55a(b)(3)(ii)(C) to require, when applying Mandatory Appendix III, that licensees categorize MOVs according to their safety significance using the methodology described in OM Code Case OMN-3, “Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing of LWR Power Plants,” subject to the conditions discussed in RG 1.192, or using an MOV risk ranking methodology accepted by the NRC on a plant-specific or industry-wide basis in accordance with the conditions in the applicable safety evaluation. Paragraph III-3720 in Appendix III to the OM Code states that when applying risk insights, each MOV shall be evaluated and categorized using a documented risk ranking methodology. Further, Appendix III only addresses risk ranking methodologies that include two risk categories. In light of the potential extension of quarterly test intervals for high risk MOVs and the relaxation of IST activities for low risk MOVs based on risk insights, the NRC has determined that the rule should specify that plant-specific or industry-wide risk ranking methodologies must have been accepted by the NRC through RG 1.192 (which accepts OM Code Case OMN-3 with the specified conditions) or the issuance of safety evaluations. As noted in the response to public comments, the intent of this condition is to indicate that when applying Appendix III to the OM Code, licensees may use either a two-risk category approach (high or low) or a three-risk category approach (high, medium, and low), provided the risk ranking method has been accepted by the NRC.

10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time (Fourth Condition on Use of Mandatory Appendix III)

The NRC is adding § 50.55a(b)(3)(ii)(D) to require that when a licensee applies Paragraph III-3600, “MOV Exercising Requirements,” of Appendix III to the OM Code, the licensee verify that the stroke time of the MOV satisfies the assumptions in the plant's safety analyses. Previous editions and addenda of the OM Code specified that the licensee must perform quarterly MOV stroke time measurements that could be used to verify that the MOV stroke time satisfies the assumptions in the safety analyses consistent with plant TS. The need for verification of the MOV stroke time during periodic exercising is consistent with the NRC's lessons learned from the implementation of OM Code Case OMN-1. However, Paragraph III-3600 of Appendix III of the versions of the OM Code that will be incorporated by reference in this rulemaking no longer require the verification of MOV stroke time during periodic exercising. For this reason, the NRC is adopting this new condition, which will effectively retain the need to verify that the MOV stroke time during periodic exercising satisfies the assumptions in the plant's safety analyses.

Based on the discussion during the public webinar on March 2, 2016, the NRC revised the condition to clarify that it applies to MOVs referenced in the plant TS. In particular, the NRC revised the condition to indicate that when a licensee applies Paragraph III-3600 of Appendix III to the OM Code, the licensee shall verify that the stroke time of MOVs specified in plant technical specifications satisfies the assumptions in the plant's safety analyses.

10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors

The NRC is adding § 50.55a(b)(3)(iii) to apply specific conditions for IST programs applicable to licensees of new nuclear power plants in addition to the provisions of the OM Code as incorporated by reference with conditions in § 50.55a. Licensees of “new reactors” are, as identified in the paragraph: (1) Holders of operating

licenses for nuclear power reactors that received construction permits under this part on or after the date 12 months after August 17, 2017, and (2) holders of combined licenses (COLs) issued under 10 CFR part 52, whose initial fuel loading occurs on or after the date 12 months after August 17, 2017. This implementation schedule for new reactors is consistent with the NRC regulations governing inservice testing in § 50.55a(f)(4)(i).

Commission Papers SECY-90-016, “Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements;” SECY-93-087, “Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs;” SECY-94-084, “Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs;” and SECY-95-132, “Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs (SECY-94-084),” discuss IST programs for new reactors licensed under 10 CFR part 52.

In recognition of new reactor designs and lessons learned from nuclear power plant operating experience, the ASME is updating the OM Code to incorporate improved IST provisions for components used in nuclear power plants that were issued (or will be issued) construction permits, or COLs, on or following January 1, 2000 (defined in the OM Code as post-2000 plants). The first phase of the ASME effort incorporated IST provisions that specify full flow pump testing and other clarifications for post-2000 plants in the OM Code beginning with the 2011 Addenda. The second phase of the ASME effort incorporated preservice and inservice inspection and surveillance provisions for pyrotechnic-actuated (squib) valves in the 2012 Edition of the OM Code. The ASME is considering further modifications to the OM Code to address additional lessons learned from valve operating experience and new reactor issues. As described in the following paragraphs, § 50.55a(b)(3)(iii) will include four specific conditions which are discussed in the following paragraphs.

10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves

The NRC is adding § 50.55a(b)(3)(iii)(A) to require that licensees within the scope of § 50.55a(b)(3)(iii) periodically verify the capability of power-operated valves (POVs) to perform their design-basis safety functions. While Appendix III to the OM Code addresses this requirement for MOVs with the conditions specified in § 50.55a, applicable applicants and licensees will need to develop programs to periodically verify the design-basis capability of other POVs. The NRC's Regulatory Issue Summary 2000-03, “Resolution of Generic Issue 158: Performance of Safety-Related Power-Operated Valves Under Design Basis Conditions,” provides attributes for a successful long-term periodic verification program for POVs by incorporating lessons learned from MOV performance at operating nuclear power plants and research programs. Implementation of Appendix III to the OM Code as accepted in § 50.55a(b)(3)(ii) satisfies § 50.55a(b)(3)(iii)(A) for MOVs.

Section 50.55a(b)(3)(iii)(A) is consistent with the Commission policy for new reactors summarized in an NRC Staff Memorandum, “Consolidation of SECY-94-084 and SECY-95-132,” dated July 24, 1995, that (a) the design capability of safety-related POVs should be demonstrated by a qualification test prior to installation; (b) prior to initial startup, POV capability under design-basis differential pressure and flow should be verified by a pre-operational test; and (c) during the operational phase, POV capability under design-basis differential pressure and flow should be verified periodically through a program similar to that developed for MOVs in Generic Letter 89-10, “Safety-Related Motor-Operated Valve Testing and Surveillance,” dated June 28, 1989.

The NRC issued seven supplements to provide guidance for the implementation of the MOV testing program requested in Generic Letter 89-10. The supplements to Generic Letter 89-10 did not modify the substance of the MOV testing program requested in Generic Letter 89-10 to provide reasonable assurance in the capability of safety-related MOVs to perform their design-basis functions.

The condition in § 50.55a(b)(3)(iii)(A) specifies with the same level of detail as the condition in § 50.55a(b)(3)(ii) that nuclear power plant licensees must establish a program to ensure the continued capability of MOVs in performing their design-basis safety functions. When establishing the MOV periodic verification condition, the NRC provided guidance in the final rule published September 22, 1999 (64 FR 51370), for licensees to develop acceptable programs that would satisfy the MOV periodic verification condition. Similarly, the NRC staff is providing guidance herein for new reactor applicants and licensees to develop acceptable programs to periodically verify the capability of POVs to perform their design-basis safety functions.

In NUREG-2124, “Final Safety Evaluation Report [FSER] Related to the Combined Licenses for Vogtle Electric Generating Plant, Units 3 and 4,” the NRC staff found the provisions established by the COL applicant for Vogtle Units 3 and 4 in its Final Safety Analysis Report (FSAR), Revision 5, Section 3.9.6.2.2, “Valve Testing,” to periodically verify the capability of POVs (such as air-operated valves (AOVs), solenoid-operated valves (SOVs), and hydraulic-operated valves (HOVs)) to perform their design-basis safety functions to be acceptable. In particular, the Vogtle Units 3 and 4 FSAR specifies that:

Power-operated valves other than active MOVs are exercised quarterly in accordance with OM ISTC, unless justification is provided in the inservice testing program for testing these valves at other than Code mandated frequencies. Although the design basis capability of power-operated valves is verified as part of the design and qualification process, power-operated valves that perform an active safety function are tested again after installation in the plant, as required, to ensure valve setup is acceptable to perform their required functions, consistent with valve qualification. These tests, which are typically performed under static (no flow or pressure) conditions, also document the “baseline” performance of the valves to support maintenance and trending programs. During the testing, critical parameters needed to ensure proper valve setup are measured. Depending on the valve and actuator type, these parameters may include seat load, running torque or thrust, valve travel, actuator spring rate, bench set and regulator supply pressure. Uncertainties associated with performance of these tests and use of the test results (including those associated with measurement equipment and potential degradation mechanisms) are addressed appropriately. Uncertainties may be considered in the specification of acceptable valve setup parameters or in the interpretation of the test results (or a combination of both). Uncertainties affecting both valve function and structural limits are addressed. Additional testing is performed as part of the air-operated valve (AOV) program, which includes the key elements for an AOV Program as identified in the JOG AOV program document, Joint Owners Group Air Operated Valve Program Document, Revision 1, December 13, 2000 (References 203 and 204) [JOG AOV Program Document, Revision 1, December 13, 2000 (ADAMS Accession No. ML010950310), and NRC comment letter dated October 8, 1999, to Nuclear Energy Institute (ADAMS Accession No. ML020360077)]. The AOV program incorporates the attributes for a successful power-operated valve long-term periodic verification program, as discussed in Regulatory Issue Summary 2000-03, Resolution of Generic Safety Issue 158: Performance of Safety-Related Power-Operated Valves Under Design Basis

Conditions, by incorporating lessons learned from previous nuclear power plant operations and research programs as they apply to the periodic testing of air- and other power-operated valves included in the IST program.

For example, key lessons learned addressed in the AOV program include:

• Valves are categorized according to their safety significance and risk ranking.

• Setpoints for AOVs are defined based on current vendor information or valve qualification diagnostic testing, such that the valve is capable of performing its design-basis function(s).

• Periodic static testing is performed, at a minimum on high risk (high safety significance) valves, to identify potential degradation, unless those valves are periodically cycled during normal plant operation, under conditions that meet or exceed the worst case operating conditions within the licensing basis of the plant for the valve, which would provide adequate periodic demonstration of AOV capability. If required based on valve qualification or operating experience, periodic dynamic testing is performed to re-verify the capability of the valve to perform its required functions.

• Sufficient diagnostics are used to collect relevant data (

e.g.,

valve stem thrust and torque, fluid pressure and temperature, stroke time, operating and/or control air pressure, etc.) to verify the valve meets the functional requirements of the qualification specification.

• Test frequency is specified, and is evaluated each refueling outage based on data trends as a result of testing. Frequency for periodic testing is in accordance with References 203 and 204, with a minimum of 5 years (or 3 refueling cycles) of data collected and evaluated before extending test intervals.

• Post-maintenance procedures include appropriate instructions and criteria to ensure baseline testing is re-performed as necessary when maintenance on the valve, repair or replacement, have the potential to affect valve functional performance.

• Guidance is included to address lessons learned from other valve programs specific to the AOV program.

• Documentation from AOV testing, including maintenance records and records from the corrective action program are retained and periodically evaluated as a part of the AOV program.

The attributes of the AOV testing program described above, to the extent that they apply to and can be implemented on other safety-related power-operated valves, such as electro-hydraulic operated valves, are applied to those other power-operated valves.” (Vogtle Electric Generating Plant, Units 3 and 4, Updated Final Safety Analysis Report (UFSAR), Section 3.9.6, “Inservice Testing of Pumps and Valves”)

Applicable applicants and licensees may follow the method described in the Vogtle Units 3 and 4 FSAR in satisfying § 50.55a(b)(3)(iii)(A), or may establish a different method, subject to evaluation by the NRC during the licensing process or inspections.

10 CFR 50.55a(b)(3)(iii)(B) Check Valves

The NRC is adding § 50.55a(b)(3)(iii)(B) to require that licensees within the scope of § 50.55a(b)(3)(iii) perform bi-directional testing of check valves within the IST program where practicable. Nuclear power plant operating experience has revealed that testing check valves in only the flow direction can result in significant degradation, such as a missing valve disc, not being identified by the test. Nonmandatory Appendix M, “Design Guidance for Nuclear Power Plant Systems and Component Testing,” to OM Code, 2011 Addenda and 2012 Edition, includes guidance for the design of new reactors to enable bi-directional testing of check valves. New reactor designs will provide the capability for licensees of new nuclear power plants to perform bi-directional testing of check valves within the IST program. Bi-directional testing of check valves in new reactors, as required by § 50.55a(b)(3)(iii)(B), could be accomplished by valve-specific testing or condition monitoring activities in accordance with Appendix II to the OM Code as accepted in § 50.55a. The NRC is specifying this provision for bi-directional testing of check valves for new reactors in § 50.55a(b)(3)(iii)(B) to emphasize that new reactors should include the capability for bi-directional testing of check valves as part of their initial design.

10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration

In the proposed rule, the NRC proposed adding § 50.55a(b)(3)(iii)(C) to require that licensees subject to § 50.55a(b)(3)(iii) monitor flow-induced vibration (FIV) from hydrodynamic loads and acoustic resonance during preservice testing and inservice testing to identify potential adverse flow effects that might impact components within the scope of the IST program.

Nuclear power plant operating experience has revealed the potential for adverse flow effects from vibration caused by hydrodynamic loads and acoustic resonance on components in the reactor coolant, steam, and feedwater systems. Therefore, the licensee will be required to address potential adverse flow effects on safety-related pumps, valves, and dynamic restraints within the IST program in the reactor coolant, steam, and feedwater systems from hydraulic loading and acoustic resonance during plant operation. In response to public comments, the NRC revised § 50.55a(b)(3)(iii)(C) to clarify its intent that FIV monitoring of components may be conducted during preservice testing or inservice testing. This requirement will confirm that piping, components, restraints, and supports have been designed and installed to withstand the dynamic effects of steady-state FIV and anticipated operational transient conditions. As part of preservice testing activities, the initial test program may be used to verify that safety-related piping and components are properly installed and supported such that vibrations caused by steady-state or dynamic effects do not result in excessive stress or fatigue in safety-related plant systems.

In the Vogtle Units 3 and 4 FSER, the NRC staff found the provisions established by the COL applicant for Vogtle Units 3 and 4 in its FSAR, Revision 5, Section 3.9, “Mechanical Systems and Components,” Section 14.2.9, “Preoperational Test Descriptions,” and Section 14.2.10, “Startup Test Procedures,” with incorporation by reference of corresponding sections of the AP1000 Design Control Document (DCD), to monitor FIV from hydrodynamic loads and acoustic resonance during preservice testing or inservice testing to be acceptable. In particular, the NRC staff stated in the Vogtle Units 3 and 4 FSER:

AP1000 DCD Tier 2, Section 3.9.2, “Dynamic Testing and Analysis,” describes tests to confirm that piping, components, restraints, and supports have been designed to withstand the dynamic effects of steady-state FIV and anticipated operational transient conditions. Section 14.2.9.1.7, “Expansion, Vibration and Dynamic Effects Testing,” in AP1000 DCD Tier 2, Chapter 14, “Initial Test Program,” states that the purpose of the expansion, vibration and dynamic effects testing is to verify that safety-related, high energy piping and components are properly installed and supported such that, in addition to other factors, vibrations caused by steady-state or dynamic effects do not result in excessive stress or fatigue to safety-related plant systems. Nuclear power plant operating experience has revealed the potential for adverse flow effects from vibration caused by hydrodynamic loads and acoustic resonance on reactor coolant, steam, and feedwater systems. . . . In its response, SNC [Vogtle Units 3 and 4 COL applicant] stated that it intended to use the overall Initial Test Program to demonstrate that the plant has been constructed as designed and the systems perform consistent with design requirements. SNC referenced the provisions in the AP1000 DCD for vibration monitoring and testing to be implemented at VEGP. For example, the applicant notes that AP1000 DCD Tier 2, Section 3.9.2.1, “Piping Vibration, Thermal Expansion and Dynamic Effects,” specifies that the preoperational test

program for ASME BPV Code, Section III, Class 1, 2, and 3 piping systems simulates actual operating modes to demonstrate that components comprising these systems meet functional design requirements and that piping vibrations are within acceptable levels. SNC indicates that the planned vibration testing program described in AP1000 DCD Tier 2, Sections 14.2.9 and 14.2.10, with the preservice and IST programs described in AP1000 DCD Tier 2, Sections 3.9.3.4.4 and 3.9.6, will confirm component installation in accordance with design requirements, and address the effects of steady-state (flow-induced) and transient vibration to ensure the operability of valves and dynamic restraints in the IST Program. The NRC staff considers the response by SNC clarifies its application of the provisions in the AP1000 DCD to ensure that potential adverse flow effects will be addressed at VEGP. Therefore, the NRC staff considers Standard Content Open Item 3.9-5 to be resolved for the VEGP COL application.” (NUREG-2124, “Final Safety Evaluation Report Related to the Combined Licenses for Vogtle Electric Generating Plant, Units 3 and 4,” Section 3.9.6, “Inservice Testing of Pumps and Valves (Related to RG 1.206, Section C.III.1, Chapter 3, C.I.3.9.6, `Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints')”).

As clarified in the final rule in response to public comments, a licensee may monitor components for adverse FIV effects during preservice testing or IST activities.

Applicable applicants and licensees may either apply the methods described in the Vogtle Units 3 and 4 FSAR in satisfying § 50.55a(b)(3)(iii)(C) or develop their own plant-specific methods to satisfy § 50.55a(b)(3)(iii)(C) for NRC review during the licensing process.

10 CFR 50.55a(b)(3)(iii)(D) High-Risk Non-Safety Systems

The NRC is adding § 50.55a(b)(3)(iii)(D) to require that licensees within the scope of § 50.55a(b)(3)(iii) establish a program to assess the operational readiness of pumps, valves, and dynamic restraints within the scope of the Regulatory Treatment of Non-Safety Systems (RTNSS) for applicable reactor designs. As of the time of this final rule, these are designs which have been certified in a design certification rule under 10 CFR part 52. In SECY-94-084 and SECY-95-132, the Commission discusses RTNSS policy and technical issues associated with passive plant designs. Some new nuclear power plants have advanced light-water reactor (ALWR) designs that use passive safety systems that rely on natural forces, such as density differences, gravity, and stored energy to supply safety injection water and to provide reactor core and containment cooling. Active systems in passive ALWR designs are categorized as non-safety systems with limited exceptions. Active systems in passive ALWR designs provide the first line of defense to reduce challenges to the passive systems in the event of a transient at the nuclear power plant. Active systems that provide a defense-in-depth function in passive ALWR designs need not meet all of the acceptance criteria for safety-related systems. However, there should be a high level of confidence that these active systems will be available and reliable when needed. The combined activities to provide confidence in the capability of these active systems in passive ALWR designs to perform their functions important to safety are referred to as the RTNSS program. In the NRC Staff Memorandum, “Consolidation of SECY-94-084 and SECY-95-132,” dated July 24, 1995, the NRC staff provided a consolidated list of the approved policy and technical positions associated with RTNSS equipment in passive plant designs discussed in SECY-94-084 and SECY-95-132. This new paragraph specifies the need for licensees to assess the operational readiness of RTNSS pumps, valves, and dynamic restraints.

The July 24, 1995, staff memorandum summarizes the Commission policy positions related to inservice testing of RTNSS pumps and valves as follows:

The staff also concluded that additional inservice testing requirements may be necessary for certain pumps and valves in passive plant designs. The unique passive plant design relies significantly on passive safety systems, but also depends on non-safety systems (which are traditionally safety-related systems in current light-water reactors) to prevent challenges to passive systems. Therefore, the reliable performance of individual components is a very significant factor in enhancing the safety of passive plant design. The staff recommends that the following provisions be applied to passive ALWR plants to ensure reliable component performance.

1. Important non-safety-related components are not required to meet criteria similar to safety-grade criteria. However, the non-safety-related piping systems with functions that have been identified as being important by the RTNSS process should be designed to accommodate testing of pumps and valves to assure that the components meet their intended functions. Specific positions on the inservice testing requirements for those components will be determined as a part of the staff's review of plant-specific implementation of the regulatory treatment of non-safety systems for passive reactor designs.

2. . . . The vendors for advanced passive reactors, for which the final designs are not complete, have sufficient time to include provisions in their piping system designs to allow testing at power. Quarterly testing is the base testing frequency in the Code and the original intent of the Code. Furthermore, the COL holder may need to test more frequently than during cold shutdowns or at every refueling outage to ensure that the reliable performance of components is commensurate with the importance of the safety functions to be performed and with system reliability goals. Therefore, to the extent practicable, the passive ALWR piping systems should be designed to accommodate the applicable Code requirements for the quarterly testing of valves. However, design configuration changes to accommodate Code-required quarterly testing should be done only if the benefits of the test outweigh the potential risk.

3. The passive system designs should incorporate provisions (1) to permit all critical check valves to be tested for performance, to the extent practicable, in both forward- and reverse-flow directions, although the demonstration of a non-safety direction test need not be as rigorous as the corresponding safety direction test, and (2) to verify the movement of each check valve's obturator during inservice testing by observing a direct instrumentation indication of the valve position such as a position indicator or by using nonintrusive test methods.

4. . . . Similarly, to the extent practicable, the design of non-safety-related piping systems with functions under design-basis condition that have been identified as being important by the RTNSS process should incorporate provisions to periodically test power-operated valves in the system during operations to assure that the valves meet their intended functions under design-basis conditions.

5. . . . Mispositioning may occur through actions taken locally (manual or electrical), at a motor control center, or in the control room, and includes deliberate changes of valve position to perform surveillance testing. The staff will determine if and the extent to which this concept should be applied to MOVs in important non-safety-related systems when the staff reviews the implementation of the regulatory treatment of non-safety systems.” (NRC Staff Memorandum, “Consolidation of SECY-94-084 and SECY-95-132,” July 24, 1995, pages 26-28).

Consistent with the Commission policy for RTNSS equipment, § 50.55a(b)(3)(iii)(D) specifies that new reactor licensees shall assess the operational readiness of pumps, valves, and dynamic restraints within the RTNSS scope. This regulatory requirement will allow licensees flexibility in developing programs to assess operational readiness of RTNSS components that satisfy the Commission policy. Guidance on the implementation of the Commission policy for RTNSS equipment is set forth in NRC Inspection Procedure 73758, “Part 52, Functional Design and Qualification, and Preservice and Inservice Testing Programs for Pumps, Valves and

Dynamic Restraints,” dated April 19, 2013.

10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)

The NRC is revising § 50.55a(b)(3)(iv) to address Appendix II, “Check Valve Condition Monitoring Program,” provided in the 2003 Addenda through the 2012 Edition of the OM Code. In the proposed rule, the NRC proposed a condition in § 50.55a(b)(3)(iv) to provide assurance that the valve or group of valves is capable of performing its intended function(s) over the entire interval. Public comments indicated that the proposed condition could be misinterpreted. Therefore, the NRC revised the proposed condition to clarify that the implementation of Appendix II must include periodic sampling of the check valves over the maximum interval allowed by Appendix II for the check valve condition monitoring program. A new table was added to the paragraph to specify the maximum intervals between check valve condition monitoring activities when applying interval extensions.

The conditions currently specified for the use of Appendix II, 1995 Edition with the 1996 and 1997 Addenda, and 1998 Edition through the 2002 Addenda, of the OM Code remain unchanged by this final rule.

10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB

The NRC is adding a new condition, § 50.55a(b)(3)(vii), to prohibit the use of Subsection ISTB, “Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants,” in the 2011 Addenda of the OM Code. In the 2011 Addenda to the OM Code, the upper end of the “Acceptable Range” and the “Required Action Range” for flow and differential or discharge pressure for comprehensive pump testing in Subsection ISTB was raised to higher values. The NRC staff on the OM Code committee accepted the proposed increase of the upper end of the “Acceptable Range” and “Required Action Range” with the planned addition of a requirement for a pump periodic verification test program in the OM Code. However, the 2011 Addenda to the OM Code did not include the requirement for a pump periodic verification test program. Since then, the 2012 Edition of the OM Code has incorporated Mandatory Appendix V, “Pump Periodic Verification Test Program,” which supports the changes to the acceptable and required action ranges for comprehensive pump testing in Subsection ISTB. Therefore, the new § 50.55a(b)(3)(vii) prohibits the use of Subsection ISTB in the 2011 Addenda of the OM Code. Licensees will be allowed to apply Subsection ISTB with the revised acceptable and required action ranges in the 2012 Edition of the OM Code as incorporated by reference in § 50.55a.

10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE

The NRC is adding § 50.55a(b)(3)(viii) to specify that licensees who wish to implement Subsection ISTE, “Risk-Informed Inservice Testing of Components in Light-Water Reactor Nuclear Power Plants,” of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition, must request and obtain NRC approval in accordance with § 50.55a(z) to apply Subsection ISTE on a plant-specific basis as a risk-informed alternative to the applicable IST requirements in the OM Code.

In the 2009 Edition of the OM Code, the ASME included new Subsection ISTE that describes a voluntary risk-informed approach in developing an IST program for pumps and valves at nuclear power plants. If a licensee chooses to implement this risk-informed IST approach, Subsection ISTE indicates that all requirements in Subsection ISTA, “General Requirements,” Subsection ISTB, and Subsection ISTC, “Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants,” of the OM Code continue to apply, except those identified in Subsection ISTE. The ASME selected risk-informed guidance from OM Code Cases OMN-1, OMN-3, OMN-4, “Requirements for Risk Insights for Inservice Testing of Check Valves at LWR Power Plants,” OMN-7, “Alternative Requirements for Pump Testing,” OMN-11, and OMN-12, “Alternative Requirements for Inservice Testing Using Risk Insights for Pneumatically and Hydraulically Operated Valve Assemblies in Light-Water Reactor Power Plants,” in preparing Subsection ISTE of the OM Code.

During development of Subsection ISTE, the NRC staff participating on the OM Code committees indicated that the conditions specified in RG 1.192 for the use of the applicable OM Code Cases need to be considered when evaluating the acceptability of the implementation of Subsection ISTE. In addition, the NRC staff noted that several aspects of Subsection ISTE will need to be addressed on a case-by-case basis when determining the acceptability of its implementation. Therefore, the new condition in § 50.55a(b)(3)(viii) requires that licensees who wish to implement Subsection ISTE of the OM Code must request approval from the NRC to apply Subsection ISTE on a plant-specific basis as a risk-informed alternative to the applicable IST requirements in the OM Code.

Nuclear power plant applicants for construction permits under 10 CFR part 50, or combined licenses for construction and operation under 10 CFR part 52, may describe their proposed implementation of the risk-informed IST approach specified in Subsection ISTE of the OM Code for NRC review in their applications.

10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF

The NRC is adding a condition on the use of Subsection ISTF in § 50.55a(b)(3)(ix). First, the condition states that Subsection ISTF, 2011 Addenda, is prohibited for use. Second, the condition specifies that licensees applying Subsection ISTF, “Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants—Post-2000 Plants,” in the 2012 Edition of the OM Code shall satisfy the requirements of Mandatory Appendix V, “Pump Periodic Verification Test Program,” of the OM Code, 2012 Edition.

As previously discussed regarding the new condition in § 50.55a(b)(3)(vii), the upper end of the “Acceptable Range” and the “Required Action Range” for flow and differential or discharge pressure for comprehensive pump testing in Subsection ISTB in the OM Code was raised to higher values in combination with the incorporation of Mandatory Appendix V, “Pump Periodic Verification Test Program.” However, the 2011 Addenda of the OM Code does not include Appendix V. In addition, Subsection ISTF in the 2011 Addenda and 2012 Edition of the OM Code does not include a requirement for a pump periodic verification test program. Therefore, the new condition in § 50.55a(b)(3)(ix) requires that the provisions of Appendix V be applied when implementing Subsection ISTF of the 2012 Edition of the OM Code to support the application of the upper end of the Acceptable Range and the Required Action Range for flow and differential or discharge pressure for inservice pump testing in Subsection ISTF.

10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication

The NRC is adding § 50.55a(b)(3)(xi) to emphasize the provisions in OM Code, 2012 Edition, Subsection ISTC-3700, “Position Verification Testing,” to verify that valve obturator position is accurately indicated. Subsection ISTC-3700 of the OM Code requires that

valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated. Subsection ISTC-3700 states that where practicable, this local observation should be supplemented by other indications, such as the use of flow meters or other suitable instrumentation to verify obturator position. Subsection ISTC-3700 also states that where local observation is not possible, other indications shall be used for verification of valve operation. Nuclear power plant operating experience has revealed that reliance on indicating lights and stem travel are not sufficient to satisfy the requirement in ISTC-3700 to verify that valve operation is accurately indicated. Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR part 50 requires that where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency, and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. This new condition specifies that when implementing OM Code, Subsection ISTC-3700, licensees shall verify that valve operation is accurately indicated by supplementing valve position indicating lights with other indications, such as flow meters or other suitable instrumentation, to provide assurance of proper obturator position. The OM Code specifies obturator movement verification in order to detect certain internal valve failure modes consistent with the definition of `exercising' found in ISTA-2000, “Definitions,” (

i.e.,

demonstration that the moving parts of a component function). Verification of the ability of an obturator to change or maintain position is an essential element of valve operational readiness determination, which is a fundamental aspect of the OM Code.

The NRC initially emphasized the ASME OM Code requirement for valve position indication in 1995 in the original issuance of NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plants,” paragraph 4.2.5. The NRC's position is further elaborated in NUREG-1482 (Revision 2), “Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants,” paragraph 4.2.7. As discussed in NUREG-1482 (Revision 2), ISTC-3700 allows flexibility to licensees in verifying that operation of valves with remote position indicators is accurately indicated. For example, NUREG-1482 refers to various methods to verify valve operation, such as nonintrusive techniques, flow initiation or absence of flow, leak testing, and pressure testing. The extent of verification necessary for valve operation to satisfy ISTC-3700 will depend on the type of valve, the sophistication of the diagnostic equipment used in testing the valve, possible failure modes of the valve, and the operating history of the valve and similar valve types. To satisfy ISTC-3700, the licensee is responsible for developing and implementing a method to provide reasonable assurance that valve operation is accurately indicated.

The NRC is requiring this condition for the implementation of the 2012 Edition of the OM Code for the 120-month IST interval in order to allow additional time for licensees to comply with this condition.

10 CFR 50.55a(f): Preservice and Inservice Testing Requirements

The NRC is revising the introductory text of § 50.55a(f) to indicate that systems and components must meet the requirements for “preservice and inservice testing” in the applicable ASME Codes and that both activities are referred to as “inservice testing” in the remainder of paragraph (f). The change clarifies that the OM Code includes provisions for preservice testing of components as part of its overall provisions for IST programs. No expansion of IST program scope was intended by this clarification.

In the proposed rule, the NRC included references to the OM Code in § 50.55a(f)(3)(iii)(A), Class 1 Pumps and Valves: First Provision; § 50.55a(f)(3)(iii)(B), Class 1 Pumps and Valves: Second Provision; § 50.55a(f)(3)(iv)(A), Class 2 and 3 Pumps and Valves: First Provision; and § 50.55a(f)(3)(iv)(B): Second Provision; to align the regulatory language with the current ASME OM Code used for IST programs. Because § 50.55a(f)(3)(iii) and (iv) specifically reference Class 1, 2, or 3 pumps and valves, the proposed changes to these paragraphs referencing the OM Code are unnecessary and have not been adopted in this final rule.

10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for Operating Plants

The NRC is revising § 50.55a(f)(4) to clarify that this paragraph is applicable to pumps and valves that are within the scope of the OM Code. This revision aligns the scope of pumps and valves for inservice testing with the scope defined in the OM Code.

Public comments on the alignment of the IST program scope in § 50.55a(f)(4) indicated that the nuclear industry is addressing the requirements in 10 CFR part 50, appendices A and B, to establish an IST program for safety-related pumps and valves that are not classified as ASME BPV Code Class 1, 2, or 3 components through either the OM Code provisions or augmented IST programs. For example, one public commenter indicated that generally, augmented IST programs are designed to meet the OM Code where practicable, but relief requests are not required when alternate testing is necessary. The NRC regulations in § 50.55a address the concept of augmented IST programs for pumps and valves at nuclear power plants. For example, § 50.55a(f)(6)(ii), “Augmented IST requirements,” indicates that the licensee may follow an augmented IST program for pumps and valves for which the NRC deems that added assurance of operational readiness is necessary. The NRC finds that an augmented IST program as addressed in § 50.55a(f)(6)(ii) is acceptable for safety-related pumps and valves that are not classified as ASME BPV Code Class 1, 2, or 3 components.

Public commenters were concerned that the alignment of the scope of the OM Code and § 50.55a would cause a potential paperwork burden for the submittal of relief or alternative requests for safety-related pumps and valves that are not classified as ASME BPV Code Class 1, 2, or 3 components. In response to these comments, the NRC included a provision in § 50.55a(f)(4) that the IST requirements for pumps and valves that are within the scope of the OM Code but are not classified as ASME BPV Code Class 1, Class 2, or Class 3 may be satisfied as an augmented IST program in accordance with § 50.55a(f)(6)(ii) without requesting relief under § 50.55a(f)(5) or alternatives under § 50.55a(z). This use of an augmented IST program may be acceptable provided the basis for deviations from the OM Code, as incorporated by reference in this section, demonstrates an acceptable level of quality and safety, or that implementing the Code provisions would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, where documented and available for NRC review. This additional provision avoids the potential paperwork burden for the submittal of relief or alternative requests by allowing the licensee to maintain the documentation demonstrating an acceptable level of quality and safety on site for NRC review, as appropriate. The

documentation and availability of the basis for deviations from the OM Code for NRC review are acceptable for pumps and valves within the scope of the OM Code but not classified as ASME BPV Code Class 1, 2, or 3, based on their lower safety significance in comparison to ASME BPV Code Class 1, 2, and 3 pumps and valves.

10 CFR 50.55a(g)(4) Inservice Inspection Standards Requirement for Operating Plants

The NRC recognizes that updating an Appendix VIII program is a complex and time-consuming process. The NRC also recognizes that licensees would face the possibility of needing to maintain multiple Appendix VIII programs if units were to update their ISI programs on different dates. Maintaining certifications to multiple Appendix VIII programs would be very complicated, while not improving the effectiveness of the programs. Based on public comments, and to assist licensees in updating and coordinating their ISI programs, the NRC is adding two options to the regulations. First, the NRC is revising § 50.55a(g)(4)(i) and (ii) to clarify that a licensee whose ISI interval commences during the 12- to 18-month period after the approval date of this final rule, may delay the update of their Appendix VIII program by up to 18 months after the approval date of this final rule. This will provide licensees with enough time to incorporate the changes for the new Appendix VIII program. Second, the NRC is adding the option for licensees to update their ISI program to use the latest edition and addenda of Appendix VIII incorporated by reference in § 50.55a(a)(1) at any time in the licensee's ten-year interval. Licensees can normally update their ISI programs using all or portions of newer versions of ASME BPV Code Section XI under § 50.55a(g)(4)(iv), subject to NRC review and approval. While some requests to use portions of ASME BPV Code Section XI require a detailed review by the NRC, a licensee asking to use the entire latest incorporated-by-reference version of Appendix VIII would certainly be approved by the NRC staff in this process. This provision will, therefore, allow licensees to use the latest incorporated version of Appendix VIII, as long as it is coupled with the same edition and addenda of Appendix I, without the NRC review and approval process. This will allow licensees to coordinate their ISI programs and use the latest approved version of Appendix VIII without the delay imposed by submitting a relief request under § 50.55a (g)(4)(iv).

D. ASME Code Cases

Administrative Changes to References in § 50.55a to NRC Regulatory Guides Identifying ASME Code Cases Approved for Use by the NRC

The NRC is removing the revision number of the three RGs currently approved by the Office of the Federal Register for incorporation by reference throughout the substantive provisions of § 50.55a addressing the ASME Code Cases,

i.e.,

paragraphs (b) through (g). The revision numbers for the RGs approved for incorporation by reference (currently, RG 1.84, RG 1.147, and RG 1.192) will be retained in § 50.55a(a)(3)(i) through (iii), where the RGs are listed by full title, including revision number. These changes simplify the regulatory language containing cross-references to these RGs and reduce the possibility of NRC error in preparing future amendments to § 50.55a with respect to these RGs. These changes are administrative in nature and do not change substantive requirements with respect to the RGs and the Code Cases listed in the RGs.

Administrative Changes To Comply With Requirements for Incorporation by Reference

The NRC is revising § 50.55a(a)(1)(iii) to maintain the ASME Code Cases in alphanumeric order.

Organization of NRC's Discussion of the Six ASME Code Cases Incorporated by Reference in This Final Rule

The discussions under the following headings address four of the six ASME Code Cases being incorporated by reference in this rulemaking (N-729-4, N-770-2, N-824, and OMN-20). A fifth ASME Code Case, N-852, is discussed in Section II.A, “ASME BPV Code, Section III,” because the NRC's approval of that Code Case relates to a provision of Section III, which is addressed in § 50.55a(b)(1)(ix). The sixth ASME Code Case, N-513-3, is discussed in Section II.B, “ASME BPV Code, Section XI,” because the NRC's approval of that Code Case relates to a provision of Section XI, which is addressed in § 50.55a(b)(2)(xxxiv).

ASME BPV Code Case N-729-4

On September 10, 2008, the NRC issued a final rule to update § 50.55a to the 2004 Edition of the ASME BPV Code (73 FR 52730). As part of the final rule, § 50.55a(g)(6)(ii)(D) implemented an augmented ISI program for the examination of pressurized water reactor RPV upper head penetration nozzles and associated partial penetration welds. The program required the implementation of ASME BPV Code Case N-729-1, with certain conditions.

The application of ASME BPV Code Case N-729-1 was necessary because the inspections required by the 2004 Edition of the ASME BPV Code, Section XI were not written to address degradation of the RPV upper head penetration nozzles and associated welds by primary water stress corrosion cracking (PWSCC). The safety consequences of inadequate inspections can be significant. The NRC's determination that the ASME BPV Code required inspections are inadequate is based upon operating experience and analysis. The absence of an effective inspection regime could, over time, result in unacceptable circumferential cracking, or the degradation of the RPV upper head or other reactor coolant system components by leakage assisted corrosion. These degradation mechanisms increase the probability of a loss-of-coolant accident.

Examination frequencies and methods for RPV upper head penetration nozzles and welds are provided in ASME BPV Code Case N-729-1. The use of code cases is voluntary, so these provisions were developed, in part, with the expectation that the NRC would incorporate the code case by reference into the CFR. Therefore, the NRC adopted rule language in § 50.55a(g)(6)(ii)(D) requiring implementation of ASME BPV Code Case N-729-1, with conditions, in order to enhance the examination requirements in the ASME BPV Code, Section XI for RPV upper head penetration nozzles and welds. The examinations conducted in accordance with ASME BPV Code Case N-729-1 provide reasonable assurance that ASME BPV Code allowable limits will not be exceeded and that PWSCC will not lead to failure of the RPV upper head penetration nozzles or welds. However, the NRC concluded that certain conditions were needed in implementing the examinations in ASME BPV Code Case N-729-1. These conditions are set forth in § 50.55a(g)(6)(ii)(D).

On June 22, 2012, the ASME approved the fourth revision of ASME BPV Code Case N-729 (N-729-4). This revision changed certain requirements based on a consensus review of inspection techniques and frequencies. These changes were deemed necessary by the ASME to supersede the previous requirements under N-729-1 to establish an effective long-term inspection program for the RPV upper head penetration nozzles and associated welds in pressurized water reactors. The

major changes included incorporation of previous NRC conditions in the CFR. Minor changes were also made to address editorial issues, to correct figures or to add clarity.

The NRC is updating the requirements of § 50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV Code Case N-729-4, with conditions. One existing condition on ASME BPV Code Case N-729-1 has been modified, four existing conditions are being deleted in this final rule, one existing condition is being redesignated without substantive change, and two new conditions—in § 50.55a(g)(6)(ii)(D)(

) and (

)—are adopted in this final rule in order to address the changes in ASME BPV Code Case N-729-4. The NRC's revisions to the conditions are discussed under the next three headings. As discussed earlier, this final rule incorporates by reference ASME BPV Code Case N-729-4 into § 50.55a(a)(1)(iii)(C).

10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation

The NRC is revising § 50.55a(g)(6)(ii)(D)(

) to change the version of ASME BPV Code Case N-729 from N-729-1 to N-729-4 for the reasons previously set forth. Due to the incorporation of N-729-4, the date to establish applicability for licensed pressurized water reactors will be changed to the effective date of this final rule.

10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6) (Removed)

The NRC is removing the existing conditions in § 50.55a(g)(6)(ii)(D)(

) through (

) and redesignating the condition currently in § 50.55a(g)(6)(ii)(D)(

) as § 50.55a(g)(6)(ii)(D)(

) without any substantive change. The existing conditions in § 50.55a(g)(6)(ii)(D)(

) through (

) have all been incorporated either verbatim or more conservatively in the revisions to ASME BPV Code Case N-729, up to version N-729-4. Therefore, there is no reason to retain these conditions in § 50.55a.

10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency (New Condition)

The NRC is adopting a new condition in § 50.55a(g)(6)(ii)(D)(

) to modify the option in ASME BPV Code Case N-729-4 to extend bare metal visual inspections of the RPV upper head surface beyond the frequency listed in Table 1 of the Code Case. Previously, upper heads aged with less than eight effective degradation years were considered to have a low probability of initiating PWSCC, the cracking mechanism of concern. This ranking of effective degradation years was based on a simple time at temperature correlation. All of the upper heads within this category, with the exception of new heads using Alloy 600 penetration nozzles, were considered to have lower susceptibility to cracking due to the upper heads being at or near the cold leg operating temperature of the reactor coolant system. Therefore, these plants were referred to as having “cold heads.” All of the upper heads that had experienced cracking prior to 2006 were near the hot leg operating temperature of the reactor coolant system, which validated the time at temperature model.

In 2006, one of the 21 “cold head” plants identified two indications within a penetration nozzle and the associated partial penetration weld. Then, between 2006 and 2013, five of the 21 “cold head” plants identified multiple indications within fifteen different penetration nozzles and the associated partial penetration welds. None of these indications caused leakage, and volumetric examination of the penetration nozzles showed that no flaws in the nozzle material had grown through-wall; however, this increasing trend creates a reasonable safety concern.

Recent operational experience has shown that the volumetric inspection of penetration nozzles, at the current inspection frequency, is adequate to identify indications in the nozzle material prior to leakage; however, volumetric examinations cannot be performed on the partial penetration welds. Therefore, given the additional cracking identified at cold leg temperatures, the NRC staff has concerns about the adequacy of the partial penetration weld examinations.

Leakage from a partial penetration weld into the annulus between the nozzle and head material can cause corrosion of the low alloy steel head. While initially limited in leak rate, due to limited surface area of the weld being in contact with the annulus region, corrosion of the vessel head material can expose more of the weld surface to the annulus, allowing a greater leak rate. Since an indication in the weld cannot be identified by a volumetric inspection, a postulated crack through the weld, just about to cause leakage, could exist as a plant performed its last volumetric and/or bare metal visual examination of the upper head material. This gives the crack years to breach the surface and leak prior to the next scheduled visual examination.

Only a surface examination of the wetted surface of the partial penetration weld can reliably detect flaws in the weld. Unfortunately, this examination cannot size the flaws in the weld, and, if performed manually, requires significant radiological dose to examine all of the partial penetration welds on the upper head. As such, the available techniques are only able to detect a flaw after it has caused leakage. These techniques are a bare metal visual examination or a volumetric leak path assessment performed on the frequency of the volumetric examination.

Volumetric leak path examinations are only done during outages when a volumetric examination of the nozzle is performed. Therefore, under the current requirements allowed by Note 4 of ASME BPV Code Case N-729-4, leakage from a crack in the weld of a “cold head” plant could start and continue to grow for the 5 years between the required bare metal visual examinations to detect leakage through the partial penetration weld.

Given the additional cracking identified at cold leg temperatures of upper head penetration nozzles and associated welds, the NRC finds limited basis to continue to categorize these “cold head” plants as having a low susceptibility to crack initiation. The NRC is increasing the frequency of the bare metal visual examinations of “cold heads” to identify potential leakage as soon as reasonably possible due to the volumetric examination limitations. Therefore, the NRC is conditioning Note 4 of ASME BPV Code Case N-729-4 to require a bare metal visual exam during each outage in which a volumetric exam is not performed. The NRC also will allow “cold head” plants to extend their bare metal visual inspection frequency from once each refueling outage, as stated in Table 1 of N-729-1, to once every 5 years, but only if the licensee performed a wetted surface examination of all of the partial penetration welds during the previous volumetric examination. Applying the conditioned bare metal visual inspection frequency or a volumetric examination each outage will allow licensees to identify any potential leakage through the partial penetration welds prior to significant degradation of the low alloy steel head material, thereby providing reasonable assurance of the structural integrity of the reactor coolant pressure boundary.

These issues, including the operational experience, the fact that volumetric examination is not available to interrogate the partial penetration welds, and potential regulatory options, were discussed publicly at multiple ASME BPV Code meetings, at the annual Materials Programs Technical Information Exchange public meeting

held at the NRC Headquarters in June 2013, and at the 2013 NRC Regulatory Information Conference.

10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria (New Condition)

The NRC is adopting a new condition in § 50.55a(g)(6)(ii)(D)(

) to define surface examination acceptance criteria. Paragraph -3132(b) of ASME BPV Code Case N-729-4 sets forth the acceptance criteria for surface examinations. In general, throughout Section XI of the ASME BPV Code, the acceptance criteria for surface examinations default to Section III, Paragraph NB-5352, “Acceptance Standards.” Typically, for rounded indications, the indication was only unacceptable if it was greater than

3/16

-inch in size. The NRC requested that the code case authors include a requirement that any size rounded indication causing nozzle leakage is unacceptable due to operating experience identifying PWSCC under rounded indications less than

3/16

-inch in size.

Recently, the ASME BPV Code Committee approved an interpretation of the language in Paragraph -3132(b), which implied that any size rounded indication is acceptable unless there is relevant indication of nozzle leakage, even those greater than

3/16

-inch. The NRC does not agree with the interpretation and maintains its original position on rounded indications that any size rounded indication is unacceptable if there is an indication of leakage. Since the adoption of ASME BPV Code Case N-729-1 into § 50.55a(g)(6)(ii)(D), all licensees have used the NRC's position in implementing Paragraph -3132(b), even after the recent ASME BPV Code Committee interpretation approval over NRC objection.

Therefore, in order to ensure compliance with the previous and ongoing requirement, the NRC is revising condition § 50.55a(g)(6)(ii)(D)(

) to include clarity within the acceptance criteria for surface examinations. The current edition requirements of NB-5352 of ASME BPV Code, Section III for the licensee's ongoing 10-year inservice inspection interval shall be met.

ASME BPV Code Case N-770-2

On June 21, 2011 (76 FR 36232), the NRC issued a final rule, which included § 50.55a(g)(6)(ii)(F) that requires the implementation of ASME BPV Code Case N-770-1, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS N86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,” with certain conditions.

On June 9, 2011, the ASME approved the second revision of ASME BPV Code Case N-770 (N-770-2). The major changes from N-770-1 to N-770-2 included establishing new ASME BPV Code Case, Table 1, inspection item classifications for optimized weld overlays and allowing alternatives when complete inspection coverage cannot be met. Minor changes were also made to address editorial issues, to correct figures, or to add clarity. The NRC found that the updates and improvements in N-770-2 are sufficient to update § 50.55a(g)(6)(ii)(F).

The NRC, therefore, is updating the requirements of § 50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code Case N-770-2, with conditions. The NRC conditions have been modified to address the changes in ASME BPV Code Case N-770-2 and to ensure that this regulatory framework will provide adequate protection of public health and safety. The following sections discuss each of the NRC's changes to the conditions on ASME BPV Code Case N-770-2. As discussed earlier, this final rule incorporates by reference ASME BPV Code Case N-770-2 into § 50.55a(a)(1)(iii)(D).

10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation