# American Society of Mechanical Engineers 2015-2017 Code Editions Incorporation by Reference

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URL: https://www.frixlaw.com/law-library/documents/fr%3A2020-08855

## Record

- **Collection:** Federal Register
- **Document type:** Rule
- **Published:** May 4, 2020
- **Citation:** 85 FR 26540

## Text

NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[NRC-2016-0082]
RIN 3150-AJ74
American Society of Mechanical Engineers 2015-2017 Code Editions Incorporation by Reference

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Final rule.

SUMMARY:

The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to incorporate by reference the 2015 and 2017 Editions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and the 2015 and 2017 Editions of the American Society of Mechanical Engineers Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code: Section IST, for nuclear power plants. The NRC is also incorporating by reference two revised American Society of Mechanical Engineers code cases. This action is in accordance with the NRC's policy to periodically update the regulations to incorporate by reference new editions of the American Society of Mechanical Engineers 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 June 3, 2020. The incorporation by reference of certain publications listed in the regulation is approved by the Director of the Federal Register as of June 3, 2020.

ADDRESSES:

Please refer to Docket ID NRC-2016-0082 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 Website:
Go to
http://www.regulations.gov
and search for Docket ID NRC-2016-0082. 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.

•
Attention:
The Public Document Room (PDR), where you may examine and order copies of public documents is currently closed. You may submit your request to the PDR via email at
PDR.Resource@nrc.gov
or call 1-800-397-4209 between 8:00 a.m. and 4:00 p.m. (EST), Monday through Friday, except Federal holidays.

FOR FURTHER INFORMATION CONTACT:

James G. O'Driscoll, Office of Nuclear Material Safety and Safeguards, telephone: 301-415-1325, email:
James.O'Driscoll@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 the 2015 and 2017 Editions of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code) and the 2015 and 2017 Editions of the ASME Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code: Section IST (OM Code), for nuclear power plants. The NRC is also incorporating by reference two ASME code cases.

The ASME periodically revises and updates its codes for nuclear power plants by issuing new editions; this final rule is in accordance with the NRC's practice to incorporate those new editions into the NRC's regulations. This rule maintains the safety of nuclear power plants, makes NRC activities more effective and efficient, and allows nuclear power plant licensees and applicants to take advantage of the latest ASME Codes. The ASME is a voluntary consensus standards organization, and the ASME Codes are voluntary consensus standards. The NRC's use of the ASME Codes is consistent with applicable requirements of the National Technology Transfer and Advancement Act (NTTAA). See also Section XIV of this document, “Voluntary Consensus Standards.”

B. Major Provisions

Major provisions of this final rule include:

• Incorporation by reference of ASME Codes (2015 and 2017 Editions of the BPV Code and the OM Code) into NRC regulations and delineation of NRC requirements for the use of these codes, including conditions.

• Incorporation by reference of two revised ASME Code Cases and delineation of NRC requirements for the use of these code cases, including conditions.

• Incorporation by reference of Electric Power Research Institute (EPRI), Topical Report, “Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement” (MRP-335, Revision 3-A), which provides requirements for the mitigation of primary water stress corrosion cracking on reactor vessel head penetrations and dissimilar metal butt welds.

C. Costs and Benefits

The NRC prepared a regulatory analysis to determine the expected costs and benefits of this final rule. The regulatory analysis identifies costs and benefits in both a quantitative fashion as well as in a qualitative fashion.

Based on the analysis, the NRC concludes that this final rule results in a net quantitative averted cost to the industry and the NRC. This final rule, relative to the regulatory baseline, results in a net averted cost for industry of $3.06 million based on a 7 percent net present value (NPV) and $3.29 million based on a 3 percent NPV. The estimated incremental industry averted cost per reactor unit ranges from $34,000 based on a 7 percent NPV to $36,600 based on a 3 percent NPV. The rulemaking alternative benefits the NRC by averting costs for reviewing and approving requests to use alternatives to the codes on a plant-specific basis under § 50.55a(z) of title 10 of the
Code of Federal Regulations
(10 CFR). The NRC net benefit ranges from $3.17 million (based on a 7 percent NPV) to $3.61 million (based on a 3 percent NPV).

Qualitative factors that were considered include regulatory stability and predictability, regulatory efficiency, and consistency with the provisions of the NTTAA. The regulatory analysis includes a discussion of the costs and benefits that were considered qualitatively. If the results of the regulatory analysis were based solely on quantified costs and benefits, the regulatory analysis would show that the rulemaking is justified because the total

quantified benefits of the regulatory action equal or exceed the costs of the action. When the qualitative benefits (including the safety benefit, cost savings, and other non-quantified benefits) are considered together with the quantified benefits, the benefits outweigh the identified quantitative and qualitative impacts.

The NRC has a decades-long practice of approving and/or mandating the use of certain parts of editions and addenda of the ASME Codes in § 50.55a. Continuing this practice in this final rule ensures regulatory stability and predictability. The practice also provides consistency across the industry and assures the industry and the public that the NRC will continue to support the use of the most updated and technically sound techniques developed by the ASME to provide adequate protection to the public. In this regard, the ASME Codes are voluntary consensus standards developed by technical committees composed of mechanical engineers and others who represent the broad and varied interests of their industries, from manufacturers and installers to insurers, inspectors, distributors, regulatory agencies, and end users. The standards have undergone extensive external review before being considered to be incorporated by reference by the NRC. Finally, the NRC's use of the ASME Codes is consistent with the NTTAA, which directs Federal agencies to adopt voluntary consensus standards instead of developing “government-unique” (
i.e.,
Federal agency-developed) standards, unless inconsistent with applicable law or otherwise impractical.

For more information, please see the regulatory analysis in ADAMS under Accession No. ML19098A807.

TABLE OF CONTENTS

I. Background

II. Discussion

A. ASME BPV Code, Section III

B. ASME BPV Code, Section XI

C. ASME OM Code

D. ASME Code Cases

III. Public Outreach

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. Environmental Assessment and Final Finding of No Significant Environmental Impact

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 American Society of Mechanical Engineers develops and publishes the BPV Code, which contains requirements for the design, construction, and inservice inspection (ISI) of nuclear power plant components; and the ASME Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code: Section IST (OM Code),
1

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 ASME OM Code. Starting in 2012, the ASME decided to issue editions of its BPV and OM Codes (no addenda) every 2 years with the BPV Code to be issued on the odd years (
e.g.,
2013, 2015, etc.) and the OM Code to be issued on the even years
2

(
e.g.,
2012, 2014, etc.). The new editions and addenda typically revise provisions of the 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 code philosophy or approach.

1
The editions and addenda of the ASME
Code for Operation and Maintenance of Nuclear Power Plants
have had different titles from 2005 to 2017 and are referred to collectively in this rule as the “OM Code.”

2
The 2014 Edition of the ASME OM Code was delayed and was designated the 2015 Edition. Similarly, the 2016 Edition of the OM Code was delayed and was designated the 2017 Edition.

The NRC's practice is 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 BPV and OM Codes (ASME Codes) in § 50.55a of title 10 of the
Code of Federal Regulations
(10 CFR). The NRC approves 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 Codes' 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 NRC's regulations in a final rule dated July 18, 2017 (82 FR 32934) and amended January 18, 2018 (83 FR 2525).

The ASME Codes are consensus standards developed by participants, including the NRC and licensees of nuclear power plants, who have broad and varied interests. 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 the NRC's 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 determines 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 conditions the use of that provision when it incorporates by reference that ASME Code edition or addenda into its regulations. In some instances, 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 section. In a Staff Requirements Memorandum dated September 10, 1999, (ADAMS Accession No. ML003755050) 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 final rule, the provisions of the 2015 and 2017 Editions of Section III, Division 1; and the 2015 and 2017 Editions of Section XI, Division 1, of the ASME BPV Code; and the 2015 and 2017 Editions of the ASME OM Code that the NRC is not adopting, or is only 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 code cases that are the subject of this final rule 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 the NRC's compliance with the NTTAA is set forth in Section XIV of this document, “Voluntary Consensus Standards.”

II. Discussion

The NRC regulations incorporate by reference ASME 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. This final rule 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 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 subjects 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's 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 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 the 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 2015 and 2017 Editions to the ASME BPV Code, Section III, Division 1 and Section XI, Division 1, with conditions on their use.

• The 2015 and 2017 Editions to Division 1 of the ASME OM Code, with conditions on their use.

• ASME BPV Code Case N-729-6, “Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1,” ASME approval date: March 3, 2016, with conditions on its use.

• ASME BPV Code Case N-770-5, “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: November 7, 2016, with conditions on its use.

• “Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement” (MRP-335, Revision 3-A), EPRI approval date: November 2016, with conditions on its use.

The current regulations in § 50.55a(a)(1)(i) incorporate by reference ASME BPV Code, Section III, 1963 Edition through the 1970 Winter Addenda; and the 1971 Edition (Division 1) through the 2013 Edition (Division 1), subject to the conditions identified in current § 50.55a(b)(1)(i) through (b)(1)(ix). This final rule revises § 50.55a(a)(1)(i) to incorporate by reference the 2015 and 2017 Editions (Division 1) of the ASME BPV Code, Section III.

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 2013 Edition (Division 1), subject to the conditions identified in current § 50.55a(b)(2)(i) through (b)(2)(xxxvii). This final rule revises § 50.55a(a)(1)(ii) to remove exclusions from the incorporation by reference of specific paragraphs of the 2011a Addenda and the 2013 Edition of ASME BPV Code, Section XI, as explained in this final rule. This final rule also revises § 50.55a(a)(1)(ii) to incorporate by reference 2015 and 2017 Editions (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 final rule.

The current regulations in § 50.55a(a)(1)(iv) incorporate by reference ASME OM Code, 1995 Edition through the 2012 Edition, subject to the conditions currently identified in § 50.55a(b)(3)(i) through (b)(3)(xi). The NRC is revising § 50.55a(a)(1)(iv) to incorporate by reference the 2015 and 2017 Editions of Division 1 of the ASME OM Code. As a result, the NRC regulations incorporate by reference in § 50.55a the 1995 Edition through the 2017 Edition of the ASME OM Code.

The NRC is revising § 50.55a(a)(4) to include the Electric Power Research Institute, Materials Reliability Program, 3420 Hillview Avenue, Palo Alto, CA 94304-1338; telephone: 1-650-855-200;
http://www.epri.com,
as a new source for a standard incorporated by reference in § 50.55a.

The NRC reviewed changes to the Codes in the editions of the Codes identified in this final rule, 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 (83 FR 56156; November 9, 2018). 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 of the Codes, 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 of this document. The discussions are organized under the applicable ASME Code and Section.

The two ASME Code Cases being incorporated by reference in this final rule (N-729-6 and N-770-5) are discussed in Section II.D of this document, “ASME Code Cases.”

A. ASME BPV Code, Section III

10 CFR 50.55a(a)(1)(i)(E) Rules for Construction of Nuclear Facility Components—Division 1

The NRC is revising § 50.55a(a)(1)(i)(E) to incorporate by reference the 2015 and 2017 Editions of the ASME BPV Code, Section III, including Subsection NCA and Division 1 Subsections NB through NH (for the 2015 Edition) and Subsections NB through NG (for the 2017 Edition) and Appendices. As stated in § 50.55a(a)(1)(i), the Nonmandatory Appendices are excluded and not incorporated by reference. The Mandatory Appendices are incorporated by reference because they include information necessary for Division 1.

However, the Mandatory Appendices also include material that pertains to other Divisions that have not been reviewed and approved by the NRC. Although this information is included in the sections and appendices being incorporated by reference, the NRC notes that the use of Divisions other than Division 1 has not been approved, nor are these Divisions required by NRC regulations and, therefore, such information is not relevant to current applicants and licensees. Therefore, this rule clarifies that current applicants and licensees may only use the sections of the Mandatory Appendices that pertain to Division 1. The NRC is not taking a position on the non-Division 1 information in the appendices and is including it in the incorporation by reference only for convenience.

10 CFR 50.55a(b)(1)(v) Section III Condition: Independence of Inspection

The 1995 Edition through the 2009b Addenda of the 2007 Edition of ASME BPV Code, Section III, Subsection NCA, endorsed the NQA-1-1994 Edition (Nuclear Quality Assurance-1) in NCA-4000, “Quality Assurance.” Paragraph (a) of NCA-4134.10, “Inspection,” states, “The provisions of NQA-1 Basic Requirement 10 and Supplement 10S-1, shall apply, except for paragraph 3.1, and the requirements of Inservice Inspection.” Paragraph 3.1, “Reporting Independence,” of Supplement 10S-1, of NQA-1, states, “Inspection personnel shall not report directly to the immediate supervisors who are responsible for performing the work being inspected.” In the 2010 Edition through the latest ASME BPV Code Editions of NCA, the Code removed the paragraph 3.1 exception for reporting independence.

Based on the above changes to the Code, the NRC is revising the condition to limit the condition so that it is applicable only for the 1995 Edition through 2009b Addenda of the 2007 Edition, where the NQA-1-1994 Edition is referenced.

In response to public comments on the proposed revision to this condition, the NRC is revising the condition to clarify that that the condition applies to only paragraph 3.1 of Supplement 10S-1 of NQA-1-1994 Edition.

10 CFR 50.55a(b)(1)(vi) Section III Condition: Subsection NH

The NRC is revising the existing condition since Subsection NH of Section III Division 1 no longer exists in the 2017 Edition of ASME BPV Code, Section III Division 1. The change is to reflect that Subsection NH existed from the 1995 Addenda through 2015 Edition of Section III Division 1. In 2015, Subsection NH contents also were included in Section III Division 5 Subpart B. In the 2017 Edition of the ASME Code, Subsection NH was deleted from Division 1 of Section III and became part of Division 5 of Section III. Division 5 of Section III is not incorporated by reference in § 50.55a. Therefore, the NRC is revising the condition to make it applicable to the 1995 Addenda through all Editions and addenda up to and including the 2013 Edition.

10 CFR 50.55a(b)(1)(x) Section III Condition: Visual Examination of Bolts, Studs, and Nuts

Visual examination is one of the processes for acceptance of a bolt, stud, or nut to ensure its structural integrity and its ability to perform its intended function. The 2015 Edition of the ASME Code contains this requirement; however, the 2017 Edition does not require these visual examinations to be performed in accordance with NX-5100 and NX-5500. Therefore, the NRC is adding two conditions to ensure adequate procedures remain and qualified personnel remain capable of determining the structural integrity of these components.

10 CFR 50.55a(b)(1)(x) Section III Condition: Visual Examination of Bolts, Studs, and Nuts, First Provision

The NRC is adding § 50.55a(b)(1)(x) to condition the provisions of NB-2582, NC-2582, ND-2582, NE-2582, NF-2582, NG-2582 in the 2017 Edition of Section III. The condition is that visual examinations are required to be performed in accordance with procedures qualified to NB-5100, NC-5100, ND-5100, NE-5100, NF-5100, and NG-5100, and personnel qualified to NB-5500, NC-5500, ND-5500, NE-5500, NF-5500, and NG-5500, respectively. The 2015 Edition of the ASME Code contains this requirement. The visual examination is one of the processes for acceptance of the final product to ensure its structural integrity and its ability to perform its intended function. The 2017 Edition does not require these visual examinations to be performed in accordance with NX-5100 and NX-5500. All other final examinations (magnetic particle testing (MT), liquid penetrant testing (PT), ultrasonic testing (UT) and radiographic testing (RT)) for acceptance of the final product in the 2017 Edition require the procedures and personnel to be qualified to NX-5100 and NX-5500.

Therefore, the NRC is adding § 50.55a(b)(1)(x)(A) to condition the provisions of NB-2582, NC-2582, ND-2582, NE-2582, NF-2582,and NG-2582 in the 2017 Edition of Section III to require that procedures are qualified to NB-5100, NC-5100, ND-5100, NE-5100, NF-5100, and NG-5100, and personnel are qualified to NB-5500, NC-5500, ND-5500, NE-5500, NF-5500, and NG-5500, respectively, in order to ensure adequate procedures and personnel remain capable of determining the structural integrity of these components. This is particularly important for small bolting, studs, and nuts that only receive a visual examination. As stated in NX-4123 of Section III, only inspections performed in accordance with Article NX-4000 (
e.g.,
marking, dimensional measurement, fitting, alignment) are exempted from NX-5100 and NX-5500, and may be qualified in accordance with the Certificate Holder's Quality Assurance Program.

10 CFR 50.55a(b)(1)(x) Section III Condition: Visual Examination of Bolts, Studs, and Nuts, Second Provision

The 2017 Edition requires that the final surfaces of threads, shanks, and heads be visually examined for workmanship, finish, and appearance in accordance with ASTM F788, for bolting material, and ASTM F812, for nuts. This examination is for acceptance of the final product to ensure its structural integrity, especially for small bolting that only receives a visual examination. However, performing an inspection for workmanship or appearance to the bolting specification is not necessarily sufficient to ensure the integrity of the bolts and nuts for their intended function in a reactor. The visual examination in Section III for bolting and nuts is intended to determine structural integrity for its intended function, which may entail quality requirements more stringent than the bolting specifications. As specified in the 2015 Edition of Section III: “discontinuities such as laps, seams, or cracks that would be detrimental to the intended service are unacceptable.”

Therefore, the NRC is adding § 50.55a(b)(1)(x)(B) to condition the provisions of NB-2582, NC-2582, ND-2582, NE-2582, NF-2582, and NG-2582 in the 2017 Edition of Section III, to require that bolts, studs, and nuts must be visually examined for discontinuities including cracks, bursts, seams, folds, thread lap, voids and tool marks.

10 CFR 50.55a(b)(1)(xi) Section III Condition: Mandatory Appendix XXVI

The NRC is adding a new paragraph with conditions on the use of ASME BPV Code, Section III, Appendix XXVI, for installation of high density

polyethylene (HDPE) pressure piping. This Appendix is new in the 2015 Edition of Section III, and electrofusion joining was added to this Appendix in the 2017 Edition of Section III. The 2015 Edition of Section III is the first time the ASME Code has provided standards for the use of polyethylene piping. The NRC has determined that the conditions that follow in § 50.55a(b)(1)(xi)(A) through (C) are necessary in order to use polyethylene piping in Class 3 safety-related applications. The conditions in § 50.55a(b)(1)(xi)(A) and (B) pertain to butt fusion joints and apply to both the 2015 and 2017 Editions of Section III. The conditions in § 50.55a(b)(1)(xi)(C) pertain to electrofusion joints and apply only to the 2017 Edition of Section III.

Both NRC and industry-funded independent research programs have shown that joint failure is the most likely cause of structural failure in HDPE piping systems. Poorly manufactured joints are susceptible to early structural failure driven by “slow crack growth,” a form of subcritical creep crack growth that is active in HDPE. The following three provisions are aimed at ensuring the highest quality for joints in HDPE systems and reducing the risk of poor joint fabrication. These provisions minimize the risk of joint structural failure and the resulting potential loss of system safety function.

10 CFR 50.55a(b)(1)(xi)(A) Mandatory Appendix XXVI: First Provision

The NRC is adding a new paragraph (b)(1)(xi)(A), which specifies the essential variables to be used in qualifying fusing procedures for butt fusion joints in polyethylene piping installed in accordance with ASME Section III, Mandatory Appendix XXVI. The NRC does not endorse the use of a standardized fusing procedure specification. A fusion procedure specification will need to be generated for each butt fusion joint with the essential variables, as listed.

Per ASME BPV Code Section IX, QF-252, essential variables are those that will affect the mechanical properties of the fused joint, if changed, and require requalification of the Fusing Procedure Specification (FPS), Standard Fusing Procedure Specification (SFPS), or Manufacturer Qualified Electrofusion Procedure Specification (MEFPS) when any change exceeds the specified limits of the values recorded in the FPS for that variable. Fourteen essential variables for HDPE butt fusion joints for nuclear applications have been identified by NRC and industry experts through extensive research and field experience. Ten of these essential variables are the same as those identified in ASME BPV Code, Section IX, Table QF-254, which applies to all HDPE butt fusions and is not limited to nuclear applications. The other four variables deemed essential by the NRC are: Diameter, cross-sectional area, ambient temperature, and fusing machine carriage model. These four additional variables are recognized by industry experts as being essential for butt fusion joints in nuclear safety applications and have been included in a proposal to list essential variables for butt fusion in the 2019 Edition of ASME BPV Code, Section III, Mandatory Appendix XXVI.

For nuclear applications, the use of HDPE is governed by ASME BPV Code, Section III, Mandatory Appendix XXVI. The NRC has determined that to ensure butt fusion joint quality is adequate for nuclear safety applications, referencing ASME BPV Code, Section IX in ASME BPV Code, Section III, Mandatory Appendix XXVI is not sufficient, because ASME BPV Code, Section IX is not incorporated into NRC regulations. Therefore, the NRC is including the essential variables for HDPE butt fusion as a condition on the use of ASME BPV Code Section III, Mandatory Appendix XXVI. This provision addresses the fact that the essential variables for HDPE butt fusion are not listed in the 2015 and 2017 Editions of ASME BPV Code, Section III, Mandatory Appendix XXVI. Proposals to incorporate these essential variables for butt fusion in the 2019 Edition of the Code have already been drafted and circulated within the ASME Code Committees. In the meantime, the NRC is adding this provision to ensure butt fusion joint quality for nuclear safety applications.

10 CFR 50.55a(b)(1)(xi)(B) Mandatory Appendix XXVI: Second Provision

The NRC is adding a new paragraph (b)(1)(xi)(B), which requires bend tests or high speed tensile impact testing to qualify fusing procedures for joints in polyethylene piping installed in accordance with ASME BPV Code, Section III, Mandatory Appendix XXVI.

Based on limited confirmatory research on the inservice behavior of HDPE butt fusion joints, as well as research results from The Welding Institute in the UK, the NRC has determined the need to add a condition to ensure the quality of butt fusion joints. When performing procedure qualification for high speed tensile impact testing of butt fusion joints in accordance with XXVI-2300 or XXVI-4330, breaks in the specimen that are away from the fusion zone must be retested. When performing fusing operator qualification bend tests of butt fusion joints in accordance with XXVI-4342, guided side bend testing must be used for all thicknesses greater than 1.25 inches.

10 CFR 50.55a(b)(1)(xi)(C) Mandatory Appendix XXVI: Third Provision

The NRC is adding a new paragraph (b)(1)(xi)(C), which specifies the essential variables to be used in qualifying fusing procedures for electrofusion of fusion joints in polyethylene piping that is to be installed in accordance with ASME BPV Code, Section III, Mandatory Appendix XXVI. The NRC does not endorse the use of a standardized fusing procedure specification. A fusion procedure specification will need to be generated for each electrofusion joint with the essential variables as listed.

Per ASME BPV Code, Section IX, QF-252: “Essential variables are those that will affect the mechanical properties of the fused joint, if changed, and require requalification of the FPS, SFPS, or MEFPS when any change exceeds the specified limits of the values recorded in the FPS for that variable.” Sixteen essential variables for HDPE electrofusion for nuclear applications have been identified by NRC and industry experts through extensive research and field experience. Twelve of these essential variables are the same as those identified in ASME BPV Code, Section IX Table QF-255, which applies to all HDPE electrofusion and is not limited to nuclear applications. The other four variables deemed essential by the NRC are: Fitting polyethylene material, pipe wall thickness, power supply, and processor. These four additional variables are recognized by industry experts as being essential for electrofusion joints in nuclear safety applications and have been included in a proposal to list essential variables for electrofusion in the 2019 Edition of ASME BPV Code, Section III Mandatory Appendix XXVI.

For nuclear applications, the use of HDPE is governed by ASME BPV Code, Section III Mandatory Appendix XXVI. The NRC has determined that, to ensure electrofusion joint quality is adequate for nuclear safety applications, referencing ASME BPV Code, Section IX in ASME BPV Code, Section III Mandatory Appendix XXVI is not sufficient, because ASME BPV Code, Section IX is not incorporated into NRC regulations. Therefore, the NRC is including the essential variables for HDPE electrofusion as a condition on the use of ASME Section III, Mandatory Appendix XXVI. This provision addresses the fact that the essential

variables for HDPE electrofusion are not listed in the 2015 and 2017 Editions of ASME BPV Code, Section III, Mandatory Appendix XXVI. Proposals to incorporate these essential variables for electrofusion in the 2019 Edition of the Code have already been drafted and circulated within the ASME Code Committees. In the meantime, the NRC is adding this provision to ensure electrofusion joint quality for nuclear safety applications.

10 CFR 50.55a(b)(1)(xii) Section III Condition: Certifying Engineer

The NRC is adding a new condition § 50.55a(b)(1)(xii) Section III Condition:
Certifying Engineer.
In the 2017 Edition of ASME BPV Code, Section III, Subsection NCA, the following Subsections were updated to replace the term “Registered Professional Engineer,” with term “Certifying Engineer” to be consistent with ASME BPV Code Section III Mandatory Appendix XXIII.

• NCA-3255 “Certification of the Design Specifications”

• NCA-3360 “Certification of the Construction Specification, Design Drawings, and Design Report”

• NCA-3551.1 “Design Report”

• NCA-3551.2 “Load Capacity Data Sheet”

• NCA-3551.3 “Certifying Design Report Summary” and

• NCA-3555 “Certification of Design Report”

• Table NCA-4134.17-2, “Nonpermanent Quality Assurance Records”

• NCA-5125, “Duties of Authorized Nuclear Inspector Supervisors”

• NCA-9200, “Definitions”

The NRC reviewed these changes and has determined that the use of a Certifying Engineer instead of a Registered Professional Engineer applies only to non-U.S. nuclear facilities. The NRC has determined that a Certifying Engineer, who is also a Registered Professional Engineer licensed in one of the states of the United States, is acceptable for U.S. nuclear facilities regulated by the NRC. As a result, the NRC is adding a new condition to § 50.55a(b)(1), that would not allow applicants and licensees to use a Certifying Engineer who is not also a Registered Professional Engineer for code-related activities that are applicable to U.S. nuclear facilities regulated by the NRC.

B. ASME BPV Code, Section XI

10 CFR 50.55a(b)(2) Conditions on ASME BPV Code, Section XI

The NRC is amending the regulations in § 50.55a(b)(2) to incorporate by reference the 2015 and the 2017 Editions (Division 1) of the ASME BPV Code, Section XI. The current regulations in § 50.55a(b)(2) incorporate by reference ASME BPV Code, Section XI, 1970 Edition through the 1976 Winter Addenda; and the 1977 Edition (Division 1) through the 2013 Edition (Division 1), subject to the conditions identified in current § 50.55a(b)(2)(i) through (xxix). This final rule revises the introductory text to § 50.55a(b)(2) to incorporate by reference the 2015 Edition (Division 1) and the 2017 Edition (Division 1) of the ASME BPV Code, Section XI, clarifies the wording, and revises or provides some additional conditions.

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

The NRC is removing the existing condition § 50.55a(b)(2)(vi). A final rule was published in the
Federal Register
(61 FR 41303) on August 8, 1996, which incorporated by reference the ASME BPV Code, Section XI, Subsection IWE and Subsection IWL for the first time. The associated statements of consideration for that rule identified the 1992 Edition with 1992 Addenda of Subsection IWE and Subsection IWL as the earliest version that the NRC found acceptable. A subsequent rule published on September 22, 1999 (64 FR 51370), included the 1995 Edition with the 1996 Addenda as an acceptable edition of the ASME BPV Code. The statements of considerations for a later rule published on September 26, 2002 (67 FR 60520), noted that the 1992 Edition with the 1992 Addenda, or the 1995 Edition with the 1996 Addenda of Subsection IWE and IWL must be used when implementing the initial 120-month interval for the ISI of Class MC and Class CC components, and that successive 120-month interval updates must be implemented in accordance with § 50.55a(g)(4)(ii).

This requirement was in place to expedite the initial containment examinations in accordance with Subsections IWE and IWL, which were required to be completed during the 5-year period from September 6, 1996, to September 9, 2001. Now that there is an existing framework in place for containment examinations in accordance with Subsections IWE and IWL, there is no need for a condition specific to the initial examination interval. The examinations conducted during the initial interval can be conducted in accordance with § 50.55a(g)(4).

10 CFR 50.55a(b)(2)(vii) Section XI Condition: Section XI References to OM Part 4, OM Part 6, and OM Part 10 (Table IWA-1600-1).

The NRC is removing the existing condition § 50.55a(b)(2)(vii). This paragraph describes the editions and addenda of the ASME OM Code to be used with the Section XI references to OM Part 4, OM Part 6, and OM Part 10 in Table IWA-1600-1 of Section XI. The condition is applicable to the ASME BPV Code, Section XI, Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition. Paragraph (g)(4)(ii) requires that a licensee's successive 120-month inspection intervals comply with the requirements of the latest edition and addenda of the Code incorporated by reference in § 50.55a(b)(2). Because licensees are no longer using these older editions and addenda of the Code referenced in this paragraph, this condition can be removed.

10 CFR 50.55a(b)(2)(ix) Metal Containment Examinations

The NRC is revising § 50.55a(b)(2)(ix), to require compliance with new condition § 50.55a(b)(2)(ix)(K). The condition ensures containment leak-chase channel systems are properly inspected in accordance with the applicable requirements. The NRC specifies the application of this condition to all editions and addenda of Section XI, Subsection IWE, of the ASME BPV Code, prior to the 2017 Edition, that are incorporated by reference in paragraph (b) of § 50.55a.

10 CFR 50.55a(b)(2)(ix)(K) Metal Containment Examinations

The NRC is adding § 50.55a(b)(2)(ix)(K) to ensure containment leak-chase channel systems are properly inspected.

Regulations in § 50.55a(g), “Inservice Inspection Requirements,” require that licensees implement the inservice inspection program for pressure retaining components and their integral attachments of metal containments and metallic liners of concrete containments in accordance with Subsection IWE of Section XI of the applicable edition and addenda of the ASME Code, incorporated by reference in paragraph (b) of § 50.55a and subject to the applicable conditions in paragraph (b)(2)(ix). The regulatory condition in § 50.55a(b)(2)(ix)(A) or equivalent provision in Subsection IWE of the ASME Code (2006 and later editions and addenda only) requires that licensees 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.

The containment floor weld leak-chase channel system forms a metal-to-metal interface with the containment shell or liner, the test connection end of which is at the containment floor level. Therefore, the leak-chase system provides a pathway for potential intrusion of moisture that could cause corrosion degradation of inaccessible embedded areas of the pressure-retaining boundary of the basemat containment shell or liner within it. In addition to protecting the test connection, the cover plates and plugs and accessible components of the leak-chase system within the access box are also intended to prevent intrusion of moisture into the access box and into the inaccessible areas of the shell/liner within the leak-chase channels, thereby protecting the shell and liner from potential corrosion degradation that could affect leak-tightness.

The containment ISI program required by § 50.55a to be implemented in accordance with Subsection IWE, of the ASME Code, Section XI, subject to regulatory conditions, requires special consideration of areas susceptible to accelerated corrosion degradation and aging, and barriers intended to prevent intrusion of moisture and water accumulation against inaccessible areas of the containment pressure-retaining metallic shell or liner. The containment floor weld leak-chase channel system is one such area subject to accelerated degradation and aging if moisture intrusion and water accumulation is allowed on the embedded shell and liner within it. Therefore, the leak-chase channel system is subject to the inservice inspection requirements of § 50.55a(g)(4).

The NRC Information Notice (IN) 2014-07, “Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner,” (ADAMS Accession No. ML14070A114) discusses examples of licensees that did not conduct the required inservice inspections. The IN also summarizes the NRC's basis for including the leak-chase components within the scope of Subsection IWE, of the ASME Code, Section XI, and how licensees could fulfill the requirements. The NRC guidance explains that 100 percent of the accessible components of the leak-chase system should be inspected during each inspection period. There are three inspection periods in one ten-year inspection interval.

After issuance of IN 2014-07, the NRC received feedback during a public meeting between NRC and ASME management, held on August 22, 2014 (ADAMS Accession No. ML14245A003), noting that the IN guidance appeared to be in conflict with ASME Section XI Interpretation XI-1-13-10. In response to the comment during the public meeting, the NRC issued a letter to ASME (ADAMS Accession No. ML14261A051), which stated that the NRC found the provisions in the IN to be consistent with the requirements in the ASME Code; and the NRC may consider adding a condition to § 50.55a to clarify the expectations. The ASME responded to the NRC's letter (ADAMS Accession No. ML15106A627) and noted that a condition in the regulations may be appropriate to clarify the NRC's position.

Based on the operating experience summarized in IN 2014-07, and the industry feedback, the NRC has determined that a new condition is necessary in § 50.55a(b)(2)(ix) to clarify the NRC's expectations and to ensure steel containment shells and liners receive appropriate examinations. In the 2017 Edition of the ASME Code, a provision was added that clearly specifies the examination of leak-chase channels. The provision requires 100 percent examination of the leak-chase channel closures over a ten-year inspection interval, as opposed to 100 percent during each inspection period. Although the examination frequency is relaxed compared to the NRC's position as identified in IN 2014-07, the NRC finds the provision in the 2017 Edition acceptable because the examination includes provisions for scope expansion and examinations of additional closures if degradation is identified within an inspection period. The NRC chose to align the condition with the acceptable provision in the latest approved edition of the ASME Code.

This condition is applicable to all editions and addenda of the ASME Code prior to the 2017 Edition. The condition is being applied to all previous editions to clarify the NRC's position in the regulation. Licensees that are using a previous edition (
i.e.,
an edition prior to the 2015 Edition that has been incorporated by reference previously) of the ASME Code for their current IWE inspection program interval may continue to conduct the required inspections in accordance with the NRC's position identified in IN 2014-17 (
i.e.,
100 percent examination every inspection period), or licensees may implement the condition as described in this rule, as long as they can demonstrate that 100 percent of the inspections have been, or will be, completed within the current interval, as required by the condition.

10 CFR 50.55a(b)(2)(xvii) Section XI Condition: Reconciliation of Quality Requirements

The NRC is removing the condition found in the current § 50.55a(b)(2)(xvii). This paragraph describes requirements for reconciliation of quality requirements when purchasing replacement items. When licensees use the 1995 Addenda through 1998 Edition of ASME BPV Code, Section XI, this condition required replacement items to be purchased in accordance with the licensee's quality assurance program description required by § 50.34(b)(6)(ii), in addition to the reconciliation provisions of IWA-4200. The NRC has accepted without conditions the content of IWA-4200 in versions of the Code since the 1999 Addenda of Section XI. Paragraph 50.55a(g)(4)(ii) requires that licensee's successive 120-month inspection intervals comply with the requirements of the latest edition and addenda of the Code incorporated by reference in § 50.55a(b)(2). Subsequently, licensees are no longer using these older editions and addenda of the Code referenced in this paragraph therefore this condition can be removed. Section 50.55a(b)(2)(xvii) is designated as [Reserved].

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

The NRC is amending the condition found in § 50.55a(b)(2)(xviii) to extend the applicability of the condition through the latest edition incorporated by reference in paragraph (a)(1)(ii) of this section of ASME BPV Code, Section XI. This current condition prohibits those licensees which use ASME BPV Code, Section XI, 2011 Addenda through the 2013 Edition from using Appendix VII, Table VII-4110-1 and Appendix VIII, Subarticle VIII-2200. The condition requires licensees and applicants using these versions of Section XI to use the prerequisites for ultrasonic examination personnel certifications in Appendix VII, Table VII-4110-1 and Appendix VIII, Subarticle VIII-2200 in the 2010 Edition. This condition was added when the 2010 through the 2013 Edition was incorporated by reference. When ASME published the 2015 and 2017 Editions, Appendix VII, Table VII-4110-1 and Appendix VIII, Subarticle VIII-2200 of ASME BPV Code, Section XI were not modified in a way that would make it possible for the NRC to remove this condition. Therefore, the NRC is amending this condition to extend the applicability to the latest

edition incorporated by reference in paragraph (a)(1)(ii) of § 50.55a.

10 CFR 50.55a(b)(2)(xx)(B) System Leakage Tests: Second Provision

The NRC is amending the condition found in § 50.55a(b)(2)(xx)(B) to clarify the NRC's expectations related to the nondestructive examination (NDE) required when a system leakage test is performed (in lieu of a hydrostatic test) following repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest edition and addenda of ASME BPV Code, Section XI incorporated by reference in paragraph (a)(1)(ii) of § 50.55a. Industry stakeholders have expressed confusion as to what Code edition/addenda the requirements for NDE and pressure testing were required to satisfy. The NRC is modifying the condition to clarify that the NDE method (
e.g.,
surface, volumetric, etc.) and acceptance criteria of the 1992 Edition or later of ASME BPV Code, Section III shall be met. The actual nondestructive examination and pressure testing may be performed using procedures and personnel meeting the requirements of the licensee's/applicant's current ISI code of record. This condition was first put in place by the NRC in a final rule that became effective October 10, 2008 (73 FR 52730). The NRC determined the condition was necessary because the ASME BPV Code eliminated the requirement to perform the Section III NDE when performing a system leakage test in lieu of a hydrostatic test following repairs and replacement activities performed by welding or brazing on a pressure retaining boundary in the 2003 Addenda of ASME BPV Code, Section XI. When ASME published the 2015 Edition and the 2017 Editions, IWA-4520 was not modified in a way that would make it possible for the NRC to remove this condition. Therefore, the NRC is amending this condition to extend the applicability to the latest edition incorporated by reference in paragraph (a)(1)(ii) of § 50.55a.

10 CFR 50.55a(b)(2)(xx)(C) System Leakage Tests: Third Provision

The NRC is adding § 50.55a(b)(2)(xx)(C) to provide two conditions for the use of the alternative Boiling Water Reactor (BWR) Class 1 system leakage test described in IWB-5210(c) and IWB-5221(d) of the 2017 Edition of ASME Section XI. The first condition addresses a prohibition against the production of heat through the use of a critical reactor core to raise the temperature of the reactor coolant and pressurize the reactor coolant pressure boundary (RCPB) (sometimes referred to as nuclear heat). The second condition addresses the duration of the hold time when testing non-insulated components to allow potential leakage to manifest itself during the performance of system leakage tests.

The alternative BWR Class 1 system leakage test was intended to address concerns that performing the ASME-required pressure test for BWRs under shutdown conditions, (1) places the unit in a position of significantly reduced margin, approaching the fracture toughness limits defined in the Technical Specification Pressure-Temperature (P-T) curves, and (2) requires abnormal plant conditions/alignments, incurring additional risks and delays, while providing little added benefit beyond tests, which could be performed at slightly reduced pressures under normal plant conditions. However, due to restrictions imposed by the pressure control systems, most BWRs cannot obtain reactor pressure corresponding to 100 percent rated power during normal startup operations at low power levels that would be conducive to performing examinations for leakage. The alternative test would be performed at slightly reduced pressures and normal plant conditions, which the NRC finds will constitute an adequate leak examination and would reduce the risk associated with abnormal plant conditions and alignments.

However, the NRC has had a longstanding prohibition against the production of heat through the use of a critical reactor core to raise the temperature of the reactor coolant and pressurize the RCPB for the purpose of pressure testing. A letter dated February 2, 1990, from James M. Taylor, Executive Director for Operations, NRC, to Messrs. Nicholas S. Reynolds and Daniel F. Stenger, Nuclear Utility Backfitting and Reform Group (ADAMS Accession No. ML14273A002), established the NRC's position with respect to use of a critical reactor core to raise the temperature of the reactor coolant and pressurize the RCPB for the purpose of pressure testing. In summary, the NRC's position is that testing under these conditions involves serious impediments to careful and complete inspections and therefore creates inherent uncertainty with regard to assuring the integrity of the RCPB. Further, the practice is not consistent with basic defense-in-depth safety principles.

The NRC's position, established in 1990, was reaffirmed in IN No. 98-13, “Post-Refueling Outage Reactor Pressure Vessel Leakage Testing Before Core Criticality,” dated April 20, 1998. The IN was issued in response to a licensee that had conducted an ASME BPV Code, Section XI, leakage test of the reactor pressure vessel (RPV) and subsequently discovered that it had violated 10 CFR part 50, appendix G, paragraph IV.A.2.d. This regulation states that pressure tests and leak tests of the reactor vessel that are required by Section XI of the ASME Code must be completed before the core is critical. The IN references NRC Inspection Report 50-254(265)-97027 (ADAMS Accession No. ML15216A276), which documents that licensee personnel performing VT-2 examinations of the drywell at one BWR plant covered 50 examination areas in 12 minutes, calling into question the adequacy of the VT-2 examinations.

The bases for the NRC's historical prohibition of pressure testing with the core critical are summarized as follows:

1. Nuclear operation of a plant should not commence before completion of system hydrostatic and leakage testing to verify the basic integrity of the RCPB, a principal defense-in-depth barrier to the accidental release of fission products. In accordance with the defense-in-depth safety precept, the nuclear power plant design provides for multiple barriers to the accidental release of fission products from the reactor.

2. Hydrotesting must be done essentially water solid (
i.e.,
free of pockets of air, steam or other gases) so that stored energy in the reactor coolant is minimized during a hydrotest or leaktest.

3. The elevated reactor coolant temperatures, associated with critical operation, result in a severely uncomfortable and difficult working environment in plant spaces where the system leakage inspections must be conducted. The greatly increased stored energy in the reactor coolant, when the reactor is critical, increases the hazard to personnel and equipment in the event of a leak. As a result, the ability for plant workers to perform a comprehensive and careful inspection becomes greatly diminished.

However, the NRC has determined that pressure testing with the core critical is acceptable under the following conditions: when performed after repairs of a limited scope; where only a few locations or a limited area needs to be examined; and when ASME Code Section XI, Table IWB-2500-1, Category B-P (the pressure test required once per cycle of the entire RCPB) has been recently performed verifying the integrity of the overall RCPB. The NRC

also notes the alternative BWR Class 1 system leakage test does not allow for the use of the alternative test pressure following repairs/replacements on the RPV; therefore, it does not violate 10 CFR part 50, appendix G. The NRC has determined that the risk associated with nuclear heat at low power is comparable with the risk to the plant when the test is performed without nuclear heat (with the core subcritical) during mid-cycle outages, when decay heat must be managed. Performing the pressure test under shutdown conditions at full operating pressure without nuclear heat requires securing certain key pressure control, heat removal, and safety systems. It is more difficult to control temperature and pressure when there is significant production of decay heat (
e.g.,
after a mid-cycle outage), and may reduce the margin available to prevent exceeding the plant pressure-temperature limits.

When the pressure test is conducted using nuclear heat, the scope of repairs should be relatively small in order to minimize the personnel safety risk and to avoid rushed examinations. The alternative BWR Class 1 system leakage test does not place any restrictions on the size or scope of the repairs for which the alternative may be used, provided the alternative test pressure is not used to satisfy pressure test requirements following repair/replacement activities on the reactor vessel. It is impractical to specify a particular number of welded or mechanical repairs that would constitute a “limited scope.” However, if the plant is still in a refueling outage and has already performed the ASME Section XI Category B-P pressure test of the entire RCPB, it is likely that subsequent repairs would be performed only on an emergent basis, and would generally be of a limited scope. Additionally, the overall integrity of the RCPB will have been recently confirmed via the Category B-P test. For mid-cycle maintenance outages, the first condition allows the use of nuclear heat to perform the test, if the outage duration is 14 days or less. This would tend to limit the scope of repairs, and also limit the use of the code case to outages where there is a significant production of decay heat. Therefore, the first condition on the alternative BWR Class 1 system leakage test states: “The use of nuclear heat to conduct the BWR Class 1 system leakage test is prohibited (
i.e.
the reactor must be in a non-critical state), except during refueling outages in which the ASME Section XI Category B-P pressure test has already been performed, or at the end of mid-cycle maintenance outages fourteen (14) days or less in duration.”

With respect to the second condition and adequate pressure test hold time, the technical analysis supporting the alternative BWR Class 1 system leakage test indicates that the lower test pressure provides more than 90 percent of the flow that would result from the pressure corresponding to 100 percent power. However, a reduced pressure means a lower leakage rate, so additional time is required in order for there to be sufficient leakage to be observed by inspection personnel. Section XI, paragraph IWA-5213, “Test Condition Holding Time,” does not require a holding time for Class 1 components, once test pressure is obtained. To account for the reduced pressure, the alternative BWR Class 1 system leakage test would require a 15-minute hold time for non-insulated components. The NRC has determined that 15 minutes does not allow for an adequate examination because it is not possible to predict the entire range of scenarios or types of defects that could result in leakage. Some types of defects could result in immediate leakage, such as an improperly torqued bolted connection; however other types of defects, such as weld defects or tight cracks, could present a more torturous path for leakage and result in delayed leakage. Due to the uncertainty in the amount of time required for leakage to occur to an extent that it would be readily detectable by visual examination, the NRC has determined that it is appropriate to conservatively specify a longer hold time of 1 hour for non-insulated components. Therefore, the second condition for the alternative BWR Class 1 system leakage test requires a 1 hour hold time for non-insulated components.

10 CFR 50.55a(b)(2)(xxi) Section XI Condition: Table IWB-2500-1 Examination Requirements

The NRC is removing the condition found in § 50.55a(b)(2)(xxi)(A) to allow licensees to use the current editions of ASME BPV Code, Section XI, Table IWB 2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 (Inspection Program A) and Items B3.120 and B3.140 (Inspection Program B). These inspection categories concern pressurizer and steam generator nozzle inner radius section examinations. Previously, the condition required licensees to use the 1998 Edition, which required examination of the nozzle inner radius when using the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of § 50.55a. As these inspection requirements were removed in the ASME BPV Code in 1999, this change eliminates the requirement to examine the nozzle inner radii in steam generators and pressurizers.

The requirements for examinations of inner nozzle radii in several components were developed in the ASME BPV Code in reaction to the discovery of thermal fatigue cracks in the inner radius section of boiling water reactor feedwater nozzles in the late 1970's and early 1980's. As described in NUREG/CR-7153, “Expanded Materials Degradation Assessment (EMDA),” (ADAMS Accession Nos. ML14279A321, ML14279A461, ML14279A349, ML14279A430, and ML14279A331), and NUREG-0619-Rev-1, “BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking: Resolution of Generic Technical Activity A-10 (Technical Report),” (ADAMS Accession No. ML031600712), the service-induced flaws that have been observed are cracks at feedwater nozzles associated with mixing of lower-temperature water with hot water in a BWR vessel with rare instances of underclad and shallow cladding cracking appearing in pressurized water reactor (PWR) nozzles. Feedwater nozzle inner radius cracking has not been detected since the plants changed operation of the low flow feedwater controller. Significant inspections and repairs were required in the late 1970s and early 1980s to address these problems. The redesign of safe end/thermal sleeve configurations and feedwater spargers, coupled with changes in operating procedures, has been effective to date. No further occurrences of nozzle fatigue cracking have been reported for PWRs or BWRs.

When the new designs and operating procedures appeared to have mitigated the nozzle inner radius cracking, the ASME BPV Code, Section XI requirements to inspect steam generator and pressurizer nozzle inner radii were removed in the 1999 Addenda of ASME BPV Code, Section XI. Since the NRC imposed the condition requiring that these areas be inspected in 2002, no new cracking has been identified in steam generator or pressurizer nozzle inner radii. The NRC finds that the complete absence of cracking since the operational change provides reasonable assurance that the observed cracking was the result of operational practices that have been discontinued. Because the inner radius inspections were instituted solely based on the observed cracking and since the cracking mechanism has now been resolved through changes in operation, the NRC

finds that the intended purpose of the steam generator and pressurizer inner radius exams no longer exists and that the exams can be discontinued. In addition to operating experience, the NRC has reviewed the nozzle inner radii examinations as part of approving alternatives and granting relief requests concerning inspections of the pressurizer and steam generator nozzle inner radii. In the safety evaluations for proposed alternatives, the NRC has concluded that the fatigue analysis for a variety of plants shows that there is reasonable assurance that there will not be significant cracking at the steam generator or pressurizer nozzle inner radii before the end of the operating licenses of the nuclear power plants.

Therefore, based on the design changes, operating experiences, and analysis done by industry and the NRC, the NRC is removing § 55.55a(b)(2)(xxi)(A), which requires the inspection of pressurizer and steam generator nozzle inner radii.

10 CFR 50.55a(b)(2)(xxi)(B) Table IWB-2500-1 Examination Requirements

The NRC is adding a new paragraph (b)(2)(xxi)(B) that places conditions on the use of the provisions of IWB-2500(f) and (g) and Notes 6 and 7 of Table IWB-2500-1 of the 2017 Edition of ASME BPV Code, Section XI. These provisions allow licensees of BWRs to reduce the number of Item Number B3.90 and B3.100 components to be examined from 100 percent to 25 percent. These conditions require licensees using the provisions of IWB-2500(f) to maintain the evaluations that determined the plant satisfied the criteria of IWB-2500(f) as records in accordance with IWA-1400. The conditions prohibit use of a new provision in Section XI, 2017 Edition, Table 2500-1 Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.90 and B3.100, specific to BWR nuclear power plants with renewed operating licenses or renewed combined licensees in accordance with 10 CFR part 54. The final condition does not allow the use of these provisions to eliminate preservice or inservice volumetric examinations of plants with a Combined Operating License pursuant to 10 CFR part 52, or a plant that receives its operating license after October 22, 2015.

The addition of these provisions addresses the incorporation of Code Case N-702, “Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds Section XI, Division 1,” into the Code. The conditions are consistent with those in Regulatory Guide (RG) 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,” Revision 19.

The NRC finds that eliminating the volumetric preservice or inservice examination, as allowed by implementing the provisions of IWB-2500(g) and Note 7 of Table IWB-2500-1, is predicated on good operating experience for the existing fleet, which has not found any inner radius cracking in the nozzles within scope of the code case. New reactor designs do not have any operating experience; therefore, the condition ensures that new reactors will perform volumetric examinations of nozzle inner radii to gather operating experience.

10 CFR 50.55a(b)(2)(xxv) Section XI Condition: Mitigation of Defects by Modification

The NRC is amending the condition found in § 50.55a(b)(2)(xxv) to allow the use of IWA-4340 of ASME BPV Code, Section XI, 2011 Addenda through 2017 Edition with conditions. The modification of § 50.55a(b)(2)(xxv) adds paragraph (A) and continues the prohibition of IWA-4340 for Section XI editions and addenda prior to the 2011 Addenda. It adds paragraph (B), which contains the five conditions that the NRC is proposing to place on the use of IWA-4340 of Section XI, 2011 Addenda through 2017 Edition. In response to public comments, the NRC modified the third condition and added the fourth and fifth conditions.

10 CFR 50.55a(b)(2)(xxv)(A) Mitigation of Defects by Modification: First Provision

The NRC is adding paragraph (b)(2)(xxv)(A), which continues the prohibition of IWA-4340 for Section XI editions and addenda prior to the 2011 Addenda. IWA-4340 as originally incorporated into Section XI, Subsubarticle IWA-4340 did not include critical requirements that were incorporated into later editions of Section XI such as: (a) Characterization of the cause and projected growth of the defect; (b) verification that the flaw is not propagating into material credited for structural integrity; (c) prohibition of repeated modifications where a defect area grew into the material required for the modification; and (d) pressure testing. Therefore, the NRC prohibited the use of IWA-4340 in its original form. This new paragraph is necessary to maintain the prohibition because the NRC, as described in the following paragraph, is allowing the use of IWA-4340 of Section XI, 2011 Addenda through 2017 Edition.

10 CFR 50.55a(b)(2)(xxv)(B) Mitigation of Defects by Modification: Second Provision

The NRC is adding paragraph (b)(2)(xxv)(B) to allow the use of IWA-4340 of Section XI, 2011 Addenda through 2017 Edition with five conditions. The NRC finds that IWA-4340 as incorporated into later editions of Section XI was improved with requirements such as: (a) Characterization of the cause and projected growth of the defect; (b) verification that the flaw is not propagating into material credited for structural integrity; (c) prohibition of repeated modifications where a defect area grew into the material required for the modification; and (d) pressure testing. With inclusion of these requirements and those stated in the following conditions, the NRC concludes that there are appropriate requirements in place to provide reasonable assurance that the modification will provide an adequate pressure boundary, even while considering potential growth of the defect. The conditions and the basis for each are as follows:

• The first condition prohibits the use of IWA-4340 on crack-like defects or those associated with flow accelerated corrosion. The design requirements and potentially the periodicity of follow-up inspections might not be adequate for crack-like defects that could propagate much faster than defects due to loss of material. Therefore, the NRC is prohibiting the use of IWA-4340 on crack-like defects. Loss of material due to flow accelerated corrosion is managed by licensee programs based on industry standards. The periodicity of follow-up inspections is best managed by plant-specific flow accelerated corrosion programs. In addition, subparagraph IWA-4421(c)(2) provides provisions for restoring minimum required wall thickness by welding or brazing, including loss of material due to flow accelerated corrosion.

• The second condition requires the design of a modification that mitigates a defect to incorporate a loss of material rate either 2 times the actual measured corrosion rate in the location, or 4 times the estimated maximum corrosion rate for the piping system. Corrosion rates are influenced by local conditions (
e.g.,
flow rate, discontinuities). The condition to extrapolate a loss of material rate either 2 times the actual measured corrosion rate in the location, or 4 times the estimated maximum corrosion rate for the system is consistent with ASME Code Cases N-786-1, “Alternative Requirements for

Sleeve Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping,” and N-789, “Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service.”

• The third condition requires the licensee to perform a wall thickness examination in the vicinity of the modification and relevant pipe base metal during each refueling outage cycle to detect propagation of the defect into the material credited for structural integrity of the item, unless the examinations in the two refueling outage cycles subsequent to the installation of the modification are capable of validating the projected flaw growth. Where the projected flaw growth has been validated, the modification shall be examined at half its expected life or once per interval whichever is smaller. The NRC concludes that the provision allowed by subparagraph IWA-4340(g) to conduct follow-up wall thickness measurements only to the extent that they demonstrate that the defect has not propagated into the material credited for structural integrity is not sufficient because it does not provide a verification of the projected flaw growth. Subparagraph IWA-4340(h) does not fully address the NRC's concern because it allows for projected flaw growth to be based on “prior Owner or industry experiences with the same conditions” instead of specific measurements in the location of the modification. The condition allows for only conducting examinations in the two refueling outages subsequent to the installation of the modification, consistent with subparagraph IWA-4340(g), if the measurements are capable of projecting the flaw growth. In response to public comments on the proposed condition, the NRC recognized that the requirement in IWA-4340(i) to conduct an examination at the modification location every interval could be interpreted to not be required based on the “practicality” statement in the cross referenced IWA-4340(g). The NRC has concluded that even if the flaw growth has been confirmed, and as a result, refueling outage interval inspections are not being conducted, over time, flaw growth rates could possibly accelerate. Although there is significant margin in the analyses, the NRC added a requirement to this condition to examine the modification at half its expected life or once per interval, whichever is smaller, to ensure that the potential effect of varying flaw growth rates is managed.

In response to public comments on the proposed condition, the NRC recognized that it may be onerous to perform follow-up examinations every refueling outage for modifications installed in inaccessible locations. The NRC is adding exceptions to the condition for buried pipe locations at potentially reduced intervals. One exception allows wall thickness measurements at a comparable accessible piping location where loss of material has occurred due to internal corrosion and the second addresses loss of material due to external corrosion.

• For buried pipe locations where the loss of material has occurred due to internal corrosion, the refueling outage interval wall thickness examinations may be conducted at a different location in the same system as long as: (a) Wall thickness measurements were conducted at the different location at the same time as installation of the modification; (b) the flow profile and flow characteristics are similar at the different location; (c) the piping configuration is the same (
e.g.,
straight run of pipe, elbow, tee), and (d) if pitting occurred at the modification location, but not at the different location, wall loss values must be multiplied by four. Where wall loss values are greater than that assumed during the design of the modification, the structural integrity of the modification shall be reanalyzed. Additionally, if the extent of degradation is different (
i.e.,
through wall, percent wall loss plus or minus 25 percent) or the corrosion mechanism (
e.g.,
general, pitting) is not the same at the different location as at the modification location, the modification must be examined at half its expected life or 10 years, whichever is smaller.

• For buried pipe locations where loss of material has occurred due to external corrosion, the modification must be examined at half its expected life or 10 years, whichever is smaller. The NRC staff included this condition because for external corrosion, there is no comparable accessible location.

10 CFR 50.55a(b)(2)(xxvi) Section XI Condition: Pressure Testing Class 1, 2 and 3 Mechanical Joints

The NRC is amending the condition found in § 50.55a(b)(2)(xxvi) to clarify the NRC's expectations related to the pressure testing of ASME BPV Code Class 1, 2, and 3 mechanical joints disassembled and reassembled during the performance of an ASME BPV Code, Section XI activity. Industry stakeholders have expressed confusion with the current regulatory requirements with regard to when a pressure test is required and which year of the Code the pressure testing should be in compliance with in accordance with this condition. The NRC is modifying the condition to clarify that all mechanical joints in Class 1, 2 and 3 piping and components greater than NPS-1 that are disassembled and reassembled during the performance of a Section XI activity (
e.g.,
a repair/replacement activity requiring documentation on a Form NIS-2) shall be pressure tested in accordance with IWA-5211(a). The pressure testing shall be performed using procedures and personnel meeting the requirements of the licensee's/applicant's current code of record. This condition was first put in place by the NRC in the final rule effective November 1, 2004 (69 FR 58804). The NRC determined that the condition was necessary because the ASME BPV Code eliminated the requirements to pressure test Class 1, 2, and 3 mechanical joints undergoing repair and replacement activities in the 1999 Addenda. The NRC finds that pressure testing of mechanical joints affected by repair and replacement activities is necessary to ensure and verify the leaktight integrity of the system pressure boundary.

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

The NRC is amending the condition found in § 50.55a(b)(2)(xxxii) to address the use of Owner Activity Reports. Through the 2009 Edition of ASME BPV Code, Section XI, Owners were required to prepare Summary Reports of preservice and inservice examinations and repair replacement activities. This condition was added when the 2010 and the 2013 Edition was incorporated by reference because up until that time, Owners were required to submit these reports to the regulatory authority having jurisdiction of the plant site. The 2010 Edition removed the requirement for submittal from IWA-6240(c), to state that submittal was only mandatory if required by the authority. The NRC added the condition in paragraph (b)(2)(xxxii) to require submittal of Summary Reports. In the 2015 Edition of ASME BPV Code, Section XI the title of these reports was changed from Summary Reports to Owner Activity Reports. Therefore, the NRC is amending the condition to also require the submittal of Owner Activity Reports.

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

The NRC is amending the requirements in current paragraph (b)(2)(xxxiv) to make the condition applicable to the latest edition incorporated by reference in paragraph (a)(1)(ii) of § 50.55a. The current condition in paragraph (b)(2)(xxxiv)(A)

requires repair and replacement activities temporarily deferred under the provisions of Nonmandatory Appendix U to be performed during the next scheduled refueling outage. This condition was added when the 2013 Edition was incorporated by reference. When ASME published the 2015 Edition and the 2017 Edition, Nonmandatory Appendix U was not modified in a way that would make it possible for the NRC to remove this condition. Therefore, the NRC is modifying this condition to make it apply to the latest edition incorporated by reference in paragraph (a)(1)(ii) of § 50.55a.

The current condition in paragraph (b)(2)(xxxiv)(B) requires a mandatory appendix in ASME Code Case N-513-3 to be used as the referenced appendix for paragraph U-S1-4.2.1(c). This condition was also added when the 2013 Edition was incorporated by reference. The omission that made this condition necessary was remedied in the 2017 Edition. Therefore, the NRC is modifying this condition to make it apply only to the 2013 and the 2015 Editions.

10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RT
T0
in the K
Ia
and K
Ic
Equations

The NRC is redesignating the requirements in current paragraph (b)(2)(xxxv), that address the use of the 2013 Edition of ASME BPV Code, Section XI, Appendix A, paragraph A-4200, as (b)(2)(xxxv)(A). The ASME BPV Code has addressed the NRC concern related to this condition in the 2015 Edition; however, it is still relevant to licensees/applicants using the 2013 Edition. The NRC is adding a new paragraph (b)(2)(xxv)(B) to condition the use of 2015 Edition of ASME BPV Code, Section XI, Appendix A, paragraph A-4200(c), to require the use of the equation RT
KIa
= T
0
+ 90.267 exp(−0.003406T
0
) for U.S. Customary Units (U.S. Units) in lieu of the equation shown in the Code. Paragraph A-4200(c) was added in the 2015 Edition to provide for an alternative method in establishing a fracture-toughness-based reference temperature, RT
T0
, for pressure retaining materials, using fracture toughness test data. The equation shown for the International System of Units (SI Units) was derived from test data. The equation shown for U.S. Units was a converted version of the equation shown for the SI Units. Unfortunately, an error was made in the conversion, which makes the equation shown for U.S. Units incorrect. The equation shown above in this paragraph for RT
KIa
is the correct formula for U.S. Units.

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

The NRC is amending the condition found in § 50.55a(b)(2)(xxxvi) to extend the applicability to use of the 2015 and 2017 Editions of ASME BPV Code, Section XI. This current condition requires licensees using ASME BPV Code, Section XI, 2013 Edition, Appendix A, paragraph A-4400, to obtain NRC approval before using irradiated T
0
and the associated RT
T0
in establishing fracture toughness of irradiated materials. This condition was added when the 2013 Edition was incorporated by reference because the newly introduced A-4200(b) could mislead the users of Appendix A into adopting methodology that is not accepted by the NRC. When ASME published the 2015 Edition and the 2017 Edition, Appendix A of the ASME BPV Code, Section XI was not modified in a way that would make it possible for the NRC to remove this condition. Therefore, the NRC is modifying this condition to make it apply to the 2015 and 2017 Editions.

10 CFR 50.55a(b)(2)(xxxviii) Section XI Condition: ASME Code Section XI Appendix III Supplement 2

The NRC is adding § 50.55a(b)(2)(xxxviii) to condition ASME BPV Code, Section XI Appendix III Supplement 2. Supplement 2 is closely based on ASME Code Case N-824, “Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,” which was incorporated by reference with conditions in § 50.55a(b)(2)(xxxvii). The conditions on ASME BPV Code, Section XI Appendix III Supplement 2 are consistent with the conditions on ASME Code Case N-824, published in July 18, 2017 (82 FR 32934).

The conditions are derived from research into methods for inspecting Cast Austenitic Stainless Steel (CASS) components; these methods are published in NUREG/CR-6933, “Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods,” (ADAMS Accession Nos. ML071020410 and ML071020414), and NUREG/CR-7122, “An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds,” (ADAMS Accession No. ML12087A004). These NUREG/CR reports show that CASS materials less than 1.6 inches thick can be reliably inspected for flaws 10 percent through wall or deeper if encoded phased array examinations are performed using low ultrasonic frequencies and a sufficient number of inspection angles. Additionally, for thicker welds, flaws greater than 30 percent through wall in depth can be detected using low-frequency encoded phased array ultrasonic inspections.

The NRC, using NUREG/CR-6933 and NUREG/CR-7122, has determined that sufficient technical basis exists to condition ASME BPV Code, Section XI, Appendix III Supplement 2. The NUREG/CR reports show that CASS materials produce high levels of coherent noise and that the noise signals can be confusing and mask flaw indications. The optimum inspection frequencies for examining CASS components of various thicknesses as described in NUREG/CR-6933 and NUREG/CR-7122 are reflected in condition § 50.55a(b)(2)(xxxviii)(A). As NUREG/CR-6933 shows that the grain structure of CASS can reduce the effectiveness of some inspection angles, the NRC finds sufficient technical basis for the use of ultrasound using angles including, but not limited to, 30 to 55 degrees, with a maximum increment of 5 degrees. This is reflected in condition § 50.55a(b)(2)(xxxviii)(B).

10 CFR 50.55a(b)(2)(xxxix)(A) Defect Removal: First Provision

The NRC is adding § 50.55a(b)(2)(xxxix)(A) to place conditions on the use of ASME BPV Code, Section XI, IWA-4421(c)(1). The condition establishes that the final configuration of the item will be in accordance with the original Construction Code, later editions and addenda of the Construction Code, or a later different Construction Code, as well as meeting the Owner's Requirements or revised Owner's Requirements. This condition ensures that welding, brazing, fabrication, and installation requirements, as well as design requirements for material, design or configuration changes, are consistent with the Construction Code and Owner's Requirements. This condition retains the intent of the revision to Section XI that: (a) Replacements in kind are acceptable; (b) replacements with alternative configurations are acceptable as long as Construction Code and Owner's Requirements are met; and (c) defect removal is required; however, this can be accomplished by replacing all or a portion of the item containing the defect.

10 CFR 50.55a(b)(2)(xxxix)(B) Defect Removal: Second Provision

The NRC is adding § 50.55a(b)(2)(xxxix)(B) to place conditions on the use of ASME BPV Code, Section XI, IWA-4421(c)(2). The inclusion of subparagraph IWA-4421(c)(2) is intended to address wall thickness degradation where the missing wall thickness is restored by weld metal deposition. This repair activity restores the wall thickness to an acceptable condition; however, it does not “remove” the degraded wall thickness (
i.e.,
the defect); rather, restoration of wall thickness by welding or brazing mitigates the need to remove the defect. The NRC finds that increasing the wall thickness of an item to reclassify a crack from a defect to a flaw
3

is not acceptable because there are no provisions in subparagraph IWA-4421(c)(2) for analyses and ongoing monitoring of potential crack growth. Therefore, this condition prohibits the use of subparagraph IWA-4421(c)(2) rather than replacement for crack-like defects.

3
As defined in ASME BPV Code, Section XI, Article IWA-9000, a “flaw” is as an imperfection or unintentional discontinuity that is detectable by nondestructive examination and a “defect” is defined as a flaw of such size, shape, orientation, location, or properties as to be rejectable.

10 CFR 50.55a(b)(2)(xl) Section XI Condition: Prohibitions on Use of IWB-3510.4(b)

The NRC is adding § 50.55a(b)(2)(xl) to prohibit the use of ASME BPV Code, Section XI, 2017 Edition, Subparagraphs IWB-3510.4(b)(4) and IWB-3510.4(b)(5), which allow use of certain acceptance standard tables for high yield strength ferritic materials because they are not supported by the fracture toughness data.

The ASME BPV Code, Section XI, Subarticle IWB-3500 provides acceptance standards for pressure retaining components made of ferritic steels. Subparagraph IWB-3510.4 specifies material requirements for ferritic steels for application of the acceptance standards. In prior editions of the ASME BPV Code, Section XI, the material requirements for ferritic steels for which the acceptance standards of IWB-3500 apply are included in a note under the title of tables that specify allowable flaw sizes (
e.g.,
Table IWB-3510-1 “Allowable Planar Flaws”). Subparagraph IWB-3510.4 separates ferritic materials into three groups: (a) Those with a minimum yield strength of 50 ksi or less, (b) five ferritic steels with these material designations: SA-508 Grade 2 Class 2 (former designation: SA-508 Class 2a), SA-508 Grade 3 Class 2 (former designation: SA-508 Class 3a), SA-533 Type A Class 2 (former designation: SA-533 Grade A Class 2), SA-533 Type B Class 2 (former designation: SA-533 Grade B Class 2), and SA-508 Class 1, and (c) those with greater than 50 ksi but not exceeding 90 ksi. The material requirements for ferritic steels with a minimum yield strength of 50 ksi or less and those with greater than 50 ksi but not exceeding 90 ksi are explicitly specified. However, there are no material requirements for the five ferritic steels identified above.

The NRC finds Subparagraph IWB-3510.4(a) acceptable because it is consistent with the current material requirements for ferritic steels having a minimum yield strength of 50 ksi or less. The NRC finds Subparagraph IWB-3510.4(c) acceptable because it is consistent with the current material requirements for ferritic steels having a minimum yield strength of greater than 50 ksi to 90 ksi.

The NRC does not find subparagraphs IWB-3510.4(b)(4) and (5) acceptable for the following reasons. The NRC plotted the ASME BPV Code, Section XI static plain-strain fracture toughness (K
Ic
) curve in relevant figures in an ASME conference paper, PVP2010-25214, “Fracture Toughness of Pressure Boundary Steels with Higher Yield Strength” that shows dynamic fracture toughness (K
Id
) data for materials listed in IWB-3510.4 (b)(1) to IWB-3510.4 (b)(4). The NRC confirmed that the materials listed in IWB-3510.4 (b)(1) and IWB-3510.4 (b)(3) are acceptable because the data are above the K
IC
curve with adequate margin to compensate for the limited size of the data set. Additionally, the NRC has approved the use of the materials listed in IWB-3510.4 (b)(1) and IWB-3510.4 (b)(3) in a licensing and a design certification application. For the material listed in IWB-3510.4 (b)(2), K
Id
data was demonstrated to be above the crack arrest fracture toughness (K
Ia
). The NRC has previously determined the K
Ia
fracture toughness standard to be acceptable. Hence, the materials listed in IWB-3510.4 (b)(2) are acceptable. However, the technical basis document does not provide sufficient data to support exclusion of the fracture toughness requirements for the materials specified in Subparagraphs IWB-3510.4(b)(4) and IWB-3510.4(b)(5).

This condition does not change the current material requirements because licensees/applicants may continue to use testing to show that the two prohibited materials meet the material requirements.

10 CFR 50.55a(b)(2)(xli) Section XI Condition: Preservice Volumetric and Surface Examinations Acceptance

The NRC is adding § 50.55a(b)(2)(xli) to prohibit the use of ASME BPV Code, Section XI, Subparagraphs IWB-3112(a)(3) and IWC-3112(a)(3) in the 2013 through 2017 Edition. The NRC is prohibiting these items consistent with a final rule that approved ASME BPV Code Cases for use, dated January 17, 2018, (83 FR 2331).

During the review of public comments that were submitted on the proposed rule, dated March 2, 2016, (81 FR 10780), the NRC identified inconsistencies between Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use,” Revision 5, and a then concurrent proposed rule to incorporate by reference the 2009-2013 Editions of the ASME BPV Code (80 FR 56819), dated December 2, 2015.

Specifically, conditions that pertain to the NRC's disapproval of Code Case N-813, “Alternative Requirements for Preservice Volumetric and Surface Examination,” in the ASME BPV Code Regulatory Guide 1.193 proposed rule were not included in the ASME BPV 2009-2013 Editions proposed rule; however, the content of Code Case N-813 had been incorporated in the 2013 Edition of the ASME Code, Section XI. In order to resolve this conflict, the NRC excluded from the incorporation by reference those applicable portions of Section IX in the 2011a Addenda and the 2013 Edition, in § 50.55a(a)(1)(ii)(C)(
52
) and (
53
) respectively. This allowed the NRC to develop an appropriate regulatory approach for the treatment of these provisions that is consistent with the ASME BPV Code Regulatory Guide 1.193 rulemaking, in which the NRC found the acceptance of preservice flaws by analytical evaluation unacceptable.

Code Case N-813 is a proposed alternative to the provisions of the 2010 Edition of the ASME Code, Section XI, paragraph IWB-3112. Paragraph IWB-3112 does not allow the acceptance of flaws detected in the preservice examination by analytical evaluation. Code Case N-813 would allow the acceptance of these flaws through analytical evaluation. Per paragraph IWB-3112, any preservice flaw that exceeds the acceptance standards of Table IWB-3410-1 must be removed. While it is recognized that operating experience has shown that large through-wall flaws and leakages have developed in previously repaired welds as a result of weld residual stresses, the NRC has the following concerns

regarding the proposed alternative in Code Case N-813:

(1) The requirements of paragraph IWB-3112 were developed to ensure that defective welds were not placed in service. The NRC finds that a preservice flaw detected in a weld that exceeds the acceptance standards of Table IWB-3410-1 demonstrates poor workmanship and/or inadequate welding practice and procedures. The NRC finds that such an unacceptable preservice flaw needs to be removed and the weld needs to be repaired before it is placed in service.

(2) Under Code Case N-813 Paragraph B-3112(a)(3), large flaws would be allowed to remain in service because paragraph IWB-3132.3, via paragraph IWB-3643, allows a flaw up to 75 percent through-wall to remain in service. The NRC finds that larger flaws could grow to an unacceptable size between inspections, reducing structural margin and potentially challenging the structural integrity of safety-related Class 1 and Class 2 piping.

Paragraph C-3112(a)(3) of Code Case N-813, provides the same alternatives for Class 2 piping as that of Paragraph B-3112(a)(3). The NRC has the same concerns for Class 2 piping as for Class 1 piping.

Therefore, for the acceptance of preservice flaws by analytical evaluation, the NRC is adding a condition that prohibits the use of IWB-3112(a)(3) and IWC-3112(a)(3) in the 2013 Edition of ASME BPV Code Section XI through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of § 50.55a.

10 CFR 50.55a(b)(2)(xlii) Section XI Condition: Steam Generator Nozzle-to-Component Welds and Reactor Vessel Nozzle-to-Component Welds

The NRC is adding § 50.55a(b)(2)(xlii) to require that the examination of steam generator nozzle-to-component welds and reactor vessel nozzle-to-component welds must be a full volume examination and that the ultrasonic examination procedures, equipment, and personnel must be qualified by performance demonstration in accordance with Mandatory Appendix VIII of ASME Code, Section XI. These conditions are consistent with the conditions on ASME Code Case N-799 in Revision 19 of RG 1.147.
4

4
The NRC notes that one condition, requiring the examination volume to include 100 percent of the weld, was not reflected in RG 1.147 Revision 18, that accompanied that rule. That condition was developed in response to a public comment as described in the
Federal Register
notice for the rule, but the associated regulatory guide was not revised as intended due to an administrative error. The NRC has corrected that error in RG 1.147 Revision 19.

This code case was developed for new construction of recent reactor designs to provide examination requirements for weld configurations (
i.e.,
component-to-component welds). Specifically, the examination requirements described in Code Case N-799 apply to the weld configurations for the steam generator nozzle-to-reactor coolant pump casing weld in the AP1000 design and the reactor vessel-to-recirculation pump weld in the Advanced Boiling Water Reactor design. These weld configurations and the associated examination requirements did not exist in Section XI and have now been incorporated into ASME Code, Section XI, IWB-2500. The NRC is authorizing the use of these examination requirements incorporated into Section XI, IWB-2500, with similar conditions to those on the use of Code Case N-799. The NRC is simplifying the conditions in Revision 19 of RG 1.147 by combining them and also relaxing the condition concerning removing or repairing defects that are examined by procedures qualified to detect or depth size defects.

The first simplified condition in this rule combines the part of the first condition from Revision 19 of RG 1.147 concerning qualification with the second and third conditions, which also addresses qualification. This consolidation of the conditions from Revision 19 of RG 1.147 simplifies the qualification requirements by reducing the length and number of conditions.

The second simplified condition in this rule combines part of the first condition from Revision 19 of RG 1.147 concerning full volume examination with the fourth condition, which requires flaws (cracks) detected but not sized to the requirements of ASME Code, Section XI, Appendix VIII be repaired or removed. This simplified second condition relaxes the conditions from Revision 19 of RG 1.147 by allowing acceptance of flaws based on a flaw evaluation for the portion of the weld volume that is not examined by a qualified ultrasonic examination in accordance with ASME Code, Section XI, Appendix VIII.

The NRC recognizes that factors exist that may limit the ultrasonic examination volume that can be qualified by performance demonstration. For example, the qualified volume would be limited in components with wall thicknesses beyond the crack detection and sizing capabilities of a through wall ultrasonic performance-based qualification. To address the scenario in which the examination volume that can be qualified by performance demonstration is less than 100 percent of the volume, the NRC is allowing an ultrasonic examination of the qualified volume, provided that a flaw evaluation is performed to demonstrate the integrity of the examination volume that cannot be qualified by performance demonstration. The flaw evaluation should be of the largest hypothetical crack that could exist in the volume not qualified for ultrasonic examination. The licensee's revised examination plan is subject to prior NRC approval as an alternative in accordance with § 50.55a(z). The NRC determines that this relaxed condition provides assurance that the integrity of the welds in question will be maintained, despite a limited examination capability.

Therefore, in order to ensure that the examinations of steam generator nozzle-to-component welds and reactor vessel nozzle-to-component welds will be examinations of the full volume of the welds and that the ultrasonic examination procedures, equipment, and personnel are qualified by performance demonstration, in accordance with Mandatory Appendix VIII of ASME Code, Section XI, the NRC is adding conditions to the provisions of Table IWB-2500-1, Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, Item B5.11 (Reactor Vessel, NPS 4 or Larger Nozzle-to-Component Butt Welds) of the 2013 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of § 50.55a. The NRC is also adding similar conditions to the provision of Table IWB-2500-1, Item B5.71 (Steam Generator, NPS 4 or Larger Nozzle-to-Component Butt Welds) of the 2011 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of § 50.55a.

The NRC edited this condition from the proposed rule for clarity. Section 50.55a(b)(2)(xlii) of this final rule reflects this change.

C. ASME OM Code

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

The new Appendix IV in the 2017 Edition of the ASME OM Code provides improved preservice testing (PST) and IST of active pneumatic-operated valves (AOVs) within the scope of the ASME OM Code. Appendix IV specifies quarterly stroke-time testing of AOVs, where practicable. These are similar to the current requirements in Subsection

ISTC, “Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants,” of the ASME OM Code. In addition, Appendix IV specifies a preservice performance assessment test for AOVs with low safety significance, and initial and periodic performance assessment testing for AOVs with high safety significance on a sampling basis over a maximum 10-year interval.

The ASME developed the improved PST and IST provisions for AOVs in Appendix IV to the ASME OM Code in response to lessons learned from operating experience and test programs for AOVs and other power-operated valves (POVs) used at nuclear power plants. Over the years, the NRC has issued numerous generic communications to address weaknesses with AOVs and other POVs in performing their safety functions. For example, the NRC issued Generic Letter (GL) 88-14, “Instrument Air Supply System Problems Affecting Safety-Related Equipment,” to request that licensees verify that AOVs will perform as expected in accordance with all design-basis events. The NRC provided the results of studies of POV issues in several documents, including NUREG/CR-6654, “A Study of Air-Operated Valves in U.S. Nuclear Power Plants” (ADAMS Accession No. ML003691872). The NRC has issued several information notices to alert licensees to IST experience related to POV performance, including IN 86-50, “Inadequate Testing to Detect Failures of Safety-Related Pneumatic Components or Systems;” and IN 85-84, “Inadequate Inservice Testing of Main Steam Isolation Valves.” The NRC issued IN 96-48, “Motor-Operated Valve Performance Issues,” which described lessons learned from motor-operated valve (MOV) programs that are applicable to other POVs. Based on operating experience with the capability of POVs to perform their safety functions, the NRC established Generic Safety Issue 158, “Performance of Safety-Related Power-Operated Valves Under Design-Basis Conditions,” to evaluate whether additional regulatory actions were necessary to address POV performance issues. In Regulatory Issue Summary 2000-03, “Resolution of Generic Safety Issue (GSI) 158, `Performance of Safety Related Power-Operated Valves Under Design-Basis Conditions',” dated March 15, 2000, the NRC closed GSI-158 by specifying attributes for an effective POV testing program that incorporates lessons learned from MOV research and testing programs. More recently, the NRC issued IN 2015-13, “Main Steam Isolation Valve Failure Events,” to alert nuclear power plant applicants and licensees to examples of operating experience where deficiencies in licensee processes and procedures can contribute to the failure of main steam isolation valves (MSIVs), which may be operated by air actuators or combined air/hydraulic actuators. The NRC considers that the improved IST provisions specified in Appendix IV to the ASME OM Code will address the POV performance issues identified by operating experience with AOVs, including MSIVs, at nuclear power plants.

Paragraph IV-3800, “Risk-Informed AOV Inservice Testing,” allows the establishment of risk-informed AOV IST that incorporates risk insights in conjunction with functional margin to establish AOV grouping, acceptance criteria, exercising requirements, and testing intervals. Risk-informed AOV IST includes initial and periodic performance assessment testing of high safety significant AOVs with the results of that testing used to confirm the capability of low-safety significant AOVs within the same AOV group. For example, paragraph IV-3600, “Grouping of AOVs for Performance Assessment Testing,” states that test results shall be evaluated for all AOVs in a group. Paragraph IV-6500, “Performance Assessment Test Corrective Action,” specifies that correction action be taken in accordance with the Owner's corrective action requirements if AOV performance is unacceptable. The NRC considers that these provisions in Appendix IV will provide assurance that all AOVs within the scope of Appendix IV will be assessed for their operational readiness initially and on a periodic basis. The NRC revised the last sentence of § 50.55a(b)(3) to specify that when implementing the ASME OM Code, conditions are applicable only as specified in (b)(3).

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

The NRC is amending § 50.55a(b)(3)(ii) to specify that the condition applies to the latest edition and addenda of the ASME OM Code incorporated by reference in § 50.55a(a)(1)(iv). This allows future rulemakings to revise § 50.55a(a)(1)(iv) to incorporate the latest edition of the ASME OM Code without the need to revise § 50.55a(b)(3)(ii).

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

The NRC is amending § 50.55a(b)(3)(iv) to accept the use of Appendix II, “Check Valve Condition Monitoring Program,” in the 2017 Edition of the ASME OM Code without conditions based on its updated provisions. For example, Appendix II in the 2017 Edition of the ASME OM Code incorporates Table II, “Maximum Intervals for Use When Applying Interval Extensions,” as well as other conditions currently specified in § 50.55a(b)(3)(iv). The NRC is also revising § 50.55a(b)(3)(iv) to apply Table II to Appendix II of the ASME OM Code, 2003 Addenda through the 2015 Edition. Further, the NRC is removing the outdated conditions in paragraphs (b)(3)(iv)(A) through (D) based on their application to older editions and addenda of the ASME OM Code that are no longer applied at nuclear power plants, and on the incorporation of those conditions in recent editions and addenda of the ASME OM Code.

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

The NRC is amending § 50.55a(b)(3)(viii) to specify that the condition on the use of Subsection ISTE, “Risk-Informed Inservice Testing of Components in Light-Water Reactor Nuclear Power Plants,” applies to the latest edition and addenda of the ASME OM Code incorporated by reference in § 50.55a(a)(1)(iv). This allows future rulemakings to revise § 50.55a(a)(1)(iv) to incorporate the latest edition of the ASME OM Code without the need to revise § 50.55a(b)(3)(viii).

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

The NRC is amending § 50.55a(b)(3)(ix) to specify that Subsection ISTF, “Inservice Testing of Pumps in Water-Cooled Reactor Nuclear Power Plants—Post-2000 Plants,” of the ASME OM Code, 2017 Edition, is acceptable without conditions. The NRC is also amending § 50.55a(b)(3)(ix) to specify that licensees applying Subsection ISTF in the 2015 Edition of the ASME OM Code shall satisfy the requirements of Mandatory Appendix V, “Pump Periodic Verification Test Program,” of the ASME OM Code, in addition to the current requirement to satisfy Appendix V when applying Subsection ISTF in the 2012 Edition of the ASME OM Code. Subsection ISTF in the 2017 Edition of the ASME OM Code has incorporated the provisions from Appendix V such that this condition is not necessary for the 2017 Edition of the ASME OM Code.

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

The NRC is amending § 50.55a(b)(3)(xi) for the implementation

of paragraph ISTC-3700, “Position Verification Testing,” in the ASME OM Code to apply to the 2012 Edition through the latest edition and addenda of the ASME OM Code incorporated by reference in § 50.55a(a)(1)(iv). This allows future rulemakings to revise § 50.55a(a)(1)(iv) to incorporate the latest edition and addenda of the ASME OM Code without the need to revise § 50.55a(b)(3)(xi). In addition, the NRC is clarifying this condition to apply to all valves with remote position indicators within the scope of Subsection ISTC, “Inservice Testing of Valves in Water-Cooled Reactor Nuclear Power Plants,” including MOVs within the scope of Mandatory Appendix III, “Preservice and Inservice Testing Active Electric Motor-Operated Valve Assemblies in Water-Cooled Reactor Nuclear Power Plants.” ISTC-3700 references Mandatory Appendix III for valve position testing of MOVs. The development of Mandatory Appendix III was intended to verify valve position indication as part of the diagnostic testing (rather than exercising) performed at the intervals established by the appendix.

In response to public comments, the NRC is clarifying § 50.55a(b)(3)(xi) to refer to Subsection ISTC including its mandatory appendices and their verification methods and frequencies. This clarification will ensure that verification of valve position indication is understood to apply to all valves with remote position indication addressed in Subsection ISTC and all of its mandatory appendices. The NRC notes that licensees may request an NRC authorized alternative to this condition under 10 CFR 50.55a(z).

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

The NRC regulations in § 50.55a(f) specify that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing of the ASME BPV Code and ASME OM Code. Paragraph (f) in § 50.55a states that the requirements for inservice inspection of Class 1, Class 2, Class 3, Class MC, and Class CC components (including their supports) are located in paragraph (g) in § 50.55a. Applicants and licensees should note that requirements for inservice examination and testing of dynamic restraints (snubbers) are located in paragraph (b)(3)(v) in § 50.55a. The NRC is considering this clarification of the location of inservice examination and testing requirements for dynamic restraints in § 50.55a(f) and (g) for a future rulemaking.

A stakeholder submitted a public comment recommending that the NRC add a statement that pressure relief devices requiring testing per § 50.55a(f)(4) shall be limited to valves and rupture disks installed in piping systems designed to the ASME BPV Codes or ASME B31 standards. The NRC agrees that the ASME OM Code applies to pumps, valves, and dynamic restraints (snubbers) in piping systems. For example, the ASME OM Code does not apply to blowout panels in structures. However, such a clarification was not included in the proposed rule, and the NRC does not see an immediate need to clarify the applicability of the ASME OM Code in this regard. The NRC understands that the ASME OM Code committee is preparing a clarification to the ASME OM Code to specify its application to piping systems. Therefore, the NRC will evaluate the need for clarification of ASME OM Code to piping systems in a future rulemaking.

10 CFR 50.55a(f)(4)(i): Applicable IST Code: Initial 120-Month Interval

Several stakeholders submitted public comments on the § 50.55a 2009-2013 proposed rule requesting that the time schedule for complying with the latest ASME Code edition and addenda in § 50.55a(f)(4)(i) and (g)(4)(i) for the IST and ISI programs, respectively, be relaxed from the current time interval of 12 months to a new time interval of 24 months prior to the applicable milestones in those paragraphs. The ASME reiterated this request during an NRC/ASME management public teleconference that was held on March 16, 2016. During that teleconference, ASME discussed the challenges associated with meeting the 12-month time schedule in order to submit timely relief or alternative requests for NRC review. These comments were outside the scope of the proposed § 50.55a ASME 2009-2013 rule. However, the NRC indicated that the request would be considered in a future rulemaking.

In evaluating the suggested change, the NRC determined that the primary benefit from the relaxation of this § 50.55a(f)(4)(i) requirement is that licensees of new nuclear power plants will have more time to prepare their initial IST program and procedures and any proposed relief or alternative requests to the applicable edition of the ASME OM Code. The NRC determined that relaxation of the time schedule for satisfying the latest edition of the ASME OM Code for the initial 120-month IST interval is appropriate. However, the NRC considered that a 24-month time schedule would be contrary to the intent of the requirement to apply the latest edition of the ASME OM Code that is published every 24 months because it could result in licensees applying an outdated edition in the initial 120-month IST interval. Therefore, the NRC is extending the time schedule to satisfy the latest edition and addenda of the ASME OM Code from the current 12 months to 18 months for the initial 120-month IST interval.

10 CFR 50.55a(f)(4)(ii): Applicable IST Code: Successive 120-Month Intervals

As discussed in the previous section, several stakeholders submitted public comments on the § 50.55a 2009-2013 proposed rule, requesting that the time schedule for complying with the latest ASME Code edition in § 50.55a(f)(4)(ii) and (g)(4)(ii) for the IST and ISI programs, respectively, be relaxed from the current time period of 12 months to a new time period of 24 months prior to the applicable milestones in those paragraphs. The ASME reiterated this request during an NRC/ASME management public teleconference that was held on March 16, 2016. During that teleconference, ASME discussed the challenges associated with meeting the 12-month time schedule in order to submit timely relief or alternative requests for NRC review. These comments were outside the scope of the proposed § 50.55a ASME 2009-2013 rule. However, the NRC staff indicated that the proposed change would be considered for a future rulemaking. The NRC determined that the primary benefit from the relaxation of this § 50.55a(f)(4)(ii) requirement is that licensees of nuclear power plants will have more time to update their successive IST programs and procedures, and to prepare any proposed relief or alternative requests to the applicable edition of the ASME OM Code. In addition, licensees of each nuclear power plant will not need to review ASME OM Code editions incorporated by reference in § 50.55a after the relaxed 18-month time period before the start of the IST program interval compared to the 12-month time period required by the current regulations. The NRC determined that relaxation of the time schedule for satisfying the latest edition of the ASME OM Code for the successive 120-month IST interval is appropriate. However, the NRC considered that a 24-month time schedule would be contrary to the intent of the requirement to apply the latest edition of the ASME OM Code that is published every 24 months. Therefore, the NRC is extending the time schedule to satisfy the latest edition and addenda of the ASME OM

Code from the current 12 months to 18 months for successive 120-month IST intervals.

10 CFR 50.55a(f)(7) Inservice Testing Reporting Requirements

The NRC proposed adding § 50.55a(f)(7) to require nuclear power plant applicants and licensees to submit their IST Plans and interim IST Plan updates related to pumps and valves, and IST Plans and interim Plan updates related to snubber examination and testing to NRC Headquarters.

The ASME OM Code states in paragraph (a) of ISTA-3200, “Administrative Requirements,” that IST Plans shall be filed with the regulatory authorities having jurisdiction at the plant site. The NRC needs these IST Plans for use in evaluating relief and alternative requests, and deferral of quarterly testing to cold shutdowns and refueling outages. However, the ASME is planning to remove this provision from the ASME OM Code in a future edition because this provision is more appropriate as a regulatory requirement rather than a Code requirement. This change was proposed rather than in a future rulemaking to ensure that there will not be a period of time when this requirement is not in effect. Therefore, the condition would be an administrative change that would relocate the provision from the ASME OM Code to § 50.55a. However, in response to public comments discussed below, the NRC removed § 50.55a(f)(7) in this final rule. The NRC will reconsider this condition if the requirement is removed from a future Edition of the ASME OM Code.

10 CFR 50.55a(g)(4)(i): Applicable ISI Code: Initial 120-Month Interval

The NRC is amending § 50.55a(g)(4)(i) to relax the time schedule for complying with the latest edition of the ASME BPV Code for the initial 120-month ISI program interval, respectively, from 12 months to 18 months. The basis for the relaxation of the time schedule discussed previously for the requirement in § 50.55a(f)(4)(i) to comply with the latest edition and addenda of ASME BPV Code, Section XI, for the initial 120-month ISI program is also applicable to the relaxation of the time period for complying with the latest edition and addenda of the ASME BPV Code for the initial 120-month ISI program.

10 CFR 50.55a(g)(4)(ii): Applicable ISI Code: Successive 120-Month Intervals

The NRC is amending § 50.55a(g)(4)(ii) to relax the time schedule for complying with the latest edition and addenda of the ASME BPV Code for the successive 120-month ISI program intervals, respectively, from 12 months to 18 months. The basis for the relaxation of the time schedule discussed above for the requirement in § 50.55a(f)(4)(ii) to comply with the latest edition and addenda of the ASME BPV Code, Section XI, for the successive 120-month ISI programs is also applicable to the relaxation of the time period for complying with the latest edition and addenda of the ASME BPV Code for the successive 120-month ISI programs. The NRC is amending the regulation in § 50.55a(g)(4)(ii) to provide up to an 18-month period for licensees to update their Appendix VIII program for those licensees whose ISI interval commences during the 12 through 18-month period after June 3, 2020.

10 CFR 50.55a(g)(6)(ii)(C): Augmented ISI Requirements: Implementation of Appendix VIII to Section XI

The NRC is removing the language found in § 50.55a(g)(6)(ii)(C) from the current regulations. This paragraph describes requirements for initial implementation of older supplements in ASME BPV Code, Section XI Appendix VIII. Because the implementation dates have passed, and because licensees are no longer using these older editions and addenda of the Code that are referenced in this paragraph, the NRC is removing the condition.

D. ASME Code Cases

ASME BPV Code Case N-729-6

On September 10, 2008, the NRC issued a final rule to update § 50.55a to incorporate by reference 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 inservice inspection program for the examination of 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 de

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A2020-08855. Public record. Not legal advice.