[Federal Register Volume 76, Number 119 (Tuesday, June 21, 2011)]
[Rules and Regulations]
[Pages 36232-36279]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-14652]



[[Page 36231]]

Vol. 76

Tuesday,

No. 119

June 21, 2011

Part III





Nuclear Regulatory Commission





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10 CFR Part 50





American Society of Mechanical Engineers (ASME) Codes and New and 
Revised ASME Code Cases; Final Rule

  Federal Register / Vol. 76 , No. 119 / Tuesday, June 21, 2011 / Rules 
and Regulations  

[[Page 36232]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AI35
[NRC-2008-0554]


American Society of Mechanical Engineers (ASME) Codes and New and 
Revised ASME Code Cases

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The NRC is amending its regulations to incorporate by 
reference the 2005 Addenda (July 1, 2005) and 2006 Addenda (July 1, 
2006) to the 2004 ASME Boiler and Pressure Vessel Code, Section III, 
Division 1; 2007 ASME Boiler and Pressure Vessel Code, Section III, 
Division 1, 2007 Edition (July 1, 2007), with 2008a Addenda (July 1, 
2008); 2005 Addenda (July 1, 2005) and 2006 Addenda (July 1, 2006) to 
the 2004 ASME Boiler and Pressure Vessel Code, Section XI, Division 1; 
2007 ASME Boiler and Pressure Vessel Code, Section XI, Division 1, 2007 
Edition (July 1, 2007), with 2008a Addenda (July 1, 2008); and 2005 
Addenda, ASME OMa Code-2005 (approved July 8, 2005) and 2006 Addenda, 
ASME OMb Code-2006 (approved July 6, 2006) to the 2004 ASME Code for 
Operation and Maintenance of Nuclear Power Plants (OM Code). The NRC is 
also incorporating by reference (with conditions on their use) ASME 
Boiler and Pressure Vessel Code Case N-722-1, ``Additional Examinations 
for PWR Pressure Retaining Welds in Class 1 Components Fabricated with 
Alloy 600/82/182 Materials, Section XI, Division 1,'' Supplement 8, 
ASME approval date: January 26, 2009, and ASME Boiler and Pressure 
Vessel Code Case N-770-1, ``Alternative Examination Requirements and 
Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt 
Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material 
With or Without Application of Listed Mitigation Activities, Section 
XI, Division 1,'' ASME approval date: December 25, 2009.

DATES: This rule is effective July 21, 2011. The incorporation by 
reference of certain publications listed in the rule is approved by the 
Director of the Office of the Federal Register as of July 21, 2011.

ADDRESSES: You can access publicly available documents related to this 
document using the following methods:
     NRC's Public Document Room (PDR): The public may examine 
and have copied for fee publicly available documents at the NRC's PDR, 
Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): Publicly available documents created or received at the NRC 
are available electronically at the NRC's Library at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain 
entry into ADAMS, which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the NRC's PDR 
reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to 
[email protected].
     Federal Rulemaking Web Site: Public comments and 
supporting materials related to this final rule can be found at http://www.regulations.gov by searching on Docket ID: NRC-2008-0554.

FOR FURTHER INFORMATION CONTACT: L. Mark Padovan, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, telephone 301-415-1423, e-mail [email protected].

SUPPLEMENTARY INFORMATION: 
I. Background
II. Response to Public Comments
    A. Overview of Public Comments
    B. NRC Responses to Public Comments
III. Discussion of NRC Approval of New Edition and Addenda to the 
Code, ASME Code Cases N-722-1 and N-770-1, and Other Changes to 10 
CFR 50.55a
    --Quality Standards, ASME Codes and Institute of Electrical and 
Electronics Engineers (IEEE) Standards, and Alternatives
    -- Applicant/Licensee-Proposed Alternatives to the Requirements 
of 10 CFR 50.55a
    -- Standards Approved for Incorporation by Reference
    -- ASME B&PV Code, Section III
    -- ASME B&PV Code, Section XI
    -- ASME OM Code
    -- Reactor Coolant Pressure Boundary, Quality Group B 
Components, and Quality Group C Components
    -- Inservice Testing Requirements
    -- Inservice Inspection Requirements
    -- Substitution of the Term ``Condition'' in 10 CFR 50.55a
IV. Paragraph-by-Paragraph Discussion
V. Generic Aging Lessons Learned Report
VI. Availability of Documents
VII. Voluntary Consensus Standards
VIII. Finding of No Significant Environmental Impact: Environmental 
Assessment
IX. Paperwork Reduction Act Statement
    Public Protection Notification
X. Regulatory and Backfit Analysis
XI. Regulatory Flexibility Certification
XII. Congressional Review Act

I. Background

    The ASME develops and publishes the ASME Boiler and Pressure Vessel 
Code (B&PV Code), which contains requirements for the design, 
construction, and inservice inspection (ISI) of nuclear power plant 
components; and the ASME OM Code, which contains requirements for 
inservice testing (IST) of nuclear power plant components. The ASME 
issues new editions of the ASME B&PV Code every 3 years and issues 
addenda to the editions yearly, except in years when a new edition is 
issued. Periodically, the ASME publishes new editions and addenda of 
the ASME OM Code. The new editions and addenda typically revise 
provisions of the 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 Codes do not significantly 
change Code philosophy or approach.
    It has been the NRC's practice to establish requirements for the 
design, construction, operation, ISI (examination) and IST of nuclear 
power plants by approving the use of editions and addenda of the ASME 
B&PV and OM Codes (ASME Codes) in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.55a. The NRC approves and/or mandates 
the use of certain parts of editions and addenda of these ASME Codes in 
10 CFR 50.55a through the rulemaking process of ``incorporation by 
reference.'' Upon incorporation by reference of the ASME Codes into 10 
CFR 50.55a, the provisions of the ASME Codes are legally-binding NRC 
requirements as delineated in 10 CFR 50.55a, and subject to the 
conditions on certain of the ASME Codes' provisions which are set forth 
in 10 CFR 50.55a. The editions and addenda of the ASME B&PV and OM 
Codes were last incorporated by reference into the regulations in a 
final rule dated September 10, 2008 (73 FR 52730), as corrected on 
October 2, 2008 (73 FR 57235), incorporating Section III of the 2004 
Edition of the ASME B&PV Code, Section XI of the 2004 Edition of the 
ASME B&PV Code, and the 2004 Edition of the ASME OM Code, subject to 
NRC conditions.
    The ASME Codes are consensus standards developed by participants 
with broad and varied interests (including the NRC and licensees of 
nuclear power plants). The ASME's adoption of new editions of and

[[Page 36233]]

addenda to the ASME Codes does not mean that there is unanimity on 
every provision in the ASME Codes. There may be disagreement among the 
technical experts, including NRC representatives on the ASME Code 
committees and subcommittees, regarding the acceptability or 
desirability of a particular Code provision included in an ASME-
approved code edition or addenda. If the NRC believes that there is a 
significant technical or regulatory concern with a provision in an 
ASME-approved code edition or addenda being considered for 
incorporation by reference, then the NRC conditions the use of that 
provision when it incorporates by reference that ASME Code edition or 
addenda. In some cases, the condition increases the level of safety 
afforded by the ASME code provision, or addresses a regulatory issue 
not considered by the ASME. In other instances, where research data or 
experience has shown that certain Code provisions are unnecessarily 
conservative, the condition may provide that the Code provision need 
not be complied with in some or all respects. The NRC's conditions are 
included in 10 CFR 50.55a, typically in paragraph (b) of that 
regulation. In an SRM dated September 10, 1999, the Commission 
indicated that NRC rulemakings adopting (incorporating by reference) a 
voluntary consensus standard must identify and justify each part of the 
standard which is not adopted. For this rulemaking, the provisions of 
the 2005 Addenda through 2008 Addenda of Section III, Division 1, and 
the 2005 Addenda through 2008 Addenda of Section XI, Division 1, of the 
ASME B&PV Code; and the 2005 Addenda and 2006 Addenda of the ASME OM 
Code that the NRC is not adopting, or partially adopting, are 
previously identified in Section III of this statement of 
considerations, and in the regulatory and backfit analysis for this 
rulemaking. The provisions of the ASME B&PV Code, OM Code, and Code 
Cases N-722-1 and N-770-1 that the NRC finds to be conditionally 
acceptable, along with the conditions under which they may be applied, 
are also identified in Section III of this statement of considerations 
and the regulatory and backfit analysis for this rulemaking.
    The ASME Codes are voluntary consensus standards, and the NRC's 
incorporation by reference of these Codes is consistent with applicable 
requirements of the National Technology Transfer and Advancement Act 
(NTTAA). Additional discussion on NRC's compliance with the NTTAA is 
set forth in Section VII of this document, Voluntary Consensus 
Standards.

II. Response to Public Comments

A. Overview of Public Comments

    The NRC published a proposed rule for public comments in the 
Federal Register on May 4, 2010 (75 FR 24324). The public comment 
period for the proposed rule closed on July 19, 2010. The NRC received 
22 letters and e-mails from the following commenters (listed in order 
of receipt), providing about 454 comments on the proposed rule:

1. South Carolina Electric and Gas Company
2. Private citizen, Charles Wirtz
3. Private citizen, Gerry C. Slagis
4. Duke Energy
5. Electric Power Research Institute
6. Nextera Energy
7. IHI Southwest Technologies
8. Private citizen, Gary G. Elder
9. Performance Demonstration Initiative
10. Exelon Corporation
11. American Society of Mechanical Engineers
11a. American Society of Mechanical Engineers
12. Westinghouse
13. U.S. Department of Energy
14. Westinghouse
15. Progress Energy
16. PWR Owners Group
17. Nuclear Energy Institute
18. Entergy Operations, Inc. and Entergy Nuclear Operations, Inc.
19. Tennessee Valley Authority
20. Exelon Corporation
21. Dominion Resources Services, Inc.
22. Strategic Teaming and Resource Sharing (STARS)

    The number assigned to each commenter is used to identify the 
sponsor of the comment in the NRC's comment summary in Part B, ``NRC 
Responses to Public Comments,'' of this document. Most of these 
comments pertained to the following:
     Suggested revising or rewording conditions to make them 
more clear.
     Supported incorporation of Code Case N-770 or N-770-1 into 
10 CFR 50.55a.
     Supported the proposed changes to add or remove 
conditions.
     Opposed proposed conditions.
     Supplied additional information for NRC consideration.
     Proposed rewriting/renumbering of paragraphs.
     Asked questions or requested information from the NRC.
    Due to the large number of comments received and the length of the 
NRC's responses, this statement of considerations (SOC) addresses: (i) 
Responses to the three questions raised by the NRC in the proposed 
rule; (ii) comments resulting in changes to the proposed regulations; 
and (iii) comments raising important issues of interest to stakeholders 
but which the NRC declined to adopt. A discussion of all comments and 
the NRC responses is available electronically at the NRC's Library, 
ADAMS Accession No. ML110280240.

B. NRC Responses to Public Comments

Responses to Specific Requests for Comments
    The NRC requested comments on three NRC questions associated with 
its implementing 10 CFR 50.55a rulemaking process improvements to make 
incorporating by reference ASME B&PV Code editions and addenda into 10 
CFR 50.55a more predictable and consistent:
    NRC Question 1. What should the scope of the ASME B&PV Code edition 
and addenda rulemaking be (i.e., how many editions and addenda should 
be compiled into a single rulemaking)?
    Comment: One commenter stated that the NRC should address every 
other edition of the ASME Code in subsequent rulemakings (begin 
rulemaking once every 4 years) as the NRC's current 2-year rulemaking 
cycle is ambitious, and previous rulemakings have not occurred on this 
schedule. Three commenters indicated that starting with the 2013 
Edition, editions of these Code sections will be published every 2 
years (without addenda), and that future rulemakings should occur on a 
2-year schedule, starting with the 2013 Edition of these Codes. [4-2, 
11a-1; 14-1a; 19-1]
    NRC Response: The NRC has decided that future 10 CFR 50.55a 
rulemakings should incorporate only one later edition of the B&PV and 
OM Codes at a time, starting with the 2013 Editions of the ASME B&PV 
Code and the ASME OM Code.
    NRC Question 2. What should the frequency of ASME B&PV Code edition 
and addenda rulemaking be (i.e., how often should the NRC incorporate 
by reference Code editions and addenda into 10 CFR 50.55a)?
    Comment: The regulation currently requires compliance with the 
latest ASME Section XI Code incorporated by reference in 10 CFR 50.55a 
just 12 months prior to the start date of subsequent inspection 
interval. A 4-year publication schedule for 10 CFR 50.55a final rules 
would be beneficial for the following reasons:
    a. This schedule would not be overly burdensome for the NRC, and 
this may allow for a more predictable process and publication schedule 
for 10 CFR 50.55a. A 4-year publication schedule would allow for more 
licensees to use the same

[[Page 36234]]

Code of Record for multiple units at each site. This is particularly 
true for those sites where multiple units were completed within 4 years 
of the first unit. Use of a common Code of Record at each plant reduces 
administrative burden for licensees and reduces the risks associated 
with having to apply different Code requirements simultaneously at the 
same plant This recommendation would also benefit the NRC because fewer 
licensees would request relief to allow the use of a common Code of 
Record. [4-2]

    NRC Response: The NRC disagrees that a 4-year publication schedule 
to incorporate ASME B&PV Code edition and addenda into 10 CFR 50.55a is 
necessary for a more predictable process. The NRC performed a Lean Six 
Sigma review of its 10 CFR 50.55a rulemaking process and implemented 
improvements to make this rulemaking process more consistent and 
predictable. The NRC now believes that it can consistently and 
predictably publish 10 CFR 50.55a rulemakings on a 2-year interval.
    The NRC agrees in principal that a 4-year review cycle could 
possibly reduce the number of requests for relief when licensees use a 
common code of record for multiple units at a site, and that it is less 
of an administrative burden to have a common code of record at multiple 
unit sites. However, reducing the number of requests would depend on 
the timing of when a particular plant was required to update its 
inservice inspection (ISI) program in accordance with Sec.  
50.55a(g)(4). The option of using a common code of record at multiple 
units is still available through the use of an alternative in 
accordance with Sec.  50.55a(a)(3), and the NRC has approved the use of 
alternatives many times in the past for this purpose.
    Comment: As indicated in the draft rule, NRC rulemaking activities 
are currently on a 2-year cycle. In order for each rulemaking to 
incorporate by reference the latest published ASME Code editions, this 
cycle should be maintained and the next NRC new rulemaking would have 
to begin immediately upon publication of this proposed rule as a final 
10 CFR 50.55a rule. [11a-1, 14-1b]
    NRC Response: The NRC agrees that future 10 CFR 50.55a rulemakings 
should occur on a 2-year schedule, starting with the 2013 Editions of 
the ASME B&PV Code and the ASME OM Code. However, the NRC disagrees 
that it should begin the next NRC rulemaking upon publication of this 
final 10 CFR 50.55a rule. In order to assure that these rulemakings 
occur consistently and predictably, the NRC is initiating a pilot 
program to begin the next rulemaking when the camera-ready version of 
the 2011 Addenda to the 2010 Edition of Sections III and XI of the ASME 
B&PV Code becomes available. This start date is expected to be about 4 
months earlier than the ASME's July 2011 publishing date, and should 
contribute towards assuring that the NRC is able to publish the 
rulemaking on a 2-year interval (from ASME's July publication date).
    NRC Question 3. In what ways should the NRC communicate the scope, 
schedule for publishing the rulemakings in the Federal Register, and 
status of 10 CFR 50.55a rulemakings to external users?
    Comment: Four commenters stated that the industry would benefit 
from a predictable publication schedule for final 10 CFR 50.55a rules, 
regardless of the frequency of subsequent rulemakings. One of these 
commenters also indicated that, as an alternative, the NRC could 
consider one of the following options to establishing a predictable 
publication schedule:
     10 CFR 50.55a could be amended to allow the use of a 
limited number of Code editions that have been incorporated by 
reference in 10 CFR 50.55a, instead of only the latest, provided all 
applicable conditions are met when using the chosen Code edition.
     10 CFR 50.55a could be amended to require that licensees 
update their programs to comply with the latest Code of Record 
incorporated by reference into 10 CFR 50.55a no more than 36 months 
prior to the start of the subsequent 120-month inspection interval. [4-
2, 11a-1, 14-1c, 19-1]

    NRC Response: The NRC acknowledges the industry's representation 
that it would benefit from a predictable publication schedule for final 
10 CFR 50.55a rules. As discussed, the NRC now believes that it can 
consistently and predictably publish 10 CFR 50.55a rulemakings on a 2-
year interval. Thus, the NRC need not consider at this time the 
alternative options presented by one of the commenters.
    Comment: If the NRC believes that a predictable schedule for 
publication of final 10 CFR 50.55a rules cannot be accomplished, the 
NRC may want to consider whether the provisions in 10 CFR 
50.55a(f)(4)(ii) and (g)(4)(ii) should be amended to allow Owners/
Licensees to update their programs to comply with the latest edition 
and addenda of the Code incorporated by reference as much as 24 months 
before the start of a subsequent 120 month interval. [11-1]
    NRC Response: The NRC believes it can publish 10 CFR 50.55a 
rulemakings on a predictable schedule as a result of implementing 
rulemaking process improvements. Therefore, the NRC need not consider 
the commenter's proposal at this time.
Re-Designating 10 CFR 50.55a Paragraphs
    The NRC proposed that several paragraphs under 10 CFR 50.55a(b)(2) 
be removed, which would cause gaps in the numbering between the 
remaining paragraphs. To address the creation of these gaps, the NRC 
proposed to re-designate (renumber) the remaining paragraphs under 10 
CFR 50.55a(b)(2). These proposed re-designations are outlined in Table 
1 of this document.
    Comment: The proposed renumbering of paragraphs should not be 
adopted. Renumbering all of the paragraphs, while helping to reduce the 
number of pages in the rulemaking, does not consider the effort it will 
take for each end user to update their procedures to reflect the new 
numbering sequence. Many implementing programs and procedures will 
include references to the specific paragraph for implementation. 
Renumbering them will cause many documents to be revised. Recommend 
that this type of cleanup be considered under a total rewrite of 10 CFR 
50.55a rather than doing it under this proposed rule. Suggest that 
those paragraphs where conditions are removed be designated as 
``reserved.'' [4-1, 4-11a, 11-2, 14-2, 19-1, 20-1]
    NRC Response: The NRC acknowledges the comments representing that 
the proposed renumbering of paragraphs under 10 CFR 50.55a(b)(2) will 
require end users to expend resources to update their procedures to 
reflect the new numbering sequence. Accordingly, the NRC did not 
renumber these paragraphs under 10 CFR 50.55a(b)(2) in the final rule. 
Where the NRC removed paragraphs in the final rule, those paragraphs 
are designated as ``Reserved.'' To assist readers in understanding the 
regulatory history of this final rule, Table 1 gives a cross-reference 
of proposed, current and final regulation paragraph numbering.

[[Page 36235]]



                       Table 1--Cross Reference of Proposed, Current and Final Regulations
----------------------------------------------------------------------------------------------------------------
                                                     Description of proposed
      Proposed regulation        Current regulation       redesignations               Final regulation
----------------------------------------------------------------------------------------------------------------
Paragraph (b)(2)(i)............  Paragraph           Redesignate paragraph    Paragraph (b)(2)(ii).
                                  (b)(2)(ii).         (b)(2)(ii) as
                                                      paragraph (b)(2)(i).
Paragraph (b)(2)(ii)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(vi).
                                  (b)(2)(vi).         (b)(2)(vi) as
                                                      paragraph (b)(2)(ii).
Paragraph (b)(2)(iii)..........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(vii).
                                  (b)(2)(vii).        (b)(2)(vii) as
                                                      paragraph (b)(2)(iii).
Paragraph (b)(2)(iv)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(viii).
                                  (b)(2)(viii).       (b)(2)(viii) as
                                                      paragraph (b)(2)(iv).
Paragraph (b)(2)(v)............  Paragraph           Redesignate paragraph    Paragraph (b)(2)(ix).
                                  (b)(2)(ix).         (b)(2)(ix) as
                                                      paragraph (b)(2)(v).
Paragraph (b)(2)(vi)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(x).
                                  (b)(2)(x).          (b)(2)(x) as paragraph
                                                      (b)(2)(vi).
Paragraph (b)(2)(vii)..........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xi).
                                  (b)(2)(xi).         (b)(2)(xi) as
                                                      paragraph (b)(2)(vii).
Paragraph (b)(2)(viii).........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xii).
                                  (b)(2)(xii).        (b)(2)(xii) as
                                                      paragraph (b)(2)(viii).
Paragraph (b)(2)(ix)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xiii).
                                  (b)(2)(xiii).       (b)(2)(xiii) as
                                                      paragraph (b)(2)(ix).
Paragraph (b)(2)(x)............  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xiv).
                                  (b)(2)(xiv).        (b)(2)(xiv) as
                                                      paragraph (b)(2)(x).
Paragraph (b)(2)(xi)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xv).
                                  (b)(2)(xv).         (b)(2)(xv) as
                                                      paragraph (b)(2)(xi).
Paragraph (b)(2)(xii)..........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xvi).
                                  (b)(2)(xvi).        (b)(2)(xvi) as
                                                      paragraph (b)(2)(xii).
Paragraph (b)(2)(xiii).........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xvii).
                                  (b)(2)(xvii).       (b)(2)(xvii) as
                                                      paragraph (b)(2)(xiii).
Paragraph (b)(2)(xiv)(A).......  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xviii)(A).
                                  (b)(2)(xviii)(A).   (b)(2)(xviii)(A) as
                                                      paragraph
                                                      (b)(2)(xiv)(A).
Paragraph (b)(2)(xiv)(B).......  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xviii)(B).
                                  (b)(2)(xviii)(B).   (b)(2)(xviii)(B) as
                                                      paragraph
                                                      (b)(2)(xiv)(B).
Paragraph (b)(2)(xiv)(C).......  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xviii)(C).
                                  (b)(2)(xviii)(C).   (b)(2)(xviii)(C) as
                                                      paragraph
                                                      (b)(2)(xiv)(C).
Paragraph (b)(2)(xv)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xix).
                                  (b)(2)(xix).        (b)(2)(xix) as
                                                      paragraph (b)(2)(xv).
Paragraph (b)(2)(xvi)..........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xx).
                                  (b)(2)(xx).         (b)(2)(xx) as
                                                      paragraph (b)(2)(xvi).
Paragraph (b)(2)(xvii).........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxi).
                                  (b)(2)(xxi).        (b)(2)(xxi) as
                                                      paragraph (b)(2)(xvii).
Paragraph (b)(2)(xviii)........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxii).
                                  (b)(2)(xxii).       (b)(2)(xxii) as
                                                      paragraph
                                                      (b)(2)(xviii).
Paragraph (b)(2)(xix)..........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxiii).
                                  (b)(2)(xxiii).      (b)(2)(xxiii) as
                                                      paragraph (b)(2)(xix).
Paragraph (b)(2)(xx)...........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxiv).
                                  (b)(2)(xxiv).       (b)(2)(xxiv) as
                                                      paragraph (b)(2)(xx).
Paragraph (b)(2)(xxi)..........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxv).
                                  (b)(2)(xxv).        (b)(2)(xxv) as
                                                      paragraph (b)(2)(xxi).
Paragraph (b)(2)(xxii).........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxvi).
                                  (b)(2)(xxvi).       (b)(2)(xxvi) as
                                                      paragraph (b)(2)(xxii).
Paragraph (b)(2)(xxiii)........  Paragraph           Redesignate paragraph    Paragraph (b)(2)(xxvii).
                                  (b)(2)(xxvii).      (b)(2)(xxvii) as
                                                      paragraph
                                                      (b)(2)(xxiii).
Paragraph (b)(2)(xxiv).........  NA................  New Paragraph..........  Paragraph (b)(2)(xxviii).
Paragraph (b)(2)(xxv)..........  NA................  New Paragraph..........  Eliminated.
Paragraph (b)(2)(xxvi).........  NA................  New Paragraph..........  Paragraph (b)(2)(xxix).
----------------------------------------------------------------------------------------------------------------

Significant Public Comments on the Proposed Rule
    A summary of the significant comments, and the NRC's responses to 
those comments for each 10 CFR 50.55a section or paragraph are set 
forth in this document. A more comprehensive summary of the comments 
and the NRC responses are set forth in the NRC's Analysis of Public 
Comments document (ADAMS Accession No. ML110280240).
10 CFR 50.55a(b)(1)(iii) Seismic Design of Piping
    Comment: The NRC received comments from a number of external 
stakeholders that stated the proposed condition in Sec.  
50.55a(b)(1)(A) should be deleted. The comments' bases for deleting the 
proposed condition included the results of extensive research efforts 
on ferritic steels operating at high temperature. The results of this 
research were intended to provide sufficient bases to eliminate the 
NRC's concern on the B2' stress indices for Class 1 elbows 
and tees, on which the proposed condition) would have centered. [11-6b; 
14-6b; 19-1]
    NRC Response: Based on the NRC's review of the information provided 
in the public comment, the NRC is not including the proposed condition 
in Sec.  50.55a(b)(1)(A) on the B2' stress index for Class 1 
elbows and tees in this final rule. The information presented by the 
commenters adequately absolves the NRC's previously held concerns on 
the use of these stress indices in the seismic design of Class 1 
piping.

[[Page 36236]]

    Comment: A minor modification to the proposed condition in Sec.  
50.55a(b)(1)(iii)(B should be adopted to provide specificity on how the 
condition should be applied. [14-6c]
    NRC Response: The NRC agrees with the comment and the final rule 
language includes the modification suggested by the comment. The NRC 
agrees with the comment given that the modification eliminates 
potential ambiguity by clearly articulating when the NRC's condition in 
Sec.  50.55a(b)(1)(iii)(A) of the final rule language applies, with 
respect to the use of the provisions of Subarticle NB-3200 of the ASME 
Code.
    Comment: The comments received on the proposed addition of the 
condition 10 CFR 50.55a(b)(1)(iii)(C) pertained to the Do/t 
limitation for the seismic design of piping. The scope of the proposed 
condition in Sec.  50.55a(b)(1)(iii)(e) should be limited based on the 
fact that the ASME Code inherently captures the proposed condition in 
many instances in its current revision. [11-6d; 14-6d; 19-1]
    NRC Response: The NRC agrees with the comments based on the fact 
that the Do/t limitation is apparent throughout a majority 
of the affected ASME Code sections. In the final rule, paragraph 
(b)(1)(iii)(C) is modified to limit the scope of the proposed condition 
to those portions of the ASME Code which do not provide the inherent 
limitation on maintaining Do/t to a value of less than 40.
10 CFR 50.55a(b)(1)(vii) Capacity Certification and Demonstration of 
Function of Incompressible-Fluid Pressure-Relief Valves
    Comment: The NRC should reconsider its position to prohibit the use 
of paragraph NB-7742. The commenter pointed out that NB-7742 addresses 
test pressures that will exceed the test facility limits and reduces 
the number of functional tests for specific valve designs. With 
advances in technology, specialty valves were being developed that 
would be a specific size, operate at a specific set pressure, and have 
a required capacity. When only one such valve is installed in a nuclear 
power plant, the manufacturer would have to build at least two 
additional production valves so three valves could be tested per NB-
7732.2, and/or a multi-million dollar test facility would have to be 
built that had the required test pressure capability. Since NB-7732.2 
covers a range of conditions/applications for valve testing, the need 
to address specialty valves that did not have a range in size and set 
pressure, or had minimal range became evident. NB-7742(a)(1) and NB-
7742(a)(2) were added to address these applications. Manufacturing 
unnecessary production valves and building new test facilities are not 
economical options for the nuclear power industry. Therefore, the 
commenter requested that the NRC reconsider its position to prohibit 
the use of paragraph NB-7742. [14-8]
    NRC Response: Upon reconsideration, the NRC agrees in general with 
the comment that NB-7742 provides an acceptable methodology to test 
incompressible-fluid, pressure-relief valves that will exceed the test 
facility limits and addresses reducing the number of functional tests 
for specific valve designs. The NRC has identified no issues with 
performing tests at less than the highest value of the set-pressure 
range for incompressible-fluid, pressure-relief valves and finds these 
new requirements for Class 2 and 3 components acceptable as described 
in paragraphs NC-7742 and ND-7742. However, the NRC has identified 
words that were inadvertently left out of the Code during final 
printing of paragraph NB-7742 for Class 1 components. The parallel 
structure of the counterpart paragraphs (NC-7742 and ND-7742) reveal 
that the words ``for the design and the maximum set pressure'' are 
missing for paragraph NB-7742(a)(2). Without these words, paragraph NB-
7742(a)(2) is confusing, illogical, and could lead to a non-
conservative interpretation of the required test pressure for the new 
Class 1 incompressible-fluid, pressure-relief valve designs. For these 
reasons, paragraph (b)(1)(vii) of the final rule reflects a change to 
include a condition allowing use of paragraph NB-7742 when the 
corrected language intended by the Code is used.
10 CFR 50.55a(b)(2)(viii) Examination of Concrete Containments 
(Proposed Rule Paragraph (b)(2)(iv))
    Comment: Proposed rule paragraphs (b)(2)(iv)(B), (b)(2)(iv)(C), 
(b)(2)(iv)(D)(1), and (b)(2)(iv)(D)(2) should be deleted since they are 
not mandated by the introductory text of paragraph (b)(2)(iv). [20-4]
    NRC Response: The NRC disagrees with the comment. The proposed rule 
inadvertently removed the language in the introductory text of 
paragraph (b)(2)(iv) that mandates the conditions in the mentioned 
paragraphs. Final rule paragraph 10 CFR 50.55a(b)(2)(viii) added back 
the removed language in the introductory text to correct this 
unintended administrative error.
10 CFR 50.55a(b)(2)(ix) Examination of Metal Containments and the 
Liners of Concrete Containments (Proposed Rule Paragraph (b)(2)(v))
    Comment: The first part of the condition in the proposed rule 
paragraph (b)(2)(v)(A) should not be applied to the 2006 through the 
2008 Addenda, which incorporated requirements into IWE-2420(c) for 
evaluating the acceptability of inaccessible areas when conditions 
existed in accessible areas that could indicate the presence or result 
in degradation to such inaccessible areas. Only the second part of the 
condition requiring specific information relative to inaccessible areas 
be submitted in the ISI Summary Report should apply to these addenda. 
[11-15b; 14-15b; 19-1]
    NRC Response: The NRC agrees with the comment since the first part 
of the condition in proposed rule paragraph (b)(2)(v)(A) has been 
incorporated into the 2006 Addenda through 2008 Addenda of the Code. As 
a result of the comment, in final rule paragraph (b)(2)(ix)(A), the NRC 
has restructured the condition into two separate paragraphs designated 
(b)(2)(ix)(A)(1) and (b)(2)(ix)(A)(2) and revised the introductory text 
such that the condition in paragraph (b)(2)(ix)(A)(1) that addresses 
the requirement for the evaluation of inaccessible areas, is not 
required to be applied to Subsection IWE, 2006 Addenda through the 2008 
Addenda.
    Comment: The new condition in the proposed rule paragraph 
(b)(2)(v)(J), applicable to the use of IWE-5000 of the 2007 Edition 
with the 2008 Addenda, should not apply to metallic shell and 
penetration liners of Class CC components because these liners do not 
serve a structural integrity function which, for Class CC containments, 
is provided by the reinforced or post-tensioned concrete structure. The 
containment pressure test requirements in IWL-5000 are sufficient to 
ensure that the structural integrity of the Class CC component is 
demonstrated following major modifications. [4-12c; 4-12f; 11-15c; 11-
15g; 14-15c; 14-15g; 19-1]
    NRC Response: The NRC agrees with the basis of the comment that the 
system pressure test requirements of IWL-5000 are adequate to 
demonstrate both structural and leak-tight integrity of the repaired 
Class CC containment pressure retaining components following a major 
modification. Specifically, the requirements in IWL-5200 to perform a 
containment pressure test at design basis accident pressure, and to 
perform surface examinations of the repaired

[[Page 36237]]

area and specified additional/extended examinations and response 
measurements, will demonstrate structural integrity of the repaired 
Class CC concrete containment. The leakage test requirements in IWL-
5230 will demonstrate leak-tight integrity of the repaired area of the 
metallic shell or penetration liner of Class CC containments. As a 
result of the comment, the final rule paragraph (b)(2)(ix)(J) is 
revised to indicate that the condition applies only to Class MC 
pressure-retaining components and not to Class CC components.
    Comment: The new condition in proposed rule paragraph (b)(2)(v)(J), 
applicable to use of IWE-5000 of the 2007 Edition with the 2008 Addenda 
for major containment modifications, allows for an alternative to an 
Appendix J Type A test required by the condition following ``major'' 
modifications. However, performing a ``short-duration structural test'' 
as proposed would satisfy the condition in 10 CFR 50.55a, but would not 
satisfy the requirements imposed by 10 CFR Part 50, Appendix J, Option 
A. As a result, a ``short duration structural test'' cannot be 
performed in lieu of a Type A Test, unless a licensee seeks an 
exemption from the Appendix J test requirement, or 10 CFR part 50, 
Appendix J, Option A is revised to address the proposed alternative 
``short-duration structural test.'' [4-12b; 11-15i; 14-15i; 19-1]
    NRC Response: The NRC agrees with the comment to the extent that 
when a licensee is implementing Option A of 10 CFR part 50, Appendix J, 
the alternative short duration structural test in the new condition in 
proposed rule paragraph (b)(2)(v)(J) cannot be performed in lieu of the 
Type A test required by the condition without seeking an exemption. The 
NRC's agreement is based on the fact that an inconsistency would exist 
between the requirement in the proposed rule paragraph (b)(2)(v)(J) and 
the existing requirements under Special Testing Requirements in 
paragraph IV.A of 10 CFR part 50, Appendix J, Option A. This 
inconsistency would exist due to the fact that the current requirements 
in Appendix J, Option A, would require a Type A test following a major 
containment modification, while the proposed requirement would also 
allow an alternative ``short duration structural test.'' The latter 
cannot be performed in lieu of a Type A test, thus leading to an 
inconsistency which could only be reconciled by an exemption. Paragraph 
IV.A of 10 CFR part 50, Appendix J, Option A does not specify any 
alternative structural test because the Type A test would demonstrate 
both structural and leak tight integrity of the repaired containment.
    The NRC disagrees with the comment, in part, given that for the 
vast majority of licensees implementing Option B of 10 CFR part 50, 
Appendix J, the argument could be made that containment modifications 
are implemented under the Inservice Inspection Program in accordance 
with ASME Section XI, Subsection IWE (for Class MC containments) 
pursuant to 10 CFR 50.55a(g)(4). Therefore, it could be argued that the 
system pressure testing requirements in IWE-5000 apply following 
containment modifications and not those in paragraph IV.A of 10 CFR 
part 50, Appendix J, Option A. Prior to the 2007 Edition of Section XI 
of the ASME B&PV Code, Article IWE-5000 referenced paragraph IV.A of 10 
CFR part 50, Appendix J, Option A, for the leakage test requirements 
following containment modifications. By referencing the Appendix J, 
Option A, requirements, Article IWE-5000 indirectly required a Type A 
test to be performed following a major containment modification. Since 
the Type A test requires pressurization of the entire containment to 
the design basis accident pressure (Pa), it would provide a 
verification of both the leakage integrity and structural integrity of 
repaired containment. However, Article IWE-5000, as modified in the 
2007 Edition and later addenda, provides a licensee the option of 
performing only a local bubble test of the brazed joints and welds 
affected by the repair even for major modifications. This provides a 
verification of local leak-tightness of the repaired area, but does not 
provide a verification of global structural integrity of the repaired 
structure, and hence, the need for the new condition to perform a Type 
A test following a major modification.
    Based on this discussion, the NRC has determined that the new 
condition in the final rule paragraph (b)(2)(ix)(J) only addresses the 
deficiency identified in Article IWE-5000, and does not include the 
provisions for an alternate short-duration structural test in the new 
condition.
    Comment: The actions specified in (b)(2)(v)(J)(1), (b)(2)(v)(J)(2) 
and (b)(2)(v)(J)(3), as part of the alternate short duration structural 
test, of the new condition in the proposed rule paragraph (b)(2)(v)(J), 
applicable to the use of IWE-5000 of the 2007 Edition with the 2008 
Addenda for Class MC components, should be modified as below.
     The actions described in (b)(2)(v)(J)(1) should not apply 
to the 2007 Edition with the 2008 Addenda of ASME Code, Section XI.
     The condition in (b)(2)(v)(J)(2) should not apply because 
IWE-5223 and IWE-5224 already provide adequate test requirements to 
assure essentially zero leakage.
     The actions described in (b)(2)(v)(J)(3) would prohibit 
the conduct of the pressure test at a pressure less than Pa. The 10 CFR 
part 50, Appendix J, Type A Test is permitted to be conducted at a test 
pressure of at least 0.96Pa. [4-12d, 4-12e, 11-15d, 11-15e, 11-15f, 14-
15d, 14-15e, 14-15f, 19-1]

    NRC Response: The NRC agrees with the comment because:
    (i) The nondestructive examination of the repair welds specified in 
paragraph (b)(2)(v)(J)(1) is typically required to be performed as part 
of the repair process;
    (ii) The provisions of IWE-5223 and IWE-5224 of the 2007 Edition 
with the 2008 Addenda include the soap bubble or equivalent leakage 
test specified in paragraph (b)(2)(v)(J)(2) and are adequate to assure 
essentially zero leakage through the repair welds or joints; and
    (iii) The action specified in paragraph (b)(2)(v)(J)(3) required 
the entire containment to be pressurized to the peak calculated design 
basis accident pressure (Pa) whereas a Type A test conducted 
in accordance with ANSI/ANS 56.8 may be performed at a pressure between 
0.96Pa and 1.1Pa.
    However, the testing provisions of IWE-5223 and IWE-5224 of the 
2007 Edition with the 2008 Addenda are not adequate to demonstrate 
global structural integrity of the repaired Class MC containment, which 
is essentially the deficiency that is sought to be addressed by the new 
condition. In the context of IWE-5000, it is the Type A test that would 
provide a verification of both structural and leak-tight integrity 
following a major modification. As such, the NRC determined that the 
new condition only addresses the deficiency in the provisions of 
Article IWE-5000 and did not include the proposed alternate short-
duration structural test provision in the condition in the final rule.
    Comment: The new condition in proposed rule paragraph (b)(2)(v)(J) 
provides a general definition of ``major'' containment modifications as 
repair/replacement activities such as replacing a large penetration, 
cutting a large opening in the containment pressure boundary to replace 
major equipment such as steam generators, reactor vessel heads, 
pressurizers, or other similar modifications. This new condition does

[[Page 36238]]

not clearly define what constitutes a ``major'' modification or repair/
replacement activity for containment structures and that the lack of a 
clear definition will cause potential confusion and possible conflict 
with requirements of 10 CFR part 50, Appendix J, paragraph IV.A. [4-
12a, 11-15h, 14-15h, 19-1]
    NRC Response: The NRC disagrees with the comment. The proposed rule 
paragraph (b)(2)(v)(J) provides a definition of a ``major'' 
modification, which is qualitative but based on citing specific 
examples of repair/replacement activities that have typically been 
performed extensively among operating power reactors historically and 
have been consistently considered as major modifications by the NRC 
staff as well as licensees. The NRC acknowledges that the definition 
provided for ``major'' modification in the proposed rule is somewhat 
more explicit than the language used in 10 CFR part 50, Appendix J, 
Option A, paragraph IV.A, in that the cited paragraph IV.A simply uses 
the term ``major modification'' without any explicit description, but 
the intent is consistent. Based on this discussion, the NRC has 
retained the qualitative definition of major modifications in the final 
rule. No change was made to the final rule as a result of this comment.
10 CFR 50.55a(b)(2)(xi) (Proposed Rule Paragraph (b)(2)(vii))
    Comment: Referencing later versions of Appendix VIII should be 
delayed and replaced with a mandatory, industry wide, version and 
implementation date. In a December public meeting with the one of the 
commenters (PDI), the commenter clarified his comment as requesting the 
NRC to delay by 18 months the date on which Appendix VIII of the 2007 
Edition and 2008 Addenda becomes effective for purposes of updating 
licensees' 10-year inservice inspection interval. The commenter 
explained that an 18-month delay is necessary to avoid an undue burden 
on those licensees who have only 12 months to update their inservice 
inspection program for the next 10-year inservice inspection interval 
(as is required under Sec.  50.55a). [9-1; 9-2; 10-1; 10-2; 20-2]
    NRC Response: The NRC agrees with the comments that there may be an 
undue burden on those licensees who have only 12 months to update their 
inservice inspection program to comply with Appendix VIII for the next 
10-year inservice inspection interval. Accordingly, the NRC is revising 
the language of the final rule to provide at least 18 months for a 
specified set of licensees to update and begin implementation of the 
2007 Edition and 2008 Addenda versions of Appendix VIII in their next 
inservice inspection interval. This set of licensees are those whose 
next inservice inspection interval must begin to be implemented during 
the period between 12 through 18 months after the effective date of the 
final rule, and therefore would otherwise be required to implement the 
2007 Edition and 2008 Addenda versions of Appendix VIII (providing them 
less than 18 months to comply with the provisions of the 2007 Edition 
and 2008 Addenda versions of Appendix VIII). For these licensees, the 
final rule provides a delay of 6 months in the implementation of 
Appendix VIII only (i.e., these licensees will still be required to 
update and implement the inservice inspection program during the next 
inspection interval without delay). Other licensees, whose next 
inservice inspection interval commences more than 18 months after the 
final date of the rule, will have sufficient time to develop their 
programs for the next inservice inspection interval and are not 
affected by this provision of the final rule.
    The NRC disagrees with the portions of the comments requesting that 
the NRC mandate the use of later versions of Appendix VIII for all 
licensees. The comments did not provide a technical or regulatory 
justification for imposing such a backfit (a uniform date of 
implementation would be regarded as a backfit because it departs from 
the current regulatory approach of a ten-year inservice inspection 
program interval). In addition, the NRC notes that Sec.  
50.55a(g)(4)(iv) currently allows licensees to voluntarily comply with 
the inservice inspection requirements of more recent editions and 
addenda which the NRC has approved (via incorporation by reference into 
Sec.  50.55a). Accordingly, the NRC declines to adopt the proposal. No 
change was made to the final rule as a result of this portion of the 
comment.
    Comment: The requirements for scanning from the austenitic side of 
the weld should be revised to accommodate certain exceptions such as 
austenitic welds with no austenitic sides or austenitic welds attached 
to cast austenitic components. [20-3]
    NRC Response: NRC agrees that paragraph (b)(2)(xv)(A)(2) should 
address the case of an austenitic weld which has no austenitic base 
material side. An austenitic weld with no austenitic sides cannot be 
qualified from an austenitic side. However, qualification from the 
austenitic side of the weld demonstrates a higher degree of proficiency 
than from the ferritic side of the weld. Therefore, an existing ASME 
Code, Section XI, Appendix VIII, Supplement 10, Qualification 
Requirements for Dissimilar (DM) Metal Welds, qualification may be 
expanded for austenitic welds with no austenitic sides. This expansion 
of the Supplement 10 qualification would require implementing a 
separate performance demonstration add-on to include samples where the 
austenitic weld is flanked by ferritic base material. The NRC disagrees 
that special consideration should be given to components with cast 
austenitic material on one side because single-side examination of 
austenitic welds attached to cast stainless steel components is outside 
the scope of the current qualification program. For these reasons, 
paragraph (b)(2)(xv)(A)(2) in the final rule is revised to include an 
add-on qualification for austenitic welds with no austenitic side to an 
existing Supplement 10 qualification.
10 CFR 50.55a(b)(2)(xii) (Proposed Rule Paragraph (b)(2)(viii))
    Comment: The condition on Appendix VIII single-side ferritic vessel 
and piping and stainless steel piping examinations was addressed in the 
2005 Addenda of ASME Code and should be removed. [11-17; 14-17a; 19-1]
    NRC Response: The NRC agrees that the condition should not apply to 
the 2007 Edition and 2008 Addenda because the condition was fully 
addressed in the 2007 Edition of Section XI. However, the condition is 
necessary through the 2006 Addenda because of changes within referenced 
Supplements 5 and 7 in I-3000. For these reasons, paragraph (b)(2)(xvi) 
is revised in this final rule to remove the condition from the 2007 
Edition and 2008 Addenda but retains the condition through the 2006 
Addenda.
10 CFR 50.55a(b)(2)(xiv)(C) (Proposed Rule Paragraph (b)(2)(x))
    Comment: 10 CFR 50.55a(b)(2)(xiv)(C) should be revised to read: 
``When applying editions and addenda prior to the 2005 Addenda of 
Section Xl licensees qualifying visual examination personnel for VT-3 
visual examination under paragraph IWA-2317 of Section Xl.'' The basis 
for this recommendation is that IWA-2317 of the 2004 Edition does not 
contain the requirements to demonstrate the proficiency of the training 
by administering an initial qualification examination and administering 
subsequent examinations on a 3-year interval. [20-5]
    NRC Response: The NRC agrees with the commenter that the 2004 
Edition

[[Page 36239]]

and earlier editions and addenda do not contain the requirements to 
demonstrate the proficiency of the training and the commenter's 
proposed wording is clearer. Paragraph (b)(2)(xviii)(C) of the final 
rule has been revised to reflect the commenter's proposed wording.
10 CFR 50.55a(b)(2)(xv) (Proposed Rule Paragraph (b)(2)(xi))
    Comment: Substitution of ultrasonic (UT) examinations performed in 
accordance with Section XI, Appendix VIII for radiographic (RT) 
examinations should be acceptable for repairs. ASME Code has already 
approved three Code Cases for UT in lieu of RT and is in the process of 
approving a fourth Code Case. [4-16; 7-1; 11-20b; 14-20; 19-1]
    NRC Response: The NRC disagrees with the comment. Section III RT 
examinations are for verifying the soundness of the full weld volume. 
In Section XI, some welds do not have defined examination volumes, and 
for the welds having defined examination volumes, only portions of the 
volume are examined. Appendix VIII qualifications are demonstrated on 
the weld volume defined in Section XI; the qualifications are tailored 
for detection and sizing cracks propagating from the inner vessel or 
pipe surfaces. The NRC's concerns with UT in lieu of RT are presented 
in the statement of considerations published in the Federal Register on 
October 27, 2006, (71 FR 62947) pertaining to Code Case N-659 which was 
not approved for use in Regulatory Guide (RG) 1.193, Revision 2. The 
NRC did not review the other two ASME approved code cases. The NRC will 
review the fourth code case and associated documentation after ASME 
approval. No change was made to the final rule as a result of this 
comment.
    Comment: The proposed rule implied UT was better suited for 
detecting planar flaws associated with inservice degradation than 
volumetric flaws, and not effective for volumetric flaws with large 
openings. Further, few studies have been done to demonstrate 
effectiveness of RT in a manner comparable to the way the effectiveness 
of UT has been demonstrated via ASME, Section XI, Appendix VIII. [7-2]
    NRC Response: The NRC agrees that few studies have been done to 
demonstrate the effectiveness of RT in a manner comparable to the way 
the effectiveness of UT has been demonstrated via ASME, Section XI, 
Appendix VIII. In particular, there are limited studies that compare 
the effectiveness of UT vs. RT on fabrication type flaws vs. service-
induced flaws for welds found in nuclear power plants. Until such time 
as studies are complete, the NRC will remain silent on the ability of 
UT to detect fabrication type (i.e., volumetric) flaws, as well as 
comparing the abilities of UT and RT. No change was made to the final 
rule as a result of this comment.
    Comment: UT should be allowed for materials where it is as 
effective, or more effective, than RT. The comment is specifically 
targeted at UT on cast stainless steel components. [7-3]
    NRC Response: Based on a recent study PNNL-19086, ``Replacement of 
Radiography with Ultrasonics for the Nondestructive Inspection of 
Welds--Evaluation of Technical Gaps--An Interim Report,'' (ADAMS 
Accession No. ML101031254), the NRC believes that the effectiveness of 
UT in lieu of RT has not been established. To address the NRC's 
concerns, the NRC believes research must be conducted to:
     Compare the flaw detection capabilities of UT and RT;
     Assess parameters such as false call rates;
     Assess qualification and acceptance standards;
     Assess the effectiveness and reliability of UT and RT for 
construction, preservice and inservice inspection;
     Assess the interchangeability of UT and RT; and
     Determine the state-of-the-art with regard to digital 
radiography.
    Therefore, no change was made to the final rule as a result of this 
comment.
    Comment: While UT requires more access and may require more weld 
surface preparation area than RT, consideration should be given to 
peripheral benefits of using UT associated with less work area 
restrictions, no risk of radiation exposure, no RT source storage 
issues, and reduced examination time. [7-4]
    NRC Response: The NRC disagrees with this comment. While benefits 
may exist, the NRC believes that examination and qualifications 
concerns must be addressed first to establish effectiveness and 
reliability of UT in lieu of RT. No change was made to the final rule 
as a result of this comment.
    Comment: UT systems needing to undergo a Section XI, Appendix VIII-
style demonstration and qualification program for construction flaws 
prior to use is illogical for replacing RT systems that have not been 
subjected to a similar demonstration and qualification program. [7-5]
    NRC Response: The NRC disagrees with the comment. Based on study 
PNNL-19086, the NRC believes that the effectiveness of UT in lieu of RT 
has not been established. Accordingly, the NRC will be conducting 
research as explained in the NRC response to comment 7-3. Though RT is 
not subject to a rigorous qualification program at this time, 
implementation of RT on new construction or repair welds in conjunction 
with application of the qualified UT often performed for pre-service 
inspections, provides a greater assurance of quality and safety than if 
only one examination technique was implemented. Until such time as the 
NRC has completed its evaluation of UT and RT for nuclear power plant 
components, the NRC will not allow substitution of UT when RT is 
prescribed for the examination. No change was made to the final rule as 
a result of this comment.
    Comment: V-path application with UT examination may not be 
applicable for all metals where UT examinations are allowed. The NRC 
should consider approving the substitution of UT for RT with specific 
conditions or limitations, such as:
    (1) UT may not be used in lieu of RT for examination of cast 
stainless steel or austenitic stainless steels and nickel alloys where 
only single-sided access is available;
    (2) When UT is used in lieu of RT, the acceptance standards of ASME 
Section XI IWA-3000 shall be used in lieu of the construction code 
acceptance standards; and
    (3) Encoded or automated UT shall be used to create a permanent 
record which would allow multiple analysis reviews as well as document 
the results for comparison with future examinations. [7-6]

    NRC Response: The NRC believes that the effectiveness of UT in lieu 
of RT has not been established. Industry studies have been initiated to 
evaluate NRC concerns with UT in lieu of RT. The NRC will consider the 
results from these studies in future reviews. Therefore, proposed 
paragraph (b)(2)(xv) pertaining to IWA-4520(b)(2) and IWA-4521 is 
adopted without change in final rule paragraph (b)(2)(xix). No change 
was made to the final rule as a result of this comment.
    Comment: With regard to paragraph (b)(2)(xv), clarify whether the 
substitution of ASME Section V ultrasonic examination method by an 
Appendix VIII ultrasonic examination method is allowed by the 
provisions of IWA-2240 of the 1997 Addenda as specified in this 
paragraph's condition. [20-6]
    NRC Response: The NRC disagrees with the comment, because it is not 
the NRC's regulatory responsibility to clarify the ASME Code. No change 
was made to the final rule as a result of this comment.

[[Page 36240]]

10 CFR 50.55a(b)(2)(xvii)(B) (Proposed Rule Paragraph (b)(2)(xiii))
    Comment: Consideration should be given to deleting this condition 
entirely as it is inconsistent with the unconditional approval of Code 
Case N-652-1 in NRC RG 1.147, Rev 15, which does not include Item B7.80 
or any provisions for examination of CRD bolting. [2-2]
    NRC Response: The NRC agrees that Item No. B7.80 was deleted in the 
1995 Addenda of Section XI. The NRC also agrees that the existing 
condition is inconsistent with the NRC unconditional approval of Code 
Case N-652-1 which eliminates Item No. B7.80 requirements. The NRC also 
believes that Examination Category B-G-2 contains examination 
requirements for all Class 1 pressure retaining bolting 2 inches and 
less in diameter to provide reasonable assurance of their structural 
integrity. For these reasons the NRC agrees with the comment. Final 
rule paragraph (b)(2)(xxi) reflects a change to eliminate the condition 
that provisions of Table IWB-2500-1, Examination Category B-G-2, Item 
B7.80, that are in the 1995 Edition are applicable only to reused 
bolting when using the 1997 Addenda through the latest edition and 
addenda incorporated by reference in paragraph (b)(2) of this section.
10 CFR 50.55a(b)(2)(xxiv) (Proposed Rule Paragraph (b)(2)(xx))
    Comment: The NRC condition, which would place conditions on the use 
of Equation (2) in A-4300(b)(1) of Nonmandatory Appendix A of Section 
XI, should be removed because the condition would result in more 
conservative crack growth rates to be computed when R-ratio (i.e., 
Kmin/Kmax) is negative. The basis for 1.12 
Sf factor was established from lab data for R < 0 and 
considers crack closure effects. [11-23; 14-23; 19-1]
    NRC Response: The NRC disagrees with the comment. The NRC has 
reviewed the laboratory test data upon which this provision was based, 
and concludes that it is insufficient to firmly establish the Section 
XI, Appendix A approach when the R-ratio is negative.
    The test data reported in the 1977 ASME Pressure Vessels and Piping 
Conference paper, ``High Stress Crack Growth--Part II, Predictive 
Methodology Using a Crack Closure Model,'' which serves as the basis 
for the ASME Code, Section XI, Appendix A approach, consists of only 10 
test data points for -1.5 < R < 0, and one of those data points shows a 
trend opposite of the others. Although this data was produced from 
tests covering a limited R value range, it is used to support the 
application of the ASME Code, Section XI, Appendix A approach for a 
much wider range of R, (i.e., all R < 0).
    Further, in ASME Code, Section XI, Appendix A applications, the 
generic, lower-bound material property values from ASME Code, Section 
II may be used. If the lower bound ASME Code, Section II generic flow 
stress ([sigma]f) for a material is less than the material's 
actual [sigma]f, the calculation in accordance with ASME 
Code, Section XI, Appendix A for R < 0 will show that Kmax - 
Kmin <= 1.12 [sigma]f [radic]([pi]a) and prompt a 
wrongful reduction of [Delta]KI where full 
[Delta]KI should be used. This potential non-conservatism in 
the use of the ASME Code, Section XI, Appendix A approach, along with 
the issues cited above regarding the available test data, calls into 
question the generic applicability of the ASME Code, Section XI, 
Appendix A approach.
    For these reasons, the NRC disagrees with the comment. No change 
was made to the final rule as a result of the comment.
10 CFR 50.55a(b)(2)(xxv) (Proposed Rule Paragraph (b)(2)(xxi))
    Comment: Qualitative arguments based on a deterministic approach 
stated the current provision in Table E-2 for a crack size up to 1 inch 
deep is sufficient based on:
    (1) Real flaw sizes in vessels are closer to a depth of 
approximately 0.10 inch deep or less based on actual vessel inspection 
data;
    (2) Use of ASME Code, Section XI, Appendix VIII, and Electric Power 
Research Institute (EPRI) Performance Demonstration Initiative (PDI) 
provides continuous verification that the beltline region welds are 
either free of defects larger than approximately 0.10 inch or that they 
are documented and recorded, and;
    (3) Additional conservatism exists in the use of a lower bound 
reference toughness curve for prevention of crack initiation for these 
reference flaws.

[11-24; 11-24; 16-17;16-18; 16-19; 16-20; 17-2; 17-3; 17-4; 17-5; 17-9; 
17-11; 19-1; 20-8; 20-11; 20-12; 20-13; 21-2; 21-3; 21-4; 21-5; 21-6 
and 21-7]

    Quantitative results based on a probabilistic approach demonstrate 
that the current Appendix E approach provides an appropriate 
conservative methodology following an unanticipated transient. The 
Pressurized Water Reactor Owners Group (PWROG) has provided a risk-
informed assessment of Appendix E, which indicated that by setting the 
core damage frequency (CDF) to 1E-6, the resulting pressure versus (T-
RTNDT) curve bounds the corresponding Appendix E curve for 
both the PWR unanticipated isothermal pressure events and the 
pressurized cool-down events, where T is the reactor pressure vessel 
(RPV) coolant temperature and RTNDT is the nil-ductility 
reference temperature of the limiting RPV material. [16-21]
    NRC Response: The commenter's qualitative arguments based on the 
deterministic approach involve extensive discussions. However, the 
bottom line is the same as for Comments 11 and 14. Hence, the NRC will 
respond to only selective parts of the comments based on the 
deterministic approach to clarify its position. This is appropriate 
because the NRC's final position is not based on the qualitative, 
deterministic fracture mechanics (FM) arguments, but on the 
quantitative, probabilistic fracture mechanics (PFM) results provided 
by the PWROG.
    The NRC agrees with most of the qualitative arguments based on the 
deterministic FM approach. However, the NRC's final position to accept 
ASME Code, Section XI, Appendix E without the proposed conditions is 
not because of these arguments, but rather because of the supporting 
quantitative PFM results provided by the PWROG.
    Although most of the qualitative arguments based on the 
deterministic FM approach have merit, they can only demonstrate that 
the probability of having a flaw close to 1/4T in size is very low. 
They cannot rule out that such a large flaw could exist. This 
observation is consistent with a key statement regarding a large flaw 
in NUREG-1806, ``Technical Basis for Revision of the Pressurized 
Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61).'' 
NUREG-1806 states ``It should also be noted that the empirical data 
used as the primary evidence to establish the distribution of embedded 
weld flaws do not, and cannot, provide any information about the 
maximum size a flaw can be.''
    The final PTS rule (75 FR 13) published on January 4, 2010, is 
based on a PFM analysis using a weld flaw distribution with a cutoff 
flaw depth close to 1/4T of the RPV wall, indicating that although the 
1/4T flaw has a low probability of existence it is prudent to still 
consider it.
    The FM analyses in both ASME Code, Section XI, Appendix G and ASME 
Code, Section XI, Appendix E are based on postulated flaws using linear 
elastic FM in a deterministic approach. It is appropriate to assume 
different margins for these two types of analyses to

[[Page 36241]]

account for the very different occurrence frequencies of the two 
events. However, it is too aggressive to change the fundamental flaw 
size assumption simply based on different event frequencies. Further, 
both appendices are for all RPVs, including the one with the worst 
combination of transients (for the Appendix E analysis), largest 
undetected flaw size, and worst degradation in fracture toughness. 
Therefore, unless a PFM approach is used which accounts for a large 
size flaw with its low probability, it is prudent that the fundamental 
flaw size assumption remains the same in these two deterministic FM 
analyses. The PWROG provided such a PFM approach in its response.
    The PWROG performed a risk-informed assessment of Appendix E using 
the Fracture Analysis of Vessels--Oak Ridge (FAVOR) Code; the same tool 
used in the PFM analyses supporting the final PTS rule. Based on a 
selected PWR and BWR RPV having the highest RTNDT of the 
limiting RPV material and a typical beltline fluence model, the PWROG 
generated a pressure versus (T-RTNDT) curve for each of the 
two RPVs by setting the CDF to 1E-6. The analytical results showed that 
the PWROG's PFM results bounds the corresponding Appendix E curve for 
both the unanticipated isothermal pressure events and the pressurized 
cool-down events. Since (1) the PFM methodology is consistent with the 
PTS rule's underlying methodology, in which large flaws are considered 
statistically, and (2) the resulting pressure versus (T-
RTNDT) curve bounds the corresponding curve based on the 
current Appendix E approach, the NRC concludes that the current 
Appendix E methodology, without the NRC's proposed condition, provides 
an appropriate conservative methodology for evaluating RPV integrity 
following an unanticipated transient that exceeds the operational 
limits in PWR plant operating procedures.
    For these reasons, the NRC agrees with the comment based on the PFM 
analyses that the current ASME Code, Section XI, Appendix E analysis is 
appropriate. The proposed conditions placed on the use of ASME Code, 
Section XI, Appendix E in the proposed rule are, therefore, not 
included in the final rule.
    Comment: Section E-1200 is useful and conservative as is, and 
prohibiting the use of Section E-1200 will ultimately result in added 
utility burden or loss of generation because of the additional time 
required to perform analysis under Section E-1300. It is estimated that 
a Section E-1200 evaluation can be completed in hours while a Section 
E-1300 evaluation may require days or weeks. Furthermore, use of a 1/4T 
flaw size can produce unacceptable analytical results, even though 
crack initiation has not occurred, thereby complicating the resolution 
process following a fairly minor thermal transient or overpressure 
event. [11-24, 14-24, 17-11, 19-1, 21-7]
    NRC Response: The NRC agrees with this comment based on the PFM 
Analysis provided by the PWROG. The final rule does not include the 
condition of paragraph (b)(2)(xxv) from the proposed rule.
    Comment: The NRC should reconsider the change specifying ``* * * 
that Section E-1200 is not acceptable.'' The intent of Section E-1200 
is to provide licensees a conservative and yet simple screening method 
that can be used to immediately judge whether a reactor vessel can be 
returned to service or whether a more in-depth analysis is needed prior 
to returning the reactor vessel to service following an unanticipated 
event. The evaluation procedures in Appendix E, Paragraphs E-1200 and 
E-1300, provide adequate safety margins for evaluating reactor pressure 
vessel integrity following an unanticipated event that results in 
pressures and temperatures outside the limits established for normal 
operation. Additionally, Appendix E is consistent with risk-informed 
acceptance criteria for normal operating and unanticipated events. 
Consequently, modifying Appendix E as proposed is unnecessary and 
disallowing use of Section E-1200 will result in an undue hardship 
without any compensating increase in safety. [20-7]
    NRC Response: The NRC agrees with this comment based on the PFM 
Analysis provided by the PWROG. The final rule does not include the 
condition of paragraph (b)(2)(xxv) from the proposed rule.
10 CFR 50.55a(b)(2)(xxvi) (Proposed Rule Paragraph (b)(2)(xxii))
    Comment: If the NRC intends to require that Risk-Informed ISI (RI-
ISI) Programs comply with RG 1.178, RG 1.200, and NRC Standard Review 
Plan 3.9.8, then in lieu of the proposed condition in paragraph 
(b)(2)(xxvi), the proposed condition should specify that use of 
Nonmandatory Appendix R is acceptable, provided licensees comply with 
these applicable RGs and the Standard Review Plan 3.9.8. [4-18; 11-25; 
14-25; 19-1]
    NRC Response: The NRC disagrees with the comment and believes that 
RI-ISI programs developed in accordance with Nonmandatory Appendix R 
should continue to be submitted as alternatives in accordance with 10 
CFR 50.55a(a)(3). The NRC has not generically approved RI-ISI 
application because the code-approved guidance to date has not 
addressed inspection strategy for existing augmented and other 
inspection programs such as intergranular stress corrosion cracking 
(IGSCC), flow assisted corrosion (FAC), microbiological corrosion 
(MIC), and pitting or provided system-level guidelines for change in 
risk evaluation to ensure that the risk from individual system failures 
will be kept small and dominant risk contributors will not be created. 
Furthermore, allowing the use of Nonmandatory Appendix R without 
requiring submittal of an alternative would allow plants being licensed 
and constructed in accordance with 10 CFR part 52 to implement 
Nonmandatory Appendix R. The NRC believes at this time that the use of 
Nonmandatory Appendix R at plants licensed under 10 CFR part 52 plants 
is something that requires additional review of plant specific 
applications. For these reasons the NRC disagrees with the comment. No 
change was made to the final rule as a result of the comment.
10 CFR 50.55a(b)(3)(v) Subsection ISTD. Article IWF-5000, ``Inservice 
Inspection Requirements for Snubbers''
    Comment: 10 CFR 50.55a(b)(3)(v) should be revised as follows for 
clarification:
    (v) Subsection ISTD. Article IWF-5000, ``Inservice Inspection 
Requirements for Snubbers,'' of the ASME B&PV Code, Section XI, must be 
used when performing inservice inspection examinations and tests of 
snubbers at nuclear plants, except as modified in (A) and (B) below. 
[11-27; 14-27a; 17-12; 19-1]
    NRC Response: The NRC agrees that paragraph (b)(3)(v) should be 
clarified, and revised it to include references to paragraphs 
(b)(3)(v)(A) and (b)(3)(v)(B). The recommended change provides clarity 
between the selection of paragraph (b)(3)(v)(A) or (b)(3)(v)(B). The 
final rule is revised to add the suggested references.
10 CFR 50.55a(b)(3)(v)(A)
    Comment: It is unclear whether the intent of paragraph (b)(3)(v) is 
that, after licensees have updated their programs to comply with the 
2006 Addenda and later editions and addenda of the ASME B&PV Code and 
the equivalent endorsed edition and addenda of the ASME OM Code, 
Subsection ISTD, preservice and inservice examinations need not be 
performed using a VT-3 visual

[[Page 36242]]

examination method as described in IWA-2213. [14-27b; 17-13]
    NRC Response: The NRC agrees with this comment to the extent that, 
as described in paragraph (b)(3)(v)(A), VT-3 visual examination must be 
used while using ASME OM Subsection ISTD in lieu of the requirements 
for snubbers in the editions and addenda up to the 2005 Addenda of the 
ASME Section XI, IWF-5200(a) and (b), and IWF-5300(a) and (b). 
Paragraph (b)(3)(v)(B) states that licensees using the 2006 Addenda and 
later editions of the ASME OM Code Subsection ISTD are not required to 
use VT-3 visual examination, because in the ASME OM Code snubber (pin-
to-pin) visual examination VT-3 requirements have been replaced with 
the Owner's defined visual examination. However, removing VT-3 
requirements for snubbers does not remove VT-3 requirements of support 
structure(s) and attachments as defined in IWF of ASME Section XI.
    The proposed rulemaking would not change the intent of the current 
paragraph (b)(3)(v). The proposed rulemaking would split paragraph 
(b)(3)(v) into (b)(3)(v)(A) and (b)(3)(v)(B), because snubber inservice 
examination and testing requirements have been deleted in the 2006 
addenda and later Editions of ASME Section XI. Up to, and including, 
the 2005 Addenda, both ASME Section XI and ASME OM Code contained 
snubber examination and testing requirements. Now, in the 2006 Addenda, 
the ASME OM Code is the only Code which contains the inservice 
examination and testing requirements of snubbers. The paragraph 
(b)(3)(v)(A) option is for licensees using ASME Section XI up to the 
2005 Addenda, which is similar to the current paragraph (b)(3)(v). The 
paragraph (b)(3)(v)(B) option is for the licensees using the 2006 
Addenda or the later edition of ASME Section XI, where the licensee 
will not find any snubber requirements in ASME Section XI; therefore, 
the ASME OM Code must be used.
    The intent of current paragraph (b)(3)(v) is based on the ASME 
Section XI, IWF-5000 and ASME OM, Subsection ISTD requirements. The 
ASME Section XI up to the 2005 Addenda does not clearly distinguish the 
boundary between the support structure, attachments and the snubber. 
The inservice examination of the support structure and attachments is 
performed using VT-3 as required by Subsection IWF of Section XI, and 
IWF-5000 requires that snubber examination must be performed using VT-3 
visual examination as described in IWA-2213. Subsection ISTD of the 
ASME OM Code does not address inspection of the support structure and 
attachments. Therefore, to be consistent with the Section XI 
requirements, VT-3 visual examination is required when using Subsection 
ISTD of the OM Code in lieu of the IWF-5000 requirements of ASME 
Section XI, up to the 2005 Addenda. The proposed VT-3 requirement is 
consistent with the current requirement to ensure that an appropriate 
visual examination method was used for integral and non-integral 
snubber supports and attachments such as lugs, bolting, and clamps when 
using ISTD of the ASME OM Code in lieu of the ASME Section XI, 2005 
Addenda.
    In the 2006 Addenda and later edition of ASME Section XI, the 
inservice examination and testing requirements of snubbers have been 
deleted, and a Figure IWF-1300-1(f) has been added to clarify the 
boundary of a snubber (pin-to-pin) and its support structure and 
attachments. Figure IWF-1300-1(f) defines that a snubber (pin-to-pin) 
examination is excluded from Section XI, and the support structure and 
attachments, etc. are still under the scope of ASME Section XI. ASME 
Section XI, IWF-1220 in the 2006 Addenda and later edition states that 
inservice examination and testing of snubbers are outside the Scope of 
IWF, and can be found in the ASME OM Code. Subsection IWF requires that 
the inservice examination of support structure and attachments are to 
be performed using VT-3 visual examination, whereas the ASME OM Code 
requires that snubber (pin-to-pin) visual examination is to be 
performed using the Owner's qualified procedures and methods. However, 
if licensees prefer, the VT-3 visual examination method still can be 
used for snubber (pin-to-pin) inservice examination, while using ASME 
OM Code requirements. No change was made to the final rule as a result 
of this comment.
10 CFR 50.55a(b)(3)(v)(B)
    Comment: The examination boundary for a snubber examination as 
defined in ISTD is the snubber unit out to the pins that hold it in 
place. Commenters request that the NRC clarify in the final rule 
whether the pin-to-pin ISTD examination of the snubber unit should be a 
VT-3, even though a VT-3 examination is a Section XI requirement. [14-
27c; 17-13]
    NRC Response: The NRC clarifies that the licensees are required to 
meet the snubber (pin-to-pin) visual examination requirements as 
specified in the Subsection ISTD of the ASME OM Code when using the 
2006 Addenda and later editions and addenda of Section XI of the ASME 
B&PV Code, as defined in paragraph (b)(3)(v)(B). Subsection ISTD of the 
ASME OM Code, 2006 Addenda and later editions requires that snubber 
(pin-to-pin) visual examination is to be performed using the Owner's 
qualified procedures and methods, whereas licensees must use VT-3 for 
integral and non-integral structure and attachments as required by ASME 
Section XI. However, licensees may use VT-3 visual examination method 
for snubber (pin-to-pin) inservice examination, while using ASME OM 
Code, 2006 Addenda and later editions.
    When using the 2005 Addenda or earlier editions and addenda of the 
ASME OM Code, Subsection ISTD in lieu of the ASME Section XI, IWF-5000 
as defined in paragraph (b)(3)(v)(A), licensees must use VT-3 visual 
examination for snubbers (pin-to-pin) and integral and non-integral 
structure and attachments as required by ASME Section XI.
Inservice Testing
10 CFR 50.55a(f)(5)(iv)
    Comment: The words ``and is not included in the revised inservice 
test program as permitted by paragraph (f)(4) of this section'' seem to 
imply that a licensee need not seek relief if the inservice test 
program is revised to identify the impractical test requirement. If 
this is the intent of these words, licensees may not need to submit 
relief requests for IST Program impracticality if the IST Program is 
updated. If this is not the intent of these words, then the phrase 
``and is not included in the revised inservice test program as 
permitted by paragraph (f)(4) of this section'' should be removed from 
paragraph (f)(5)(iv). [4-22]
    NRC Response: The NRC does not agree with the comment. The proposed 
amendment states that where a pump or valve test requirement by the 
code or addenda is determined to be impractical by the licensee and is 
not included in the revised inservice test program, the basis for this 
determination must be submitted for NRC review and approval not later 
than 12 months after the expiration of the initial 120-month interval 
of operation. Therefore, a licensee has to submit relief requests for 
inservice testing (IST) Program impracticality if the IST Program is 
updated. No change was made to the final rule as a result of this 
comment.

[[Page 36243]]

Inservice Inspection
10 CFR 50.55a(g)(2), (g)(3)(i), (g)(3)(ii), (g)(4)(i) and (g)(4)(ii)
    Comment: The introductory text and other applicable sections should 
state that licensees use the provisions for examination and testing of 
snubbers in Subsection ISTD of the ASME OM Code or the requirements in 
plant Technical Specifications (TS). [1-1; 17-6]
    NRC Response: The NRC does not agree with the commenter to include 
the optional provision of TS requirements for inservice examination and 
testing of snubbers along with Subsection ISTD of the ASME OM Code.
    Paragraph (g) establishes the ISI requirements that licensees must 
use when performing ISI of components (including supports). 
Additionally, paragraph (g)(4)(iv) states that ISI of components 
(including supports) may meet the requirements set forth in subsequent 
editions to the ``Code of Record'' and addenda that are incorporated by 
reference in 10 CFR 50.55a(b), subject to limitations and modifications 
listed in 10 CFR 50.55a(b) and subject to NRC approval.
    The requirements at 10 CFR 50.55a do not define any documents 
beyond ``Code of Record'' to control the snubber inservice examination 
and testing program. Licensees have the option to control the ASME 
Code-required ISI and testing of snubbers through their TS or other 
licensee-controlled documents (e.g. technical requirements manual, 
etc.). For facilities using their TS to govern ISI and testing of 
snubbers, paragraph (g)(5)(ii) requires that if a revised ISI program 
for a facility conflicts with the TS, the licensee shall apply to the 
NRC for amendment of the TS to conform the TS to the revised program. 
Therefore, the regulation does not state the type of documents to be 
used by the licensees to meet the snubber inservice examination and 
testing requirements as specified in the ASME Code, but TS must meet 
the ``Code of Record'' requirements. For a particular facility, the 
snubber inservice examination and testing may be controlled by its TS, 
including the applicable snubber inservice examination and testing 
requirements as specified in the ASME Code. No change was made to the 
final rule as a result of this comment.
10 CFR 50.55a(g)(5)(iii) and (g)(5)(iv)
    Comment: The proposed rule adds extra burden on licensees to submit 
relief requests within 12 months of examinations where code 
requirements were determined to be impractical and the proposed rule 
language would put paragraph (g)(5)(iii) in conflict with paragraph 
(g)(5)(iv). [2-3; 4-25; 11-31a-g; 14-31; 17-7; 17-10; 18-1; 20-14; 21-
1; 22-1]
    NRC Response: The NRC agrees with the comments that paragraph 
(g)(5)(iii) would place an extra burden on the licensee by requiring 
that requests for relief made in accordance with paragraph (g)(5)(iii) 
must be submitted to the NRC no later than 12 months after the 
examination has been attempted. This requirement could increase the 
number of submittals licensees need to submit for code requirements 
determined to be impractical. Rather than submitting one request for 
relief at the end of the interval for all requirements determined to be 
impractical throughout the 10-year interval as currently allowed, 
licensees would be required to prepare a submittal within 12 months of 
every examination that determined a requirement was impractical. This 
could result in the licensee preparing numerous submittals for relief 
requests where under the current rules only one submittal is required 
at the end of the interval. This requirement is revised in this final 
rule to align with paragraph (g)(5)(iv) to require submittal of these 
requests no later than 12 months after the expiration of the initial or 
subsequent 120-month inspection interval for which relief is sought.
    Comment: Paragraph (g)(5) in general, and this proposed change to 
paragraph (g)(5)(iii) in particular, could also have a direct impact on 
examinations associated with welds and weld repairs performed during 
the course of a repair/replacement activity. Based on the proposed 
change to paragraph (g)(5)(iii), it could be argued that a relief 
request does not have to be submitted until after performance of a weld 
repair and alternative NDE or NDE with limited coverage. If the intent 
is to exclude NDE associated with welds and weld repairs (i.e., repair/
replacement activities), then the proposed change to paragraph 
(g)(5)(iii) should be revised to make this clarification. [17-8; 17-14; 
18-2]
    NRC Response: If a licensee proposes to use a different inspection 
technique (e.g., UT vs. RT), an alternative must be submitted under the 
provisions of 10 CFR 50.55a(a)(3), regardless of what amount of 
coverage they would achieve with either technique. If the licensee has 
knowledge of the fact that the inspection using the different 
inspection technique will be limited, it is the NRC's expectation that 
such information will be included as an integral part of the requested 
alternative. The alternative that would be approved would be based on 
the technique and the amount of coverage the licensee expects to 
achieve. If the requested alternative is approved and the licensee 
achieves less coverage using the alternative inspection technique than 
that stipulated in the original alternative request, the licensee would 
need to submit a request for relief based on 10 CFR 50.55a(g)(5). No 
change was made to the final rule as a result of this comment.
    Comment: The requirement to submit the relief request after the 
examination has been attempted may in fact be a clarification of the 
NRC's intent, but the requirement to submit the relief request within 
12 months of the attempt is certainly not a clarification, it is a new 
requirement. [2-3]
    NRC Response: The NRC agrees that submitting the relief request 
within 12 months of the attempted examination would be a new 
requirement, which was not the NRC's intent. This paragraph is revised 
in this final rule to align with paragraph (g)(5)(iv).
    Comment: The words ``and is not included in the revised inservice 
inspection program as permitted by paragraph (g)(4) of this section'' 
seem to imply that a licensee need not seek relief if the inservice 
inspection program is revised to identify the impractical requirement. 
If this is the intent of these words, licensees may not need to submit 
relief requests for ISI Program impracticality if the ISI Program is 
updated. If this is not the intent of these words, then the phrase 
``and is not included in the revised inservice inspection program as 
permitted by paragraph (g)(4) of this section'' should be removed from 
10 CFR 50.55a(g)(5)(iv). [4-26]
    NRC Response: The NRC agrees the phrase, ``and is not included in 
the revised inservice inspection program as permitted by paragraph 
(g)(4) of this section,'' could cause confusion, because paragraph 
(g)(4) does not address the basis for the determination of an 
examination requirement's impracticality. The submittal of the basis 
for determination of the impracticality of an examination requirement 
is required by (g)(5)(iii) and the timing of this submittal is 
discussed in (g)(5)(iv). Therefore, paragraph (g)(5)(iv) of the final 
rule is revised to remove the wording ``and is not included in the 
revised inservice inspection program as permitted by paragraph (g)(4) 
of this section.''
10 CFR 50.55a(g)(6)(ii)(F)(1)
    Comment: The final rule should incorporate by reference Code Case 
N-770-1, approved by ASME on Dec. 25, 2009, in lieu of Code Case N-770. 
In

[[Page 36244]]

Code Case N-770-1, ``cladding'' was changed to ``onlay'' to eliminate 
confusion and misapplication in either installation requirements or 
examination/evaluation requirements, or both. The confusion and 
misapplication could result from someone applying the existing Code 
rules for ``cladding,'' which is not the intent when ``cladding 
mitigation'' in N-770 is used. [4-4; 4-27a; 11-3; 11a-34a; 14-3; 14-
34a; 19-1]
    NRC Response: The NRC agrees that incorporating by reference Code 
Case N-770-1 into the final rule could eliminate a number of the 
proposed conditions. Many of the conditions the NRC proposed to impose 
on the use of Code Case N-770 have been incorporated into Code Case N-
770-1, as discussed in specific comments related to Code Case N-770. 
Therefore, the final rule incorporates by reference Code Case N-770-1, 
and does not include most of the conditions on the use of Code Case N-
770 that were included in the proposed rule. The NRC agrees that the 
term ``cladding,'' as used by Section XI, does not apply to mitigation 
in the context of Code Case N-770. ``Onlay'' is the terminology used in 
the code case. The incorporation of Code Case N-770-1 in the final rule 
addresses the commenters' recommendation that the final rule use the 
terminology ``onlay'' instead of ``cladding.''
10 CFR 50.55a(g)(6)(ii)(F)(2)
    Comment: The NRC has typically approved the application of pressure 
boundary weld mitigation techniques on a case-by-case basis. All 
mitigation techniques discussed in Code Case N-770, with the exception 
of Mechanical Stress Improvement Process (MSIP), are the subject of 
separate code cases which will be subject to approval by the NRC. MSIP 
meets the requirements of Appendix I of Code Case N-770 and has been 
separately approved by the NRC. If approved mitigation techniques are 
employed, a separate review of the reclassification of the welds as 
proposed by the condition in paragraph (g)(6)(ii)(F)(2) should not be 
required. [5-2; 8-1; 11a-34b; 14-34b; 16-1; 17-16; 18-4; 19-1; 20-16; 
21-8]
    NRC Response: The NRC disagrees that a separate NRC review of the 
reclassification of welds should not be required for mitigation 
techniques approved in ASME code cases. It is the NRC's position that a 
separate review of the reclassification of welds will be required 
unless NRC-approved mitigation techniques are employed. This condition 
provides clarity for the licensee and inspectors for the classification 
of each weld. Under the condition, unless there is NRC approval of a 
mitigation technique, whether generic or plant specific, such welds 
will be classified as category items A-1, A-2 or B of Table 1 of ASME 
Code Case N-770-1. All mitigation techniques discussed in Code Case N-
770, with the exception of MSIP, are covered by separate code cases in 
various stages of development. These code cases are subject to approval 
by the NRC. As ASME completes these mitigation code cases, the NRC will 
review and approve them, if appropriate, possibly with conditions. The 
NRC uses RG 1.147, which is incorporated by reference in 10 CFR 50.55a, 
to endorse approved code cases for generic use. Based on the wording of 
paragraph (g)(6)(ii)(F)(2), as the NRC endorses mitigation code cases 
in the RG, the rule permits licensees to categorize mitigated welds in 
the corresponding Inspection Items in Code Case N-770-1, without a 
separate NRC review of the classification or reclassification. No 
change to paragraph (g)(6)(ii)(F)(2) was made in the final rule as a 
result of this comment.
    Comment: The proposed condition in paragraph (g)(6)(ii)(F)(2) is 
not consistent with the other proposed conditions in paragraphs 
(g)(6)(ii)(F)(6) and (g)(6)(ii)(F)(7) or Code Case N-770. Paragraph 
(g)(6)(ii)(F)(6) requires that a weld that has been mitigated by inlay 
or corrosion resistant cladding, and then is found to be cracked, be 
reclassified and inspected using the frequencies of Inspection Item A-
I, A-2, or B. This indicates that an uncracked weld that has been 
mitigated by inlay or corrosion resistant cladding would not be 
categorized as Inspection Items A-1, A-2 or B following an acceptable 
pre-service examination. Additionally, paragraph (g)(6)(ii)(F)(7) 
requires that a weld mitigated by inlay or corrosion resistant cladding 
be examined each interval if at hot-leg temperatures and as part of a 
25-percent sample plan on a 20-year frequency if at cold-leg 
temperatures, which is not consistent with Inspection Item A-1, A-2, or 
B. [5-2; 8-1; 11a-34b; 14-34b; 16-1; 17-16; 18-4; 19-1; 20-16; 21-8]
    NRC Response: The NRC agrees with the first point about the 
inconsistency between paragraphs (g)(6)(ii)(F)(2) and (g)(6)(ii)(F)(6), 
but disagrees with the second point about an inconsistency between 
paragraphs (g)(6)(ii)(F)(2) and (g)(6)(ii)(F)(7). Proposed paragraph 
(g)(6)(ii)(F)(6) referred to welds mitigated by inlay or cladding 
rather than referring to welds in Inspection Items G, H, J, and K. The 
wording in proposed paragraph (g)(6)(ii)(F)(6) overlooked the step 
required by paragraph (g)(6)(ii)(F)(2) to obtain NRC authorization for 
an alternative classification of welds as Inspection Items G, H, J, or 
K. However, paragraph (g)(6)(ii)(F)(6) of the proposed rule is not 
included in the final rule because Code Case N-770-1 addresses the 
NRC's concern that was contained in this condition, and Code Case N-
770-1 is incorporated by reference in the final rule.
    The NRC disagrees with the commenters' second point. Paragraph 
(g)(6)(ii)(F)(7) in the proposed rule correctly referred to, and would 
apply to, welds in Inspection Items G, H, J and K. Before welds can be 
categorized as Inspection Items G, H, J, or K, the categorization would 
first have to be authorized by the NRC under the condition in paragraph 
(g)(6)(ii)(F)(2). Therefore, paragraph (g)(6)(ii)(F)(7) in the proposed 
rule would be consistent with paragraph (g)(6)(ii)(F)(2). No change to 
paragraph (g)(6)(ii)(F)(7) of the proposed rule was made in the final 
rule as a result of this comment.
10 CFR 50.55a(g)(6)(ii)(F)(3)
    Comment: The proposed condition in paragraph (g)(6)(F)(3) should 
not be applied. The final rule approval timing for some plants may be 
such that there would not be time to plan and prepare for the required 
baseline inspection under the proposed condition in paragraph 
(g)(6)(ii)(F)(3) and prepare repair contingencies, (e.g., approval of 
the rule in June and the next refueling outage for a plant is in 
September). By providing a window of the next two refueling outages, 
the required planning and preparation can be accommodated.
    Additionally, for baseline examinations already completed to the 
requirements of the industry guidance, any condition applied should 
recognize these examinations as acceptable for compliance to N-770 and 
the NRC Conditions. [5-3; 8-2; 11a-34c; 14-34c; 16-2; 17-17; 18-5; 19-
1; 20-17; 21-9]
    NRC Response: The NRC agrees that more time may be needed after the 
rule becomes effective for licensees to complete the baseline 
examinations, but does not agree that the condition should not be 
included in the final rule. The NRC believes that there are welds 
within the scope of Code Case N-770 that have not been examined under 
the industry program MRP-139, ``Primary System Piping Butt Weld 
Inspection and Evaluation Guideline.'' There may also be welds that 
received less than complete ASME Code, Section XI, examination coverage 
under the MRP-139 program. Paragraph (g)(6)(ii)(F)(3) is necessary to 
ensure that adequate

[[Page 36245]]

baseline examinations have been performed on all welds within the scope 
of Code Case N-770, since these welds are susceptible to PWSCC. The 
need for ensuring the integrity of these welds, beginning with baseline 
examinations, has been recognized by the NRC and industry groups for a 
number of years. The NRC included paragraph (g)(6)(ii)(F)(3) in the 
proposed rule because it believes that the code case requirement 
allowing two refueling outages after adoption of the code case to 
complete the baseline examinations is inconsistent with the safety 
significance of performing the initial inspections of these welds.
    The NRC recognizes that the timing in paragraph (g)(6)(ii)(F)(3) as 
proposed would, in some cases, constrain planning and preparation 
efforts for the required baseline examination. Therefore, for butt 
welds that were not in the scope of MRP-139 and did not receive a 
baseline examination, the timing in paragraph (g)(6)(ii)(F)(3) in the 
final rule is extended to require that these baseline examinations be 
completed at the next refueling outage that occurs more than 6 months 
from the effective date of the final rule. This change in the condition 
would give licensees at least 6 months to plan and prepare for the 
baseline examination. If a baseline examination cannot be performed by 
the licensee to meet the requirements of paragraph (g)(6)(ii)(F), then 
the licensee is required to obtain NRC authorization of alternative 
examination requirements in accordance with paragraphs (a)(3)(i) or 
(a)(3)(ii).
    In response to the comment regarding using examinations performed 
prior to issuance of the final rule as baseline examinations for Code 
Case N-770, the NRC revised paragraph (g)(6)(ii)(F)(3) in the final 
rule to address this option. Previous examinations of these welds can 
be credited for baseline examinations if they were performed using 
Section XI, Appendix VIII requirements and met the Code-required 
examination volume for axial and circumferential flaws of essentially 
100 percent. For butt welds that received a MRP-139 examination that 
did not fully meet Section XI, Appendix VIII requirements, or achieve 
essentially 100-percent coverage, licensees can re-perform the baseline 
examination to meet these requirements or obtain NRC authorization of 
alternative examination requirements in accordance with paragraph 
(a)(3)(i) or (a)(3)(ii) by the end of next refueling outage that occurs 
after 6 months from the effective date of the final rule. This 
provision acknowledges previous examinations that could satisfy the 
Code Case N-770-1 baseline requirement, with NRC authorization of 
alternative examination requirements within a reasonable time frame.
    A licensee may also choose to use previous inspections of 
dissimilar-metal butt welds performed under the plant's ASME Code, 
Section XI, Inservice Inspection program to count as meeting the 
paragraph (g)(6)(ii)(F)(3) baseline requirement. This is acceptable, 
provided the previous inspection falls within the re-inspection period 
for welds in ASME Code Case N-770-1, Table 1, Inspection Items A-1, A-
2, and B. Additionally, the NRC-approved alternative examination 
coverage for these welds during the current 10-year inservice 
inspection interval remain applicable. In all of these cases, the 
previously-approved alternative will continue to apply for the duration 
authorized by the NRC as the final rule does not revoke previous NRC-
approved alternatives or relief requests.
    In the final rule, paragraph (g)(6)(ii)(F)(3) is revised to require 
baseline examinations for welds in Table 1, Inspection Items A-1, A-2, 
and B, to be performed at the next refueling outage that occurs later 
than 6 months after the effective date of the final rule. The rule 
allows previous examinations of these welds to be credited for baseline 
examinations if they were performed (1) within the re-inspection period 
for the weld item in Table 1, and (2) using Section XI, Appendix VIII 
requirements and met the Code-required examination volume of 
essentially 100 percent. The rule allows other previous examinations 
that do not meet these requirements to be used to meet the baseline 
examination requirement, provided NRC authorization of alternative 
inspection requirements in accordance with 10 CFR 50.55a(a)(3)(i) or 
(a)(3)(ii) is granted prior to the end of the next refueling outage 
that occurs later than 6 months after the effective date of the final 
rule.
10 CFR 50.55a(g)(6)(ii)(F)(5)
    Comment: In Code Case N-770-1, approved by the ASME on December 25, 
2009, Paragraph--3132.3(b) has been modified, so the adoption of Code 
Case N-770-1 would make the proposed condition in paragraph 
(g)(6)(ii)(F)(5) no longer necessary. [5-5; 8-4; 11-34e; 14-34e; 16-4; 
19-1; 20-19; 21-11]
    NRC Response: The NRC agrees with this comment for several reasons. 
Code Case N-770, Paragraph --3132.3(b) states that a ``flaw is not 
considered to have grown if the size difference (from a previous 
examination) is within the measurement accuracy of the NDE technique 
employed.'' Use of this terminology may have resulted in a departure 
from the past practice when applying the ASME Code, Section XI. 
Paragraph (g)(6)(ii)(F)(5) of the proposed rule stated that a flaw is 
not considered to have grown if a previously evaluated flaw has 
remained essentially unchanged. This wording is consistent with the 
requirements and practice of NDE under Section XI. Paragraph--3132.3(b) 
of Code Case N-770-1 uses the same wording as contained in paragraph 
(g)(6)(ii)(F)(5) of the proposed rule. The revised requirement of Code 
Case N-770-1 fully addresses the NRC's concern contained in paragraph 
(g)(6)(ii)(F)(5) of the proposed rule. Because the final rule 
incorporates by reference Code Case N-770-1, the final rule does not 
include the condition of paragraph (g)(6)(ii)(F)(5) from the proposed 
rule.
10 CFR 50.55a(g)(6)(ii)(F)(6)
    Comment: Code Case N-770-1, approved by the ASME on Dec. 25, 2009, 
modified Note 16(c), so the adoption of Code Case N-770-1 would make 
the proposed condition in 10 CFR 50.55a(g)(6)(ii)(F)(6) no longer 
necessary. [5-6; 8-5; 11a-34f; 14-34f; 16-5; 19-1; 20-20; 21-12]
    NRC Response: The NRC agrees with this comment for several reasons. 
Code Case N-770 would permit welds mitigated by inlay or cladding 
(i.e., onlay) in Inspection Items G, H, J, and K, to remain in those 
Inspection Items if cracking were to occur that penetrates through the 
thickness of the inlay or onlay. The purpose of an inlay or onlay is to 
provide a corrosion-resistant barrier between reactor coolant and the 
underlying Alloy 82/182 weld material that is susceptible to PWSCC. If 
cracking penetrates through the thickness of an inlay or onlay, the 
inspection frequencies of Inspection Items G, H, J, and K would no 
longer be appropriate even after satisfying the successive examination 
requirements of Paragraph--2420. Paragraph (g)(6)(ii)(F)(6) would 
require welds in Inspection Items G, H, J, or K, with cracking that 
penetrates beyond the thickness of the inlay or cladding, be 
reclassified as Inspection Item A-1, A-2, or B, as appropriate, until 
corrected by repair/replacement activity in accordance with IWA-4000 or 
by corrective measures beyond the scope of Code Case N-770. A new 
sentence added to Note (16)(c) of Code Case N-770-1 states that ``if 
cracking penetrates beyond the thickness of the inlay or onlay, the 
weld shall be reclassified as Inspection Item A-1, A-2, or B, as 
appropriate, until corrected by repair/replacement activity in 
accordance with

[[Page 36246]]

IWA-4000 or by corrective measures beyond the scope of this Case (e.g., 
stress improvement).'' The revision of Note (16)(c) in Code Case N-770-
1 fully addresses the NRC concerns contained in paragraph 
(g)(6)(ii)(F)(6) of the proposed rule. Because the final rule 
incorporates by reference Code Case N-770-1, the final rule does not 
include the condition of paragraph (g)(6)(ii)(F)(6) from the proposed 
rule.
10 CFR 50.55a(g)(6)(ii)(F)(7)
    Comment: The proposed condition is appropriate because the Appendix 
VIII supplement has not yet been developed to demonstrate the detection 
of flaws in the thin inlay or cladding when the examination is 
performed from the outside surface. Code Case N-770-1, approved by the 
ASME on Dec. 25, 2009, modified the ``Extent and Frequency of 
Examination'' column for Inspection Items G, H, J, and K in Table 1, so 
adoption of Code Case N-770-1 would allow the NRC to modify the 
proposed condition in paragraph (g)(6)(ii)(F)(7). [5-7; 8-6; 11a-34g; 
14-34g; 16-6; 19-1; 20-21; 21-13]
    NRC Response: The NRC agrees with this comment. In Code Case N-770, 
the Table 1 column titled ``Extent and Frequency of Examination'' for 
Inspection Items G, H, J, and K (welds mitigated by inlay or cladding) 
only requires a surface examination for welds in Inspection Items G, H, 
J, and K if a volumetric examination is performed from the weld inside-
diameter surface. The NRC proposed adding paragraph (g)(6)(ii)(F)(7) on 
welds in Inspection Items G, H, J, and K, which would have required 
that the ISI surface examination requirements of Table 1 apply whether 
the inservice volumetric examinations are performed from the weld 
outside diameter or the weld inside diameter. A volumetric examination 
performed from the weld outside-diameter surface would not be capable 
of detecting flaws in an inlay or onlay. In Code Case N-770-1, the 
Table 1 column titled ``Extent and Frequency of Examination'' for 
Inspection Items G, H, J, and K contains revised requirements to 
perform a surface examination from the weld inside surface and a 
volumetric examination performed from either the inside or outside 
surface. The revised requirement of Code Case N-770-1 for surface 
examination of welds in Inspection Items G, H, J, and K is the same 
requirement that was contained in paragraph (g)(6)(ii)(F)(7) of the 
proposed rule. Because the final rule incorporates by reference Code 
Case N-770-1, the final rule does not include the surface examination 
requirement of paragraph (g)(6)(ii)(F)(7) from the proposed rule.
10 CFR 50.55a(g)(6)(ii)(F)(8)
    Comment: Code Case N-770-1, approved by the ASME on Dec. 25, 2009, 
modified Notes 11(b)(1) and (2), so adoption of Code Case N-770-1 would 
make the proposed condition in paragraph (g)(6)(ii)(F)(8) no longer 
necessary. [5-9; 8-8; 11a-34h; 16-8; 19-1; 20-23; 21-15]
    NRC Response: The NRC agrees with this comment for several reasons. 
Inspection Items D, G, and H pertain to mitigation of uncracked butt 
welds by stress improvement, weld inlay, and weld onlay, respectively. 
Code Case N-770 does not explicitly preclude deferral of the first 
examination of Items D, G, and H following mitigation to the end of the 
interval. Therefore, the NRC proposed paragraph (g)(6)(ii)(F)(8) to 
ensure that the initial examinations of welds in Inspection Items D, G, 
and H take place on an appropriate schedule to verify the effectiveness 
of the mitigation process. Note (11), which pertains to deferral of the 
first examinations after mitigation, was revised in Code Case N-770-1. 
The revised requirements of Code Case N-770-1, Note (11), indicate that 
the first examinations following mitigation are to be performed within 
10 years following mitigation for Item D butt welds, but can be 
performed any time within the 10 years. The revised requirements of 
Code Case N-770-1, Note (11), indicate that the first examinations 
following mitigation are to be performed as specified in Table 1 for 
Items G and H butt welds. The revised requirements of Code Case N-770-1 
preclude deferral of the first examinations of Item D butt welds beyond 
the 10 years allowed by Table 1, and preclude deferral of the first 
examinations for Item G and H butt welds to the end of an interval, if 
that is later than the specified time in Table 1. The revision of Note 
(11) in Code Case N-770-1 addresses the NRC's concerns in paragraph 
(g)(6)(ii)(F)(8) of the proposed rule. Because the final rule 
incorporates by reference Code Case N-770-1, the final rule does not 
include the condition of paragraph (g)(6)(ii)(F)(8) from the proposed 
rule.
10 CFR 50.55a(g)(6)(ii)(F)(9)
    Comment: Code Case N-770-1, approved by the ASME on Dec. 25, 2009, 
modified paragraph I-1.1, so adoption of Code Case N-770-1 would make 
the proposed condition in paragraph (g)(6)(ii)(F)(9) no longer 
necessary. [5-10; 8-9; 11-34i; 14-34i; 16-9; 19-1; 20-24; 21-16]
    NRC Response: The NRC agrees with this comment for several reasons. 
Code Case N-770, Appendix I, Measurement or Quantification Criteria I-
1.1, requires an analysis that assumes the pre-stress-improvement, 
residual-stress condition resulting from a construction weld repair 
from the inside diameter to a depth of 50-percent of the weld 
thickness. Code Case N-770 does not specify the circumferential extent 
of the weld repair that must be assumed. Paragraph (g)(6)(ii)(F)(9) of 
the proposed rule would require that in applying Measurement or 
Quantification Criterion I-1.1, the weld repair be assumed to extend 
360[deg] around the weld. Code Case N-770-1 specifies in Measurement or 
Quantification Criterion I-1.1 that the weld repair be assumed to 
extend 360[deg] around the weld. The addition of the circumferential 
extent of the assumed weld repair in Appendix I of Code Case N-770-1 
fully addresses the NRC's concern contained in paragraph 
(g)(6)(ii)(F)(9) of the proposed rule. Because the final rule 
incorporates by reference Code Case N-770-1, the final rule does not 
include the condition of paragraph (g)(6)(ii)(F)(9) from the proposed 
rule.
10 CFR 50.55a(g)(6)(ii)(F)(10)
    Comment: Code Case N-770-1, approved by the ASME on Dec. 25, 2009, 
modified paragraph I-2.1, so adoption of Code Case N-770-1 in lieu of 
N-770 in the final rule would allow the NRC to remove this condition. 
[5-11; 8-10; 11-34j; 14-34j; 16-10; 19-1; 20-25; 21-17]
    NRC Response: The NRC agrees with this comment for several reasons. 
Code Case N-770, Appendix I, Measurement or Quantification Criterion I-
2.1, requires that an analysis or demonstration test account for load 
combinations that could cause plastic ratcheting. This wording is 
inappropriate since this criterion pertains to the permanence of a 
mitigation process by stress improvement, and ``shakedown'' rather than 
``ratcheting'' is the phenomenon that could lead to lack of permanence 
of the mitigation. Paragraph (g)(6)(ii)(F)(10) of the proposed rule 
would require that the last sentence of Measurement or Quantification 
Criterion I-2.1 be replaced with a sentence that uses the correct 
terminology. Code Case N-770-1 of Appendix I, Measurement or 
Quantification Criterion I-2.1, requires that an analysis or 
demonstration test account for load combinations that could relieve 
stress due to shakedown. The revised requirement of Code Case N-770-1 
fully addresses the NRC's

[[Page 36247]]

concern contained in paragraph (g)(6)(ii)(F)(10) of the proposed rule. 
Because the final rule incorporates by reference Code Case N-770-1, the 
final rule does not include the condition of paragraph 
(g)(6)(ii)(F)(10) from the proposed rule.
10 CFR 50.55a(g)(6)(ii)(F)(11)
    Comment: The NRC proposes to add a condition to require that in 
applying Measurement or Quantification Criterion I-7.1 of Appendix I, 
an analysis be performed using IWB-3600 evaluation methods and 
acceptance criteria to verify that the mitigation process will not 
cause any existing flaws to grow. However, measurement or 
Quantification Criterion I-7.1 permits the growth of existing flaws in 
welds mitigated by stress improvement recognizing that flaw growth can 
also be caused by fatigue crack growth, which cannot be precluded. 
Criterion I-7.1, however, also includes the requirement that the 
mitigation process will not cause any existing flaws to become 
unacceptable.
    Code Case N-770-1 modified paragraph 1-7.1, so adoption of Code 
Case N-770-1 would allow the NRC to remove proposed condition 10 CFR 
50.55a(g)(6)(ii)(F)(11). [5-12; 8-11; 11a-34k; 14-34k; 16-11; 19-1; 20-
26; 21-18]
    NRC Response: The NRC agrees with this comment for several reasons. 
Code Case N-770, Appendix I, Performance Criteria I-7, requires that 
the stress intensity factor at the depth of the flaw (the flaw tip) be 
determined using combined residual and operating stresses, and shall be 
zero. Under paragraph I-7, no flaw growth could occur if the stress 
intensity factor is zero at the flaw tip using the combined residual 
and operating stresses. The following section of the code case, 
Measurement or Quantification Criteria I-7.1, requires that an analysis 
be performed to verify that the mitigation process will not cause any 
existing flaws to become unacceptable. The NRC proposed adding 
paragraph (g)(6)(ii)(F)(11), because it appeared that, contrary to the 
requirements of I-7, the analysis required by the Mitigation or 
Quantification Criteria may have allowed flaw growth, even growth by 
primary-water stress corrosion cracking.
    The revised requirements of Code Case N-770-1, Appendix I, 
Performance Criteria I-7, state that the stress intensity factor at the 
depth of the flaw shall be determined using combined residual and 
steady-state operating stresses, and shall not be greater than zero. By 
adding the words ``steady-state'' in I-7 of Code Case N-770-1, and 
maintaining the stress intensity factor at the flaw tip to zero or 
less, primary-water stress corrosion cracking would not be expected to 
occur. The next section of the Code Case N-770-1, Measurement or 
Quantification Criteria I-7.1, requires that an analysis be performed, 
using IWB-3600 evaluation methods and acceptance criteria, to verify 
that the mitigation process will not result in any existing flaws 
becoming unacceptable. The revised wording in I-7 and I-7.1 would only 
allow flaw growth under non-steady-state operating stresses (fatigue) 
and would ensure that standard ASME Code analysis methods are used to 
limit any fatigue growth to acceptable levels. Code Case N-770-1, 
Appendix I, uses different wording than proposed in paragraph 
(g)(6)(ii)(F)(11). However, the revised requirements in Code Case N-
770-1 fully address the NRC's concern that the criteria of Code Case N-
770, Appendix I, were contradictory and may have permitted flaw growth 
by PWSCC. Because the final rule incorporates by reference Code Case N-
770-1, the final rule does not include the condition of paragraph 
(g)(6)(ii)(F)(11) from the proposed rule.
10 CFR 50.55a(g)(6)(ii)(F)(13)
    Comment: Code Case N-770-1 modified the wording of the Extent and 
Frequency of Examination for Inspection Items C and F, so adoption of 
Code Case N-770-1 would allow removal of the proposed condition in 10 
CFR 50.55a(g)(6)(ii)(F)(13). [5-14; 8-13; 11-34m; 14-34m; 16-13; 19-1; 
20-28; 21-19]
    NRC Response: The NRC agrees with this comment. Inspection Items C 
and F pertain to butt welds mitigated by full structural weld overlays. 
Note (10) of Code Case N-770 requires that welds in Inspection Items C 
and F that are not included in the 25-percent sample be examined prior 
to the end of the mitigation evaluation period if the plant is to be 
operated beyond that time. Proposed paragraph (g)(6)(ii)(F)(13) was 
written because Code Case N-770 does not contain a similar requirement 
to inspect prior to the end of the mitigation evaluation period for 
welds that are included in the 25-percent sample. Code Case N-770-1, 
Table 1, requires that for welds in the Inspection Items C and F 25-
percent inspection sample that have a design life of less than 10 
years, at least one inservice inspection shall be performed prior to 
exceeding the life of the overlay. The revised requirements in Code 
Case N-770-1 fully address the NRC concern that Inspection Item C and F 
welds in the 25-percent inspection sample may not have been inspected 
prior to the end of the life of the overlay. Because the final rule 
incorporates by reference Code Case N-770-1, the final rule does not 
include the condition of paragraph (g)(6)(ii)(F)(13) from the proposed 
rule.
10 CFR 50.55a(g)(6)(ii)(F)(14)
    Comment: The change in the dimension to be used in determining the 
thickness ``t'' in the acceptance criteria should be adopted, but the 
NRC-proposed condition should not be adopted, for the following reason.
    The proposed condition in paragraph (g)(6)(ii)(F)(14) would cause a 
conflict in the definition of the required examination volume A-B-C-D, 
with Figures 2(a) and 5(a) showing the correct definition of the 
required volume and Figures 2(b) and 5(b) combined with the NRC's 
proposed condition defining a larger and unintended examination volume 
(by extending the examination volume of an overlay in both axial 
directions).
    Code Case N-770-1 removed the \1/2\-inch (13 mm) dimension shown in 
Figures 2(b) and 5(b) of Code Case N-770 and replaced them with 
dimensions ``X'' and ``Y''. The notes beneath each figure define 
dimensions ``X'' and ``Y''.
    Concurrent with the change in the \1/2\-inch dimension, Code Case 
N-770-1 also removed the examination volume A-B-C-D from Figures 2(b) 
and 5(b). This change was made to clarify that Figures 2(b) and 5(b) 
were not defining any examination volume, but were only defining the 
thicknesses to use in applying IWB-3514 acceptance standards. The 
thickness ``t2'' in Figures 2(b) and 5(b) was also revised/corrected in 
Code Case N-770-1 to reflect the total thickness of the original pipe 
plus the overlay at the location of the flaw.
    The adoption of Code Case N-770-1 in lieu of N-770 in the final 
rule would allow the NRC to remove the proposed condition in paragraph 
(g)(6)(ii)(F)(14). If Code Case N-770-1 is not adopted in the final 
rule, the proposed NRC condition needs to be revised to either require 
the use of Figures 2(b) and 5(b) in Code Case N-770-1, or provide 
specific figures to use with the condition that are identical to 
Figures 2(b) and 5(b) in Case N-770-1. [11a-34n]
    NRC Response: The NRC agrees with this comment for several reasons. 
Code Case N-770, Figures 2(b) and 5(b), contain information on 
component thicknesses to be used in application of the acceptance 
standards of ASME Code, Section XI, lWB-3514, to evaluate flaws 
detected during preservice and inservice inspection of weld overlays. 
The \1/2\-inch (13 mm) dimensions shown

[[Page 36248]]

in Figures 2(b) and 5(b) could have resulted in a non-conservative 
application of the acceptance standards. The appropriate dimensions are 
a function of the nominal thickness of the nozzle and pipe being 
overlaid rather than a single, specified value (\1/2\-inch) on either 
side of the weld for all pipes and nozzles. The revision in Code Case 
N-770-1 of the \1/2\-inch dimension in Figures 2(b) and 5(b) to be used 
in determining the thickness ``t'' in the acceptance standards is 
consistent with paragraph (g)(6)(ii)(F)(14) of the proposed rule. 
Concurrent with the change in the \1/2\ inch dimension, Code Case N-
770-1 also removed the examination volume A-B-C-D from Figures 2(b) and 
5(b). This change was made to clarify that Figures 2(b) and 5(b) were 
not defining an examination volume, but were defining the thicknesses 
to use in applying IWB-3514 acceptance standards, that is, the 
locations in the weld overlay where each of the two thicknesses, ``t1'' 
and ``t2'', would apply to flaws. The thickness ``t2'' in Figures 2(b) 
and 5(b) was also corrected in Code Case N-770-1 to reflect the total 
thickness of the original pipe plus the overlay at the location of the 
flaw. The changes to Figures 2(b) and 5(b) that are reflected in Code 
Case N-770-1 address the NRC's concern regarding non-conservative 
application of acceptance standards during preservice inspection. The 
NRC agrees that the other changes made to Figures 2(b) and 5(b) in Code 
Case N-770-1 correct errors in these figures in Code Case N-770. 
Because the final rule incorporates by reference Code Case N-770-1, the 
final rule does not include the condition of paragraph 
(g)(6)(ii)(F)(14) from the proposed rule.
10 CFR 50.55a(g)(6)(ii)(F)(15)
    Comment: The condition as proposed will not accomplish what was 
intended. As proposed, for a flaw in the original nozzle/weld material 
we would have to use ``t'' equal to the inlay/onlay thickness to 
determine the acceptable size per IWB-3514. Nothing would be acceptable 
under that condition. For flaws that are not contained within the 
inlay/onlay/cladding, the value of ``t'' used should be the full 
structural wall thickness. If the NRC feels that there still needs to 
be a condition specified in this area, it needs to be re-structured to 
specify appropriate ``t'' values for flaws that are contained within 
the inlay/onlay, and t values for flaws that are contained in the 
original structural material. [11a-34o; 14-34o; 17-20; 18-9; 19-1]
    NRC Response: The NRC agrees that the condition in paragraph 
(g)(6)(ii)(F)(15) of the proposed rule would be more effective if it 
were revised as recommended. The condition in paragraph 
(g)(6)(ii)(F)(15) of the proposed rule dealt with the value of ``t'' to 
use for flaws found in an inlay or onlay. Although a value of ``t'' 
equal to the full structural wall thickness is inferred by the code 
case, the condition did not address the value of ``t'' to be used for 
flaws that are not contained within the inlay or onlay material. In the 
final rule this condition has been revised to clarify that for 
Inspection Items G, H, J, and K, when applying the acceptance standards 
of ASME B&PV Code, Section XI, IWB-3514, for planar flaws that are not 
contained within the inlay or onlay material, the thickness ``t'' in 
IWB-3514 is the combined thickness of the inlay or onlay and the 
dissimilar metal weld.

III. Discussion of NRC Approval of New Edition and Addenda to the 
Codes, ASME Code Cases N-722-1 and N-770-1, and Other Changes to 10 CFR 
50.55a

    The NRC is amending its regulations to incorporate by reference the 
2005 Addenda through 2008 Addenda of Section III, Division 1, and 
Section XI, Division 1 of the ASME B&PV Code; and the 2005 Addenda and 
2006 Addenda of the ASME OM Code into 10 CFR 50.55a. The NRC also is 
incorporating by reference Code Case N-770-1, and revision 1 to Code 
Case N-722, which was incorporated by reference into the NRC's 
regulations on September 10, 2008 (73 FR 52729).
    The NRC follows a three-step process to determine acceptability of 
new provisions in new editions and addenda to the Codes, and the need 
for conditions on the uses of these Codes. This process was employed in 
the review of the Codes that are the subjects of this rule. First, 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 in 
support of each new or revised Code. Second, the NRC committee 
representatives discuss the Codes and technical justifications with 
other cognizant NRC staff to ensure an adequate technical review. 
Finally, the NRC position on each Code is reviewed and approved by NRC 
management as part of the rule amending 10 CFR 50.55a to incorporate by 
reference new editions and addenda of the ASME Codes, and conditions on 
their use. This regulatory process, when considered together with the 
ASME's own process for developing and approving ASME Codes, provides 
reasonable assurances that the NRC approves for use only those new and 
revised Code edition and addenda (with conditions as necessary) that 
provide reasonable assurance of adequate protection to public health 
and safety and that do not have significant adverse impacts on the 
environment.
    The NRC reviewed changes to the Codes in the editions and addenda 
of the Codes identified in this rulemaking. The NRC concluded, in 
accordance with the process for review of changes to the Codes, that 
each of the editions and addenda of the Codes, and the 1994 Edition of 
NQA-1, are technically adequate, consistent with current NRC 
regulations, and approved for use with the specified conditions.
    The following paragraphs contain the NRC's evaluation of the 
changes to the Code editions and addenda (including new Code 
provisions) and Code Cases N-722-1 and N-770-1, where the NRC added 
new, revised existing, or removed conditions in 10 CFR 50.55a.

Quality Standards, ASME Codes and Institute of Electrical and 
Electronics Engineers (IEEE) Standards, and Alternatives

10 CFR 50.55a(a)
    The NRC is amending Sec.  50.55a(a) to add a new paragraph heading 
entitled ``Quality standards, ASME Codes and IEEE standards, and 
alternatives.'' This will be consistent with paragraph headings 
throughout 10 CFR 50.55a.

Applicant/Licensee-Proposed Alternatives to the Requirements of 10 CFR 
50.55a

10 CFR 50.55a(a)(3)
    The NRC is amending Sec.  50.55a(a)(3) to clarify that an 
alternative must be submitted to, and authorized by, the NRC prior to 
implementing the alternative. Licensees have misinterpreted Sec.  
50.55a(a)(3) and erroneously concluded that it is permissible to obtain 
NRC authorization of an alternative after its implementation. The final 
rule requires that alternatives to the requirements of Sec. Sec.  
50.55a(c), (d), (e), (f), (g), and (h) must be submitted to, and 
authorized by, the NRC prior to implementing the alternatives.

Standards Approved for Incorporation by Reference

10 CFR 50.55a(b)
    The NRC is amending Sec.  50.55a(b) to add a new paragraph heading 
entitled ``Standards approved for incorporation by reference.'' This 
will be consistent with paragraph headings throughout 10 CFR 50.55a.

[[Page 36249]]

    The question has arisen many times in the past of whether 
Subsection NE, ``Class MC Components;'' Subsection NF, ``Supports;'' 
Subsection NG, ``Core Support Structures;'' and Appendices of the ASME 
B&PV Code, Section III, are NRC requirements. The NRC is clarifying in 
this section how the regulations in 10 CFR 50.55a apply to these 
Section III subsections and appendices. This discussion sets forth the 
NRC's views regarding the applicable NRC requirements, clarifies which 
portions of Section III are approved for use by applicants and 
licensees, identifies which portions of Section III are NRC 
requirements, and identifies which portions of Section III are not 
covered by the regulations in 10 CFR 50.55a. The requirements of 
Subsection NH, ``Class 1 Components in Elevated Temperature Service,'' 
of Section III are already addressed in Sec.  50.55a(b)(1)(vi), and the 
bases for these requirements have been discussed in the final rule (69 
FR 58804) issued on October 1, 2004, that amended 10 CFR 50.55a to 
incorporate by reference the 2001 Edition up to and including the 2003 
Addenda of the ASME Code, Section III.
    First, it should be noted that in 10 CFR 50.55a, the NRC mandates 
the use of Section III, Division 1, rules for ASME Code Class 1, 2, and 
3 components in 10 CFR 50.55a(c), (d) and (e), respectively. 
Specifically, 10 CFR 50.55a(c), (d) and (e) state that for applicants 
constructing a nuclear power plant, those components which are part of 
the reactor coolant pressure boundary must meet the requirements for 
Class 1 components in Section III (e.g., Subsection NB, ``Class 1 
Components''); components classified as Quality Group B must meet the 
requirements for Class 2 components (e.g., Subsection NC, ``Class 2 
Components''); and components classified as Quality Group C must meet 
the requirements for Class 3 components (e.g., Subsection ND, ``Class 3 
Components''). The NRC considers the rules of Subsection NCA and 
Section III mandatory appendices to be mandated as well, but only as 
they apply to Class 1, 2, and 3 components because the language in 10 
CFR 50.55a(c), (d) and (e) also covers general requirements in 
Subsection NCA and mandatory appendices in Section III that are 
applicable to Class 1, 2, and 3 components.
    In addition, the introductory text of 10 CFR 50.55a(b) states, in 
part, that the ASME Code, Section III, is approved for incorporation by 
reference by the Director of the Federal Register pursuant to 5 U.S.C. 
552(a) and 1 CFR part 51. However, the regulatory language does not 
identify specific subsections in Section III that are incorporated by 
reference, and one can only assume that all of Section III (including 
all subsections, appendices and Division 2 and 3 rules) are 
incorporated by reference. Although it is clear that Subsections NB, NC 
and ND are regulatory requirements because they are mandated by 10 CFR 
50.55a(c), (d) and (e) as discussed in this document, the lack of 
specific rule language in 10 CFR 50.55a mandating the use of 
Subsections NE, NF, NG, and the Section III mandatory (roman numeral) 
appendices has created confusion about the regulatory requirements 
applicable to Subsections NE, NF, and NG, and the Section III mandatory 
appendices. Subsection NE provides rules for constructing metal 
containment components (Class MC). Subsection NF provides rules for 
constructing supports for Class 1, 2, 3, and MC components. Subsection 
NG provides rules for constructing reactor core support structures. The 
Section III mandatory appendices are used in conjunction with the 
aforementioned subsections. In this sense, ``constructing'' is an all-
inclusive term that comprises the design, fabrication, installation, 
examination, testing, inspection and selection of materials for nuclear 
power plant components.
    The NRC is, therefore, clarifying that when Subsections NE, NF, NG, 
and the Section III mandatory appendices are incorporated by reference, 
but not mandated, these subsections are not NRC requirements. Rather, 
the NRC considers Subsections NE, NF, NG and the Section III mandatory 
appendices to be approved by the NRC for use by applicants and 
licensees of nuclear power plants by virtue of the NRC's overall 
approval of Section III, Division 1 rules without condition. In this 
manner, approval of the rules in Subsections NE, NF, NG, and the 
Section III mandatory appendices is similar to regulatory guidance 
provided in NRC RGs in that it provides an acceptable method for 
meeting NRC requirements and, in this particular case, in 10 CFR part 
50, Appendix A, General Design Criterion (GDC) 1, ``Quality standards 
and records.'' Applicants and licensees may propose means other than 
those specified by the provisions in Subsections NE, NF, NG, and the 
Section III mandatory appendices for meeting the applicable regulation. 
It should be noted that the NRC reviews an applicant's proposed means 
of meeting the requirements of GDC 1 as part of its review of an 
application for each manufacturing license, standard design approval, 
standard design certification and combined license under 10 CFR part 52 
and for each construction permit and operating license under 10 CFR 
part 50 using the guidelines of NRC NUREG-0800, ``Standard Review Plan 
[SRP] for the Review of Safety Analysis Reports for Nuclear Power 
Plants--LWR Edition,'' and applicable regulatory guides. During its 
review of new reactor designs under 10 CFR part 52, the NRC is 
reviewing the criteria and extent of compliance of standard plant 
designs and combined licenses with the rules of the specific edition 
and addenda to Subsections NE, NF, NG, and the associated Section III 
mandatory appendices for applicability to these new reactor designs. 
The process being used by the NRC in the review of Subsections NE, NF, 
NG, and the Section III mandatory appendices for new reactors as 
described in this document is essentially the same process used by the 
NRC for the licensing of all nuclear power plants since the SRP was 
first issued in 1975. Therefore, this clarification does not establish 
new positions or requirements in the regulatory application of 
Subsections NE, NF, NG, and the Section III mandatory appendices to the 
construction of nuclear power plants.
    Because the NRC staff participates on the ASME Code committees in 
the development of any revisions to Subsections NE, NF, NG, and the 
Section III mandatory appendices, the NRC is cognizant of the 
acceptability of the Code rules applicable to Subsections NE, NF, NG 
and the Section III mandatory appendices. NRC's use of consensus 
technical standards meets the requirements of Public Law 104-113, 
National Technology Transfer and Advancement Act of 1995. Additional 
discussion on NRC's compliance with the NTTAA is set forth in Section 
VII, ``Voluntary Consensus Standards,'' of this document.
    Consistent with this discussion, the NRC did not substantially 
change the language in the introductory text to 10 CFR 50.55a(b). The 
NRC is modifying the regulatory language in the introductory text of 10 
CFR 50.55a(b) to clarify that non-mandatory appendices are excluded 
from Section III rules that are incorporated by reference because the 
NRC does not review the acceptability of non-mandatory Section III 
appendices. Similarly, the NRC is clarifying in the introductory text 
of 10 CFR 50.55a(b) that only Division 1 rules of Section III and 
Section XI are incorporated by reference (i.e., Divisions 2 and 3 rules 
are not incorporated by reference). The NRC also is

[[Page 36250]]

incorporating by reference ASME Code Case N-722-1, ``Additional 
Examinations for PWR Pressure Retaining Welds in Class 1 Components 
Fabricated With Alloy 600/82/182 Materials Section XI, Division 1,'' 
and Code Case N-770-1, ``Alternative Examination Requirements and 
Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt 
Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material 
with or without Application of Listed Mitigation Activities.''

ASME B&PV Code, Section III

Introductory Text to 10 CFR 50.55a(b)(1)
    The NRC is amending the introductory text of Sec.  50.55a(b)(1) to 
clarify that references to Section III refer to Section III of the ASME 
Boiler and Pressure Vessel Code.
10 CFR 50.55a(b)(1)(ii) Weld-Leg Dimensions
    The NRC is amending Sec.  50.55a(b)(1)(ii) in order to apply the 
conditions currently in Sec.  50.55a(b)(1)(ii) to the latest Edition 
and Addenda incorporated by reference in this rulemaking. The current 
regulations in Sec.  50.55a(b)(1)(ii) outline the conditions on the use 
of stress indices used for welds in piping design under Subarticles NB-
3600, NC-3600, and ND-3600 of the ASME B&PV Code. The current 
regulations are based on the NRC's concern about the undersized weld-
leg dimension of less than 1.09tn, which results in a weld 
which is weaker than the pipe to which it is adjoined. The reasons for 
the current condition in 10 CFR 50.55a(b)(1)(ii) are articulated in a 
previous NRC rulemaking (64 FR 51370; September 22, 1999).
    In the proposed rule, the NRC proposed a revision to the condition 
identified in Sec.  50.55a(b)(1)(ii) to address the NRC concerns with 
the undersized welds (Cx=0.75 tn), which are not 
acceptable because the current ASME Code design rules would result in a 
circumferential, fillet-welded or socket-welded joint where the weld 
size is smaller than the adjoining pipe wall thickness, which makes the 
weld weaker than the pipe. The proposed rule also included an editorial 
addition of a condition on the use of paragraph NB-3683.4(c)(2). The 
proposed rule indicated that the use of paragraph NB-3683.4(c)(1) is 
currently not allowed and would continue to be prohibited in the 
proposed rulemaking. The addition of the condition on the use of 
paragraph NB-3683.4(c)(2) is purely editorial in nature since, by 
imposing a condition on the use of NB-3683.4(c)(1), the regulations 
would inherently impose a condition on the use of NB-3683.4(c)(2) given 
their use within Section III of the ASME B&PV Code. Therefore, this 
condition in the proposed rule was not new from a technical standpoint. 
Also, an editorial correction was proposed regarding Footnote 11, which 
should be Footnote 13 for the 2004 Edition through the 2008 Addenda in 
Figure NC-3673.2(b)-1 and Figure ND-3673.2(b)-1.
    For licensees and applicants using the 1989 Addenda through the 
latest edition and addenda of Section III of the ASME B&PV Code 
incorporated by reference in Sec.  50.55a(b)(1), the final rule 
prohibits applicants and licensees from applying the following ASME 
Code provisions: subparagraphs NB-3683.4(c)(1) and NB-3683.4(c)(2) and 
Footnote 11 from the 1989 Addenda through the 2003 Addenda, or Footnote 
13 from the 2004 Edition through the 2008 Addenda, to Figures NC-
3673.2(b)-1 and ND-3673.2(b)-1. The final rule requires applicants and 
licensees to adhere to these prohibitions when considering welds with 
leg size less than 1.09tn.
    The NRC received a number of public comments regarding the proposed 
modification to Sec.  50.55a(b)(1)(ii), all of which disagreed with the 
proposed rule language. The disagreements were based on the assertion 
that the proposed rule language was not referencing the correct ASME 
B&PV Code provisions on weld sizes. However, the NRC disagreed with 
these public comments due to the fact that the language in the proposed 
rule was merely a modification to a current condition in the existing 
regulations and none of the public comments received on the proposed 
modification to Sec.  50.55a(b)(1)(ii) present any new arguments or 
information that would cause the NRC to revisit its determination 
described in the previous rulemaking. As previously stated, the reasons 
for the current condition in 10 CFR 50.55a(b)(1)(ii) are articulated in 
a previous NRC rulemaking (64 FR 51370; September 22, 1999). Therefore, 
no change was made to paragraph Sec.  50.55a(b)(1)(ii) of the final 
rule as a result of these comments. The complete bases for making no 
modifications to the proposed rule are found in the public comment 
response document.
10 CFR 50.55a(b)(1)(iii) Seismic Design of Piping
    The NRC is amending Sec.  50.55a(b)(1)(iii) to explicitly prohibit 
the use of Subarticles NB-3200, NB-3600, NC-3600 and ND-3600 from the 
1994 Addenda through the 2005 Addenda of Section III of the ASME B&PV 
Code for the seismic design of piping. However, the amendment to Sec.  
50.55a(b)(1)(iii) does permit the use of Subarticle NB-3200 from the 
2004 Edition through the 2008 Addenda of the ASME Code for the seismic 
design of piping, subject to the new condition identified as Sec.  
50.55a(b)(1)(iii)(A). The amendment to Sec.  50.55a(b)(1)(iii) also 
permits the use of Subarticles NB-3600, NC-3600 and ND-3600 from the 
2006 Addenda through the 2008 Addenda of Section III of the ASME B&PV 
Code for the seismic design of piping, subject to a new condition 
identified as Sec.  50.55a(b)(1)(iii)(B).
    The current requirements regarding piping seismic rules in Section 
III of the ASME B&PV Code were first introduced in the 1994 Addenda to 
the ASME B&PV Code. These rules were subsequently modified in the 2001 
Edition and 2002 Addenda to the ASME B&PV Code. The current regulations 
in Sec.  50.55a(b)(1)(iii) only allow the use of Subarticles NB-3200, 
NB-3600, NC-3600, and ND-3600 from the 1993 Addenda and earlier 
editions and addenda of the ASME B&PV Code, Section III for the seismic 
design of piping.
    As noted, the amendment to Sec.  50.55a(b)(1)(iii) includes the 
addition of a new condition identified as Sec.  50.55a(b)(1)(iii)(A). 
The condition in Sec.  50.55a(b)(1)(iii)(A) resolves an issue 
identified by the NRC regarding the inclusion of reversing dynamic 
loads when calculating the primary bending stresses for Level B service 
limits. Also, the amendment to Sec.  50.55a(b)(1)(iii) includes the 
addition of a new condition identified as Sec.  50.55a(b)(1)(iii)(B). 
The condition in Sec.  50.55a(b)(1)(iii)(B) relates to the use of the 
Do/t ratio and material requirements of NB-3656(b) when 
applying the 2006 Addenda through the 2008 Addenda of Section III of 
the ASME B&PV Code to the seismic design of piping.
    In the proposed rule, the NRC proposed an amendment to Sec.  
50.55a(b)(1)(iii) which would have allowed the use of the latest 
edition and addenda of Section III of the ASME B&PV Code, incorporated 
by reference in this rulemaking, subject to three new conditions 
identified as Sec.  50.55a(b)(1)(iii)(A), (b)(1)(iii)(B), and 
(b)(1)(iii)(C). These additional requirements would have provided three 
conditions on the use the latest edition and addenda of Section III of 
the ASME B&PV Code incorporated by reference in the current rulemaking, 
as they apply to the seismic design of piping. As a result of public 
comments received, the final rule retains only two of the original 
three conditions with respect to the use of the editions and

[[Page 36251]]

addenda of Section III of the ASME B&PV Code incorporated by reference 
in Sec.  50.55a(b)(1) for the seismic design of piping.
    In the proposed rule, the NRC proposed an additional paragraph 
identified as Sec.  50.55a(b)(1)(iii)(A) which addressed the NRC's 
position regarding the B2' indices in paragraph NB-3656 of 
Section III of the ASME B&PV Code. This condition would have stipulated 
that the value of B2' should be no less than 
0.75B2 (from Table NB-3681(A)-1) when applying the 2006 
Addenda through the 2008 Addenda of Section III of the ASME B&PV Code 
for the seismic design of piping. The NRC proposed this condition to 
address the possibility that ferritic steels may exhibit lower margins 
and a decrease in toughness at higher temperatures due to dynamic 
strain aging.
    A number of public comments were received regarding the proposed 
condition on the B2' indices, all of which cited ASME 
Position Paper STP-NU-008, issued on November 6, 2009, as the bases for 
eliminating the proposed condition. This position paper presents 
information demonstrating that dynamic strain aging at typical seismic 
strain rates is insignificant and that adequate margin exists between 
the ASME Section III code criteria and the ultimate moment under 
dynamic cyclic loading (``adequate margin'' refers to the margin 
recommended in Appendix III of NUREG/CR-5361). The NRC agreed with the 
comments, and considers the previous concerns regarding the possible 
reduction in margin due to dynamic strain aging effectively resolved 
based on the information found in the aforementioned ASME position 
paper. Therefore, as a result of public comments received, the final 
rule does not include this condition. Additionally, as a result of the 
deletion of this condition from the final rule, the paragraphs which 
were identified as Sec.  50.55a(b)(1)(iii)(B) and Sec.  
50.55a(b)(1)(iii)(C) in the proposed rule are identified as Sec.  
50.55a(b)(1)(iii)(A) and Sec.  50.55a(b)(1)(iii)(B) in the final rule. 
A more comprehensive discussion regarding the bases for this change can 
be found in the public comment response document.
    In the proposed rule, the NRC proposed an additional paragraph 
identified as Sec.  50.55a(b)(1)(iii)(B) which addressed the NRC's 
position regarding Note (1) of Figure NB-3222-1 of Section III of the 
ASME B&PV Code. The NRC proposed this condition based on the premise 
that while the inclusion of reversing dynamic loads in the calculation 
of primary bending stresses for Level B service limits may not be 
warranted when the Operating Basis Earthquake is not included in the 
design basis for the facility, at other times these loads must be 
considered. Such is the case when a licensee's Operating Basis 
Earthquake level is more than one-third the value of the Safe Shutdown 
Earthquake. However, the current wording of Note (1) in Figure NB-3222-
1 of Section III of the ASME B&PV Code does not account for this 
situation.
    Multiple public comments were received regarding this proposed 
condition and most generally concurred with the proposed language. 
However, all of the public comments received indicated that additional 
specificity should be provided in the condition by adding the words, 
``by NB-3223(b)'' immediately after the word, ``required'' in the 
proposed wording for Sec.  50.55a(b)(1)(iii)(B). The NRC agreed with 
the public commenters based on the fact that the suggestion within the 
comment results in a more direct application of the proposed condition 
in that there is no ambiguity as to how the condition applies with 
respect to the seismic design of piping. The final rule includes 
additional information regarding the applicability of this condition by 
noting the specific subparagraph (NB-3223(b)) for which this condition 
applies when the 2006 Addenda through the 2008 Addenda of Section III 
of the ASME B&PV Code are used for the seismic design of piping as a 
result of public comments received regarding this condition. 
Additionally, as a result of public comments, the final rule regarding 
this condition is identified as Sec.  50.55a(b)(1)(iii)(A). The 
complete bases for this change can be found in the public comment 
response document.
    In the proposed rule, the NRC proposed an additional paragraph 
identified as Sec.  50.55a(b)(1)(iii)(C) which addressed the NRC's 
position regarding the limitation on the Do/t ratio of ASME 
Class 1, 2 and 3 piping when applying Subarticles NB-3600, NC-3600 and 
ND-3600 in the 2006 Addenda through the 2008 Addenda of Section III of 
the ASME B&PV Code. This proposed addition would have placed a 
condition on the Do/t ratio by requiring this value to be no 
greater than 40 when applying Subarticles NB-3600, NC-3600, or ND-3600 
in the 2006 Addenda through the 2008 Addenda of Section III of the ASME 
B&PV Code for the seismic design of piping.
    The public comment responses received regarding this proposed 
condition all indicated that the condition which the NRC was proposing 
already existed within the code, except for one anomaly. Specifically, 
the comments noted that the limitation on the Do/t ratio is 
already contained in NB-3656(b), NC/ND-3653.1(b), NC/ND-3655(b), and, 
by reference to the Level D requirements, NB-3655.2(b) and NC/ND-
3654.2(b). However, the comments also noted that the Do/t 
ratio limitation is not inherent or explicit for Level B service limits 
in Class 1 piping. As such, all of the comments suggested that the 
focus of the proposed condition be narrowed to capture the condition 
where it is not already included within the ASME Code provisions. The 
NRC agreed with these comments.
    The final rule includes a provision for the seismic design of Class 
1 piping which requires the material and Do/t requirements 
of NB-3656(b) to be met for all Service Limits when the Service Limits 
include reversing dynamic loads, and the alternative rules for 
reversing dynamic loads are used as a result of the public comments 
received on this condition. Additionally, as a result of public 
comments, the final rule regarding the condition on the Do/t 
requirements is identified as Sec.  50.55a(b)(1)(iii)(B). The complete 
bases for this change can be found in the public comment response 
document.
10 CFR 50.55a(b)(1)(iv) Quality Assurance
    The NRC is amending Sec.  50.55a(b)(1)(iv) to be consistent with a 
revised quality assurance provision in the 2006 Addenda of the ASME 
B&PV Code, Section III, Subsection NCA. The final rule allows the use 
of 1994 Edition of NQA-1, ``Quality Assurance Requirements for Nuclear 
Facility Applications,'' when using the 2006 Addenda of Section III of 
the ASME B&PV Code and later editions and addenda. The reference to 
ASME NQA-1 in Article 4000 of the ASME B&PV Code, Section III was 
updated to a later edition of NQA-1 in the 2006 Addenda. NCA-4110(b) 
was revised to require that the N-Type Certificate Holders comply with 
the Basic Requirements and Supplements of the ASME NQA-1-1994 Edition. 
Previous editions/addenda of the ASME B&PV Code, Section III referenced 
earlier editions and addenda of ASME NQA-1. There are no significant 
differences between of NQA-1-1994 Edition and the editions and addenda 
of NQA-1 currently referenced in the regulation. The NRC has reviewed 
and found the changes to Subsection NCA that reference the 1994 Edition 
of NQA-1 to be acceptable.

[[Page 36252]]

10 CFR 50.55a(b)(1)(vii) Capacity Certification and Demonstration of 
Function of Incompressible-Fluid Pressure-Relief Valves
    The NRC is amending Sec.  50.55a(b)(1) to add a new paragraph 
(b)(1)(vii) to modify the requirements in Subsection NB of the ASME 
B&PV Code, Section III, for certifying the capacity of incompressible-
fluid, pressure-relief valves when the testing facility has less than 
the full range of pressure capability necessary for achieving valve 
set-pressure conditions during the testing. The NRC has identified no 
issues with performing tests at less than the highest value of the set-
pressure range for incompressible-fluid, pressure-relief valves and 
finds these new requirements for Class 2 and 3 components acceptable as 
described in paragraphs NC-7742 and ND-7742. However, the NRC has 
identified words that were inadvertently left out of the Code during 
the final printing of paragraph NB-7742 for Class 1 components. The 
parallel structure of the counterpart paragraphs (NC-7742 and ND-7742) 
reveal that the words ``for the design and the maximum set pressure'' 
are missing from paragraph NB-7742(a)(2). Without these words, 
paragraph NB-7742(a)(2) is confusing, illogical, and could lead to a 
non-conservative interpretation of the required test pressure for the 
new Class 1 incompressible-fluid, pressure-relief valve designs. For 
these reasons, the final rule includes a condition in paragraph 
(b)(1)(vii) allowing use of paragraph NB-7742 when the corrected 
language intended by the Code is used.

ASME B&PV Code, Section XI

    The regulations in Sec.  50.55a(b)(2) incorporate by reference ASME 
B&PV Code, Section XI, 1970 Edition through the 1976 Winter Addenda; 
and the 1977 Edition (Division 1) through the 2004 Addenda (Division 
1), subject to the conditions identified in Sec.  50.55a(b)(2)(i) 
through (b)(2)(xxvii). The NRC is amending the introductory text to 
Sec.  50.55a(b)(2) to incorporate by reference the 2005 Addenda 
(Division 1) through the 2008 Addenda (Division 1) of the ASME B&PV 
Code, Section XI, clarify the wording, and remove or revise some of the 
conditions as explained in this document.
    The question has arisen in the past of whether Appendices of the 
ASME B&PV Code, Section XI, are NRC requirements. The NRC is clarifying 
in this section how the regulations in 10 CFR 50.55a apply to the 
Section XI subsections and appendices. This discussion sets forth the 
NRC's views regarding the applicable NRC requirements, clarifies which 
portions of Section XI are approved for use by applicants and 
licensees, identifies which portions of Section XI are NRC 
requirements, and identifies which portions of Section XI are not 
covered by the regulations in 10 CFR 50.55a.
    First, it should be noted that in 10 CFR 50.55a, the NRC mandates 
in 10 CFR 50.55a(g)(4) that throughout the service life of a boiling or 
pressurized water-cooled nuclear power facility, components (including 
supports) which are classified Class 1, 2, 3, MC and CC meet the 
requirements of Section XI (with some exceptions). Specifically, within 
Section XI, Subsection IWB provides the requirements for Class 1 
components, Subsection IWC provides the requirements for Class 2 
components, Subsection IWD provides the requirements for Class 3 
components, Subsection IWE provides the requirements for Class MC 
components, and Subsection IWL provides the requirements for Class CC 
components. The NRC considers the rules of Subsection IWA and Section 
XI Mandatory Appendices to be mandated as well, because the language in 
IWA and the Mandatory Appendices covers general requirements that could 
apply to the inservice inspection of Class 1, 2, 3, MC and CC 
components.
    The NRC is clarifying that the Section XI non-mandatory appendices 
which are incorporated by reference into 10 CFR 50.55a are approved for 
use, but are not mandated. The non-mandatory appendices may be used by 
applicants and licensees of nuclear power plants (subject to the 
conditions in 10 CFR 50.55a(b)(2)).
Introductory Text of 10 CFR 50.55a(b)(2)
    The NRC is amending the introductory text of Sec.  50.55a(b)(2) to 
clarify that references to Section XI refer to Section XI of the ASME 
Boiler and Pressure Vessel Code.
10 CFR 50.55a(b)(2)(i) Limitations on Specific Editions and Addenda
    The NRC is amending Sec.  50.55a(b)(2) to remove Sec.  
50.55a(b)(2)(i) from the regulations and is designating that paragraph 
as ``Reserved.'' This paragraph specified which addenda may be used 
when applying the 1974 and 1977 Editions of Section XI of the ASME B&PV 
Code. Section 50.55a(g)(4)(ii) requires that licensees' successive 120-
month inspection intervals comply with the requirements of the latest 
edition and addenda of the code incorporated by reference in Sec.  
50.55a(b)(2). Subsequently, licensees are no longer using these older 
editions (1974 and 1977 Editions) and addenda of the ASME B&PV Code, 
and therefore the NRC removed this paragraph.
10 CFR 50.55a(b)(2)(iii) Steam Generator Tubing
    The NRC is amending Sec.  50.55a(b)(2) to remove Sec.  
50.55a(b)(2)(iii) from the regulations and is designating that 
paragraph as ``Reserved.'' The current regulations in Sec.  
50.55a(b)(2)(iii) state that if the technical specifications of a 
nuclear power plant include surveillance requirements for steam 
generators different than those in Section XI, Article IWB-2000, the 
ISI program of steam generator tubing is governed by the requirements 
in the technical specifications. The 1989 Edition through the 2008 
Addenda of Section XI IWB-2413, ``Inspection Program for Steam 
Generator Tubing,'' state that ``the examinations shall be governed by 
the plant Technical Specification.'' Because the condition in Sec.  
50.55a(b)(2)(iii) is redundant to the 1989 Edition through the 2008 
Addenda of Section XI, the NRC is removing this condition.
10 CFR 50.55a(b)(2)(iv) Pressure-Retaining Welds in ASME Code Class 2 
Piping
    The NRC is amending Sec.  50.55a(b)(2) to remove Sec.  
50.55a(b)(2)(iv) from the regulations and is designating that paragraph 
as ``Reserved.'' This paragraph states how to select appropriate Code 
Class 2 pipe welds in residual heat removal systems, emergency core 
cooling systems, and containment heat removal systems when applying 
editions and addenda up to the 1983 Edition through the Summer 1983 
Addenda of Section XI of the ASME B&PV Code. Section 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 Sec.  50.55a(b)(2). Subsequently, 
licensees are no longer using these older editions and addenda of the 
code (editions and addenda up to the 1983 Edition through the Summer 
1983 Addenda of Section XI) and, therefore, the NRC is removing Sec.  
50.55a(b)(2)(iv).
10 CFR 50.55a(b)(2)(v) Evaluation Procedure and Acceptance Criteria for 
Austenitic Piping
    The NRC is amending Sec.  50.55a(b)(2) to remove Sec.  
50.55a(b)(2)(v) from the regulations and is designating that paragraph 
as ``Reserved.'' This paragraph deals with evaluation procedures and 
acceptance criteria for austenitic piping when applying the

[[Page 36253]]

Winter 1983 Addenda and the Winter 1984 Addenda of Section XI. Section 
50.55a(g)(4)(ii) requires that licensees' successive 120-month 
inspection intervals comply with the requirements of the latest edition 
and addenda of the code incorporated by reference in Sec.  
50.55a(b)(2). Subsequently, licensees are no longer using these older 
editions and addenda of the code (editions and addenda up to the 1983 
Edition through the Summer 1983 Addenda of Section XI), and therefore, 
the NRC is removing Sec.  50.55a(b)(2)(v).
10 CFR 50.55a(b)(2)(vi) Effective Edition and Addenda of Subsection IWE 
and Subsection IWL, Section XI
    The NRC is amending Sec.  50.55a(b)(2)(vi) to stipulate the 
editions and addenda of Subsection IWE and Subsection IWL of Section XI 
of the ASME B&PV Code which are approved for use when licensees are 
implementing the initial 120-month inspection interval for containment 
inservice inspection requirements found in Section XI of the Code. The 
final rule also requires that the use of these applicable editions and 
addenda is subject to the conditions found in Sec.  50.55a(b)(2)(viii) 
and (b)(2)(ix) for Subsection IWL and Subsection IWE, respectively. 
Additionally, the NRC is amending Sec.  50.55a(b)(2)(vi) to change the 
words ``modified and supplemented'' to ``conditioned'' for 
clarification.
10 CFR 50.55a(b)(2)(viii) Examination of Concrete Containments
    This paragraph stipulates the conditions that apply to the 
inservice examination of concrete containments using Subsection IWL of 
various editions and addenda of the ASME B&PV Code, Section XI, 
incorporated by reference in Sec.  50.55a(b)(2). The regulations, in 
part, require that licensees applying Subsection IWL, 2001 Edition 
through the 2004 Edition shall apply the conditions in Sec.  
50.55a(b)(2)(viii)(E) through (b)(2)(viii)(G). The NRC is amending 
Sec.  50.55a(b)(2)(viii) to remove the conditions in Sec.  
50.55a(b)(2)(viii)(F) and (b)(2)(viii)(G) in the final rule when 
applying Subsection IWL of the 2007 Edition with 2008 Addenda of the 
ASME B&PV Code, Section XI because the intent of these conditions has 
been incorporated into the 2007 Edition with the 2008 Addenda of the 
ASME B&PV Code, as explained in this document. Accordingly, the final 
rule requires that licensees applying Subsection IWL, 2007 Edition with 
the 2008 Addenda shall apply only the condition in Sec.  
50.55a(b)(2)(viii)(E). Further, in the final rule, the conditions in 
Sec.  50.55a(b)(2)(viii)(E) through (b)(2)(viii)(G) remain applicable 
to licensees applying Subsection IWL, 2004 Edition through the 2006 
Addenda.
    The condition in Sec.  50.55a(b)(2)(viii)(F) relates to 
qualification of personnel that examine containment concrete surfaces 
and tendon hardware, wires, or strands. This condition states that 
personnel that examine containment concrete surfaces and tendon 
hardware, wires, or strands must meet the qualification provisions in 
IWA-2300, and that the ``owner-defined'' personnel qualification 
provisions in IWL-2310(d) are not approved for use. IWA-2300 stipulates 
qualification provisions for personnel performing nondestructive 
examination, including VT-1, VT-2, and VT-3 visual examinations. 
Paragraph IWA-2312(c) requires training, qualification, and 
certification of visual examination personnel to comply with the 
requirements of Appendix VI of the Code, which makes reference to ANSI/
ASNT CP-189, and allows for limited certification (for personnel who 
are restricted to performing examinations of limited or specific scope, 
i.e., limited operations or limited techniques) per IWA-2350.
    In Subsection IWL of the 2007 Edition, the ASME revised paragraph 
IWL-2100 to state, in part, that except as noted in IWL-2320, the 
requirements of IWA-2300 do not apply. Also, the 2007 Edition deleted 
subparagraphs IWL-2310(d) and IWL-2310(e), which allowed certain 
requirements (i.e., requirements for personnel qualification and 
requirements for visual examination of concrete and tendon anchorage 
hardware, wires, or strands) to be owner-defined. Further, the 2007 
Edition with 2008 Addenda added a new paragraph IWL-2320 ``Personnel 
Qualifications'' and re-designated the former IWL-2320 ``Responsible 
Engineer'' as IWL-2330 ``Responsible Engineer.''
    The new paragraph IWL-2320 stipulates specific plant experience, 
training, written and practical examination and frequency of 
administration to demonstrate training proficiency, and vision test 
requirements for qualification of personnel approved by the Responsible 
Engineer for performing general or detailed visual examinations of 
structural concrete, reinforcing steel and post-tensioning system 
components (i.e., wires, strands, anchorage hardware, corrosion 
protection medium and free water) of Class CC containments. The 
provision requires documentation of qualification requirements in the 
Employer's written practice. The Responsible Engineer is responsible 
for approval, instruction and training of personnel performing general 
and detailed visual examinations. The new provision also provided the 
requisite detailed requirements for the instruction material to be used 
to qualify personnel performing IWL inspections. Specifically, the 
addition included requirements for preservice and inservice inspections 
for concrete (references American Concrete Institute 201.1R), 
reinforcing steel, and post-tensioning items such as wires, strands, 
anchorage hardware, corrosion protection medium, and free water. Thus, 
the qualification requirements adequately include the areas and extent 
of required plant experience, instructional topics for class room 
training in IWL requirements and plant-specific IWL visual examination 
procedures, and require the vision test requirements of IWA-2321. The 
new paragraph IWL-2320, ``Personnel Qualifications,'' details specific 
guidance for personnel qualification for containment concrete and 
reinforcing steel and post-tensioning system visual inspections that 
provide an acceptable level of quality and safety similar to the 
requirements in IWA-2300 and therefore, addressed the intent of the 
conditions in Sec.  50.55a(b)(2)(viii)(F) of the current regulations. 
Therefore, the condition in Sec.  50.55a(b)(2)(viii)(F) is not required 
to be applied for licensees using Subsection IWL, 2007 Edition with the 
2008 Addenda. It is noted that the NRC's acceptance of the new code 
provision IWL-2320, ``Personnel Qualifications,'' is based on paragraph 
IWL-2320 of the 2007 Edition as supplemented by the addition by errata 
in the 2008 addenda.
    The condition in Sec.  50.55a(b)(2)(viii)(G) of the final rule 
requires that corrosion protection material be restored following 
concrete containment post-tensioning system repair and replacement 
activities in accordance with the quality assurance program 
requirements specified in IWA-1400.'' In the 2007 Edition of Subsection 
IWL, the following revisions were made related to corrosion protection 
medium for post-tensioning systems:
    1. The revised paragraph IWL-4110 added footnote 1 which states 
that the corrosion protection medium is exempt from the requirements of 
IWL-4000. However, the corrosion protection medium must be restored in 
accordance with IWL-2526 following concrete containment post-tensioning 
system repair/replacement activities.

[[Page 36254]]

    2. The revised Line Item L2.40 ``Corrosion Protection Medium'' of 
Table IWL-2500-1 added reference to paragraph IWL-2526 in the columns 
for Test or Examination Requirement, Test or Examination Method, and 
Extent of Examination.
    3. In the revised paragraph IWL-2526, subparagraph (b) requires 
that following the completion of tests and examinations required by 
Examination Category L-B, Items L2.10, L.2.20, and L2.30, the corrosion 
protection medium must be replaced to ensure sufficient coverage of 
anchorage hardware, wires, and strands. The total amount replaced in 
each tendon sheath must be recorded and differences between amount 
removed and amount replaced must be documented.
    4. In the revised paragraph IWL-2526, subparagraph (d) requires 
that the Responsible Engineer specify the method for corrosion 
protection medium.
    With the understanding that the Responsible Engineer (who per IWL-
2320 is a Registered Professional Engineer) will ensure that the 
corrosion protection medium is restored in accordance with the 
applicable Quality Assurance Program, the revised paragraphs IWL-
4110(b)(3) [with footnote 1] and IWL-2526, and revised line item L2.40 
in Table IWL-2500-1 of Subsection IWL, 2007 Edition through the 2008 
Addenda adequately incorporated the intent of the condition in Sec.  
50.55a(b)(2)(viii)(G) of the current regulations and is acceptable to 
the NRC. Therefore, the condition in Sec.  50.55a(b)(2)(viii)(G) is not 
required to be applied for licensees using Subsection IWL, 2007 Edition 
through the 2008 Addenda.
10 CFR 50.55a(b)(2)(ix) Examination of Metal Containments and the 
Liners of Concrete Containments
    This paragraph stipulates the conditions that apply to the 
inservice examination of metal containments and liners of concrete 
containments using Subsection IWE of various editions and addenda of 
the ASME B&PV Code, Section XI, incorporated by reference in Sec.  
50.55a(b)(2). As a result of public comments, the NRC is amending Sec.  
50.55a(b)(2)(ix)(A) to divide the existing condition in Sec.  
50.55a(b)(2)(ix)(A) into paragraphs (b)(2)(ix)(A)(1) and 
(b)(2)(ix)(A)(2). The NRC is removing the conditions in Sec.  
50.55a(b)(2)(ix)(A)(1), (b)(2)(ix)(F), (b)(2)(ix)(G), (b)(2)(ix)(H) and 
(b)(2)(ix)(I) when applying the 2004 Edition with 2006 Addenda through 
the 2007 Edition with 2008 Addenda of the ASME Code, Section XI because 
these conditions have now been incorporated into the Code. The NRC is 
also removing the condition in Sec.  50.55a(b)(2)(ix)(I) when applying 
the 2004 Edition, up to and including, the 2005 Addenda. Furthermore, 
the NRC is also amending Sec.  50.55a(b)(2)(ix) to add a new condition 
as Sec.  50.55a(b)(2)(ix)(J) on the use of Article IWE-5000 of 
Subsection IWE when applying the 2007 Edition, up to and including the 
2008 Addenda of the ASME Code, Section XI. These changes are further 
explained in this document.
    The current regulations, in part, require that licensees applying 
Subsection IWE, 1998 Edition through the 2004 Edition apply the 
conditions in Sec.  50.55a(b)(2)(ix)(A), (b)(2)(ix)(B), and 
(b)(2)(ix)(F) through (b)(2)(ix)(I). In the final rule, the conditions 
in Sec.  50.55a(b)(2)(ix)(F) through (b)(2)(ix)(I) remain applicable to 
licensees applying Subsection IWL, 1998 Edition through the 2001 
Edition with the 2003 Addenda. As a minor correction to the current 
regulations, the final rule requires that licensees applying Subsection 
IWE of the 2004 Edition through the 2005 Addenda of the ASME B&PV Code, 
satisfy the requirements of Sec.  50.55a(b)(2)(ix)(A), (b)(2)(ix)(B), 
and (b)(2)(ix)(F) through (b)(2)(ix)(H). This correction is being made 
since paragraph IWE-3511.3 of the 2004 Edition of the ASME B&PV Code 
incorporated the condition in Sec.  50.55a(b)(2)(ix)(I), which requires 
that the ultrasonic examination acceptance standard specified in IWE-
3511.3 for Class MC pressure-retaining components must also be applied 
to metallic liners of Class CC pressure-retaining components. Further, 
the final rule requires that licensees applying Subsection IWE, 2004 
Edition with the 2006 Addenda through the latest edition and addenda 
incorporated by reference in Sec.  50.55a(b)(2) satisfy the 
requirements of Sec.  50.55a(b)(2)(ix)(A) and (b)(2)(ix)(B). This is 
because the intent of the conditions in Sec.  50.55a(b)(2)(ix)(F) 
through (b)(2)(ix)(H) were incorporated into Subsection IWE, 2004 
Edition with the 2006 addenda, and the condition Sec.  
50.55a(b)(2)(ix)(I) was incorporated into Subsection IWE, 2004 Edition, 
as explained in this document.
    The condition in Sec.  50.55a(b)(2)(ix)(F) of the final rule 
requires that VT-1 and VT-3 examinations be conducted in accordance 
with IWA-2200. Personnel conducting examinations in accordance with the 
VT-1 or VT-3 examination method must be qualified in accordance with 
IWA-2300, and the ``owner-defined'' personnel qualification provisions 
in IWE-2330(a) for personnel that conduct VT-1 and VT-3 examinations 
are not approved for use. This condition defines the code provision 
(IWA-2200) and personnel qualification (IWA-2300) requirements for 
personnel performing visual examinations by the VT-1 or VT-3 method, as 
specified in the conditions in Sec.  50.55a(b)(2)(ix)(G) and 
(b)(2)(ix)(H) of the rule. The condition does not allow use of the 
``owner-defined'' personnel qualification provisions in IWA-2330(a) for 
personnel that conduct VT-1 and VT-3 examinations. The revised code 
provision in IWE-2330(a) of the 2006 Addenda requires that personnel 
performing VT-1 and VT-3 visual examinations shall meet the 
qualification requirements of IWA-2300. The revised code provision in 
IWL-2100 of the 2006 Addenda states that IWA-2000 applies with the 
exception that IWA-2210 and IWA-2300 do not apply to general visual 
examination only (except as required by 2330(b) for vision test 
requirements). Therefore, the code provisions in IWA-2200 and IWA-2300 
will apply to VT-1 and VT-3 examinations. Thus, the revised code 
provisions in IWE-2330(a) and IWE-2100 of the 2006 through 2008 Addenda 
fully incorporates the condition in Sec.  50.55a(b)(2)(ix)(F). 
Therefore, the condition in Sec.  50.55a(b)(2)(ix)(F) is not required 
to be applied for licensees using Subsection IWE, 2004 Edition with the 
2006 Addenda and the 2007 Edition through the 2008 Addenda.
    The condition in Sec.  50.55a(b)(2)(ix)(G) of the final rule 
requires that the VT-3 examination method be used to conduct the 
examinations in Items E1.12 and E1.20 of Table IWE 2500-1, and the VT-1 
examination method be used to conduct the examination in Item E4.11 of 
Table IWE-2500-1. An examination of the pressure-retaining bolted 
connections in Item E1.11 of Table IWE-2500-1 using the VT-3 
examination method must be conducted once each interval. The ``owner-
defined'' visual examination provisions in IWE-2310(a) are not approved 
for use for VT-1 and VT-3 examinations. This condition, applicable in 
the current regulations to the 1998 Edition through the 2004 Edition, 
requires that the VT-3 and VT-1 examination methods be used in lieu of 
the ``General Visual'' and ``Detailed Visual'' methods, respectively, 
as specified in Table IWE-2500-1 for the Item Numbers listed in the 
condition, and that the owner-defined visual examination provisions in 
IWE-2310(a) cannot be used for VT-1 and VT-3 examinations. In the 2006 
Addenda through the 2008 Addenda, Table IWE-2500-1 was revised to 
change the examination method for Item

[[Page 36255]]

Numbers E1.12 and E1.20 to the VT-3 method and for Item E4.11 to the 
VT-1 method. Also, a new Examination Category E-G was added for 
pressure-retaining bolting with Item No. E8.10 which requires 100 
percent of each bolted connection to be examined, using the VT-1 method 
and the acceptance standard in the newly added paragraph IWE-3530, once 
during each Inspection Interval with the connection assembled and 
bolting in-place, provided the connection is not disassembled during 
the interval, or in the disassembled configuration if the connection is 
disassembled for any reason during the interval. This VT-1 examination, 
which is more stringent than the VT-3 method specified in the 
condition, is in addition to the general visual examination of 100 
percent of the pressure-retaining bolted connections during each 
inspection period required to be performed under Item No. E1.11 of 
Table IWE-2500-1. Further, the revised IWE-2310 does not have any 
owner-defined provisions for performing visual examinations including 
VT-1 and VT-3 examinations. Thus, the provisions in the revised Table 
IWE-2500-1 and the revised paragraph IWE-2310 addressed the intent of 
the condition in Sec.  50.55a(b)(2)(ix)(G). Therefore, the condition in 
Sec.  50.55a(b)(2)(ix)(G) is not required to be applied for licensees 
using Subsection IWE, 2004 Edition with the 2006 Addenda and the 2007 
Edition through the 2008 Addenda.
    The condition in Sec.  50.55a(b)(2)(ix)(H) of the final rule 
requires that containment bolted connections that are disassembled 
during the scheduled performance of the examinations in Item E1.11 of 
Table IWE-2500-1 be examined using the VT-3 examination method. Flaws 
or degradation identified during the performance of a VT-3 examination 
must be examined in accordance with the VT-1 examination method, and 
the criteria in the material specification or IWB 3517.1 must be used 
to evaluate containment bolting flaws or degradation. As an alternative 
to performing VT-3 examinations of containment bolted connections that 
are disassembled during the scheduled performance of Item E1.11, VT-3 
examinations of containment bolted connections may be conducted 
whenever containment bolted connections are disassembled for any 
reason. The condition in Sec.  50.55a(b)(2)(ix)(H) is similar to the 
condition for bolted connections in Sec.  50.55a(b)(2)(ix)(G), but 
applies only to the examination of pressure-retaining bolted 
connections that are disassembled. The condition requires flaws or 
degradation identified during the VT-3 examination to be examined using 
the VT-1 method. The NRC notes that the VT-1 (and not VT-3) examination 
method is the method specified in the new Item E8.10 for pressure-
retaining bolted connections in the revised Table IWE-2500-1 in the 
2006 Addenda through 2008 Addenda of the ASME B&PV Code. Further, the 
acceptance standard for the VT-1 examination of pressure-retaining 
bolting in the new paragraph IWE-3530 requires that the relevant 
conditions, as defined in IWA-9000, and listed in IWB-3517.1, shall be 
corrected or evaluated to meet the requirements of IWE-3122, prior to 
continued service. Therefore, the new provision for pressure-retaining 
bolting in Table IWE 2500-1, as discussed in this document, and the new 
acceptance standard specified in IWE-3530, as discussed in this 
document, fully addressed the intent of the condition in Sec.  
50.55a(b)(2)(ix)(H). Therefore, the condition in Sec.  
50.55a(b)(2)(ix)(H) is not required to be applied for licensees using 
Subsection IWE, 2004 Edition with the 2006 Addenda and the 2007 Edition 
through the 2008 Addenda.
    The condition in Sec.  50.55a(b)(2)(ix)(I) of the rule requires 
that the ultrasonic examination acceptance standard specified in IWE-
3511.3 for Class MC pressure-retaining components also be applied to 
metallic liners of Class CC pressure-retaining components. This 
condition requires that the acceptance standard in IWE-3511.3 also 
apply to the metallic shell and penetration liners of Class CC 
pressure-retaining components in the re-designated paragraph IWE-3522, 
``Ultrasonic Examination,'' in the 2004 Edition through the 2007 
Edition and 2008 Addenda. Therefore, the condition in Sec.  
50.55a(b)(2)(ix)(I) is not required to be applied for licensees using 
Subsection IWE, 2004 Edition through the 2007 Edition and the 2008 
Addenda.
    The revised paragraph IWE-2310 (IWE-2313 to be specific) and new 
subparagraphs IWE-2420(c) and IWE-2500(d), in the 2006 Addenda through 
the 2008 Addenda, address the condition in Sec.  50.55a(b)(2)(ix)(A) of 
the final rule with regard to requiring evaluation of acceptability of 
inaccessible areas when conditions exist in accessible areas that could 
indicate the presence or result in degradation to such inaccessible 
areas. However, the information specified in the condition to be 
provided in the ISI Summary Report is not explicitly addressed in the 
ASME B&PV Code. Therefore, based on a public comment, for expediency to 
remove part of the condition for certain addenda, the NRC is dividing 
the existing condition in 50.55a(b)(2)(ix)(A) into paragraphs 
(b)(2)(ix)(A)(1) and (b)(2)(ix)(A)(2). The condition in Sec.  
50.55a(b)(2)(ix)(A)(1) of the final rule, addressing the requirement 
for evaluation of inaccessible areas, is not required to be applied for 
licensees using Subsection IWE, 2006 Addenda through the 2008 Addenda. 
However, the condition in Sec.  50.55a(b)(2)(ix)(A)(2), addressing the 
information relative to evaluation of inaccessible areas to be provided 
in the ISI Summary Report, is required to be applied for licensees 
using the 2006 Addenda through the 2008 Addenda.
10 CFR 50.55a(b)(2)(ix)(J)
    The NRC is amending Sec.  50.55a(b)(2)(ix) to add a new Sec.  
50.55a(b)(2)(ix)(J) to place a condition on the use of Article IWE-
5000, ``System Pressure Tests,'' of Subsection IWE when applying the 
2007 Edition up to and including the 2008 Addenda of the ASME Code, 
Section XI, for Class MC pressure-retaining components. The revised 
Article IWE-5000 does not make a distinction between ``major'' and 
``minor'' modification (or repair/replacement) with regard to the type 
of pneumatic leakage tests specified following repair/replacement 
activities. The NRC notes that IWL-5210 provides a reasonable 
quantitative definition of a repair/replacement activity, in terms of 
meeting the design basis Construction Code requirements prior to and 
during the repair/replacement activity, that is considered major for 
Class CC containments and requiring a containment pressure test to be 
conducted at the design basis accident pressure (Pa) that would 
demonstrate structural integrity of the repaired containment. There is 
no such definition provided in IWE-5000 for Class MC containments. IWE-
5000 (2007 Edition with 2008 Addenda) requires a pneumatic leakage test 
to be performed following welding or brazing associated with repair or 
replacement activities, prior to returning the component to service. It 
also allows the test boundary for the pneumatic leak test to be limited 
to the brazed joints and welds affected by the repair/replacement 
activity, which is acceptable from the point of ensuring leak-tightness 
of the locally repaired area. However, it allows a licensee the option 
of only performing a local bubble test even for a ``major'' containment 
modification or repair/replacement, which is not sufficient to provide 
a verification of global structural integrity. Following ``major'' 
containment repair/replacement activities, it makes the

[[Page 36256]]

performance of the appropriate pneumatic leakage test (which is a Type 
A test) in accordance with 10 CFR part 50, Appendix J, optional, which 
is inconsistent with the NRC position and the provisions in 10 CFR part 
50, Appendix J, paragraph IV.A, and hence the NRC is adding a new 
condition in this rule. It is, and has been, the NRC's position that a 
10 CFR part 50, Appendix J, Type A test must be performed following a 
``major'' containment modification or repair/replacement, prior to 
returning the containment to operation. This is because a ``major'' 
containment modification such as the replacement of a large penetration 
or the creation of large construction opening(s) for replacement of 
equipment such as steam generators, reactor vessel head, pressurizers, 
etc., or other similar repair/replacement activity results in the 
breach of the containment pressure boundary that invalidates the 
periodic verification of structural and leak tight integrity provided 
by the previous Type A test as required by the Containment Leakage Rate 
Testing Program in 10 CFR part 50, Appendix J. Further, the breach of 
pressure boundary of the magnitude resulting from a ``major'' 
containment modification has a global effect on containment structural 
integrity and not a localized effect. Therefore, performing a Type A 
test prior to returning to operation, is necessary to provide a 
reasonable assurance and verification of both containment structural 
integrity and leakage integrity following restoration of a breach in 
the containment pressure boundary due to a ``major'' repair/replacement 
activity. Thus, the new condition in Sec.  50.55a(b)(2)(ix)(J) of the 
final rule requires the performance of Type A test following a 
``major'' containment modification of a Class MC containment structure.
    The new condition provides a general, qualitative definition of 
what constitutes a ``major'' modification or repair/replacement 
activity for containments consistent with what the NRC has historically 
considered as major modifications. The new condition also requires 
that, when applying IWE-5000, if a Type A, B or C test is performed in 
accordance with 10 CFR part 50, Appendix J, the test pressure and 
acceptance standard for the test shall also be in accordance with 10 
CFR part 50, Appendix J. This is because the test pressure range in 
IWE-5223.1 seems to apply even for Type B and Type C tests; and the 
acceptance standard for leakage in IWE-5223.5 is based only on Section 
V, Article 10, for any pneumatic leakage test performed when applying 
IWE-5000 of the 2007 Edition up to and including the 2008 Addenda of 
Section XI of the ASME Code. The requirement in the new condition for 
performing a Type A test prior to returning to operation following a 
major containment modification, is necessary to provide a reasonable 
assurance and verification of both containment structural and leakage 
integrity following restoration of a breach in the containment pressure 
boundary due to the ``major'' repair/replacement activity of a Class MC 
containment structure.
10 CFR 50.55a(b)(2)(xv) Appendix VIII Specimen Set and Qualification 
Requirements
    The NRC is amending Sec.  50.55a(b)(2)(xv) so the conditions in 
that paragraph would not apply to the 2007 Edition through the 2008 
Addenda of Section XI of the ASME B&PV Code. Section 50.55a(b)(2)(xv) 
has conditions that may be used to modify Appendix VIII of Section XI, 
1995 Edition through the 2001 Edition. The ASME Boiler and Pressure 
Vessel Code Committees took action to address these conditions in the 
2007 Edition of the Code and revised Appendix VIII to address the NRC's 
concerns with specimen sets and qualification requirements. Therefore, 
the final rule does not require these conditions when using the 2007 
Edition through the 2008 Addenda of the ASME B&PV Code.
10 CFR 50.55a(b)(2)(xv)(A)(2)
    The NRC is amending Sec.  50.55a(b)(2)(xv)(A)(2) to modify the 
condition to allow for an add-on qualification for austenitic welds 
with no austenitic base metal side to an existing Supplement 10 
qualification.
10 CFR 50.55a(b)(2)(xvi) Appendix VIII Single-Side Ferritic Vessel and 
Piping and Stainless Steel Piping Examinations
    The NRC is amending Sec.  50.55a(b)(2)(xvi) to modify the condition 
to only apply to those licensees using the 2006 Addenda and earlier 
editions and addenda of ASME Section XI.
10 CFR 50.55a(b)(2)(xviii) Certification of NDE Personnel
    The NRC is amending Sec.  50.55a(b)(2)(xviii)(B) so the current 
condition in that paragraph would not apply to the 2007 Edition through 
the 2008 Addenda of Section XI of the ASME B&PV Code. Section 
50.55a(b)(2)(xviii)(B) limits the activities that can be performed by 
NDE personnel certified in accordance with IWA-2316 of the 1998 Edition 
through the 2004 Addenda of the ASME B&PV Code. These personnel are 
limited to observing for leakage during system leakage and hydrostatic 
tests conducted in accordance with IWA-5211(a) and (b). The ASME Boiler 
and Pressure Vessel Code Committees took action to address this, and 
modified IWA-2316 in the 2005 Addenda and the 2007 Edition to limit the 
activities performed by personnel qualified in accordance with IWA-
2316. Therefore, the condition is not required when using the 2007 
Edition through the 2008 Addenda. Accordingly, the NRC is amending 
Sec.  50.55a(b)(2)(xviii)(B) for this condition not to apply when using 
the 2007 Edition through the 2008 Addenda of the ASME B&PV Code.
    The NRC is amending Sec.  50.55a(b)(2)(xviii)(C) so the condition 
in that paragraph would not apply to the 2005 Addenda through the 2008 
Addenda of Section XI of the ASME B&PV Code. This paragraph places 
conditions on the qualification of VT-3 examination personnel certified 
under paragraph IWA-2317 of the 1998 Edition through the 2004 Addenda. 
The regulation requires the administering of an initial qualification 
examination to demonstrate proficiency of this training, and 
administering subsequent examinations on a 3-year interval. The ASME 
Boiler and Pressure Vessel Code Committees took action to address this 
condition and modified IWA-2317 in the 2005 Addenda of the ASME B&PV 
Code to require a written examination for initial qualification and at 
least every 3 years thereafter for VT-3 qualification. Therefore, the 
final rule does not require this condition when using the 2005 Addenda 
through the 2008 Addenda. The NRC is revising the wording of the 
condition for clarity in the final rule based on public comment.
10 CFR 50.55a(b)(2)(xix) Substitution of Alternative Methods
    The NRC is amending Sec.  50.55a(b)(2)(xix) so the conditions for 
the substitution of alternative examination methods in that paragraph 
would not apply when using the 2005 Addenda through the 2008 Addenda. 
The conditions in Sec.  50.55a(b)(2)(xix) do not allow the use of 
Section XI, IWA-2240 of the 1998 Edition through the 2004 Edition of 
the ASME B&PV Code. These conditions also do not allow the use of IWA-
4520(c) of the 1997 Addenda through the 2004 Edition of Section XI of 
the ASME B&PV Code. In 2005, the ASME Boiler and Pressure Vessel Code 
Committees took action to address these conditions and modified IWA-
2240 and deleted IWA-4520(c) in the 2005 Addenda such that alternative

[[Page 36257]]

examination methods or newly developed techniques are not allowed to be 
substituted for the methods specified in the construction code. 
Therefore, these conditions are not required when using the 2005 
Addenda through the 2008 Addenda.
    The final rule also imposes the condition that paragraphs IWA-
4520(b)(2) and IWA-4521 of the 2007 Edition of Section XI, Division 1, 
of the ASME B&PV Code, with the 2008 Addenda are not approved for use. 
In the 2008 Addenda of Section XI of the ASME B&PV Code, the ASME added 
new provisions in IWA 4520(b)(2) and IWA-4521 that allow the 
substitution of ultrasonic examination (UT) for radiographic 
examination (RT) specified in the Construction Code. Substitution of UT 
for RT as addressed in paragraph IWA-4520(b)(2) of the ASME B&PV Code, 
Section XI, for the repair/replacement welds in 2008 Addenda is of a 
concern to the NRC because, depending on flaw type (i.e., volumetric or 
planar) and orientation, UT and RT are not equally effective for flaw 
detection and characterization. The NRC had originally identified 
concerns relative to the calibration blocks to be used, and developed 
two conditions that appear in RG 1.84, ``Design, Fabrication, and 
Materials Code Case Acceptability, ASME Section III, Proposed Revision 
34,'' October 2006.
    RT is effective in detecting volumetric-type flaws (e.g., slag, 
porosity, root concavity, and misalignment), planar type flaws with 
large openings (e.g., lack of fusion and large cracks in high stressed 
areas), and those flaws that are oriented in a plane parallel to the X-
ray beam. RT is effective in all materials common to the nuclear 
industry for detecting the type of flaws generated during construction 
due to workmanship issues and, therefore, ensures an acceptable level 
of weld quality and safety at the time of construction. In contrast, UT 
is most effective in detecting and sizing planar-type flaws associated 
with inservice degradation due to, for example, fatigue or stress 
corrosion cracking. Significant advances have recently been made 
regarding the use of UT to detect flaws in cast stainless steel. 
However, the ASME Code provisions addressing the inspection of cast 
stainless steels are still under development and are, therefore, not 
yet published for use. Finally, UT requires more surface scanning area 
than RT to perform examinations.
    To ensure that a UT technique would be capable of detecting typical 
construction flaws, the NRC requires a licensee to demonstrate, through 
performance-based ASME B&PV Code, Section XI, Appendix VIII-like 
requirements, its capability of identifying the construction flaws 
which are easily detected by RT. Performance-based qualifications 
require demonstrations on mockups having flaws with realistic UT 
responses and with a statistically sufficient number of representative 
flaws and non-flawed volumes to establish procedure effectiveness and 
personnel skill. The statistical approach to qualification has been 
shown to improve the reliability of inspections, to improve the 
probability of flaw detection, and to reduce the number of false calls. 
The addition of only two or three construction flaws to a demonstration 
is not sufficient to capture the variety of flaws common to 
construction or to statistically evaluate procedure effectiveness and 
personnel skills.
    The NRC is concerned that using the second leg of the ultrasound 
metal path (V-path) to achieve two direction scanning from only one 
side of the weld may not be adequate in detecting construction flaws. 
Single-side examinations have not been demonstrated for construction 
flaws for any material. Single-side examinations of welds have been 
successfully qualified for planar flaws in ferritic carbon and low 
alloy steels but have not been reliably demonstrated for austenitic 
stainless steel and nickel alloys.
    Based on this information, the NRC concludes that the substitution 
of UT for RT may not be adequate for detecting some construction flaws, 
specifically in a single-V full penetration groove welds. Therefore, 
substitution of UT for RT is not generically acceptable. This position 
is consistent with the NRC's previous position with respect to the 
review of ASME Code Case N-659-1, which is published in RG 1.193, 
Revision 2, ``ASME Code Cases not Approved for Use.'' Accordingly, the 
final rule imposes the condition that paragraphs IWA-4520(b)(2), and 
IWA-4521 of the 2007 Edition of Section XI, Division 1, with 2008 
Addenda are not approved for use.
10 CFR 50.55a(b)(2)(xxi) Table IWB-2500-1 Examination Requirements
    The NRC is amending Sec.  50.55a(b)(2)(xxi) to remove and designate 
as ``Reserved'' paragraph (b)(2)(xxi)(B) of this section because this 
condition was not consistent with the NRC's unconditional approval of 
Code Case N-652-1 in RG 1.147, Revision 15.
10 CFR 50.55a(b)(2)(xxiv) Incorporation of the Performance 
Demonstration Initiative and Addition of Ultrasonic Examination 
Criteria
    The NRC is amending Sec.  50.55a(b)(2)(xxiv) not to apply the 
condition when using the 2007 Edition through the 2008 Addenda. Section 
50.55a(b)(2)(xxiv) prohibits the use of Appendix VIII, the supplements 
of Appendix VIII and Article I-3000 of ASME B&PV Code, 2002 Addenda 
through the 2004 Edition. In 2007, the ASME Boiler and Pressure Vessel 
Code Committees took action to address this condition and modified 
Appendix VIII and its Supplements in the 2007 Edition. Therefore, the 
condition is not required when using the 2007 Edition through the 2008 
Addenda, and the final rule eliminates this condition when using the 
2007 Edition through the 2008 Addenda.
10 CFR 50.55a(b)(2)(xxv) Evaluation of Unanticipated Operating Events
    The NRC had proposed a new Sec.  50.55a(b)(2)(xxv) to condition the 
use of ASME B&PV Code, Section XI, Nonmandatory Appendix E, 
``Evaluation of Unanticipated Operating Events.'' Based on the 
Probabilistic Fracture Mechanics Analysis (PFMA) provided by 
commenters, which used the Fracture Analysis of Vessels--Oak Ridge 
(FAVOR) Code, the same tool used in the PFM analyses supporting the 
final PTS rule (75 FR 13), the NRC concludes this condition is no 
longer necessary. The PFMA showed that, based on a selected PWR and BWR 
RPV having the highest RTNDT of the limiting RPV material 
and a typical beltline fluence model, the PFMA generated a pressure 
versus (T - RTNDT) curve for each of the two RPVs by setting 
the CDF to 1E-6. The analytical results showed that the PFMA results 
bounds the corresponding Appendix E curve for both the unanticipated 
isothermal pressure events and the pressurized cool-down events. Since 
(1) the PFMA methodology is consistent with the PTS rule's underlying 
methodology, in which large flaws are considered statistically, and (2) 
the resulting pressure versus (T - RTNDT) curve bounds the 
corresponding curve based on the current Appendix E approach, the NRC 
concludes that the current Appendix E methodology, without the NRC's 
proposed condition, provides an appropriate conservative methodology 
for evaluating RPV integrity following an unanticipated transient that 
exceeds the operational limits in PWR plant operating procedures. 
Therefore, the proposed condition placed on the use of

[[Page 36258]]

ASME Code, Section XI, Appendix E in the proposed rule is not included 
in the final rule.
10 CFR 50.55a(b)(2)(xxvii) Removal of Insulation
    The NRC is amending Sec.  50.55a(b)(2)(xxvii) to refer to IWA-5242 
of the 2003 Addenda through the 2006 Addenda or IWA-5241 of the 2007 
Edition through the 2008 Addenda of Section XI of the ASME B&PV Code 
for performing VT-2 visual examination of insulated components in 
systems borated for the purpose of controlling reactivity. The 
regulations at Sec.  50.55a(b)(2)(xxvii) place specific requirements on 
when insulation must be removed to visually examine insulated 
components in accordance with IWA-5242. In the 2007 Edition of the ASME 
B&PV Code, Section XI, paragraph IWA-5242 was deleted and these 
requirements were included in paragraph IWA-5241.
10 CFR 50.55a(b)(2)(xxviii) Analysis of Flaws
    The NRC is amending Sec.  50.55a(b)(2) to add a new paragraph 
(b)(2)(xxviii) placing conditions on the use of Section XI, 
Nonmandatory Appendix A, ``Analysis of Flaws.'' The final rule places a 
condition on the use of Appendix A related to the fatigue crack growth 
rate calculation for subsurface flaws defined in paragraph A-4300(b)(1) 
when the ratio of the minimum cyclic stress to the maximum cyclic 
stress (R) is less than zero. The fatigue crack growth rate, da/dN, is 
defined as follows when using Equation (1) in paragraph A-4300(a) and 
Equation (2) in paragraph A-4300(b)(1):

da/dN = 1.99 x 10-10 S ([Delta]KI)3.07
Where S is a scaling parameter and [Delta]KI is the range 
of applied stress intensity factor.

    S and [Delta]KI are defined in A-4300 (b)(1) of the ASME 
B&PV Code, Section XI, Appendix A as follows:

For -2 <= R <= 0 and Kmax - Kmin <= 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = 
Kmax
For R < -2 and Kmax - Kmin <= 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = (1 
-R) Kmax/3
For R < 0 and Kmax - Kmin > 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = 
Kmax - Kmin

    The above guidelines permit reduction of [Delta]KI from 
the value of (Kmax - Kmin) when Kmax - 
Kmin <= 1.12 [sigma]f [radic]([pi]a). This is 
adequate if the material property [sigma]f is from test-
based data of the component material and if the geometry of the cracked 
component can be modeled as an edge crack in a half plane, so that the 
formula K = 1.12 [sigma] [radic]([pi]a) applies. In most ASME B&PV 
Code, Section XI, Appendix A applications, test-based 
[sigma]f is not available, and the generic value from the 
ASME B&PV Code tabulations is used. Further, the geometry of a 
subsurface flaw in a plate differs significantly from the model of an 
edge crack in a half plane. Consequently, for the case where full 
[Delta]KI should be used, the calculation in accordance with 
ASME B&PV Code, Section XI, Appendix A may show that Kmax-
Kmin <= 1.12 [sigma]f [radic]([pi]a) and prompt a 
wrongful reduction of [Delta]KI.
    To address the use of the generic [sigma]f value instead 
of the test-based value for the cracked component and the significant 
difference between the cracked component geometry and the cracked test-
specimen geometry on which the criterion of 1.12 [sigma]f 
[radic]([pi]a) is derived, the NRC revised the criterion of 1.12 
[sigma]f [radic]([pi]a) to 0.8 times 1.12 
[sigma]f [radic]([pi]a). By doing so, reduction of 
[Delta]KI will not take place during the range of 
Kmax - Kmin from 0.8 x 1.12 [sigma]f 
[radic]([pi]a) to 1.12 [sigma]f [radic]([pi]a), erasing the 
non-conservatism from the two sources mentioned above. Selection of a 
multiplying factor of 0.8 is based on the following:
     The 10 percent error that could be introduced for the 
subsurface flaw configurations having membrane stress correction 
factors less than 1.12 as indicated in Appendix A, Figure A-3310-1; and
     Another 10-percent error that accounts for the uncertainty 
in the [sigma]f value.
    Applying the revised criterion of 0.8 times 1.12 
[sigma]f [radic]([pi]a), results in the following condition 
on the use of the fatigue crack growth rate calculation for subsurface 
flaws defined in paragraph A-4300(b)(1) of Section XI, Nonmandatory 
Appendix A when R is less than zero:

da/dN = 1.99 x 10-10 S ([Delta]KI)\3.07\

    For R < 0, [Delta]KI depends on the crack depth, a, and 
the flow stress, [sigma]f. The flow stress is defined by 
[sigma]f = \1/2\ ([sigma]ys + 
[sigma]ult), where [sigma]ys is the yield 
strength and [sigma]ult is the ultimate tensile strength in 
units ksi (MPa) and a is in units in. (mm).

For -2 <= R <= 0 and Kmax-Kmin <= 0.8 x 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = 
Kmax
For R < -2 and Kmax - Kmin <= 0.8 x 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = (1 
-R) Kmax/3
For R < 0 and Kmax-Kmin > 0.8 x 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = 
Kmax -Kmin
10 CFR 50.55a(b)(2)(xxix) Non-Mandatory Appendix R
    The NRC is amending Sec.  50.55a(b)(2) to add a new condition in 
Sec.  50.55a(b)(2)(xxix) to condition the use of ASME B&PV Code, 
Section XI, Non-Mandatory Appendix R, ``Risk-Informed Inspection 
Requirements of Piping.'' The final rule requires licensees to submit 
an alternative in accordance with Sec.  50.55a(a)(3) and obtain NRC 
authorization of the proposed alternative prior to implementing RI-ISI 
programs under Appendix R. The 2004 Edition of the ASME B&PV Code, 
Section XI, currently incorporated by reference in the regulations, did 
not contain provisions for RI-ISI. The 2005 Addenda introduced Non-
Mandatory Appendix R into Section XI to provide requirements for the 
RI-ISI of ASME B&PV Code Class 1, 2 and 3 piping. The addition of 
Appendix R to Section XI was essentially the incorporation of ASME Code 
Cases N-577 and N-578 into the ASME B&PV Code. The NRC determined that 
ASME Code Cases N-577 and N-578 were unacceptable for use and are 
currently listed in RG 1.193,``ASME Code Cases Not Approved for Use,'' 
Revision 2. Licensees have been implementing RI-ISI requirements for 
piping as an alternative to the ASME B&PV Code, Section XI requirements 
of Tables IWB-2500-1, IWC-2500-1 and IWD-2500-1 submitted in accordance 
with Sec.  50.55a(a)(3). Adding a condition as Sec.  50.55a(b)(2)(xxvi) 
that would require licensees to submit an alternative in accordance 
with Sec.  50.55a(a)(3) and obtain NRC authorization of the proposed 
alternative prior to implementing Appendix R, RI-ISI programs would 
ensure that future RI-ISI programs continue to comply with RG 1.178, 
``An Approach for Plant-Specific Risk-Informed Decisionmaking for 
Inservice Inspection of Piping,'' RG1.200, ``An Approach for 
Determining the Technical Adequacy of Probabilistic Risk Assessment 
Results for Risk-Informed Activities,'' and NRC Standard Review Plan 
3.9.8, ``Risk-Informed Inservice Inspection of Piping.''

ASME OM Code

    The NRC is amending the introductory text in Sec.  50.55a(b)(3) to 
incorporate by reference the 2005 and 2006 Addenda of the ASME OM Code 
into 10 CFR 50.55a. The amendment to Sec.  50.55a(b)(3) also clarifies 
that Subsections ISTA, ISTB, ISTC, and ISTD, Mandatory Appendices I and 
II, and Nonmandatory Appendices A through H and J of the ASME OM Code 
are incorporated by reference.
    The conditions in Sec.  50.55a(b)(3)(i), (b)(3)(ii), and (b)(3)(iv) 
continue to apply to the 2005 and 2006 Addenda because the earlier ASME 
B&PV Code provisions that these regulations are based on were not 
revised in the 2005 and 2006 Addenda of the ASME B&PV Code to address 
the underlying issues

[[Page 36259]]

which led the NRC to impose the conditions on the ASME B&PV Code.
    The NRC is amending the current requirements in Sec.  
50.55a(b)(3)(v) to be consistent with the revised snubber ISI 
provisions in the 2006 Addenda of the ASME B&PV Code, Section XI. To 
accomplish this Sec.  50.55a(b)(3)(v) was divided into Sec.  
50.55a(b)(3)(v)(A) and Sec.  50.55a(b)(3)(v)(B). Where Sec.  
50.55a(b)(3)(v)(A) allows licensees using editions and addenda up to 
the 2005 Addenda of ASME Section XI to optionally use Subsection ISTD, 
ASME OM Code in place of the requirements for snubbers in Section XI. 
Section 50.55a(b)(3)(v)(B) requires licensees using the 2006 Addenda 
and later editions and addenda of Section XI to follow the requirements 
of Subsection ISTD of the ASME OM Code for snubbers. Provisions for the 
ISI of snubbers have been in Subsection ISTD since the ASME OM Code was 
first issued in 1990.
10 CFR 50.55a(b)(3)(v) Subsection ISTD
    Section 50.55a(b)(3)(v) allows licensees using editions and addenda 
up to the 2004 Edition of the ASME B&PV Code, Section XI to comply 
with, at their option, Subsection ISTD, ASME OM Code instead of the 
requirements for snubbers in Section XI. If the licensee chooses to 
comply with subsection ISTD, Sec.  50.55a(b)(3)(v) requires the snubber 
preservice and inservice examinations to be performed using the VT-3 
visual examination method. The NRC previously imposed this requirement 
to ensure that an appropriate visual examination method was used for 
the inspection of integral and non-integral snubber attachments, such 
as lugs, bolting, and clamps when using Subsection ISTD of the ASME OM 
Code. Section 50.55a(b)(3)(v)(A) allows licensees using editions and 
addenda up to the 2005 Addenda of ASME B&PV Code, Section XI, to 
optionally use Subsection ISTD, ASME OM Code in place of the 
requirements for snubbers in Section XI and continues to invoke the VT-
3 requirement. This option does not apply when using the 2006 Addenda 
and later editions and addenda of Section XI of the ASME B&PV Code. 
Figure IWF-1300-1 was revised in the 2006 Addenda of Section XI to 
clarify that integral and non-integral snubber attachments are in the 
scope of Section XI. Therefore, the visual examination method specified 
in the 2006 Addenda and later editions and addenda of Section XI 
applies to the examination of integral and non-integral snubber 
attachments. The NRC is thus amending Sec.  50.55a(b)(3)(v)(B) in the 
final rule to require licensees using the 2006 Addenda and later 
editions and addenda of Section XI to follow the requirements of 
Subsection ISTD of the ASME OM Code for snubbers.
10 CFR 50.55a(b)(3)(vi) Exercise Interval for Manual Valves
    The NRC is amending the current requirement for exercising manual 
valves in Sec.  50.55a(b)(3)(vi). The final rule limits the current 
requirement to the 1999 through 2005 Addenda of the ASME OM Code. The 
current requirement is not applicable to the 2006 Addenda of the ASME 
OM Code because the exercise interval in Subarticle ISTC-3540 for 
manually operated valves was revised in this Addenda to make it the 
same as the current requirement in Sec.  50.55a(b)(3)(vi).

Reactor Coolant Pressure Boundary, Quality Group B Components, and 
Quality Group C Components

    The NRC is amending Sec.  50.55a(c)(3), (d)(2), and (e)(2) to 
replace ``but--'' with ``subject to the following conditions'' at the 
end of the introductory text to each paragraph for clarity.

Inservice Testing Requirements

10 CFR 50.55a(f)(5)(iv) Requests for Relief
    The NRC is amending Sec.  50.55a(f)(5)(iv) to clarify that 
licensees are required to submit requests for relief based on 
impracticality within 12 months after the expiration of the IST 
interval for which relief is being sought. Section 50.55a(f)(5)(iv) 
describes the licensee's responsibility to demonstrate to the 
satisfaction of the NRC those items determined to be impractical and 
discusses the timeframe for this determination. The final rule 
clarifies Sec.  50.55a(f)(5)(iv) to more clearly articulate the 
requirements for licensee action when compliance with certain code 
requirements is determined to be impractical. Licensees have 
interpreted the current language in Sec.  50.55a(f)(5)(iv) in a number 
of ways, especially regarding NRC approval of their submittal within 
the specified timeframe. Since the licensee has little or no control 
over the timeliness of NRC action on their submittal, this 
interpretation is problematic.

Inservice Inspection Requirements

Snubber Examination and Testing
    Paragraphs (g)(2), (g)(3)(i), (g)(3)(ii), the introductory text of 
paragraph (g)(4), and paragraphs (g)(4)(i) and (g)(4)(ii) of 10 CFR 
50.55a reference Section XI of the ASME B&PV Code for component support 
ISI (including snubber examination and testing provisions). Section 
50.55a(b)(3)(v) allows licensees the option of complying with the 
provisions in Subsection ISTD of the ASME OM Code for snubber 
examination and testing in lieu of the ISI provisions for snubber 
examination and testing in Article IWF-5000 of Section XI of the ASME 
B&PV Code. However, Article IWF-5000 was deleted in the 2006 Addenda of 
Section XI. Therefore, the NRC is amending Sec.  50.55a(b)(3)(v) to 
require that licensees who use the 2006 Addenda and later editions and 
addenda of Section XI must use the snubber examination and testing 
provisions in Subsection ISTD of the ASME OM Code.
    The NRC is amending Sec.  50.55a(g)(2), (g)(3)(i), (g)(3)(ii), 
(g)(4)(i) and (g)(4)(ii) to require that licensees use the provisions 
for preservice and inservice examination and testing of snubbers in 
Subsection ISTD of the ASME OM Code when using the 2006 Addenda and 
later edition of Section XI. Licensees may also use optional code cases 
in RG 1.192 as approved by the NRC. The NRC is clarifying that 
preservice examination may meet preservice examination requirements in 
Section III as an alternative to preservice examination of Section XI. 
The NRC is also amending the introductory text of Sec.  50.55a(g)(4) to 
require that licensees using the ASME OM Code shall follow the 
provisions in Subsection ISTD of the ASME OM Code for examination and 
testing of snubbers instead of Article IWF-5000 of Section XI. 
Provisions for examinations and tests of snubbers have been in Article 
IWF-5000 since Subsection IWF was first issued in the Winter 1978 
Addenda of Section XI, but Article IWF-5000 was deleted in the 2006 
Addenda of Section XI. Because Article IWF-5000 was deleted, Subarticle 
IWF-1220 in the 2006 Addenda of Section XI states that the examination 
and testing requirements for snubbers are now outside the scope of 
Section XI, and that the examination and test requirements for snubbers 
can be found in Subsection ISTD of the ASME OM Code.
    The NRC is also correcting an error to reinstate rule language 
adopted in an August 2007 rulemaking (72 FR 49352; August 28, 2007), 
which was deleted in a final rule (72 FR 71750; December 19, 2007) 
whose publication closely followed the August 2007 rule. The statement 
of considerations for the December 2007 rule did not acknowledge or 
explain the reason for its removal of rule language which was adopted 
four months earlier. The NRC believes that the December 2007 removal of 
the rule language adopted in August 2007 was inadvertent, and the

[[Page 36260]]

result of the NRC's failure to revise the ``December 2007 rule language 
to reflect the newly-adopted August 2007 rule language, before the 
December 2007 rule was transmitted to the Federal Register for 
publication.
    This correction was not included in the May 4, 2010 proposed rule 
(75 FR 24324) which preceded this final rule. The NRC finds, in 
accordance with the Administrative Procedure Act, 5 U.S.C. 
553(b)(3)(B), that good cause exists for adopting this correction 
without notice in the Federal Register and an opportunity for public 
comment.
    The NRC is also amending Sec.  50.55a(g)(4)(ii) to provide at least 
18 months for a specified set of licensees to update and begin 
implementation of the 2007 Edition and 2008 Addenda versions of 
Appendix VIII in their next inservice inspection interval. This set of 
licensees are those whose next inservice inspection interval must begin 
to be implemented during the period between 12 through 18 months after 
the effective date of the final rule, and therefore would otherwise be 
required to implement the 2007 Edition and 2008 Addenda versions of 
Appendix VIII (providing them less than 18 months to comply with the 
provisions of the 2007 Edition and 2008 Addenda versions of Appendix 
VIII). For these licensees, the final rule permits a delay of 6 months 
in the implementation of Appendix VIII only (i.e., these licensees will 
still be required to update and implement the inservice inspection 
program during the next inspection interval without delay). Other 
licensees, whose next inservice inspection interval commences more than 
18 months after the final date of the rule, will have sufficient time 
to develop their programs for the next inservice inspection interval 
and are not affected by this provision of the final rule.
10 CFR 50.55a(g)(4)(iii) Surface Examinations of High-Pressure Safety 
Injection Systems
    Section 50.55a(g)(4)(iii) currently gives licensees the option of 
not performing surface examinations of high-pressure safety injection 
systems as specified in Section XI, Table IWB-2500-1, ``Examination 
Category B-J,'' Item Numbers B9.20, B9.21 and B9.22. Editions and 
addenda of Section XI after the 1995 Edition have been modified, and 
some of the Item Numbers have either changed or been deleted. The 
surface examination requirement was removed from Table IWB-2500-1 in 
the 2003 Addenda. Therefore, the final rule requires this condition to 
apply to those licensees using Code editions and addenda prior to the 
2003 Addenda.
10 CFR 50.55a(g)(5)(iii) and (g)(5)(iv) Inservice Inspection Requests 
for Relief
    Section 50.55a(g)(5)(iii) currently requires the licensee to notify 
the NRC if conformance with certain code requirements are found to be 
impractical and submit the information to support this determination to 
the NRC. Section 50.55a(g)(5)(iv) currently requires that when 
examination requirements of the code are determined to be impractical 
by the licensee, that the basis for this determination must be 
demonstrated to the satisfaction of the NRC not later than 12 months 
after the expiration of the 120-month interval during which the 
examination is determined to be impractical.
    The final rule adds a sentence to Sec.  50.55a(g)(5)(iii) to 
clarify that a request for relief must be submitted only after the 
necessary examination has been attempted during a given ISI interval 
and the ASME B&PV Code requirement determined to be impractical. In the 
past, licensees have submitted requests under Sec.  50.55a(g)(5)(iii) 
prior to performing the ASME B&PV Code examination in a given interval 
based on limited examination coverage from previous ISI 10-year 
intervals. The NRC believes that this is an inappropriate basis for a 
determination of impracticality as new examination techniques are often 
developed from one interval to the next, which could result in a 
reasonable expectation of improved results. As a result, the NRC 
believes that a licensee usually cannot make the determination that an 
examination is indeed impractical without first attempting the 
examination in the current ISI interval. In addition, if the NRC were 
to grant relief prior to the component having been examined and the 
results of the examination are less than stated in the request for 
relief, the licensee would be required to resubmit the request for 
relief to address the actual examination. This places an unnecessary 
burden on the licensee and the NRC to review the same issue twice. 
Accordingly, the final rule requires that the determination of 
impracticality should be based on actual attempts to perform a 
requirement, and that the relief request be submitted only after the 
licensee has unsuccessfully attempted to perform the inspection in the 
relevant inspection interval.
    The final rule removes the requirement that the basis for the 
licensee's determination that an examination is impractical be 
demonstrated to the satisfaction of the NRC not later than 12 months 
after the expiration of the 120-month interval during which the 
examination is determined to be impractical. The current regulatory 
language is problematic, inasmuch as the current regulations do not 
explicitly require the licensee to submit a request for relief. This 
interpretation of the current regulations was reflected in a comment 
which stated that the current regulations may be interpreted to mean 
that determinations of impracticality need not be submitted to the NRC 
for approval (i.e., the licensee merely needed to be able to justify 
the impracticality determination to the NRC's satisfaction if asked by 
the NRC). In addition, the NRC recognizes that the licensee has little 
or no control over the timeliness of NRC action on a licensee's request 
for relief. Therefore, the final rule removes the current regulatory 
language, and replaces it with language clearly stating that all 
licensee determinations of impracticality must be submitted to the NRC 
for approval.
    The proposed rule would have required that a relief request under 
Sec.  50.55a(g)(5)(iii) be submitted no later than 12 months after the 
examination has been attempted in a given ISI interval and the licensee 
has determined that the ASME Code requirement is impractical. Several 
commenters stated that this proposed change, which differs from the 
current requirement to submit a single relief request at the end of the 
ISI interval, would place additional burden on licensees by increasing 
the number of submittals licensees need to submit for code relief when 
requirements are determined to be impractical. Rather than submitting 
one request for relief at the end of the interval for all examination 
requirements determined to be impractical throughout the 10-year 
interval as currently allowed, licensees would be required to prepare a 
submittal within 12 months of every examination that determined a 
requirement was impractical. The NRC has determined that the 
administrative burden on the licensee of preparing multiple relief 
requests throughout the inspection interval, and the concomitant burden 
on the NRC to act on those relief requests, does not appear to be 
justified. Therefore, the final rule requires relief requests under 
paragraph (g)(5)(iv) to be submitted no later than 12 months after the 
expiration of the 120-month interval for which relief is sought.
10 CFR 50.55a(g)(6)(ii)(E) Reactor Coolant Pressure Boundary Visual 
Inspections
    The NRC is amending Sec.  50.55a(g)(6)(ii)(E)(1) through 
(g)(6)(ii)(E)(3) to reference Revision 1 of Code Case N-722, and is 
revising

[[Page 36261]]

footnote 1 to clarify requirements in that paragraph that pertain to 
reactor coolant pressure boundary visual inspections. In the last 
update to 10 CFR 50.55a, the NRC added new Sec.  50.55a(g)(6)(ii)(E), 
requiring all PWR licensees to augment their ISI program by 
implementing ASME Code Case N-722, subject to the conditions specified 
in Sec.  50.55a(g)(6)(ii)(E)(2) through (g)(6)(ii)(E)(4). ASME Code 
Case N-722-1, ``Additional Examinations for PWR Pressure Retaining 
Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials 
Section XI, Division 1,'' was published in Supplement 8 of the 2007 
Edition of the ASME Boiler and Pressure Vessel Code Nuclear Code Case 
book. Code Case N-722 has been updated to Revision 1 (N-722-1) and 
contains one additional note indicating that visual examination of 
Alloy 600/82/182 materials in flange seal leak-off lines is not 
required. This change eliminates the need for licensees to submit 
relief requests under Sec.  50.55a(3)(i) or 50.55a(a)(3)(ii) for flange 
seal leak-off lines which are not normally exposed to a corrosive 
environment and are inaccessible for visual examination. The NRC 
believes that the likelihood of the flange seals being degraded is 
relatively low. Therefore, the visual inspection of these flange leak-
off lines is not needed.
    The current wording in the second sentence of footnote 1 to Sec.  
50.55a(g)(6)(ii)(E) has generated some confusion, and has the 
unintended consequence of some licensees believing that they need to 
submit additional relief requests related to the percentage of 
inspections to be completed during the current interval. The second 
sentence in the current footnote was intended to provide guidance to 
licensees for the distribution of weld inspections required by Code 
Case N-722 throughout the remainder of a plant's current 10-year ISI 
period after January 1, 2009. The intent was to require licensees to 
distribute the population of weld inspections that are required only 
once in a 10-year interval to be distributed over a licensee's current 
interval and into the next interval in a manner such as that described 
in IWA-2400 of the 1994 Addenda and later editions and addenda of 
Section XI. Because the current wording was not specific, licensees 
using editions and addenda of Section XI prior to the 1994 Addenda have 
interpreted the regulation as requiring all the weld inspections 
required by Code Case N-722 to be distributed over, and inspected 
during, the remaining periods and outages in the current interval only, 
which could be less than 10 years. The final rule revises footnote 1 to 
Sec.  50.55a(g)(6)(ii)(E) to clarify this issue by directing licensees 
to use the rules of IWB-2400 of the 1994 Addenda or later editions and 
addenda of Section XI for scheduling weld inspections for Code Case N-
722-1 welds added in the middle of an interval.
10 CFR 50.55a(g)(6)(ii)(F) Examination Requirements for Class 1 Piping 
and Nozzle Dissimilar-Metal Butt Welds
    The NRC proposed adding a new Sec.  50.55a(g)(6)(ii)(F) to require 
licensees to implement ASME Code Case N-770, ``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 the Application of Listed Mitigation 
Activities, Section XI, Division 1,'' with 15 conditions. Code Case N-
770 contains baseline and ISI requirements for unmitigated butt welds 
fabricated with Alloy 82/182 material and preservice and ISI 
requirements for mitigated butt welds fabricated with Alloy 82/182 
material. On December 25, 2009, ASME approved Code Case N-770-1. The 
ASME prepared Code Case N-770-1 to address comments on Code Case N-770 
that NRC had provided to the ASME code committee. The NRC addressed 
these comments in the proposed rule as conditions on implementation of 
Code Case N-770.
    The NRC reviewed the changes made in Code Case N-770-1 to determine 
if it was appropriate for referencing in the new Sec.  
50.55a(g)(6)(ii)(F) in lieu of Code Case N-770. The NRC concluded that 
it was appropriate for referencing based on the following 
considerations. Incorporation by reference of Code Case N-770-1 in lieu 
of Code Case N-770 allows the NRC to remove eight and partially remove 
one of the 15 conditions in the proposed rule. The NRC concluded that 
removing these conditions significantly improves the rule. The basis 
for removing or modifying each of these proposed conditions is 
contained in the Analysis of Public Comments document (ADAMS Accession 
No. ML110280240).
    ASME Code Case N-770-1 has, in addition to changes to address 
proposed NRC conditions, additional changes that made no significant 
modification to the requirements from N-770. The NRC considers that the 
editorial changes improve the usability of the rule. Only one technical 
addition was made in Code Case N-770-1 that was not covered by the 
proposed rule. The technical addition provides an alternative 
examination volume for welds mitigated by optimized weld overlays. The 
NRC concluded that, with the exception of the one technical addition, 
Code Case N-770-1 was appropriate for referencing. Therefore, the NRC 
is amending its regulations to incorporate Code Case N-770-1 by 
reference instead of Code Case N-770. The NRC is adding a new condition 
to the rule to preclude the use of the technical addition made to Code 
Case N-770-1. The NRC has prepared a document, ``Review of Changes 
Between American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code Cases N-770 and N-770-1 to Support 10 CFR 50.55a Final 
Rule'' (ADAMS Accession No. ML111250292), setting forth the NRC's bases 
for approval of all of the changes made in Code Case N-770-1.
    In addition to the new condition discussed, the NRC is adding a 
condition and is modifying two conditions from the proposed rule as a 
result of public comments it received. Because a number of the proposed 
conditions were not included, many of the remaining conditions in the 
final rule have been renumbered.

Substitution of the Term ``Condition'' in 10 CFR 50.55a

    The NRC is amending 10 CFR 50.55a to substitute the word 
``condition(s)'' for the words ``limitation(s),'' ``modification(s),'' 
and ``provision(s)'' throughout 10 CFR 50.55a for consistency. The NRC 
does not believe it necessary to distinguish among different types of 
``caveats'' that it imposes on the use of the ASME Codes. Therefore, 
the NRC will now use the term ``condition'' for clarity and 
consistency.

IV. Paragraph-by-Paragraph Discussion

Quality Standards, ASME Codes and IEEE Standards, and Alternatives

10 CFR 50.55a(a)
    The NRC is amending 10 CFR 50.55a to add the title ``Quality 
standards, ASME Codes and IEEE standards, and alternatives'' to 
paragraph (a).

Applicant/Licensee Proposed Alternatives to the Requirements of 10 CFR 
50.55a

10 CFR 50.55a(a)(3)
    The NRC is amending 10 CFR 50.55a(a)(3) to clarify that a proposed 
alternative must be submitted to, and authorized by, the NRC prior to 
an applicant or licensee implementing the alternative.

[[Page 36262]]

Standards Approved for Incorporation by Reference

10 CFR 50.55a(b) Standards Approved for Incorporation by Reference
    The NRC is amending 10 CFR 50.55a(b) to add the title ``Standards 
approved for incorporation by reference'' to paragraph (b).
    The final rule also clarifies that non-mandatory appendices are 
excluded from the ASME B&PV Code, Section III requirements that are 
incorporated by reference into 10 CFR 50.55a, and clarifies that only 
Division 1 requirements of Section III and Section XI are incorporated 
by reference (not Division 2 and Division 3 requirements). The NRC is 
also incorporating by reference ASME Code Case N-722-1 and N-770-1 into 
10 CFR 50.55a.

ASME B&PV Code, Section III

10 CFR 50.55a(b)(1)
    The NRC is amending paragraph (b)(1) to incorporate by reference 
the 2005 Addenda (Division 1) through 2008 Addenda (Division 1) of 
Section III of the ASME B&PV Code into 10 CFR 50.55a, subject to the 
conditions outlined in modified paragraphs (b)(1)(i) through 
50.55a(b)(1)(vi) and paragraph (b)(vii). The paragraph modification 
also includes an editorial change to the references to Section III ASME 
B&PV Code for clarification purposes. As a result, applicants and 
licensees may use the 1974 Edition (Division 1) through the 2008 
Addenda (Division 1) of Section III of the ASME B&PV Code subject to 
the conditions contained within modified paragraphs (b)(1)(i) through 
(b)(1)(vi) and new paragraph (b)(1)(vii).
10 CFR 50.55a(b)(1)(ii) Weld-Leg Dimensions
    The NRC is applying the existing condition in paragraph (b)(1)(ii) 
regarding stress indices used for weld stresses in piping design to the 
comparable provisions in the ASME Code editions and addenda 
incorporated by reference in this final rule. The paragraph 
modification also includes the addition of a condition on the use of 
paragraph NB-3683.4(c)(2) for applicants and licensees applying the 
1989 Addenda through the latest edition and addenda of Section III of 
the ASME B&PV Code incorporated by reference in this final rule. As a 
result, this final rule prohibits applicants and licensees from using 
Footnote 13 from the 2004 Edition through the 2008 Addenda of Section 
III of the ASME B&PV Code to Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 
for welds with leg size less than 1.09 times the nominal pipe wall 
thickness (tn). Also as a result, the use of paragraph NB-
3683.4(c)(2), is not allowed for applicants and licensees applying the 
1989 Addenda through the latest edition and addenda of Section III of 
the ASME B&PV Code incorporated by reference in this final rule.
10 CFR 50.55a(b)(1)(iii) Seismic Design of Piping
    The NRC is amending paragraph (b)(1)(iii) to impose conditions on 
the seismic design of piping when licensees use the latest editions and 
addenda of the ASME B&PV Code, Section III, incorporated by reference 
in modified paragraph (b). The paragraph is also amended to include an 
editorial change to replace ``limitations and modifications'' with 
``conditions'' and ``limitation'' with ``condition.'' The final rule 
allows the use of Subarticles NB-3200, NB-3600, NC-3600, and ND-3600 
for the seismic design of piping when applying editions and addenda, up 
to and including the 1993 Addenda of the ASME B&PV Code, Section III, 
subject to the condition in modified paragraph (b)(1)(ii). The amended 
paragraph does not allow the use of Subarticles NB-3200, NB-3600, NC-
3600, and ND-3600 for the seismic design of piping when applying the 
1994 Addenda through the 2005 Addenda of Section III of the ASME B&PV 
Code except that Subarticle NB-3200 in the 2004 Edition through the 
2008 Addenda of Section III of the ASME B&PV Code may be used by 
applicants and licensees subject to the condition in new paragraph 
(b)(1)(iii)(A) (see the following discussion on this new paragraph). 
The final rule allows the use of Subarticles NB-3200, NB-3600, NC-3600, 
and ND-3600 for the seismic design of piping when applying the 2006 
Addenda through the 2008 Addenda of Section III of the ASME B&PV Code, 
subject to the two new conditions in new paragraphs (b)(1)(iii)(A) and 
(b)(1)(iii)(B).
10 CFR 50.55a(b)(1)(iii)(A)
    The NRC is amending 10 CFR 50.55a(b)(1)(iii) to add a new paragraph 
(b)(1)(iii)(A) which requires licensees and applicants using Note (1) 
of Figure NB-3222-1 in Section III of the 2004 Edition up to and 
including the 2008 Addenda of the ASME B&PV Code to include reversing 
dynamic loads in calculating primary bending stresses, if consideration 
of these loads is warranted by subparagraph NB-3223(b).
10 CFR 50.55a(b)(1)(iii)(B)
    The NRC is amending 10 CFR 50.55a(b)(1)(iii) to add a new 
paragraph(b)(1)(iii)(B) which imposes a condition on the use of 
Subarticle NB-3600 of the ASME B&PV Code, Section III when applying the 
2006 Addenda through the 2008 Addenda of Section III of the ASME B&PV 
Code by requiring the material and Do/t requirements found 
in NB-3656(b) to be adhered to for all Service Limits if the Service 
Limits include reversing dynamic loads which are not required to be 
combined with non-reversing dynamic loads, and the alternative rules 
for reversing dynamic loads are used. As such, per NB-3656(b), the 
final rule requires that licensee's adhere to a Do/t ratio 
limitation requiring this ratio to be less than 40 for all Service 
Limits when evaluating the seismic design of Class 1 piping. Paragraph 
(b)(1)(iii) specifies both whether the condition applies and the 
circumstances in which it applies.
10 CFR 50.55a(b)(1)(iv) Quality Assurance
    The NRC is amending paragraph (b)(1)(iv) to allow the use of the 
1994 Edition of NQA-1 when applying the 2006 Addenda and later editions 
of the ASME B&PV Code, Section III, up to the 2008 Addenda. Previously 
paragraph (b)(1)(iv) permitted the use of NQA-1 up to the 1992 Edition.
10 CFR 50.55a(b)(1)(vii) Capacity Certification and Demonstration of 
Function of Incompressible-Fluid Pressure-Relief Valves
    In the 2006 Addenda, new requirements were added to the ASME Code, 
Section III, that have a parallel structure in paragraphs NB-7742, NC-
7742, and ND-7742 for Class 1, 2, and 3 incompressible-fluid, pressure 
relief valves, respectively. These new paragraphs address new valve 
designs having a range of possible sizes and set-pressure conditions. 
The method described in these paragraphs for performing the tests and 
evaluation data involves performing tests at less than the highest 
value of the set-pressure range and establishing an incompressible 
fluid flow coefficient of discharge that then allows extrapolation of 
capacities to higher pressures. These new paragraphs address 
circumstances in which a certified test facility lacks the fluid-
pressure capability at the necessary flow rate for testing a new, 
incompressible-fluid, pressure-relief valve design. Due to a printing 
error in the ASME Code for paragraph NB-7742(a)(2), some words were 
omitted. The NRC is amending paragraph (b)(1)(vii) to add a condition 
to allow use of NB-7742(a)(2) when the language intended to be included 
in the Code is used.

[[Page 36263]]

ASME B&PV Code, Section XI

10 CFR 50.55a(b)(2)
    The NRC is amending the introductory text to paragraph (b)(2) to 
incorporate by reference only Subsections IWA, IWB, IWC, IWD, IWE, IWF, 
IWL, Mandatory and Non-Mandatory Appendices, of the 2005 Addenda 
through 2008 Addenda of Section XI of the ASME B&PV Code, with 
conditions, into 10 CFR 50.55a. It is also amended to make clear that 
references to Section XI are to Section XI of the ASME B&PV Code.
10 CFR 50.55a(b)(2)(i)
    The NRC is deleting the requirements of paragraph (b)(2)(i), which 
address limitations on specific editions and addenda, and is 
designating the paragraph as ``Reserved.'' Licensees are no longer 
using these older editions (1974 and 1977 Editions) and addenda of the 
ASME B&PV Code.
10 CFR 50.55a(b)(2)(iii)
    The NRC is deleting the requirements of paragraph (b)(2)(iii), 
which address steam generator tubing, and is designating this paragraph 
as ``Reserved.''
10 CFR 50.55a(b)(2)(iv)
    The NRC is deleting the requirements of paragraph (b)(2)(iv), which 
address pressure retaining welds in ASME Code Class 2 piping, and is 
designating this paragraph as ``Reserved.''
10 CFR 50.55a(b)(2)(v)
    The NRC is deleting the requirements of paragraph (b)(2)(v), which 
address the evaluation procedures and acceptance criteria for 
austenitic piping when applying the Winter 1983 Addenda and the Winter 
1984 Addenda of Section XI, and is designating this paragraph as 
``Reserved.''
10 CFR 50.55a(b)(2)(vi)
    This paragraph addresses the pertinent editions and addenda of the 
ASME B&PV Code for which licensees must utilize when implementing the 
initial inservice inspection requirements for containment structures. 
The NRC is amending paragraph (b)(2)(vi) to clarify that, in accordance 
with the paragraph, licensees may use either the 1992 Edition with the 
1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection 
IWE and Subsection IWL of the ASME B&PV Code, Section XI, for the 
initial 120-month inspection interval, subject to the conditions in 
paragraphs (b)(2)(viii) and (b)(2)(ix), including the new condition 
identified in paragraph (b)(2)(ix)(J). Following the initial 120-month 
inspection interval, successive 120-month inspection interval updates 
must be implemented in accordance with the provisions of paragraph 
(g)(4)(ii).
10 CFR 50.55a(b)(2)(viii)
    This paragraph, which addresses the inservice examination of 
concrete containments in accordance with Subsection IWL of the ASME 
B&PV Code, Section XI, is amended so that the conditions in paragraphs 
(b)(2)(viii)(F) and (b)(2)(viii)(G) do not apply when using the 2007 
Edition to the latest edition and addenda incorporated by reference 
into Sec.  50.55a (currently the 2008 Addenda of the ASME B&PV Code).
10 CFR 50.55a(b)(2)(ix)
    This paragraph addresses the examination of metal containments and 
the liners of concrete containments in accordance with Subsection IWE 
of the ASME B&PV Code, Section XI. The NRC is dividing the existing 
condition in paragraph (b)(2)(ix)(A) into paragraphs (b)(2)(ix)(A)(1) 
and (b)(2)(ix)(A)(2). The NRC is also amending the introductory text of 
this paragraph so that the conditions in paragraphs (b)(2)(ix)(F), 
(b)(2)(ix)(G), (b)(2)(ix)(H) and (b)(2)(ix)(I) do not apply when using 
the 2004 Edition with 2006 Addenda through the 2007 Edition with 2008 
Addenda of Subsection IWE of the ASME B&PV Code, Section XI. Also, the 
NRC is amending the introductory text of this paragraph so that the 
condition in paragraph (b)(2)(ix)(I) does not apply when using the 2004 
Edition, up to and including the 2005 Addenda of Subsection IWE of the 
ASME B&PV Code, Section XI.
10 CFR 50.55a(b)(2)(ix)(J)
    The NRC is amending paragraph (b)(2)(ix) to add a new paragraph 
(b)(2)(ix)(J) to address pressure testing requirements following major 
modifications of Class MC containment structures when applying Article 
IWE-5000, of Subsection IWE of the 2007 Edition to the latest edition 
and addenda incorporated by reference into Sec.  50.55a (currently the 
2008 Addenda of the ASME B&PV Code, Section XI).
10 CFR 50.55a(b)(2)(xv)
    The NRC is amending the requirements in paragraph (b)(2)(xv), which 
address Appendix VIII specimen set and qualification requirements, by 
limiting the use of the provisions described in paragraphs 
(b)(2)(xv)(A) through (b)(2)(xv)(M) to licensees using the B&PV Code 
2001 Edition and earlier editions and addenda. Additionally, paragraph 
(b)(2)(xv)(A)(2) is amended to allow a qualification for austenitic 
welds with no austenitic base metal side to be added on to an existing 
Supplement 10 qualification.
10 CFR 50.55a(b)(2)(xvi)
    The NRC is amending the requirements in paragraph (b)(2)(xvi), 
which address Appendix VIII single-sided ferritic-vessel and piping and 
stainless steel piping examination, to limit the condition to those 
licensees using the editions and addenda of ASME Section XI prior to 
the 2007 Edition on Section VIII.
10 CFR 50.55a(b)(2)(xviii)(B)
    The NRC is amending paragraph (b)(2)(xviii)(B), which addresses 
certification of NDE personnel that observe leakage during system 
leakage and hydrostatic testing, such that the condition would only 
apply to editions and addenda prior to the 2007 Edition of Section XI.
10 CFR 50.55a(b)(2)(xviii)(C)
    The NRC is amending paragraph (b)(2)(xviii)(C), which addresses 
certification of NDE personnel, such that the current conditions on the 
qualification of VT-3 examination personnel requiring initial 
qualification examinations and subsequent examinations on a 3-year 
interval would only apply to the editions and addenda prior to the 2005 
Addenda of Section XI.
10 CFR 50.55a(b)(2)(xix)
    The NRC is amending paragraph (b)(2)(xix), which addresses 
substitution of alternative methods, so the current conditions for the 
substitution of alternative examination methods in that paragraph would 
not apply when using the 2005 Addenda through the 2008 Addenda. The 
paragraph is also amended to impose the condition that paragraphs IWA-
4520(b)(2) and IWA-4521 of the 2007 Edition of Section XI, Division 1, 
with 2008 Addenda, are not approved for use.
10 CFR 50.55a(b)(2)(xxi)
    The NRC is deleting the requirements of paragraph (b)(2)(xxi)(B), 
which addressed examination requirements for Examination Category B-G-
2, Item B7.80 bolting, and designating it as ``Reserved.'' This 
condition was inconsistent with the NRC's unconditional approval of 
Code Case N-652-1, ``Alternative Requirements to Categories B-G-1, B-G-
2, and C-D Bolting Examination Methods and Selection Criteria'' in RG 
1.147, Revision 15.

[[Page 36264]]

10 CFR 50.55a(b)(2)(xxiv)
    The NRC is amending the requirements in paragraph (b)(2)(xxiv), 
which addresses incorporation of the performance demonstration 
initiative and addition of ultrasonic examination criteria, so that the 
current condition would not apply when using the 2007 Edition through 
the 2008 Addenda of Section XI of the ASME B&PV Code.
10 CFR 50.55a(b)(2)(xxvii)
    The NRC is amending the requirements in paragraph (b)(2)(xxvii), 
which address removal of insulation, to add a condition to refer to 
paragraph IWA-5241 instead of IWA-5242 for the 2007 Edition and later 
addenda of Section XI of the ASME B&PV Code.
10 CFR 50.55a(b)(2)(xxviii)
    The NRC is adding a new paragraph (b)(2)(xxviii), Analysis of 
flaws, which conditions the use of the fatigue crack growth rate 
calculation for subsurface flaws defined in paragraph A-4300(b)(1) of 
Section XI, Nonmandatory Appendix A when the ratio of the minimum 
cyclic stress to the maximum cyclic stress (R) is less than zero.
10 CFR 50.55a(b)(2)(xxix)
    The NRC is adding a new paragraph (b)(2)(xxix), which conditions 
the use of ASME B&PV Code, Section XI, Non-Mandatory Appendix R, to 
require licensees to submit an alternative in accordance with paragraph 
(a)(3) and obtain NRC authorization of the proposed alternative prior 
to implementing Appendix R, RI-ISI programs.

ASME OM Code

10 CFR 50.55a(b)(3)
    The NRC is amending the introductory text of paragraph (b)(3) to 
require that the 2004 Edition with the 2005 and 2006 Addenda of the 
ASME OM Code be used during the initial 120-month IST interval under 
paragraph (f)(4)(i) and during mandatory 120-month IST program updates 
under paragraph (f)(4)(ii). The amendment also allows users to 
voluntarily update their IST programs to the 2004 Edition with the 2005 
and 2006 Addenda of the ASME OM Code under paragraph (f)(4)(iv).
10 CFR 50.55a(b)(3)(v)
    The NRC is amending paragraph (b)(3)(v) to require that the 
provisions in Subsection ISTD of the ASME OM Code be used for the 
inservice examination and testing of snubbers when using the 2006 
Addenda and later editions and addenda of Section XI.
10 CFR 50.55a(b)(3)(vi)
    The NRC is amending paragraph (b)(3)(vi) to require that the 
current condition for exercising manual valves continue to apply when 
using the 1999 through 2005 Addenda of the ASME OM Code. This condition 
does not apply to the 2006 Addenda and later editions and addenda of 
the ASME OM Code.

Reactor Coolant Pressure Boundary, Quality Group B Components and 
Quality Group C Components

    The NRC is amending paragraphs (c)(3), (d)(2), and (e)(2) to 
replace ``but--'' with ``subject to the following conditions'' at the 
end of the introductory text to the paragraphs for clarity.

Inservice Testing Requirements

10 CFR 50.55a(f)(5)(iv)
    The NRC is amending paragraph (f)(5)(iv) to clarify that licensees 
are required to submit requests for relief based on impracticality 
within 12 months after the expiration of the IST interval for which 
relief is being sought.

Inservice Inspection Requirements

10 CFR 50.55a(g)(2), (g)(3)(i), (g)(3)(ii), and the Introductory Text 
of (g)(4)
    The NRC is amending paragraphs (g)(2), (g)(3)(i), and (g)(3)(ii) to 
require that the provisions in the ASME OM Code, and the optional ASME 
code cases listed in RG 1.192, be used for the examination and testing 
of snubbers. The NRC is amending the introductory text of paragraph 
(g)(4) to require that licensees use the provisions in the ASME OM Code 
for the examination and testing of snubbers.
10 CFR 50.55a(g)(4)(i)
    The NRC is amending paragraph (g)(4)(i) to require that the 
optional code cases listed in RG 1.192 be followed when using the ASME 
OM Code. The NRC is also correcting an earlier error which deleted rule 
language in this paragraph which is applicable to combined licenses 
under 10 CFR part 52. The restored rule language makes clear that, for 
combined license holders under 10 CFR part 52, the inservice 
examinations for the initial 120-month inspection interval must comply 
with the inservice examination requirements in the latest edition and 
addenda of the Code approved by the NRC in Sec.  50.55a on the date 12 
months before the date scheduled for initial loading of fuel under a 
combined license under 10 CFR part 52, except as allowed--as with 
operating licenses under 10 CFR part 50--under the remainder of 
paragraph (g)(4)(i).
10 CFR 50.55a(g)(4)(ii)
    The NRC is amending paragraph (g)(4)(ii) to allow the optional code 
cases listed in RG 1.192 to be followed when using the ASME OM Code. 
Paragraph (g)(4)(ii) is also amended to provide up to a 6-month delay 
in the implementation of the 2007 Edition and 2008 Addenda provisions 
of Appendix VIII for those licensees whose next inspection interval 
must be implemented in the period between 12 through 18 months after 
the effective date of the final rule. Other licensees, whose next 
inservice inspection interval commences more than 18 months after the 
final date of the rule, are not affected by this provision of the final 
rule.
10 CFR 50.55a(g)(4)(iii)
    The NRC is amending paragraph (g)(4)(iii) to provide the proper 
references to Section XI, Table IWB-2500-1, ``Examination Category B-
J,'' Item Numbers B9.20, B9.21 and B9.22, and to limit the condition's 
applicability to the editions and addenda prior to the 2003 Addenda of 
Section XI.
10 CFR 50.55a(g)(5)(iii)
    The NRC is amending paragraph (g)(5)(iii) by adding a sentence to 
clarify that a request for relief must be submitted to the NRC only 
after an examination has been attempted during a given ISI interval and 
the ASME Code requirement determined to be impractical. These requests 
for relief describing the determinations that the code requirement is 
impractical must be submitted to the NRC no later than 12 months after 
the expiration of the initial or subsequent 120-month inspection 
interval for which relief is sought.
10 CFR 55a(g)(5)(iv)
    The NRC is amending paragraph (g)(5)(iv) to clarify that licensees 
are required to submit requests for relief based on impracticality no 
later than 12 months after the end of the ISI interval for which relief 
is being sought.
10 CFR 50.55a(g)(6)(ii)(E)(1) Through (g)(6)(ii)(E)(3)
    The NRC is amending paragraphs (g)(6)(ii)(E)(1) through 
(g)(6)(ii)(E)(3) by changing the requirement to implement Code Case N-
722 to a requirement to implement Code Case N-722-1.
10 CFR 50.55a(g)(6)(ii)(F)
    The final rule incorporates ASME Code Case N-770-1 by reference in 
paragraph (g)(6)(ii)(F)(1). The NRC is not including the following 
proposed

[[Page 36265]]

conditions in this final rule, since they are addressed in Code Case N-
770-1: paragraphs (g)(6)(ii)(F)(5), (6), (8), (9), (10), (11), (13), 
and (14). The NRC is not including part of the proposed condition in 
paragraph (g)(6)(ii)(F)(7), since the part is addressed in Code Case N-
770-1. Because the NRC did not include these proposed conditions in the 
final rule, the numbering of the conditions in the final rule differs 
from that of the proposed rule.
    Paragraph (g)(6)(ii)(F)(2) pertains to obtaining NRC approval prior 
to reclassification of welds under the Inspection Items of Code Case N-
770. All mitigation techniques discussed in Code Case N-770, with the 
exception of Mechanical Stress Improvement Process, are covered by 
separate ASME Code Cases. These Code Cases are subject to approval by 
the NRC. As ASME completes these mitigation Code Cases, the NRC will 
review and approve them, if appropriate, possibly with conditions. The 
NRC uses RG 1.147, which is incorporated by reference in 10 CFR 50.55a, 
to endorse approved Code Cases for generic use. Based on the wording of 
paragraph (g)(6)(ii)(F)(2), as the NRC endorses mitigation Code Cases 
in the RG, the rule permits licensees to categorize mitigated welds in 
the corresponding Inspection Items in Code Case N-770-1, without a 
separate NRC review of the classification or reclassification. This 
condition is unchanged from the proposed rule.
    Paragraph (g)(6)(ii)(F)(3) pertains to the schedule for completing 
baseline examinations. The final rule extends the timing for completing 
baseline examinations. Previous examinations of these welds can be 
credited for baseline examinations if they were performed using Section 
XI, Appendix VIII requirements and met the Code required examination 
volume for axial and circumferential flaws of essentially 100 percent. 
For butt welds that received a MRP-139 examination that did not fully 
meet Section XI, Appendix VIII requirements or achieve essentially 100 
percent coverage, licensees can re-perform the baseline examination to 
meet these requirements or obtain NRC authorization of alternative 
examination requirements in accordance with 10 CFR 50.55a(a)(3)(i) or 
(ii) by the end of next refueling outage that occurs after six months 
from the effective date of the final rule. A licensee may choose to use 
previous inspections of dissimilar metal butt welds performed under the 
plant's ASME Code, Section XI, Inservice Inspection program to meet the 
paragraph (g)(6)(ii)(F)(3) baseline requirement. This is acceptable 
provided the previous inspection falls within the re-inspection period 
for welds in ASME Code Case N-770-1, Table 1, Inspection Items A-1, A-
2, and B. Additionally, the NRC-approved alternative examination 
coverage for these welds during the current 10-year inservice 
inspection interval remain applicable. In all of these cases the 
previously approved alternative will continue to apply for the duration 
authorized by the NRC. In the final rule the NRC modified the proposed 
condition to extend the timing for completing baseline examinations and 
to address credit for previous baseline examinations.
    Paragraph (g)(6)(ii)(F)(4) pertains to the requirement for 
satisfying axial examination coverage of welds. The discussion for 
paragraph (g)(6)(ii)(F)(4) contains guidance on satisfying the axial 
examination coverage requirement during previous baseline examinations. 
This condition is unchanged from the proposed rule.
    Paragraph (g)(6)(ii)(F)(5) requires that all hot-leg temperature 
welds in the Code Case N-770-1 Inspection Items G, H, J and K for 
inlays and onlays be inspected each interval and specifies requirements 
for sample inspection of cold leg temperature welds in these Inspection 
Items. This condition prohibits sample inspection of hot leg 
temperature welds in Inspection Items G, H, J, and K. This condition 
was part of paragraph (g)(6)(ii)(F)(7) of the proposed rule. This part 
of the condition is unchanged from the proposed rule.
    Paragraph (g)(6)(ii)(F)(6) pertains to submitting reports to the 
NRC for mitigated welds whose volumetric examination detects new flaws 
or growth of existing flaws in the required examination volume. This 
condition was included in paragraph (g)(6)(ii)(F)(12) of the proposed 
rule. This condition is unchanged from the proposed rule.
    Paragraph (g)(6)(ii)(F)(7) requires that the thickness of the inlay 
or onlay be used as the thickness ``t'' when applying the acceptance 
standards in ASME Section XI, IWB-3514, for planar flaws contained 
within the inlay or onlay in Inspection Items G, H, J, and K. This 
condition was included in paragraph (g)(6)(ii)(F)(15) of the proposed 
rule. In the final rule paragraph (g)(6)(ii)(F)(7) is expanded to 
clarify that for planar flaws in the balance of the dissimilar metal 
weld examination volume, the thickness ``t'' in IWB-3514 is the 
combined thickness of the inlay or onlay and the dissimilar metal weld.
    Paragraph (g)(6)(ii)(F)(8) prohibits sample inspection of welds 
mitigated by optimized weld overlays in Inspection Items D and E. This 
condition was included in paragraph (g)(6)(ii)(F)(16) of the proposed 
rule. This condition is unchanged from the proposed rule.
    Paragraph (g)(6)(ii)(F)(9) is a new condition as a result of public 
comments. This condition removes the requirement of Code Case N-770-1 
to spread the initial examinations of the Inspection Item D welds 
mitigated in the same inspection period throughout years 3 through 10 
following application of stress improvement. For the extent and 
frequency of examination in Table 1, the condition requires that the 
initial examination for all Inspection Item D welds shall be performed 
no sooner than the third refueling outage and no later than 10 years 
following stress improvement application. The condition addresses 
deferral of the examinations to the end of the interval by repeating 
the previous requirement, that is, to perform the initial examination 
of Inspection Item D welds no sooner than the third refueling outage 
and no later than 10 years following stress improvement application.
    Paragraph (g)(6)(ii)(F)(10) is a new condition as a result of 
incorporating Code Case N-770-1 in lieu of Code Case N-770. Note 2 of 
Figure 5(a) in Code Case N-770-1 permits the use of an alternative 
examination volume for an alternative examination volume for welds 
mitigated by optimized weld overlays. This alternative examination 
volume was not issued as part of the proposed rule and, therefore, this 
condition in the final rule prohibits the use of the alternative 
examination volume. While the NRC does not have a technical objection 
to Note 2 of Figure 5(a), licensees must obtain NRC authorization to 
use the alternative examination volume pursuant to 10 CFR 
50.55a(a)(3)(i) or (ii).
10 CFR 50.55a(g)(6)(ii)(E)(1) Through (g)(6)(ii)(E)(3)
    The NRC is amending paragraphs (g)(6)(ii)(E)(1) through 
(g)(6)(ii)(E)(3) to update the requirement to implement Code Case N-
722-1. The amendment also clarifies that for inspections conducted once 
per interval, the portion of welds to be inspected in the remaining 
portion of the interval is based on rules already established by the 
ASME B&PV Code.
Footnote 1 to 10 CFR 50.55a(g)(6)(ii)(E)
    The NRC is amending footnote 1 to paragraph (g)(6)(ii)(E) to 
clarify that for inspections conducted once per interval, the portion 
of welds to be inspected in the remaining portion of the interval be 
based on rules already

[[Page 36266]]

established by the ASME B&PV Code, Section XI, paragraph IWB-2400.

Substitution of the Term ``Condition'' in 10 CFR 50.55a

    The NRC is amending 10 CFR 50.55a to substitute the words 
``limitation(s),'' ``modification(s),'' and ``provision(s)'' with the 
word ``condition(s)'' throughout the regulations for consistency.

V. Generic Aging Lessons Learned Report

    In December 2010, the NRC issued ``Generic Aging Lessons Learned 
(GALL) Report,'' NUREG-1801, Revision 2, for applicants to use in 
preparing their license renewal applications. The GALL Report evaluates 
existing programs and documents the bases for determining when existing 
programs, without change or augmentation, are adequate for aging 
management in accordance with the license renewal rule, as given in 10 
CFR 54.21(a)(3). In Revision 2 of the GALL Report, editions of the ASME 
B&PV Code, Section XI, Subsections IWB, IWC, IWD, IWE, IWF, and IWL 
from the 1995 Edition through the 2004 Edition were evaluated and were 
found to be acceptable editions and addenda for complying with the 
requirements of 10 CFR 54.21(a)(3), unless specifically noted in 
certain sections of the GALL Report. For example, GALL Report Section 
XI.S1, ``ASME Section XI, Subsection IWE,'' specifically addresses the 
1992 Edition of ASME B&PV Code, Section XI, Subsection IWE.
    In the GALL Report, Section XI.M1, ``ASME Section XI Inservice 
Inspection, Subsections IWB, IWC, and IWD;'' Section XI.S1, ``ASME 
Section XI, Subsection IWE;'' Section XI.S2, ``ASME Section XI, 
Subsection IWL;'' and Section XI.S3, ``ASME Section XI, Subsection 
IWF'' describe the evaluation and technical bases for determining the 
adequacy of these ASME Code subsections. In addition, many other aging 
management programs (AMPs) in the GALL Report rely in part, but to a 
lesser degree, on the requirements in the ASME B&PV Code, Section XI.
    The NRC has evaluated Subsections IWB, IWC, IWD, IWE, IWF, and IWL 
of Section XI of the ASME B&PV Code, 2004 Edition with the 2005 and 
2006 Addenda through the 2007 Edition with the 2008 Addenda as part of 
the Sec.  50.55a amendment process to determine if the conclusions of 
the GALL Report also apply to AMPs that rely upon the ASME B&PV Code 
editions and addenda that are incorporated by reference into Sec.  
50.55a by this rule. The NRC finds that the 2004 Edition, inclusive of 
the 2005 and 2006 Addenda, and the 2007 Edition, inclusive of the 2008 
Addenda of Section XI of the ASME B&PV Code, Subsections IWB, IWC, IWD, 
IWE, IWF, and IWL, as subject to the conditions of this rule, are 
acceptable to be adopted as AMPs for license renewal and the 
conclusions of the GALL Report remain valid, except where specifically 
noted and augmented in the GALL Report. Accordingly, an applicant for 
license renewal may use, in its plant-specific license renewal 
application, Subsections IWB, IWC, IWD, IWE, IWF, and IWL of Section XI 
of the 2004 Edition with the 2005 and 2006 Addenda through the 2007 
Edition with the 2008 Addenda of the ASME B&PV Code, subject to 
conditions in this rule, as acceptable alternatives to the requirements 
of the 1995 Edition through the 2004 Edition of the ASME B&PV Code, 
Section XI, as referenced in Revision 2 of the GALL Report. Similarly, 
a licensee approved for license renewal that relied on the GALL AMPs 
may use Subsections IWB, IWC, IWD, IWE, IWF, and IWL of Section XI of 
the 2004 Edition inclusive of the 2005 and the 2006 Addenda through the 
2007 Edition with the 2008 Addenda of the ASME B&PV Code as acceptable 
alternatives to the AMPs described in the Revision 2 of the GALL 
report. However, a licensee must assess and follow applicable NRC 
requirements with regard to changes to its licensing basis.
    The NRC, however, notes that the GALL Report includes Subsection 
IWE AMP that is evaluated based on the requirements in the 1992 Edition 
through 2004 Edition of Section XI of the ASME B&PV Code. Also, some of 
the terminology used and some details in this AMP is based on the 1992 
Edition. Since this AMP in Revision 2 of the GALL report has a specific 
ASME B&PV Code year in the description of the AMP or in one or more of 
the ten elements, the details in the AMP based on a specific ASME B&PV 
Code edition may not be accurate for other editions.
    Revision 2 of the GALL Report includes AMPs that are based on the 
requirements in the 1995 Edition through the 2004 Edition of Section XI 
of the ASME B&PV Code but in which the AMPs may recommend additional 
augmentation of the Code requirements or the use of specific Code 
Edition or Addenda in order to achieve adequate aging management for 
license renewal. The technical or regulatory aspects of the AMPs, for 
which augmentation is recommended, also apply if using the 2004 Edition 
inclusive of the 2005 Addenda, or the 2007 Edition, inclusive of the 
2008 Addenda, of Section XI of the ASME B&PV Code to meet the 
requirements of 10 CFR 54.21(a)(3). A license renewal applicant may 
either augment its AMPs in these areas, as described in the GALL 
report, or propose alternatives (exceptions) for the NRC to review as 
part of a plant-specific program element justification for its AMP.GALL 
Revision 1, in AMP XI.M11A, provides an acceptable approach for aging 
management--through inservice inspection--of PWR nickel-alloy upper 
vessel head penetration nozzles. This inservice inspection is the same 
as the inservice inspection mandated by Order EA-03-009, ``Issuance of 
Order Establishing Interim Inspection Requirements for Reactor Pressure 
Vessel Heads at Pressurized Water Reactors (PWRs),'' as amended by the 
First Revision of the Order. GALL Revision 2, in GALL AMP XI.M11B, 
``Cracking of Nickel-Alloy Components and Loss of Material Due to Boric 
Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components 
(PWRs Only),'' provides inspection guidance for all PWR nickel-alloy 
reactor coolant pressure boundary (RCPB) components (including nickel-
alloy welds) and nickel alloy aging management review line items. Thus, 
AMP XI.M11B in GALL Revision 2 supersedes the provisions of GALL 
Revision 1 AMP XI.M11A. GALL Revision 2 AMP XI.M11B is based on, and is 
consistent with the provisions of several ASME Code Cases addressing 
inspection of nickel alloy upper vessel head penetration nozzles which 
have been endorsed by the NRC (with conditions in 10 CFR 50.55a). 
Accordingly, new or current license renewal applicants who identify 
consistency with GALL AMP XI.M11B through compliance with 10 CFR 
50.55a(g)(6)(ii)(D), (g)(6)(ii)(E), and (g)(6)(ii)(F) need not take an 
exception to the program elements in GALL AMP XI.M11B. Licensees that 
have been granted a renewed operating license will eventually update 
their ISI programs to comply with the Code Cases on inspection of 
nickel alloy upper vessel head penetration nozzles, in accordance with 
Sec.  50.55a(g). Accordingly, these licensees will eventually become 
consistent with GALL AMP XI.M11B.

VI. Availability of Documents

    The NRC is making the documents identified below available to 
interested persons through one or more of the following:
    Public Document Room (PDR): The NRC PDR is located at 11555 
Rockville Pike, Room O-1F21, Rockville, Maryland 20852.

[[Page 36267]]

    Federal rulemaking Web site: Public comments and supporting 
material related to this final rule can be found at http://regulations.gov by searching on the Docket ID NRC-2008-0554.
    The NRC's Library: The NRC's Library is located at http://www.nrc.gov/reading-rm.html.

------------------------------------------------------------------------
                                       Rulemaking web
           Document              PDR        site            Library
------------------------------------------------------------------------
Analysis of Public Comments..      X   ..............  ML110280240.
ASME B&PV Code *.............      X   ..............
ASME Code Case N-770-1 *.....      X   ..............
ASME Code Case N-722-1 *.....      X   ..............
ASME OM Code *...............      X   ..............
EPRI Report NP-5151 **,        ......  ..............
 ``Evaluation of Reactor
 Vessel Beltline Integrity
 Following Unanticipated
 Operating Events,'' April
 1987.
GALL Report, NUREG-1801,           X   ..............  ML052770419.
 Rev.1, September 2005,.
Volume 1.....................      X   ..............  ML052780376.
Volume 2.....................  ......  ..............
NQA-1 *, ``Quality Assurance
 Requirements for Nuclear
 Facilities,'' 1994 Edition.
NUREG-0800, ``Standard Review      X   ..............  reading-rm/doc-
 Plan for the Review of                                 collections/
 Safety Analysis Reports for                            nuregs/staff/
 Nuclear Power Plants--LWR                              sr0800/.
 Edition.
PNNL-19086, ``Replacement of   ......  ..............  ML1010312543.
 Radiography with Ultrasonics
 for the Nondestructive
 Inspection of Welds--
 Evaluation of Technical
 Gaps--An Interim Report''.
Public Submissions (Comments)  ......              X   ML103200546.
 on Proposed Rule.
Regulatory Analysis and            X               X   ML110320011.
 Backfit Considerations for
 Final Amendment 10 CFR
 50.55a, ``Codes and
 Standards''.
Regulatory Guide 1.178, ``An       X   ..............  ML032510128.
 Approach for Plant-Specific
 Risk-Informed Decisionmaking
 for Inservice Inspection of
 Piping''.
Regulatory Guide 1.193,            X   ..............  ML072470294.
 Revision 2, ``ASME Code
 Cases not Approved for Use''.
Regulatory Guide 1.200, ``An       X   ..............  ML090410014.
 Approach for Determining the
 Technical Adequacy of
 Probabilistic Risk
 Assessment Results for Risk-
 Informed Activities''.
``Review of Changes Between        X   ..............  ML111250292.
 American Society of
 Mechanical Engineers Boiler
 and Pressure Vessel Code
 Cases N-770 and N-770-1 to
 Support 10 CFR 50.55a Final
 Rule''.
Standard Review Plan 3.9.8,        X   ..............  ML032510135.
 ``Risk-Informed Inservice
 Inspection of Piping''
------------------------------------------------------------------------
* Available on the ASME Web site.
** Available on the EPRI Web site.

VII. Voluntary Consensus Standards

    Section 12(d)(3) of the National Technology Transfer and 
Advancement Act of 1995, Public Law 104-113 (NTTAA), and implementing 
guidance in U.S. Office of Management and Budget (OMB) Circular A-119 
(February 10, 1998), requires each Federal government agency (should it 
decide that regulation is necessary) to use a voluntary consensus 
standard instead of developing a government-unique standard. An 
exception to using a voluntary consensus standard is allowed where the 
use of such a standard is inconsistent with applicable law or is 
otherwise impractical. The NTTAA requires Federal agencies to use 
industry consensus standards to the extent practical; it does not 
require Federal agencies to endorse a standard in its entirety. Neither 
the NTTAA nor Circular A-119 prohibit an agency from adopting a 
voluntary consensus standard while taking exception to specific 
portions of the standard, if those provisions are deemed to be 
``inconsistent with applicable law or otherwise impractical.'' 
Furthermore, taking specific exceptions furthers the Congressional 
intent of Federal reliance on voluntary consensus standards because it 
allows the adoption of substantial portions of consensus standards 
without the need to reject the standards in their entirety because of 
limited provisions which are not acceptable to the agency.
    In this rulemaking, the NRC is continuing its existing practice of 
establishing requirements for the design, construction, operation, ISI 
(examination) and IST of nuclear power plants by approving the use of 
the latest editions and addenda of the ASME Codes in 10 CFR 50.55a. The 
ASME Codes are voluntary consensus standards, developed by participants 
with broad and varied interests, in which all interested parties 
(including the NRC and licensees of nuclear power plants) participate. 
Therefore, the NRC's incorporation by reference of the ASME Codes is 
consistent with the overall objectives of the NTTAA and OMB Circular A-
119.
    As discussed in Section III of this statement of considerations, in 
this final rule the NRC is conditioning the use of certain provisions 
of the 2005 Addenda through 2008 Addenda of Section III, Division 1, 
and the 2005 Addenda through 2008 Addenda of Section XI, Division 1, of 
the ASME B&PV Code; and the 2005 Addenda and 2006 Addenda of the ASME 
OM Code, and Code Cases N-722-1 and N-770-1. In addition, the final 
rule does not adopt (``excludes'') certain provisions of the ASME Codes 
and this statement of considerations, and in the regulatory and backfit 
analysis for this rulemaking. The NRC believes that this final rule 
complies with the NTTAA and OMB Circular A-119 despite these conditions 
and ``exclusions.''
    If the NRC did not conditionally accept ASME editions, addenda, and 
code cases, the NRC would disapprove these entirely. The effect would 
be that licensees and applicants would submit a larger number of 
requests for use of alternatives under Sec.  50.55a(a)(3), requests for 
relief under Sec.  50.55a(f) and (g), or requests for exemptions under 
10 CFR 50.12 and/or 10 CFR 52.7. These requests would likely include 
broad-scope requests for approval to issue the full scope of the ASME 
Code editions and addenda which would otherwise be approved in this 
final rulemaking (i.e., the request would not be simply for approval of 
a specific ASME Code provision with conditions). These requests would 
be an unnecessary additional burden for both the licensee

[[Page 36268]]

and the NRC, inasmuch as the NRC has already determined that the ASME 
Codes and Code Cases which are the subject of this final rulemaking are 
acceptable for use (in some cases with conditions). For these reasons, 
the NRC concludes that this final rule's treatment of ASME Code 
editions and addenda, and code cases and any conditions placed on them 
does not conflict with any policy on agency use of consensus standards 
specified in OMB Circular A 119.
    The NRC did not identify any other voluntary consensus standards, 
developed by US voluntary consensus standards bodies for use within the 
US, which the NRC could incorporate by reference instead of the ASME 
Codes. The NRC also did not identify any voluntary consensus standards, 
developed by multinational voluntary consensus standards bodies for use 
on a multinational basis, which the NRC could incorporate by reference 
instead of the ASME Codes. The NRC identified codes addressing the same 
subject as the ASME Codes for use in individual countries. At least one 
country, Korea, directly translated the ASME Code for use in that 
country. In other countries (e.g., Japan), ASME Codes were the basis 
for development of the country's codes, but the ASME Codes were 
substantially modified to accommodate that country's regulatory system 
and reactor designs. Finally, there are countries (e.g., the Russian 
Federation) where that country's code was developed without regard to 
the ASME Code. However, some of these codes may not meet the definition 
of a voluntary consensus standard, because they were developed by the 
state rather than a voluntary consensus standards body. NRC evaluation 
of the countries codes to determine whether each code provides a 
comparable or enhanced level of safety when compared against the level 
of safety provided under the ASME Codes would require a significant 
expenditure of agency resources. This expenditure does not seem 
justified, given that substituting another country's code for the US 
voluntary consensus standard does not appear to substantially further 
the apparent underlying objectives of the NTTAA.
    In summary, this final rulemaking satisfies the requirements of the 
Section 12(d)(3) of the NTTAA and Office of Management and Budget (OMB) 
Circular A 119.

VIII. Finding of No Significant Environmental Impact: Environmental 
Assessment

    This final rule action is in accordance with the NRC's policy to 
incorporate by reference in 10 CFR 50.55a new editions and addenda of 
the ASME B&PV and OM Codes to provide updated rules for constructing 
and inspecting components and testing pumps, valves, and dynamic 
restraints (snubbers) in light-water nuclear power plants. ASME Codes 
are national voluntary consensus standards and are required by the 
National Technology Transfer and Advancement Act of 1995, Public Law 
104-113, to be used by government agencies unless the use of such a 
standard is inconsistent with applicable law or otherwise impractical. 
The National Environmental Policy Act (NEPA) requires Federal 
government agencies to study the impacts of their ``major Federal 
actions significantly affecting the quality of the human environment,'' 
and prepare detailed statements on the environmental impacts of the 
proposed action and alternatives to the proposed action (42 U.S.C. Sec. 
4332(C); NEPA Sec. 102(C)).
    The NRC has determined under NEPA, as amended, and the NRC's 
regulations in Subpart A of 10 CFR part 51, that this final rule is not 
a major Federal action significantly affecting the quality of the human 
environment and, therefore, an environmental impact statement is not 
required. The final rulemaking does not significantly increase the 
probability or consequences of accidents; no changes are being made in 
the types of effluents that may be released off-site; and there is no 
significant increase in public radiation exposure. The NRC estimates 
the radiological dose to plant personnel performing the inspections 
required by Code Case N-770-1 would be about 3 rem per plant over a 10-
year interval, and a one-time exposure for mitigating welds of about 30 
rem per plant. As required by 10 CFR part 20, and in accordance with 
current plant procedures and radiation protection programs, plant 
radiation protection staff will continue monitoring dose rates and 
would make adjustments in shielding, access requirements, 
decontamination methods, and procedures as necessary to minimize the 
dose to workers. The increased occupational dose to individual workers 
stemming from the Code Case N-770-1 inspections must be maintained 
within the limits of 10 CFR part 20 and as low as reasonably 
achievable. Therefore, the NRC concludes that the increase in 
occupational exposure would not be significant. The final rulemaking 
does not involve non-radiological plant effluents and has no other 
environmental impact. Therefore, no significant non-radiological 
impacts are associated with this action. The determination of this 
final environmental assessment is that there will be no significant 
off-site impact to the public from this action.

IX. Paperwork Reduction Act Statement

    This final rule decreases the overall burden on licensees by 
reducing the number of relief requests licensees would have to submit 
to the NRC under 10 CFR 50.55a(f)(5) and 10 CFR 50.55a(g)(5), but adds 
burden for 69 Pressurized Water Reactors (PWRs) to revise procedures 
and programs related to ASME Code Case N-770-1. The public burden 
reduction for these information collections is estimated to average -4 
hours per response. Because the burden for this information collection 
is insignificant, Office of Management and Budget (OMB) clearance is 
not required. Existing requirements were approved by the Office of 
Management and Budget, approval number 3150-0011.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
OMB control number.

X. Regulatory Analysis and Backfitting

    The NRC prepared a document, ``Regulatory Analysis and Backfit 
Considerations for Final Amendment 10 CFR 50.55a, ``Codes and 
Standards''''. The document provides the regulatory analysis for this 
final rule. It also addresses backfitting for the final rule and 
provides the basis for the NRC's determination that the final rule does 
not constitute ``backfitting'' as defined in 10 CFR 50.109(a)(4). The 
analysis is available for review as indicated in Section VI, 
``Availability of Documents,'' of this document.

XI. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the 
NRC certifies that this final rule does not impose a significant 
economical impact on a substantial number of small entities. This final 
rule affects only the licensing and operation of commercial nuclear 
power plants. A licensee who is a subsidiary of a large entity does not 
qualify as a small entity. The companies that own these plants are not 
``small entities'' as defined in the Regulatory Flexibility Act or the 
size standards established by the NRC (10 CFR 2.810), as the companies:

[[Page 36269]]

     Provide services that are not engaged in manufacturing, 
and have average gross receipts of more than $6.5 million over their 
last 3 completed fiscal years, and have more than 500 employees;
     Are not governments of a city, county, town, township or 
village;
     Are not school districts or special districts with 
populations of less than 50; and
     Are not small educational institutions.

XII. Congressional Review Act

    In accordance with the Congressional Review Act of 1996, the NRC 
has determined that this action is not a major rule and has verified 
this determination with the Office of Information and Regulatory 
Affairs of the Office of Management and Budget.

List of Subjects in 10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
protection, Incorporation by reference, Intergovernmental relations, 
Nuclear power plants and reactors, Radiation protection, Reactor siting 
criteria, Reporting and recordkeeping requirements.

    For the reasons set forth in the preamble, and under the authority 
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting 
the following amendments to 10 CFR part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

0
1. The authority citation for part 50 continues to read as follows:

    Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); Energy Policy Act 
of 2005, Pub. L. 109-58, 119 Stat. 194 (2005).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42 
U.S.C. 5841), Section 50.10 also issued under secs. 101, 185, 68 
Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-
190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 
50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 
U.S.C. 2138).
    Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 
185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and 
Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853 
(42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 
204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 
50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 
2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 
U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68 
Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under 
sec. 187, 68 Stat. 955 (42 U.S.C. 2237).


0
2. In Sec.  50.55a:
0
a. Revise paragraph (a), the introductory text of paragraphs (b) and 
(b)(1), paragraphs (b)(1)(ii), (b)(1)(iii), and (b)(1)(iv); and add 
paragraph (b)(1)(vii);
0
b. Revise paragraph (b)(2);
0
c. Revise the introductory text of paragraph (b)(3), paragraphs 
(b)(3)(v), (b)(3)(vi), (c)(3), (d)(2), (e)(2), (f)(2), (f)(3)(v), 
(f)(4), (f)(5)(iv), (g)(2), (g)(3), (g)(4), (g)(5)(iii), (g)(5)(iv), 
(g)(6)(ii)(B), (g)(6)(ii)(E)(1), (g)(6)(ii)(E)(2), and 
(g)(6)(ii)(E)(3);
0
d. Add paragraph (g)(6)(ii)(F); and
0
e. Revise footnote 1 to this section that appears after paragraph 
(h)(3).
    The revisions and additions read as follows:


Sec.  50.55a  Codes and standards.

* * * * *
    (a) Quality standards, ASME Codes and IEEE standards, and 
alternatives.
    (1) Structures, systems, and components must be designed, 
fabricated, erected, constructed, tested, and inspected to quality 
standards commensurate with the importance of the safety function to be 
performed.
    (2) Systems and components of boiling and pressurized water-cooled 
nuclear power reactors must meet the requirements of the ASME Boiler 
and Pressure Vessel Code specified in paragraphs (b), (c), (d), (e), 
(f), and (g) of this section. Protection systems of nuclear power 
reactors of all types must meet the requirements specified in paragraph 
(h) of this section.
    (3) Proposed alternatives to the requirements of paragraphs (c), 
(d), (e), (f), (g), and (h) of this section, or portions thereof, may 
be used when authorized by the Director, Office of Nuclear Reactor 
Regulation, or Director, Office of New Reactors, as appropriate. Any 
proposed alternatives must be submitted and authorized prior to 
implementation. The applicant or licensee shall demonstrate that:
    (i) The proposed alternatives would provide an acceptable level of 
quality and safety; or
    (ii) Compliance with the specified requirements of this section 
would result in hardship or unusual difficulty without a compensating 
increase in the level of quality and safety.
    (b) Standards approved for incorporation by reference. Systems and 
components of boiling and pressurized water cooled nuclear power 
reactors must meet the requirements of the following standards 
referenced in paragraphs (b)(1), (b)(2), (b)(3), (b)(4), (b)(5), and 
(b)(6) of this section: The ASME Boiler and Pressure Vessel Code, 
Section III, Division 1 (excluding Non-mandatory Appendices), and 
Section XI, Division 1; the ASME Code for Operation and Maintenance of 
Nuclear Power Plants; NRC Regulatory Guide (RG) 1.84, Revision 35, 
``Design, Fabrication, and Materials Code Case Acceptability, ASME 
Section III'' (July 2010), RG 1.147, Revision 16, ``Inservice 
Inspection Code Case Acceptability, ASME Section XI, Division 1'' (July 
2010), and RG 1.192, ``Operation and Maintenance Code Case 
Acceptability, ASME OM Code'' (March 2003); and the following ASME Code 
Cases, approved with conditions by the NRC: N-722-1, ``Additional 
Examinations for PWR Pressure Retaining Welds in Class 1 Components 
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' 
(ASME Approval Date: January 26, 2009), in accordance with the 
requirements in paragraph (g)(6)(ii)(E) of this section; N-729-1, 
``Alternative Examination Requirements for PWR Reactor Vessel Upper 
Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, 
Section XI, Division 1'' (ASME Approval Date: March 28, 2006), in 
accordance with the requirements in paragraph (g)(6)(ii)(D) of this 
section; and N-770-1, ``Alternative Examination Requirements and 
Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt 
Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material 
With or Without Application of Listed Mitigation Activities, Section 
XI, Division 1,'' (ASME Approval Date: December 25, 2009), in 
accordance with the requirements in paragraph (g)(6)(ii)(F) of this 
section. These standards have been approved for incorporation by 
reference by the Director of the Federal Register pursuant to 5 U.S.C. 
552(a) and 1 CFR part 51. Copies of the ASME Boiler and Pressure Vessel 
Code, the ASME Code for Operation and Maintenance of Nuclear Power 
Plants, ASME Code Case N-722-1, ASME Code Case N-729-1, and ASME Code 
Case N-770-1 may be purchased from the American Society of Mechanical 
Engineers, Three Park Avenue, New York, NY 10016, phone 800-843-2763, 
or through the Web http://www.asme.org/Codes/. Single copies of NRC 
Regulatory Guides 1.84,

[[Page 36270]]

Revision 35; 1.147, Revision 16; and 1.192 may be obtained free of 
charge by writing the Reproduction and Distribution Services Section, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; or by 
fax to 301-415-2289; or by e-mail to [email protected]. 
Copies of the ASME Codes and NRC Regulatory Guides incorporated by 
reference in this section may be inspected at the NRC Technical 
Library, Two White Flint North, 11545 Rockville Pike, Rockville, MD 
20852-2738 or call 301-415-5610, or at the National Archives and 
Records Administration (NARA). For information on the availability of 
this material at NARA, call 202-741-6030, or go to: http://www.archives.gov/federal-register/cfr/ibr-locations.html.
    (1) As used in this section, references to Section III refer to 
Section III of the ASME Boiler and Pressure Vessel Code, and include 
the 1963 Edition through 1973 Winter Addenda, and the 1974 Edition 
(Division 1) through the 2008 Addenda (Division 1), subject to the 
following conditions:
* * * * *
    (ii) Weld leg dimensions. When applying the 1989 Addenda through 
the latest edition and addenda incorporated by reference in paragraph 
(b)(1) of this section, applicants or licensees may not apply 
subparagraphs NB-3683.4(c)(1) and NB-3683.4(c)(2) or Footnote 11 from 
the 1989 Addenda through the 2003 Addenda, or Footnote 13 from the 2004 
Edition through the 2008 Addenda to Figures NC-3673.2(b)-1 and ND-
3673.2(b)-1 for welds with leg size less than 1.09 tn.
    (iii) Seismic design of piping. Applicants or licensees may use 
Subarticles NB-3200, NB-3600, NC-3600, and ND-3600 for seismic design 
of piping, up to and including the 1993 Addenda, subject to the 
condition specified in paragraph (b)(1)(ii) of this section. Applicants 
or licensees may not use these subarticles for seismic design of piping 
in the 1994 Addenda through the 2005 Addenda incorporated by reference 
in paragraph (b)(1) of this section except that Subarticle NB-3200 in 
the 2004 Edition through the 2008 Addenda may be used by applicants and 
licensees subject to the condition in paragraph (b)(1)(iii)(B) of this 
section. Applicants or licensees may use Subarticles NB-3600, NC-3600 
and ND-3600 for the seismic design of piping in the 2006 Addenda 
through the 2008 Addenda subject to the conditions of this paragraph 
corresponding to these subarticles.
    (A) When applying Note (1) of Figure NB-3222-1 for Level B service 
limits, the calculation of Pb stresses must include 
reversing dynamic loads (including inertia earthquake effects) if 
evaluation of these loads is required by NB-3223(b).
    (B) For Class 1 piping, the material and Do/t 
requirements of NB-3656(b) shall be met for all Service Limits when the 
Service Limits include reversing dynamic loads, and the alternative 
rules for reversing dynamic loads are used.
    (iv) Quality assurance. When applying editions and addenda later 
than the 1989 Edition of Section III, the requirements of NQA-1, 
``Quality Assurance Requirements for Nuclear Facilities,'' 1986 Edition 
through the 1994 Edition, are acceptable for use, provided that the 
edition and addenda of NQA-1 specified in NCA-4000 is used in 
conjunction with the administrative, quality, and technical provisions 
contained in the edition and addenda of Section III being used.
* * * * *
    (vii) Capacity certification and demonstration of function of 
incompressible-fluid pressure-relief valves. When applying the 2006 
Addenda through the 2007 Edition up to and including the 2008 Addenda, 
applicants and licensees may use paragraph NB-7742, except that 
paragraph NB-7742(a)(2) may not be used, and for a valve design of a 
single size to be certified over a range of set pressures, the 
demonstration of function tests under paragraph NB-7742 must be 
conducted as prescribed in NB-7732.2 on two valves covering the minimum 
set pressure for the design and the maximum set pressure which can be 
accommodated at the demonstration facility selected for the test.
    (2) As used in this section, references to Section XI refer to 
Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, 
and include the 1970 Edition through the 1976 Winter Addenda, and the 
1977 Edition through the 2007 Edition with the 2008 Addenda, subject to 
the following conditions:
    (i) [Reserved]
    (ii) Pressure-retaining welds in ASME Code Class 1 piping (applies 
to Table IWB-2500 and IWB-2500-1 and Category B-J). If the facility's 
application for a construction permit was docketed prior to July 1, 
1978, the extent of examination for Code Class 1 pipe welds may be 
determined by the requirements of Table IWB-2500 and Table IWB-2600 
Category B-J of Section XI of the ASME B&PV Code in the 1974 Edition 
and addenda through the Summer 1975 Addenda or other requirements the 
NRC may adopt.
    (iii) [Reserved]
    (iv) [Reserved]
    (v) [Reserved]
    (vi) Effective edition and addenda of Subsection IWE and Subsection 
IWL, Section XI. Applicants or licensees may use either the 1992 
Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda 
of Subsection IWE and Subsection IWL as conditioned by the requirements 
in paragraphs (b)(2)(viii) and (b)(2)(ix) of this section when 
implementing the initial 120-month inspection interval for the 
containment inservice inspection requirements of this section. 
Successive 120-month interval updates must be implemented in accordance 
with paragraph (g)(4)(ii) of this section.
    (vii) Section XI References to OM Part 4, OM Part 6 and OM Part 10 
(Table IWA-1600-1). When using Table IWA-1600-1, ``Referenced Standards 
and Specifications,'' in the Section XI, Division 1, 1987 Addenda, 1988 
Addenda, or 1989 Edition, the specified ``Revision Date or Indicator'' 
for ASME/ANSI OM part 4, ASME/ANSI part 6, and ASME/ANSI part 10 must 
be the OMa-1988 Addenda to the OM-1987 Edition. These requirements have 
been incorporated into the OM Code which is incorporated by reference 
in paragraph (b)(3) of this section.
    (viii) Examination of concrete containments. Applicants or 
licensees applying Subsection IWL, 1992 Edition with the 1992 Addenda, 
shall apply paragraphs (b)(2)(viii)(A) through (b)(2)(viii)(E) of this 
section. Applicants or licensees applying Subsection IWL, 1995 Edition 
with the 1996 Addenda, shall apply paragraphs (b)(2)(viii)(A), 
(b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of this section. Applicants or 
licensees applying Subsection IWL, 1998 Edition through the 2000 
Addenda shall apply paragraphs (b)(2)(viii)(E) and (b)(2)(viii)(F) of 
this section. Applicants or licensees applying Subsection IWL, 2001 
Edition through the 2004 Edition, up to and including the 2006 Addenda, 
shall apply paragraphs (b)(2)(viii)(E) through (b)(2)(viii)(G) of this 
section. Applicants or licensees applying Subsection IWL, 2007 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (b)(2) of this section, shall apply paragraph (b)(2)(viii)(E) 
of this section.
    (A) Grease caps that are accessible must be visually examined to 
detect grease leakage or grease cap deformations. Grease caps must be 
removed for this examination when there is evidence of grease cap

[[Page 36271]]

deformation that indicates deterioration of anchorage hardware.
    (B) When evaluation of consecutive surveillances of prestressing 
forces for the same tendon or tendons in a group indicates a trend of 
prestress loss such that the tendon force(s) would be less than the 
minimum design prestress requirements before the next inspection 
interval, an evaluation must be performed and reported in the 
Engineering Evaluation Report as prescribed in IWL-3300.
    (C) When the elongation corresponding to a specific load (adjusted 
for effective wires or strands) during retensioning of tendons differs 
by more than 10 percent from that recorded during the last measurement, 
an evaluation must be performed to determine whether the difference is 
related to wire failures or slip of wires in anchorage. A difference of 
more than 10 percent must be identified in the ISI Summary Report 
required by IWA-6000.
    (D) The applicant or licensee shall report the following 
conditions, if they occur, in the ISI Summary Report required by IWA-
6000:
    (1) The sampled sheathing filler grease contains chemically 
combined water exceeding 10 percent by weight or the presence of free 
water;
    (2) The absolute difference between the amount removed and the 
amount replaced exceeds 10 percent of the tendon net duct volume;
    (3) Grease leakage is detected during general visual examination of 
the containment surface.
    (E) For Class CC applications, the applicant or licensee shall 
evaluate the acceptability of inaccessible areas when conditions exist 
in accessible areas that could indicate the presence of or result in 
degradation to such inaccessible areas. For each inaccessible area 
identified, the applicant or licensee shall provide the following in 
the ISI Summary Report required by IWA-6000:
    (1) A description of the type and estimated extent of degradation, 
and the conditions that led to the degradation;
    (2) An evaluation of each area, and the result of the evaluation, 
and;
    (3) A description of necessary corrective actions.
    (F) Personnel that examine containment concrete surfaces and tendon 
hardware, wires, or strands must meet the qualification provisions in 
IWA-2300. The ``owner-defined'' personnel qualification provisions in 
IWL-2310(d) are not approved for use.
    (G) Corrosion protection material must be restored following 
concrete containment post-tensioning system repair and replacement 
activities in accordance with the quality assurance program 
requirements specified in IWA-1400.
    (ix) Examination of metal containments and the liners of concrete 
containments. Applicants or licensees applying Subsection IWE, 1992 
Edition with the 1992 Addenda, or the 1995 Edition with the 1996 
Addenda, shall satisfy the requirements of paragraphs (b)(2)(ix)(A) 
through (b)(2)(ix)(E) of this section. Applicants or licensees applying 
Subsection IWE, 1998 Edition through the 2001 Edition with the 2003 
Addenda, shall satisfy the requirements of paragraphs (b)(2)(ix)(A), 
(b)(2)(ix)(B), and (b)(2)(ix)(F) through (b)(2)(ix)(I) of this section. 
Applicants or licensees applying Subsection IWE, 2004 Edition, up to 
and including the 2005 Addenda, shall satisfy the requirements of 
paragraphs (b)(2)(ix)(A), (b)(2)(ix)(B), and (b)(2)(ix)(F) through 
(b)(2)(ix)(H) of this section. Applicants or licensees applying 
Subsection IWE, 2004 Edition with the 2006 Addenda, shall satisfy the 
requirements of paragraphs (b)(2)(ix)(A)(2) and (b)(2)(ix)(B) of this 
section. Applicants or licensees applying Subsection IWE, 2007 Edition 
through the latest addenda incorporated by reference in paragraph 
(b)(2) of this section, shall satisfy the requirements of paragraphs 
(b)(2)(ix)(A)(2), (b)(2)(ix)(B) and (b)(2)(ix)(J) of this section.
    (A) For Class MC applications, the following apply to inaccessible 
areas.
    (1) The applicant or licensee shall evaluate the acceptability of 
inaccessible areas when conditions exist in accessible areas that could 
indicate the presence of or result in degradation to such inaccessible 
areas.
    (2) For each inaccessible area identified for evaluation, the 
applicant or licensee shall provide the following in the ISI Summary 
Report as required by IWA-6000:
    (i) A description of the type and estimated extent of degradation, 
and the conditions that led to the degradation;
    (ii) An evaluation of each area, and the result of the evaluation, 
and;
    (iii) A description of necessary corrective actions.
    (B) When performing remotely the visual examinations required by 
Subsection IWE, the maximum direct examination distance specified in 
Table IWA-2210-1 may be extended and the minimum illumination 
requirements specified in Table IWA-2210-1 may be decreased provided 
that the conditions or indications for which the visual examination is 
performed can be detected at the chosen distance and illumination.
    (C) The examinations specified in Examination Category E-B, 
Pressure Retaining Welds, and Examination Category E-F, Pressure 
Retaining Dissimilar Metal Welds, are optional.
    (D) This paragraph (b)(2)(ix)(D) may be used as an alternative to 
the requirements of IWE-2430.
    (1) If the examinations reveal flaws or areas of degradation 
exceeding the acceptance standards of Table IWE-3410-1, an evaluation 
must be performed to determine whether additional component 
examinations are required. For each flaw or area of degradation 
identified which exceeds acceptance standards, the applicant or 
licensee shall provide the following in the ISI Summary Report required 
by IWA-6000:
    (i) A description of each flaw or area, including the extent of 
degradation, and the conditions that led to the degradation;
    (ii) The acceptability of each flaw or area, and the need for 
additional examinations to verify that similar degradation does not 
exist in similar components, and;
    (iii) A description of necessary corrective actions.
    (2) The number and type of additional examinations to ensure 
detection of similar degradation in similar components.
    (E) A general visual examination as required by Subsection IWE must 
be performed once each period.
    (F) VT-1 and VT-3 examinations must be conducted in accordance with 
IWA-2200. Personnel conducting examinations in accordance with the VT-1 
or VT-3 examination method shall be qualified in accordance with IWA-
2300. The ``owner-defined'' personnel qualification provisions in IWE-
2330(a) for personnel that conduct VT-1 and VT-3 examinations are not 
approved for use.
    (G) The VT-3 examination method must be used to conduct the 
examinations in Items E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 
examination method must be used to conduct the examination in Item 
E4.11 of Table IWE-2500-1. An examination of the pressure-retaining 
bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 
examination method must be conducted once each interval. The ``owner-
defined'' visual examination provisions in IWE-2310(a) are not approved 
for use for VT-1 and VT-3 examinations.
    (H) Containment bolted connections that are disassembled during the 
scheduled performance of the examinations in Item E1.11 of Table IWE-
2500-1 must be examined using

[[Page 36272]]

the VT-3 examination method. Flaws or degradation identified during the 
performance of a VT-3 examination must be examined in accordance with 
the VT-1 examination method. The criteria in the material specification 
or IWB-3517.1 must be used to evaluate containment bolting flaws or 
degradation. As an alternative to performing VT-3 examinations of 
containment bolted connections that are disassembled during the 
scheduled performance of Item E1.11, VT-3 examinations of containment 
bolted connections may be conducted whenever containment bolted 
connections are disassembled for any reason.
    (I) The ultrasonic examination acceptance standard specified in 
IWE-3511.3 for Class MC pressure-retaining components must also be 
applied to metallic liners of Class CC pressure-retaining components.
    (J) In general, a repair/replacement activity such as replacing a 
large containment penetration, cutting a large construction opening in 
the containment pressure boundary to replace steam generators, reactor 
vessel heads, pressurizers, or other major equipment; or other similar 
modification is considered a major containment modification. When 
applying IWE-5000 to Class MC pressure-retaining components, any major 
containment modification or repair/replacement, must be followed by a 
Type A test to provide assurance of both containment structural 
integrity and leaktight integrity prior to returning to service, in 
accordance with 10 CFR part 50, Appendix J, Option A or Option B on 
which the applicant's or licensee's Containment Leak-Rate Testing 
Program is based. When applying IWE-5000, if a Type A, B, or C Test is 
performed, the test pressure and acceptance standard for the test must 
be in accordance with 10 CFR part 50, Appendix J.
    (x) Quality assurance. When applying Section XI editions and 
addenda later than the 1989 Edition, the requirements of NQA-1, 
``Quality Assurance Requirements for Nuclear Facilities,'' 1979 Addenda 
through the 1989 Edition, are acceptable as permitted by IWA-1400 of 
Section XI, if the licensee uses its 10 CFR part 50, Appendix B, 
quality assurance program, in conjunction with Section XI requirements. 
Commitments contained in the licensee's quality assurance program 
description that are more stringent than those contained in NQA-1 must 
govern Section XI activities. Further, where NQA-1 and Section XI do 
not address the commitments contained in the licensee's Appendix B 
quality assurance program description, the commitments must be applied 
to Section XI activities.
    (xi) [Reserved]
    (xii) Underwater welding. The provisions in IWA-4660, ``Underwater 
Welding,'' of Section XI, 1997 Addenda through the latest edition and 
addenda incorporated by reference in paragraph (b)(2) of this section, 
are not approved for use on irradiated material.
    (xiii) [Reserved]
    (xiv) Appendix VIII personnel qualification. All personnel 
qualified for performing ultrasonic examinations in accordance with 
Appendix VIII shall receive 8 hours of annual hands-on training on 
specimens that contain cracks. Licensees applying the 1999 Addenda 
through the latest edition and addenda incorporated by reference in 
paragraph (b)(2) of this section may use the annual practice 
requirements in VII-4240 of Appendix VII of Section XI in place of the 
8 hours of annual hands-on training provided that the supplemental 
practice is performed on material or welds that contain cracks, or by 
analyzing prerecorded data from material or welds that contain cracks. 
In either case, training must be completed no earlier than 6 months 
prior to performing ultrasonic examinations at a licensee's facility.
    (xv) Appendix VIII specimen set and qualification requirements. 
Licensees using Appendix VIII in the 1995 Edition through the 2001 
Edition of the ASME Boiler and Pressure Vessel Code may elect to comply 
with all of the provisions in paragraphs (b)(2)(xv)(A) through 
(b)(2)(xv)(M) of this section, except for paragraph (b)(2)(xv)(F) of 
this section, which may be used at the licensee's option. Licensees 
using editions and addenda after 2001 Edition through the 2006 Addenda 
shall use the 2001 Edition of Appendix VIII, and may elect to comply 
with all of the provisions in paragraphs (b)(2)(xv)(A) through 
(b)(2)(xv)(M) of this section, except for paragraph (b)(2)(xv)(F) of 
this section, which may be used at the licensee's option.
    (A) When applying Supplements 2, 3, and 10 to Appendix VIII, the 
following examination coverage criteria requirements must be used:
    (1) Piping must be examined in two axial directions, and when 
examination in the circumferential direction is required, the 
circumferential examination must be performed in two directions, 
provided access is available. Dissimilar metal welds must be examined 
axially and circumferentially.
    (2) Where examination from both sides is not possible, full 
coverage credit may be claimed from a single side for ferritic welds. 
Where examination from both sides is not possible on austenitic welds 
or dissimilar metal welds, full coverage credit from a single side may 
be claimed only after completing a successful single-sided Appendix 
VIII demonstration using flaws on the opposite side of the weld. 
Dissimilar metal weld qualifications must be demonstrated from the 
austenitic side of the weld, and the qualification may be expanded for 
austenitic welds with no austenitic sides using a separate add-on 
performance demonstration. Dissimilar metal welds may be examined from 
either side of the weld.
    (B) The following conditions must be used in addition to the 
requirements of Supplement 4 to Appendix VIII:
    (1) Paragraph 3.1, Detection acceptance criteria--Personnel are 
qualified for detection if the results of the performance demonstration 
satisfy the detection requirements of ASME Section XI, Appendix VIII, 
Table VIII-S4-1 and no flaw greater than 0.25 inch through wall 
dimension is missed.
    (2) Paragraph 1.1(c), Detection test matrix--Flaws smaller than the 
50 percent of allowable flaw size, as defined in IWB-3500, need not be 
included as detection flaws. For procedures applied from the inside 
surface, use the minimum thickness specified in the scope of the 
procedure to calculate a/t. For procedures applied from the outside 
surface, the actual thickness of the test specimen is to be used to 
calculate a/t.
    (C) When applying Supplement 4 to Appendix VIII, the following 
conditions must be used:
    (1) A depth sizing requirement of 0.15 inch RMS must be used in 
lieu of the requirements in Subparagraphs 3.2(a) and 3.2(c), and a 
length sizing requirement of 0.75 inch RMS must be used in lieu of the 
requirement in Subparagraph 3.2(b).
    (2) In lieu of the location acceptance criteria requirements of 
Subparagraph 2.1(b), a flaw will be considered detected when reported 
within 1.0 inch or 10 percent of the metal path to the flaw, whichever 
is greater, of its true location in the X and Y directions.
    (3) In lieu of the flaw type requirements of Subparagraph 
1.1(e)(1), a minimum of 70 percent of the flaws in the detection and 
sizing tests shall be cracks. Notches, if used, must be limited by the 
following:
    (i) Notches must be limited to the case where examinations are 
performed from the clad surface.
    (ii) Notches must be semielliptical with a tip width of less than 
or equal to 0.010 inches.

[[Page 36273]]

    (iii) Notches must be perpendicular to the surface within  2 degrees.
    (4) In lieu of the detection test matrix requirements in paragraphs 
1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain 
a representative distribution of flaw orientations, sizes, and 
locations.
    (D) The following conditions must be used in addition to the 
requirements of Supplement 6 to Appendix VIII:
    (1) Paragraph 3.1, Detection Acceptance Criteria--Personnel are 
qualified for detection if:
    (i) No surface connected flaw greater than 0.25 inch through wall 
has been missed.
    (ii) No embedded flaw greater than 0.50 inch through wall has been 
missed.
    (2) Paragraph 3.1, Detection Acceptance Criteria--For procedure 
qualification, all flaws within the scope of the procedure are 
detected.
    (3) Paragraph 1.1(b) for detection and sizing test flaws and 
locations--Flaws smaller than the 50 percent of allowable flaw size, as 
defined in IWB-3500, need not be included as detection flaws. Flaws 
which are less than the allowable flaw size, as defined in IWB-3500, 
may be used as detection and sizing flaws.
    (4) Notches are not permitted.
    (E) When applying Supplement 6 to Appendix VIII, the following 
conditions must be used:
    (1) A depth sizing requirement of 0.25 inch RMS must be used in 
lieu of the requirements of subparagraphs 3.2(a), 3.2(c)(2), and 
3.2(c)(3).
    (2) In lieu of the location acceptance criteria requirements in 
Subparagraph 2.1(b), a flaw will be considered detected when reported 
within 1.0 inch or 10 percent of the metal path to the flaw, whichever 
is greater, of its true location in the X and Y directions.
    (3) In lieu of the length sizing criteria requirements of 
Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch 
RMS must be used.
    (4) In lieu of the detection specimen requirements in Subparagraph 
1.1(e)(1), a minimum of 55 percent of the flaws must be cracks. The 
remaining flaws may be cracks or fabrication type flaws, such as slag 
and lack of fusion. The use of notches is not allowed.
    (5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test 
matrix, personnel demonstration test sets must contain a representative 
distribution of flaw orientations, sizes, and locations.
    (F) The following conditions may be used for personnel 
qualification for combined Supplement 4 to Appendix VIII and Supplement 
6 to Appendix VIII qualification. Licensees choosing to apply this 
combined qualification shall apply all of the provisions of Supplements 
4 and 6 including the following conditions:
    (1) For detection and sizing, the total number of flaws must be at 
least 10. A minimum of 5 flaws shall be from Supplement 4, and a 
minimum of 50 percent of the flaws must be from Supplement 6. At least 
50 percent of the flaws in any sizing must be cracks. Notches are not 
acceptable for Supplement 6.
    (2) Examination personnel are qualified for detection and length 
sizing when the results of any combined performance demonstration 
satisfy the acceptance criteria of Supplement 4 to Appendix VIII.
    (3) Examination personnel are qualified for depth sizing when 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws 
are sized within the respective acceptance criteria of those 
supplements.
    (G) When applying Supplement 4 to Appendix VIII, Supplement 6 to 
Appendix VIII, or combined Supplement 4 and Supplement 6 qualification, 
the following additional conditions must be used, and examination 
coverage must include:
    (1) The clad to base metal interface, including a minimum of 15 
percent T (measured from the clad to base metal interface), must be 
examined from four orthogonal directions using procedures and personnel 
qualified in accordance with Supplement 4 to Appendix VIII.
    (2) If the clad-to-base-metal-interface procedure demonstrates 
detectability of flaws with a tilt angle relative to the weld 
centerline of at least 45 degrees, the remainder of the examination 
volume is considered fully examined if coverage is obtained in one 
parallel and one perpendicular direction. This must be accomplished 
using a procedure and personnel qualified for single-side examination 
in accordance with Supplement 6. Subsequent examinations of this volume 
may be performed using examination techniques qualified for a tilt 
angle of at least 10 degrees.
    (3) The examination volume not addressed by paragraph 
(b)(2)(xv)(G)(1) of this section is considered fully examined if 
coverage is obtained in one parallel and one perpendicular direction, 
using a procedure and personnel qualified for single sided examination 
when the conditions in paragraph (b)(2)(xv)(G)(2) are met.
    (H) When applying Supplement 5 to Appendix VIII, at least 50 
percent of the flaws in the demonstration test set must be cracks and 
the maximum mis-orientation must be demonstrated with cracks. Flaws in 
nozzles with bore diameters equal to or less than 4 inches may be 
notches.
    (I) When applying Supplement 5, Paragraph (a), to Appendix VIII, 
the number of false calls allowed must be D/10, with a maximum of 3, 
where D is the diameter of the nozzle.
    (J) [Reserved]
    (K) When performing nozzle-to-vessel weld examinations, the 
following conditions must be used when the requirements contained in 
Supplement 7 to Appendix VIII are applied for nozzle-to-vessel welds in 
conjunction with Supplement 4 to Appendix VIII, Supplement 6 to 
Appendix VIII, or combined Supplement 4 and Supplement 6 qualification.
    (1) For examination of nozzle-to-vessel welds conducted from the 
bore, the following conditions are required to qualify the procedures, 
equipment, and personnel:
    (i) For detection, a minimum of four flaws in one or more full-
scale nozzle mock-ups must be added to the test set. The specimens must 
comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for 
flaw locations specified in Table VIII S6-1. Flaws may be notches, 
fabrication flaws or cracks. Seventy-five (75) percent of the flaws 
must be cracks or fabrication flaws. Flaw locations and orientations 
must be selected from the choices shown in paragraph (b)(2)(xi)(K)(4) 
of this section, Table VIII-S7-1--Modified, with the exception that 
flaws in the outer eighty-five (85) percent of the weld need not be 
perpendicular to the weld. There may be no more than two flaws from 
each category, and at least one subsurface flaw must be included.
    (ii) For length sizing, a minimum of four flaws as in paragraph 
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set. 
The length sizing results must be added to the results of combined 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The 
combined results must meet the acceptance standards contained in 
paragraph (b)(2)(xv)(E)(3) of this section.
    (iii) For depth sizing, a minimum of four flaws as in paragraph 
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set. 
Their depths must be distributed over the ranges of Supplement 4, 
Paragraph 1.1, to Appendix VIII, for the inner 15 percent of the wall 
thickness and Supplement 6, Paragraph 1.1, to Appendix VIII, for the 
remainder of the wall thickness. The depth sizing results must be 
combined with the sizing results from Supplement 4 to Appendix VIII for 
the inner 15 percent and to Supplement 6 to Appendix VIII for the 
remainder of the

[[Page 36274]]

wall thickness. The combined results must meet the depth sizing 
acceptance criteria contained in paragraphs (b)(2)(xv)(C)(1), 
(b)(2)(xv)(E)(1), and (b)(2)(xv)(F)(3) of this section.
    (2) For examination of reactor pressure vessel nozzle-to-vessel 
welds conducted from the inside of the vessel,
    (i) The clad to base metal interface and the adjacent examination 
volume to a minimum depth of 15 percent T (measured from the clad to 
base metal interface) must be examined from four orthogonal directions 
using a procedure and personnel qualified in accordance with Supplement 
4 to Appendix VIII as conditioned by paragraphs (b)(2)(xv)(B) and 
(b)(2)(xv)(C) of this section.
    (ii) When the examination volume defined in paragraph 
(b)(2)(xi)(K)(2)(i) of this section cannot be effectively examined in 
all four directions, the examination must be augmented by examination 
from the nozzle bore using a procedure and personnel qualified in 
accordance with paragraph (b)(2)(xi)(K)(1) of this section.
    (iii) The remainder of the examination volume not covered by 
paragraph (b)(2)(xv)(K)(2)(ii) of this section or a combination of 
paragraphs (b)(2)(xv)(K)(2)(i) and (b)(2)(xv)(K)(2)(ii) of this 
section, must be examined from the nozzle bore using a procedure and 
personnel qualified in accordance with paragraph (b)(2)(xv)(K)(1) of 
this section, or from the vessel shell using a procedure and personnel 
qualified for single sided examination in accordance with Supplement 6 
to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through 
(b)(2)(xv)(G) of this section.
    (3) For examination of reactor pressure vessel nozzle-to-shell 
welds conducted from the outside of the vessel,
    (i) The clad to base metal interface and the adjacent metal to a 
depth of 15 percent T, (measured from the clad to base metal interface) 
must be examined from one radial and two opposing circumferential 
directions using a procedure and personnel qualified in accordance with 
Supplement 4 to Appendix VIII, as conditioned by paragraphs 
(b)(2)(xv)(B) and (b)(2)(xv)(C) of this section, for examinations 
performed in the radial direction, and Supplement 5 to Appendix VIII, 
as conditioned by paragraph (b)(2)(xv)(J) of this section, for 
examinations performed in the circumferential direction.
    (ii) The examination volume not addressed by paragraph 
(b)(2)(xv)(K)(3)(i) of this section must be examined in a minimum of 
one radial direction using a procedure and personnel qualified for 
single sided examination in accordance with Supplement 6 to Appendix 
VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (b)(2)(xv)(G) 
of this section.
    (4) Table VIII-S7-1, ``Flaw Locations and Orientations,'' 
Supplement 7 to Appendix VIII, is conditioned as follows:

                        Table VIII-S7-1--Modified
------------------------------------------------------------------------
                     Flaw locations and orientations
-------------------------------------------------------------------------
                                        Parallel  to      Perpendicular
                                            weld             to weld
------------------------------------------------------------------------
Inner 15 percent....................                X                 X
OD Surface..........................                X   ................
Subsurface..........................                X   ................
------------------------------------------------------------------------

    (L) As a condition to the requirements of Supplement 8, 
Subparagraph 1.1(c), to Appendix VIII, notches may be located within 
one diameter of each end of the bolt or stud.
    (M) When implementing Supplement 12 to Appendix VIII, only the 
provisions related to the coordinated implementation of Supplement 3 to 
Supplement 2 performance demonstrations are to be applied.
    (xvi) Appendix VIII single side ferritic vessel and piping and 
stainless steel piping examination. When applying editions and addenda 
prior to the 2007 Edition of Section XI, the following conditions 
apply.
    (A) Examinations performed from one side of a ferritic vessel weld 
must be conducted with equipment, procedures, and personnel that have 
demonstrated proficiency with single side examinations. To demonstrate 
equivalency to two sided examinations, the demonstration must be 
performed to the requirements of Appendix VIII as conditioned by this 
paragraph and paragraphs (b)(2)(xv)(B) through (b)(2)(xv)(G) of this 
section, on specimens containing flaws with non-optimum sound energy 
reflecting characteristics or flaws similar to those in the vessel 
being examined.
    (B) Examinations performed from one side of a ferritic or stainless 
steel pipe weld must be conducted with equipment, procedures, and 
personnel that have demonstrated proficiency with single side 
examinations. To demonstrate equivalency to two sided examinations, the 
demonstration must be performed to the requirements of Appendix VIII as 
conditioned by this paragraph and paragraph (b)(2)(xv)(A) of this 
section.
    (xvii) Reconciliation of quality requirements. When purchasing 
replacement items, in addition to the reconciliation provisions of IWA-
4200, 1995 Addenda through 1998 Edition, the replacement items must be 
purchased, to the extent necessary, in accordance with the licensee's 
quality assurance program description required by 10 CFR 
50.34(b)(6)(ii).
    (xviii) Certification of NDE personnel. (A) Level I and II 
nondestructive examination personnel shall be recertified on a 3-year 
interval in lieu of the 5-year interval specified in the 1997 Addenda 
and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 
1999 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(2) of this section.
    (B) When applying editions and addenda prior to the 2007 Edition of 
Section XI, paragraph IWA-2316 may only be used to qualify personnel 
that observe leakage during system leakage and hydrostatic tests 
conducted in accordance with IWA 5211(a) and (b).
    (C) When applying editions and addenda prior to the 2005 Addenda of 
Section XI, licensee's qualifying visual examination personnel for VT-3 
visual examination under paragraph IWA-2317 of Section XI, must 
demonstrate the proficiency of the training by administering an initial 
qualification examination and administering subsequent examinations on 
a 3-year interval.
    (xix) Substitution of alternative methods. The provisions for 
substituting alternative examination methods, a combination of methods, 
or newly developed techniques in the 1997 Addenda of IWA-2240 must be 
applied

[[Page 36275]]

when using the 1998 Edition through the 2004 Edition of Section XI of 
the ASME B&PV Code. The provisions in IWA-4520(c), 1997 Addenda through 
the 2004 Edition, allowing the substitution of alternative methods, a 
combination of methods, or newly developed techniques for the methods 
specified in the Construction Code are not approved for use. The 
provisions in IWA-4520(b)(2) and IWA-4521 of the 2008 Addenda through 
the latest edition and addenda approved in paragraph (b)(2) of this 
section, allowing the substitution of ultrasonic examination for 
radiographic examination specified in the Construction Code are not 
approved for use.
    (xx) System leakage tests.
    (A) When performing system leakage tests in accordance with IWA-
5213(a), 1997 through 2002 Addenda, the licensee shall maintain a 10-
minute hold time after test pressure has been reached for Class 2 and 
Class 3 components that are not in use during normal operating 
conditions. No hold time is required for the remaining Class 2 and 
Class 3 components provided that the system has been in operation for 
at least 4 hours for insulated components or 10 minutes for uninsulated 
components.
    (B) The NDE provision in IWA-4540(a)(2) of the 2002 Addenda of 
Section XI must be applied when performing system leakage tests after 
repair and replacement activities performed by welding or brazing on a 
pressure retaining boundary using the 2003 Addenda through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section.
    (xxi) Table IWB-2500-1 examination requirements.
    (A) The provisions of Table IWB-2500-1, Examination Category B-D, 
Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 
(Inspection Program A) and Items B3.120 and B3.140 (Inspection Program 
B) of the 1998 Edition must be applied when using the 1999 Addenda 
through the latest edition and addenda incorporated by reference in 
paragraph (b)(2) of this section. A visual examination with 
magnification that has a resolution sensitivity to detect a 1-mil width 
wire or crack, utilizing the allowable flaw length criteria in Table 
IWB-3512-1, 1997 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (b)(2) of this section, with a 
limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be 
performed instead of an ultrasonic examination.
    (B) [Reserved]
    (xxii) Surface examination. The use of the provision in IWA-2220, 
``Surface Examination,'' of Section XI, 2001 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section, that allow use of an ultrasonic examination method is 
prohibited.
    (xxiii) Evaluation of thermally cut surfaces. The use of the 
provisions for eliminating mechanical processing of thermally cut 
surfaces in IWA-4461.4.2 of Section XI, 2001 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section are prohibited.
    (xxiv) Incorporation of the performance demonstration initiative 
and addition of ultrasonic examination criteria. The use of Appendix 
VIII and the supplements to Appendix VIII and Article I-3000 of Section 
XI of the ASME B&PV Code, 2002 Addenda through the 2006 Addenda is 
prohibited.
    (xxv) Mitigation of defects by modification. The use of the 
provisions in IWA-4340, ``Mitigation of Defects by Modification,'' 
Section XI, 2001 Edition through the latest edition and addenda 
incorporated by reference in paragraph (b)(2) of this section are 
prohibited.
    (xxvi) Pressure testing Class 1, 2, and 3 mechanical joints. The 
repair and replacement activity provisions in IWA-4540(c) of the 1998 
Edition of Section XI for pressure testing Class 1, 2, and 3 mechanical 
joints must be applied when using the 2001 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section.
    (xxvii) Removal of insulation. When performing visual examination 
in accordance with IWA-5242 of Section XI of the ASME B&PV Code, 2003 
Addenda through the 2006 Addenda, or IWA-5241 of the 2007 Edition 
through the latest edition and addenda incorporated in paragraph (b)(2) 
of this section, insulation must be removed from 17-4 PH or 410 
stainless steel studs or bolts aged at a temperature below 1100 [deg]F 
or having a Rockwell Method C hardness value above 30, and from A-286 
stainless steel studs or bolts preloaded to 100,000 pounds per square 
inch or higher.
    (xxviii) Analysis of flaws. Licensees using ASME B&PV Code, Section 
XI, Appendix A shall use the following conditions when implementing 
Equation (2) in A-4300(b)(1):
    For R < 0, [Delta]KI depends on the crack depth (a), and 
the flow stress ([sigma]f). The flow stress is defined by 
[sigma]f = \1/2\([sigma]ys + 
[sigma]ult), where [sigma]ys is the yield 
strength and [sigma]ult is the ultimate tensile strength in 
units ksi (MPa) and a is in units in. (mm). For -2 <= R <= 0 and 
Kmax - Kmin <= 0.8 x 1.12 [sigma]f 
[radic]([pi]a), S = 1 and [Delta]KI = Kmax. For R 
< -2 and Kmax - Kmin <= 0.8 x 1.12 
[sigma]f [radic]([pi]a), S = 1 and [Delta]KI = (1 
- R) Kmax/3. For R < 0 and Kmax - Kmin 
> 0.8 x 1.12 [sigma]f [radic]([pi]a), S = 1 and 
[Delta]KI = Kmax - Kmin.
    (xxix) Nonmandatory Appendix R. Nonmandatory Appendix R, ``Risk-
Informed Inspection Requirements for Piping,'' of Section XI, 2005 
Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(2) of this section, may not be implemented 
without prior NRC authorization of the proposed alternative in 
accordance with paragraph (a)(3)(i) of this section.
    (3) As used in this section, references to the OM Code refer to the 
ASME Code for Operation and Maintenance of Nuclear Power Plants, 
Subsections ISTA, ISTB, ISTC, and ISTD, Mandatory Appendices I and II, 
and Nonmandatory Appendices A through H and J, and include the 1995 
Edition through the 2006 Addenda subject to the following conditions:
* * * * *
    (v) Subsection ISTD. Article IWF-5000, ``Inservice Inspection 
Requirements for Snubbers,'' of the ASME B&PV Code, Section XI, must be 
used when performing inservice inspection examinations and tests of 
snubbers at nuclear power plants, except as conditioned in paragraphs 
(b)(3)(v)(A) and (b)(3)(v)(B) of this section.
    (A) Licensees may use Subsection ISTD, ``Preservice and Inservice 
Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water 
Reactor Power Plants,'' ASME OM Code, 1995 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section, in place of the requirements for snubbers in the editions 
and addenda up to the 2005 Addenda of the ASME B&PV Code, Section XI, 
IWF-5200(a) and (b) and IWF-5300(a) and (b), by making appropriate 
changes to their technical specifications or licensee-controlled 
documents. Preservice and inservice examinations must be performed 
using the VT-3 visual examination method described in IWA-2213.
    (B) Licensees shall comply with the provisions for examining and 
testing snubbers in Subsection ISTD of the ASME OM Code and make 
appropriate changes to their technical specifications or licensee-
controlled documents when using the 2006 Addenda and later editions and 
addenda of Section XI of the ASME B&PV Code.
    (vi) Exercise interval for manual valves. Manual valves must be 
exercised on a 2-year interval rather that the 5-

[[Page 36276]]

year interval specified in paragraph ISTC-3540 of the 1999 through the 
2005 Addenda of the ASME OM Code, provided that adverse conditions do 
not require more frequent testing.
* * * * *
    (c) * * *
    (3) The Code edition, addenda, and optional ASME Code cases to be 
applied to components of the reactor coolant pressure boundary must be 
determined by the provisions of paragraph NCA-1140, Subsection NCA of 
Section III of the ASME Boiler and Pressure Vessel Code, subject to the 
following conditions:
    (i) The edition and addenda applied to a component must be those 
which are incorporated by reference in paragraph (b)(1) of this 
section;
    (ii) The ASME Code provisions applied to the pressure vessel may be 
dated no earlier than the Summer 1972 Addenda of the 1971 edition;
    (iii) The ASME Code provisions applied to piping, pumps, and valves 
may be dated no earlier than the Winter 1972 Addenda of the 1971 
edition; and
    (iv) The optional Code cases applied to a component must be those 
listed in NRC Regulatory Guide 1.84 that is incorporated by reference 
in paragraph (b) of this section.
* * * * *
    (d) * * *
    (2) The Code edition, addenda, and optional ASME Code cases to be 
applied to the systems and components identified in paragraph (d)(1) of 
this section must be determined by the rules of paragraph NCA-1140, 
Subsection NCA of Section III of the ASME Boiler and Pressure Vessel 
Code, subject to the following conditions:
    (i) The edition and addenda must be those which are incorporated by 
reference in paragraph (b)(1) of this section;
    (ii) The ASME Code provisions applied to the systems and components 
may be dated no earlier than the 1980 Edition; and
    (iii) The optional Code cases must be those listed in the NRC 
Regulatory Guide 1.84 that is incorporated by reference in paragraph 
(b) of this section.
    (e) * * *
    (2) The Code edition, addenda, and optional ASME Code cases to be 
applied to the systems and components identified in paragraph (e)(1) of 
this section must be determined by the rules of paragraph NCA-1140, 
subsection NCA of Section III of the ASME Boiler and Pressure Vessel 
Code, subject to the following conditions:
    (i) The edition and addenda must be those which are incorporated by 
reference in paragraph (b)(1) of this section;
    (ii) The ASME Code provisions applied to the systems and components 
may be dated no earlier than the 1980 Edition; and
    (iii) The optional Code cases must be those listed in NRC 
Regulatory Guide 1.84 that is incorporated by reference in paragraph 
(b) of this section.
    (f) * * *
    (2) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit was issued on or after January 1, 
1971, but before July 1, 1974, pumps and valves which are classified as 
ASME Code Class 1 and Class 2 must be designed and provided with access 
to enable the performance of inservice tests for operational readiness 
set forth in editions and addenda of Section XI of the ASME Boiler and 
Pressure Vessel Code incorporated by reference in paragraph (b) of this 
section (or the optional ASME Code cases listed in NRC Regulatory Guide 
1.147, Revision 16, or Regulatory Guide 1.192 that are incorporated by 
reference in paragraph (b) of this section) in effect 6 months before 
the date of issuance of the construction permit. The pumps and valves 
may meet the inservice test requirements set forth in subsequent 
editions of this Code and addenda which are incorporated by reference 
in paragraph (b) of this section (or the optional ASME Code Cases 
listed in NRC Regulatory Guide 1.147, Revision 16, or Regulatory Guide 
1.192 that are incorporated by reference in paragraph (b) of this 
section), subject to the applicable conditions listed therein.
    (3) * * *
    (v) All pumps and valves may meet the test requirements set forth 
in subsequent editions of codes and addenda or portions thereof which 
are incorporated by reference in paragraph (b) of this section, subject 
to the conditions listed in paragraph (b) of this section.
    (4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, pumps and valves which are classified as 
ASME Code Class 1, Class 2, and Class 3 must meet the inservice test 
requirements, except design and access provisions, set forth in the 
ASME OM Code and addenda that become effective subsequent to editions 
and addenda specified in paragraphs (f)(2) and (f)(3) of this section 
and that are incorporated by reference in paragraph (b) of this 
section, to the extent practical within the limitations of design, 
geometry and materials of construction of the components.
    (i) Inservice tests to verify operational readiness of pumps and 
valves, whose function is required for safety, conducted during the 
initial 120-month interval must comply with the requirements in the 
latest edition and addenda of the Code incorporated by reference in 
paragraph (b) of this section on the date 12 months before the date of 
issuance of the operating license under this part, or 12 months before 
the date scheduled for initial loading fuel under a combined license 
under part 52 of this chapter (or the optional ASME Code cases listed 
in NRC Regulatory Guide 1.192, that is incorporated by reference in 
paragraph (b) of this section), subject to the conditions listed in 
paragraph (b) of this section.
    (ii) Inservice tests to verify operational readiness of pumps and 
valves, whose function is required for safety, conducted during 
successive 120-month intervals must comply with the requirements of the 
latest edition and addenda of the Code incorporated by reference in 
paragraph (b) of this section 12 months before the start of the 120-
month interval (or the optional ASME Code cases listed in NRC 
Regulatory Guide 1.147, Revision 16, or Regulatory Guide 1.192 that are 
incorporated by reference in paragraph (b) of this section), subject to 
the conditions listed in paragraph (b) of this section.
    (iii) [Reserved]
    (iv) Inservice tests of pumps and valves may meet the requirements 
set forth in subsequent editions and addenda that are incorporated by 
reference in paragraph (b) of this section, subject to the conditions 
listed in paragraph (b) of this section, and subject to NRC approval. 
Portions of editions or addenda may be used provided that all related 
requirements of the respective editions or addenda are met.
    (5) * * *
    (iv) Where a pump or valve test requirement by the code or addenda 
is determined to be impractical by the licensee and is not included in 
the revised inservice test program as permitted by paragraph (f)(4) of 
this section, the basis for this determination must be submitted for 
NRC review and approval not later than 12 months after the expiration 
of the initial 120-month interval of operation from start of facility 
commercial operation and each subsequent 120-month interval of 
operation during which the test is determined to be impractical.
* * * * *
    (g) * * *
    (2) For a boiling or pressurized water-cooled nuclear power 
facility whose

[[Page 36277]]

construction permit was issued on or after January 1, 1971, but before 
July 1, 1974, components (including supports) which are classified as 
ASME Code Class 1 and Class 2 must be designed and be provided with 
access to enable the performance of inservice examination of such 
components (including supports) and must meet the preservice 
examination requirements set forth in editions and addenda of Section 
III or Section XI of the ASME B&PV Code (or ASME OM Code for snubber 
examination and testing) incorporated by reference in paragraph (b) of 
this section (or the optional ASME code cases listed in NRC Regulatory 
Guide 1.147, Revision 16, that are incorporated by reference in 
paragraph (b) of this section) in effect six months before the date of 
issuance of the construction permit. The components (including 
supports) may meet the requirements set forth in subsequent editions 
and addenda of this Code which are incorporated by reference in 
paragraph (b) of this section (or the optional ASME code cases listed 
in NRC Regulatory Guide 1.147, Revision 16, when using Section XI, or 
Regulatory Guide 1.192 when using the OM Code, that are incorporated by 
reference in paragraph (b) of this section), subject to the applicable 
conditions.
    (3) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit under this part, or design 
certification, design approval, combined license, or manufacturing 
license under part 52 of this chapter, was issued on or after July 1, 
1974:
    (i) Components (including supports) which are classified as ASME 
Code Class 1 must be designed and provided with access to enable the 
performance of inservice examination of these components and must meet 
the preservice examination requirements set forth in the editions and 
addenda of Section III or Section XI of the ASME B&PV Code (or ASME OM 
Code for snubber examination and testing) incorporated by reference in 
paragraph (b) of this section (or the optional ASME code cases listed 
in NRC Regulatory Guide 1.147, Revision 16, when using Section XI, or 
Regulatory Guide 1.192 when using the OM Code, that are incorporated by 
reference in paragraph (b) of this section) applied to the construction 
of the particular component.
    (ii) Components which are classified as ASME Code Class 2 and Class 
3 and supports for components which are classified as ASME Code Class 
1, Class 2, and Class 3 must be designed and be provided with access to 
enable the performance of inservice examination of these components and 
must meet the preservice examination requirements set forth in the 
editions and addenda of Section III or Section XI of the ASME B&PV Code 
(or ASME OM Code for snubber examination and testing) incorporated by 
reference in paragraph (b) of this section (or the optional ASME code 
cases listed in NRC Regulatory Guide 1.147, Revision 16, when using 
Section XI; or Regulatory Guide 1.192 when using the OM Code, that are 
incorporated by reference in paragraph (b) of this section) applied to 
the construction of the particular component.
    (iii)-(iv) [Reserved]
    (v) All components (including supports) may meet the requirements 
set forth in subsequent editions of codes and addenda or portions 
thereof which are incorporated by reference in paragraph (b) of this 
section, subject to the conditions listed therein.
    (4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) which 
are classified as ASME Code Class 1, Class 2, and Class 3 must meet the 
requirements, except design and access provisions and preservice 
examination requirements, set forth in Section XI of editions and 
addenda of the ASME B&PV Code (or ASME OM Code for snubber examination 
and testing) that become effective subsequent to editions specified in 
paragraphs (g)(2) and (g)(3) of this section and that are incorporated 
by reference in paragraph (b) of this section, to the extent practical 
within the limitations of design, geometry and materials of 
construction of the components. Components which are classified as 
Class MC pressure retaining components and their integral attachments, 
and components which are classified as Class CC pressure retaining 
components and their integral attachments must meet the requirements, 
except design and access provisions and preservice examination 
requirements, set forth in Section XI of the ASME B&PV Code and addenda 
that are incorporated by reference in paragraph (b) of this section, 
subject to the condition listed in paragraph (b)(2)(vi) of this section 
and the conditions listed in paragraphs (b)(2)(viii) and (b)(2)(ix) of 
this section, to the extent practical within the limitation of design, 
geometry and materials of construction of the components.
    (i) Inservice examinations of components and system pressure tests 
conducted during the initial 120-month inspection interval must comply 
with the requirements in the latest edition and addenda of the Code 
incorporated by reference in paragraph (b) of this section on the date 
12 months before the date of issuance of the operating license under 
this part, or 12 months before the date scheduled for initial loading 
of fuel under a combined license under part 52 of this chapter (or the 
optional ASME Code cases listed in NRC Regulatory Guide 1.147, through 
Revision 16, when using Section XI; or Regulatory Guide 1.192 when 
using the OM Code, that are incorporated by reference in paragraph (b) 
of this section), subject to the conditions listed in paragraph (b) of 
this section.
    (ii) Inservice examination of components and system pressure tests 
conducted during successive 120-month inspection intervals must comply 
with the requirements of the latest edition and addenda of the Code 
incorporated by reference in paragraph (b) of this section 12 months 
before the start of the 120-month inspection interval (or the optional 
ASME Code cases listed in NRC Regulatory Guide 1.147, Revision 16, that 
are incorporated by reference in paragraph (b) of this section), 
subject to the conditions listed in paragraph (b) of this section. 
However, a licensee whose inservice inspection interval commences 
during the 12 through 18-month period after July 21, 2011 may delay the 
update of their Appendix VIII program by up to 18 months after July 21, 
2011.
    (iii) When applying editions and addenda prior to the 2003 Addenda 
of Section XI of the ASME B&PV Code licensees may, but are not required 
to, perform the surface examinations of high-pressure safety injection 
systems specified in Table IWB-2500-1, Examination Category B-J, Item 
Numbers B9.20, B9.21 and B9.22.
    (iv) Inservice examination of components and system pressure tests 
may meet the requirements set forth in subsequent editions and addenda 
that are incorporated by reference in paragraph (b) of this section, 
subject to the conditions listed in paragraph (b) of this section, and 
subject to Commission approval. Portions of editions or addenda may be 
used provided that all related requirements of the respective editions 
or addenda are met.
    (v) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit under this part or combined license 
under part 52 of this chapter was issued after January 1, 1956:
    (A) Metal containment pressure retaining components and their 
integral attachments must meet the inservice inspection, repair, and 
replacement requirements applicable to components

[[Page 36278]]

which are classified as ASME Code Class MC;
    (B) Metallic shell and penetration liners which are pressure 
retaining components and their integral attachments in concrete 
containments must meet the inservice inspection, repair, and 
replacement requirements applicable to components which are classified 
as ASME Code Class MC; and
    (C) Concrete containment pressure retaining components and their 
integral attachments, and the post-tensioning systems of concrete 
containments must meet the inservice inspections, repair, and 
replacement requirements applicable to components which are classified 
as ASME Code Class CC.
    (5) * * *
    (iii) If the licensee has determined that conformance with a code 
requirement is impractical for its facility, the licensee shall notify 
the NRC and submit, as specified in Sec.  50.4, information to support 
the determinations. Determinations of impracticality in accordance with 
this section must be based on the demonstrated limitations experienced 
when attempting to comply with the code requirements during the 
inservice inspection interval for which the request is being submitted. 
Requests for relief made in accordance with this section must be 
submitted to the NRC no later than 12 months after the expiration of 
the initial or subsequent 120-month inspection interval for which 
relief is sought.
    (iv) Where the licensee determines that an examination required by 
Code edition or addenda is impractical, the basis for this 
determination must be submitted for NRC review and approval not later 
than 12 months after the expiration of the initial or subsequent 120-
month inspection interval for which relief is sought.
    (6) * * *
    (ii) * * *
    (B) Licensees do not have to submit to the NRC for approval of 
their containment inservice inspection programs which were developed to 
satisfy the requirements of Subsection IWE and Subsection IWL with 
specified conditions. The program elements and the required 
documentation must be maintained on site for audit.
* * * * *
    (E) * * *
    (1) All licensees of pressurized water reactors shall augment their 
inservice inspection program by implementing ASME Code Case N-722-1 
subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) 
through (g)(6)(ii)(E)(4) of this section. The inspection requirements 
of ASME Code Case N-722-1 do not apply to components with pressure 
retaining welds fabricated with Alloy 600/82/182 materials that have 
been mitigated by weld overlay or stress improvement.
    (2) If a visual examination determines that leakage is occurring 
from a specific item listed in Table 1 of ASME Code Case N-722-1 that 
is not exempted by the ASME Code, Section XI, IWB-1220(b)(1), 
additional actions must be performed to characterize the location, 
orientation, and length of crack(s) in Alloy 600 nozzle wrought 
material and location, orientation, and length of crack(s) in Alloy 82/
182 butt welds. Alternatively, licensees may replace the Alloy 600/82/
182 materials in all the components under the item number of the 
leaking component.
    (3) If the actions in paragraph (g)(6)(ii)(E)(2) of this section 
determine that a flaw is circumferentially oriented and potentially a 
result of primary water stress corrosion cracking, licensees shall 
perform non-visual NDE inspections of components that fall under that 
ASME Code Case N-722-1 item number. The number of components inspected 
must equal or exceed the number of components found to be leaking under 
that item number. If circumferential cracking is identified in the 
sample, non-visual NDE must be performed in the remaining components 
under that item number.
* * * * *
    (F) Examination requirements for class 1 piping and nozzle 
dissimilar-metal butt welds.
    (1) Licensees of existing, operating pressurized-water reactors as 
of July 21, 2011 shall implement the requirements of ASME Code Case N-
770-1, subject to the conditions specified in paragraphs 
(g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) of this section, by the 
first refueling outage after August 22, 2011.
    (2) Full structural weld overlays authorized by the NRC staff may 
be categorized as Inspection Items C or F, as appropriate; welds that 
have been mitigated by the Mechanical Stress Improvement Process 
(MSIP\TM\) may be categorized as Inspection Items D or E, as 
appropriate, provided the criteria in Appendix I of the code case have 
been met; for ISI frequencies, all other butt welds that rely on Alloy 
82/182 for structural integrity shall be categorized as Inspection 
Items A-1, A-2 or B until the NRC staff has reviewed the mitigation and 
authorized an alternative code case Inspection Item for the mitigated 
weld, or until an alternative code case Inspection Item is used based 
on conformance with an ASME mitigation code case endorsed in Regulatory 
Guide 1.147 with conditions, if applicable, and incorporated in this 
section.
    (3) Baseline examinations for welds in Table 1, Inspection Items A-
1, A-2, and B, shall be completed by the end of the next refueling 
outage after January 20, 2012. Previous examinations of these welds can 
be credited for baseline examinations if they were performed within the 
re-inspection period for the weld item in Table 1 using Section XI, 
Appendix VIII requirements and met the Code required examination volume 
of essentially 100 percent. Other previous examinations that do not 
meet these requirements can be used to meet the baseline examination 
requirement, provided NRC approval of alternative inspection 
requirements in accordance with paragraphs (a)(3)(i) or (a)(3)(ii) of 
this section is granted prior to the end of the next refueling outage 
after January 20, 2012.
    (4) The axial examination coverage requirements of -2500(c) may not 
be considered to be satisfied unless essentially 100 percent coverage 
is achieved.
    (5) All hot-leg operating temperature welds in Inspection Items G, 
H, J, and K must be inspected each interval. A 25-percent sample of 
cold-leg operating temperature welds must be inspected whenever the 
core barrel is removed (unless it has already been inspected within the 
past 10 years) or has reached 20 years, whichever is less.
    (6) For any mitigated weld whose volumetric examination detects 
growth of existing flaws in the required examination volume that exceed 
the previous IWB-3600 flaw evaluations or new flaws, a report 
summarizing the evaluation, along with inputs, methodologies, 
assumptions, and cause of the new flaw or flaw growth is to be provided 
to the NRC prior to the weld being placed in service other than modes 5 
or 6.
    (7) For Inspection Items G, H, J, and K, when applying the 
acceptance standards of ASME B&PV Code, Section XI, IWB-3514, for 
planar flaws contained within the inlay or onlay, the thickness ``t'' 
in IWB-3514 is the thickness of the inlay or onlay. For planar flaws in 
the balance of the dissimilar metal weld examination volume, the 
thickness ``t'' in IWB-3514 is the combined thickness of the inlay or 
onlay and the dissimilar metal weld.
    (8) Welds mitigated by optimized weld overlays in Inspection Items 
D and E are not permitted to be placed into a population to be examined 
on a sample basis and must be examined once each inspection interval.

[[Page 36279]]

    (9) Replace the first two sentences of Extent and Frequency of 
Examination for Inspection Item D in Table 1 of Code Case N-770-1 with, 
``Examine all welds no sooner than the third refueling outage and no 
later than 10 years following stress improvement application.'' Replace 
the first two sentences of Note (11)(b)(2) in Code Case N-770-1 with, 
``The first examination following weld inlay, onlay, weld overlay, or 
stress improvement for Inspection Items D through K shall be performed 
as specified.''
    (10) Note (2) to Figure 5(a) of Code Case N-770-1 pertaining to 
alternative examination volume for optimized weld overlays may not be 
applied unless NRC approval is authorized under paragraphs (a)(3)(i) or 
(a)(3)(ii) of this section.
* * * * *
    Footnotes to Sec.  50.55a:

    \1\ For inspections to be conducted once per interval, the 
inspections shall be performed in accordance with the schedule in 
Section XI, paragraph IWB-2400, except for plants with inservice 
inspection programs based on a Section XI edition or addenda prior 
to the 1994 Addenda. For plants with inservice inspection programs 
based on a Section XI edition or addenda prior to the 1994 Addenda, 
the inspection shall be performed in accordance with the schedule in 
Section XI, paragraph IWB-2400, of the 1994 Addenda.
* * * * *

    Dated at Rockville, Maryland, this 27th day of May 2011.

    For the Nuclear Regulatory Commission.
Eric J. Leeds,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2011-14652 Filed 6-20-11; 8:45 am]
BILLING CODE 7590-01-P