[Federal Register Volume 73, Number 176 (Wednesday, September 10, 2008)]
[Rules and Regulations]
[Pages 52730-52750]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-20624]



[[Page 52729]]

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Part II





Nuclear Regulatory Commission





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10 CFR Part 50



Industry Codes and Standards; Amended Requirements; Final Rule

Federal Register / Vol. 73, No. 176 / Wednesday, September 10, 2008 / 
Rules and Regulations

[[Page 52730]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AH76
[NRC-2007-0003]


Industry Codes and Standards; Amended Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its 
regulations to incorporate by reference the 2004 Edition of Section 
III, Division 1, and Section XI, Division 1, of the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), 
and the 2004 Edition of the ASME Code for Operation and Maintenance of 
Nuclear Power Plants (OM Code) to provide updated rules for 
constructing and inspecting components and testing pumps, valves, and 
dynamic restraints (snubbers) in light-water nuclear power plants. The 
NRC also is incorporating by reference ASME Code Cases N-722, 
``Additional Examinations for PWR [pressurized water reactor (PWR)] 
Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 
600/82/182 Materials, Section XI, Division 1,'' and N-729-1, 
``Alternative Examination Requirements for PWR Reactor Vessel Upper 
Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, 
Section XI, Division 1,'' both with conditions. The amendment also 
removes certain obsolete requirements specified in the NRC's 
regulations. This action is in accordance with the NRC's policy to 
periodically update the regulations to incorporate by reference new 
editions and addenda of the ASME Codes and is intended to maintain the 
safety of nuclear reactors and make NRC activities more effective and 
efficient.

DATES: Effective Date: October 10, 2008. The incorporation by reference 
of certain publications listed in the regulation is approved by the 
Director of the Office of the Federal Register as of October 10, 2008.

ADDRESSES: You can access publicly available documents related to this 
document using the following methods:
    Federal e-Rulemaking Portal: Go to http://www.regulations.gov and 
search for documents filed under Docket ID [NRC-2007-0003]. Address 
questions about NRC dockets to Carol Gallagher 301-415-5905; e-mail 
Carol.Gallagher@nrc.gov.
    NRC's Public Document Room (PDR): The public may examine and have 
copied for a fee publicly available documents at the NRC's PDR, Public 
File Area O1F21, 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 electronic Reading Room 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 
pdr.resource@nrc.gov.

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 Mark.Padovan@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Background
II. Analysis of Public Comments
III. Section-by-Section Analysis
IV. Generic Aging Lessons Learned Report
V. Availability of Documents
VI. Voluntary Consensus Standards
VII. Finding of No Significant Environmental Impact: Environmental 
Assessment
VIII. Paperwork Reduction Act Statement
IX. Regulatory Analysis
X. Regulatory Flexibility Certification
XI. Backfit Analysis
XII. Congressional Review Act

I. Background

    The NRC is amending 10 CFR 50.55a to incorporate by reference the 
2004 Edition of Section III, Division 1 and Section XI, Division 1 of 
the ASME BPV Code and the 2004 Edition of the ASME OM Code. Section 
50.55a requires the use of Section III, Division 1 of the ASME BPV Code 
for the construction of nuclear power plant components; Section XI, 
Division 1 of the ASME BPV Code for the inservice inspection (ISI) of 
nuclear power plant components; and the ASME OM Code for the inservice 
testing (IST) of pumps and valves. The NRC published a proposed 
rulemaking on this subject in the Federal Register on April 5, 2007 (72 
FR 16731). The 75-day public comment period for the proposed rule 
closed on June 19, 2007.
    The introductory paragraph of Sec.  50.55a establishes the 
applicability of the conditions therein to licenses and approvals 
issued under Part 52. Specifically, that rule states the following:
     ``Each combined license for a utilization facility is 
subject to the following conditions in addition to those specified in 
Sec.  50.55, except that each combined license for a boiling or 
pressurized water-cooled nuclear power facility is subject to the 
conditions in paragraphs (f) and (g) of this section, but only after 
the Commission makes the finding under Sec.  52.103(g) of this 
chapter.''
     ``Each manufacturing license, standard design approval, 
and standard design certification application under part 52 of this 
chapter is subject to the conditions in paragraphs (a), (b)(1), (b)(4), 
(c), (d), (e), (f)(3), and (g)(3) of this section.''
    Accordingly, combined licenses, manufacturing licenses, standard 
design approvals, and standard design certifications are subject to 
these requirements.
    The ASME BPV Code and OM Code 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 is otherwise impractical. The NRC reviews new 
editions and addenda of the ASME BPV and OM Codes, and periodically 
updates Sec.  50.55a to incorporate by reference newer editions and 
addenda. New editions of the subject codes are issued every 3 years; 
addenda to the editions are issued yearly except in years when a new 
edition is issued. The editions and addenda of the ASME BPV and OM 
Codes were last incorporated by reference into the regulations in a 
final rule dated October 1, 2004 (69 FR 58804). In that rule, Sec.  
50.55a was revised to incorporate by reference the 2001 Edition, and 
2002 and 2003 Addenda, of Sections III and XI, Division 1, of the ASME 
BPV Code and the 2001 Edition, and 2002 and 2003 Addenda, of the ASME 
OM Code.
    The NRC is now incorporating by reference Section III, Division 1, 
of the 2004 Edition of the ASME BPV Code; Section XI, Division 1, of 
the 2004 Edition of the ASME BPV Code subject to modifications and 
limitations; and the 2004 Edition of the ASME OM Code.

II. Analysis of Public Comments

    The NRC received 23 letters and e-mails from the public that 
provided about 87 comments on the proposed rule. These comments were 
submitted by individuals, nuclear utilities, and nuclear industry 
organizations

[[Page 52731]]

consisting of the Nuclear Energy Institute (NEI), the Performance 
Demonstration Initiative, and the Strategic Teaming and Resource 
Sharing (STARS) organization. The NRC reviewed and considered the 
comments in its final rulemaking, as discussed in the following 
sections:

1. 10 CFR 50.55a(b)(1)

    Public Comment:
    In a letter dated June 12, 2007, G.C. Slagis Associates commented 
that the reversing dynamic load rules of the ASME BPV Code, Section 
III, should not be approved for new construction. The commenter stated 
that the draft rule language incorporated the 2004 Edition of the 
Section III piping rules (NB/NC/ND-3600) for evaluation of ``reversing 
dynamic loads,'' whereas the NRC had taken exception to these rules in 
the past. The commenter also stated that these piping rules should not 
be approved for new construction.
    NRC Response:
    The NRC has not approved the reversing dynamic load rules in the 
piping rules for the ASME BPV Code, Section III for new construction or 
existing nuclear plants. The NRC believes that the commenter's 
interpretation of the proposed rule was based on the wording contained 
in the summary of the proposed revisions to 10 CFR 50.55a (on the 
bottom of page 72 FR 16732 and top of page 72 FR 16733; April 5, 2007) 
that said ``The proposed rule would revise Sec.  50.55a(b)(1) to 
incorporate by reference the 2004 Edition of Section III of the ASME 
Boiler and Pressure Vessel (BPV) Code. The NRC does not propose to 
adopt any limitations with respect to the 2004 Edition of Section 
III.'' The wording in the second sentence contained an editorial error. 
The sentence should have read ``The NRC does not propose to adopt any 
additional limitations with respect to the 2004 Edition of Section 
III.'' The proposed rule language on page 72 FR 16740 retained the 
previous restriction regarding the piping rules. The restriction 
applies to the 1994 Edition through the 2004 Edition. To clarify this, 
the NRC revised the subject sentences in Section III, Section-by 
Section Analysis, of this document as follows:
    The final rule revises Sec.  50.55a(b)(1) in the current 
regulation to incorporate by reference the 2004 Edition of Section 
III of the ASME BPV Code into 10 CFR 50.55a. The NRC is not adopting 
any additional limitations with respect to the 2004 Edition of 
Section III.

2. 10 CFR 50.55a(b)(1)(iii)--Seismic Design of Piping

    Public Comment:
    In a letter dated June 19, 2007, Westinghouse Electric Company 
requested that the NRC clarify the current limitation specified in 
Sec.  50.55a(b)(1)(iii) regarding seismic design. The commenter stated 
that the limitations are related to the treatment of piping. However, 
as is stated in Sec.  50.55a(b)(1)(iii), the rules in Article NB-3200 
of Section III of the ASME BPV Code contain criteria applicable to the 
seismic design of components other than piping systems. The commenter 
recommended that the wording in Sec.  50.55a(b)(1)(iii) be revised to 
clarify that the limitation only applies to the seismic design of 
piping.
    NRC Response:
    The NRC agrees with the commenter, and has revised Sec.  
50.55a(b)(1)(iii) in this final rule as follows:

    Seismic design of piping. Applicants and licensees may use 
Articles NB-3200, NB-3600, NC-3600, and ND-3600 for seismic design 
of piping up to and including the 1993 Addenda, subject to the 
limitation specified in paragraph (b)(1)(ii) of this section. 
Applicants and licensees may not use these Articles for seismic 
design of piping in the 1994 addenda through the latest edition and 
addenda incorporated by reference in paragraph (b)(1) of this 
section.

3. 10 CFR 50.55a(b)(2)(xv)--Appendix VIII Specimen Set and 
Qualification Requirements

    Public Comment:
    Conflicts between Sec. Sec.  50.55a(b)(2)(xv) and 
50.55a(b)(2)(xxiv) were identified by the Performance Demonstration 
Initiative (letter dated May 11, 2007), Nuclear Management Company 
(letter dated June 19, 2007), and Mr. Michael Gothard (comment received 
on the NRC's public Web site on May 11, 2007). The proposed rule 
extends the application of Sec.  50.55a(b)(2)(xv) from the 1995 Edition 
through the 2001 Edition to the 1995 Edition through the 2004 Edition. 
10 CFR 50.55a(b)(2)(xxiv) prohibits the use of Appendix VIII of Section 
XI, 1995 Edition through the 2001 Edition, and the supplements of 
Appendix VIII and Article I-3000 of the 2002 Addenda through the latest 
edition and addenda incorporated by reference in Sec.  50.55a(b). The 
proposed change in Sec.  50.55a(b)(2)(vx) creates confusion, 
unnecessary burden, and conflicting requirements. The commentors 
proposed leaving Sec.  50.55a(b)(2)(xv) unchanged.
    NRC Response:
    The NRC agrees with the commentors that the requirements in 
Sec. Sec.  50.55a(b)(2)(xv) and 50.55a(b)(2)(xxiv) conflict. The intent 
of the proposed rule was to minimize the burden associated with 
reconciling an existing Appendix VIII of Section XI, 1995 Edition 
through the 2001 Edition, program with changes that occurred in the 
2002 Addenda and later edition and addenda. In keeping with the NRC's 
intent, Sec.  50.55a(b)(2)(xv) will reference up to, and including, the 
2001 Edition of Appendix VIII as follows:

    Appendix VIII specimen set and qualification requirements. The 
following provisions may be used to modify implementation of 
Appendix VIII of Section XI, 1995 Edition through the 2001 Edition. 
Licensees choosing to apply these provisions shall apply all of the 
following provisions under this paragraph except for those in Sec.  
50.55a(b)(2)(xv)(F) which are optional. Licensees who use later 
editions and addenda than the 2001 Edition of Section XI of the ASME 
Code shall use the 2001 Edition of Appendix VIII.

4. 10 CFR 50.55a(b)(2)(xx)--System Leakage Tests

    Public Comment:
    In a letter dated June 19, 2007, Progress Energy stated that the 
construction code requirement for a hydrostatic pressure test is not 
performed at a pressure that constitutes a challenge to the material. A 
hydrostatic test at this pressure does not contribute to safety any 
more than a pressure test at operating pressure, since both are 
conducted below the yield strength of the materials involved. 
Therefore, from a safety perspective, the hydrostatic test is not used 
to verify the structural integrity of the component or system being 
tested. It only proves leak tightness, which is also accomplished by a 
system leakage test. Hence, the end results of the hydrostatic test and 
the system leakage test are the same (leak tightness is verified). The 
additional nondestructive examination (NDE) being suggested by the NRC 
is of no value in verifying leak tightness, and thus is not related to 
the safety significance of not performing a hydrostatic test. The 
construction code NDE that is implemented by ASME Code, Section XI 
(IWA-4500, [``Examination and Testing'']), is all that is needed to 
verify any welding discontinuities that could affect the required joint 
efficiency for the required quality of the weld or brazed joint.
    NRC Response:
    Subarticle IWA-4540(a) of the 1995 Edition of the ASME BPV Code, 
Section XI, requires that after repair and replacement activities, a 
system hydrostatic pressure test be performed. The industry asserted 
that the hydrostatic pressure test creates a significant hardship. 
Subsequently, the ASME Committee developed Code Case N-416-3, 
``Alternative Pressure Test Requirements for Welded Repairs or

[[Page 52732]]

Installation of Replacement Items by Welding Class 1, 2, and 3, Section 
XI, Division 1,'' to allow the use of system leakage testing and NDE to 
replace the hydrostatic test. Later, the technical provisions of Code 
Case N-416-3 were incorporated into the 2001 Edition of ASME Section 
XI, IWA-4540(a) and maintained through the 2002 Addenda. However, the 
NDE requirements of IWA-4540(a) were eliminated from the 2003 Addenda 
of the Code. Therefore, the NRC proposed a condition in Sec.  
50.55a(b)(2)(xx) requiring Section III NDE be performed following 
repair and replacement activities if a system leakage test was to be 
used in lieu of a hydrostatic test under the 2003 Addenda through the 
latest edition and addenda incorporated by reference in 10 CFR 
50.55a(b)(2).
    The piping systems in some vintage nuclear power plants were 
fabricated in accordance with American National Standards Institute 
(ANSI)/ASME B31.1, ``Power Piping,'' Code. ANSI/ASME B31.1 does not 
require a volumetric examination for those systems that would now be 
classified as ASME Class 2 and Class 3 piping systems during original 
construction. The current ASME BPV Code, Section XI (IWA-4500), allows 
licensees to use the NDE requirement of the original construction code 
as part of repair/replacement activities. Licensees of these vintage 
plants would not need to perform volumetric examinations after repair/
replacement activities for piping classified as ASME Class 2 or Class 3 
piping for which ANSI B31.1 does not require NDE. A system pressure 
test or hydrostatic pressure test does not verify the structural 
integrity of the repaired piping components. However, it is generally 
recognized in the industry that the volumetric examinations do provide 
significant information relative to the structural integrity of the 
repaired piping components. For those Class 2 and 3 piping systems that 
may not receive a volumetric examination for the life of the systems, 
the NRC is concerned that performance of a system leakage test without 
associated volumetric examinations would not adequately ensure high 
quality welds for the repaired or replaced component. Therefore, 
performance of a Section III volumetric examination in connection with 
a system leakage test in repair/replacement activities is necessary.
    Public Comment:
    In letter dated June 13, 2007, ASME stated that Sec.  
50.55a(b)(2)(xx) does not explicitly state that the NDE shall be 
performed after the system leakage test. As written, a licensee could 
comply with this requirement by performing the required NDE before the 
system leakage test. It is common practice to perform this NDE prior to 
the system leakage test.
    NRC Response:
    The NRC agrees with the commenter that an ASME BPV Code, Section 
III, 1992 Edition, volumetric examination performed as part of the 
repair/replacement activities prior to the system leakage test can be 
accepted to fulfill the NDE requirement of Sec.  50.55a(b)(2)(xx)(B). 
The NRC's position has been, and continues to be, that the NDE 
performed as part of the repair/replacement activities satisfies the 
NDE provision of subarticle IWA-4540(a) of the 2002 Addenda of the ASME 
Code, Section XI.
    Public Comment:
    In letter dated June 19, 2007, Duke Energy stated that Sec.  
50.55a(b)(2)(xx) does not restrict a licensee from using the provisions 
of IWA-5213(a) in the 2003 Addenda of Section XI. Therefore, licensees 
may currently use the provisions of IWA-4540(a) in the 2003 Addenda 
without having to perform NDE in accordance with the requirements of 
IWA-4540(a)(2) of the 2002 Addenda after a system leakage test. Because 
the proposed change imposes additional requirements on licensees, the 
change should be evaluated to determine whether the change is a 
backfit.
    NRC Response:
    The NRC agrees with the commenter that the proposed requirement 
would result in a backfit for some licensees because this final rule 
would now require them to perform the required NDE in conjunction with 
the system leakage test in lieu of the hydrostatic test. In the October 
1, 2004 (69 FR 58804), rulemaking of the 2003 Addenda of the ASME Code, 
the NRC neglected to incorporate the above NDE requirement in 10 CFR 
50.55a(b)(2). However, the oversight needs to be corrected to ensure 
that during repair or replacement activities, the volumetric 
examination, in conjunction with a system leakage test, is performed to 
ensure structural integrity of the repaired or replaced piping system. 
The NRC discusses its backfit analysis for those licensees who may be 
affected by this rule in Section XI, Backfit Analysis, of this 
document.

5. 10 CFR 50.55a(b)(2)(xxi)(A)--Table IWB-2500-1 Examination 
Requirements

    Public Comment:
    In letter dated June 13, 2007, ASME; in letter dated June 19, 2007, 
Nuclear Energy Institute; and in letter dated June 19, 2007, Duke 
Energy disagree with modifying the limitation to require visual 
examination of Class 1 pressurizer and steam generator nozzle inner 
radius areas (ASME Code Case N-619) based on the previous reactor 
vessel nozzle inner radius limitation (ASME Code Case N-648-1). The 
commenters believe that the original limitation (to continue 
examination of the inner nozzle radius region) is unnecessary because 
of the following:
    a. Inner nozzle radius regions in Class 1 systems have been 
examined for over 25 years without detecting cracking.
    b. Structural integrity evaluations demonstrated a large tolerance 
for flaws.
    c. Risk informed evaluations demonstrated that these nozzles have a 
large tolerance for flaws.
    d. Risk informed evaluations demonstrated a low probability of 
failure under plant operating conditions.
    e. There is a negligible change in risk if inspections are 
eliminated.
    f. The term enhanced VT-1 is not defined in Code, and studies show 
that VT-1 character heights provide the same or better resolution than 
the 1 mil wire.
    NRC Response:
    The NRC disagrees with the commentors. The limitation on the visual 
examination in 10 CFR 50.55a(b)(2)(xxi)(A) did not differentiate 
between vessel components. The limitation is an alternative for 
volumetric examinations. The proposed change in the rule is to provide 
a visual examination criterion for determining fatigue crack flaw 
depth.
    With respect to Item 5.a above, the commentor's information on 25 
years of inservice ultrasonic examinations with no evidence of inner 
radius cracking on nozzles covered by the ASME Code cases is from an 
ASME document issued in 2001. At that time, ultrasonic examinations of 
pressurized-water reactors were normally performed from the inside 
surface, and were normally performed from the outside surface for 
boiling-water reactors. The NRC took issue with the effectiveness of 
ultrasonic examinations of the inner nozzle radius performed prior to 
performance-based qualification requirements. Performance-based 
examinations of all reactor pressure vessel (RPV) inner nozzle radii 
became mandatory on November 22, 2002. On July 26, 2006, the Electric 
Power Research Institute--Boiling Water Reactor Vessel & Internal 
Project (BWRVIP) provided a summary of results from inner nozzle radius 
performance-based examinations to support reducing RPV inner nozzle 
radii examination frequency by 75 percent.
    By letter dated December 19, 2007, the NRC issued a safety 
evaluation

[[Page 52733]]

accepting BWRVIP-108 which reduced the inspection frequency of reactor 
nozzle-to-vessel shell welds and nozzle inner radius for BWRs (NRC 
ADAMS Accession Number ML073600374).
    Operating conditions, such as fluctuating temperature, and 
fabricating conditions, such as work hardening can cause cracking of 
the inner nozzle radius. The ASME Code Cases (N-619 and N-648-1) are 
silent on conditions that are associated with cracking. These 
conditions may appear, or be affected, at various times during the 
operating cycle and may not be specific to vessel design. To detect 
degradation that appears during operations, NDE of inner nozzle radii 
are warranted.
    Items 5b, 5c, and 5d pertained to risk-informed computations. Of 
the risk-informed piping programs reviewed to date, none of the 
programs contained risk data for Class 1 inner nozzle radius regions. 
The NRC did not find documentation of a review on the ASME 2001 
article. Recently, the BWRVIP submitted to the NRC information on 
structural integrity and probability of failure and risk calculations 
concerning the inspection of inner nozzle radius regions to the NRC for 
review, which is ongoing.
    With respect to Item 5f, the commentors referenced proprietary 
documents that were not made available to the NRC. Therefore, the NRC 
was unable to verify the data used to validate the adequacy of VT-1 and 
of character recognition for examinations of the inner radii regions. 
While characters are useful for distinguishing shapes, NUREG/CR 6860, 
``An Assessment of Visual Testing,'' identified the crack open width 
dimension as a key variable for visually detecting cracks. In 10 CFR 
50.55a(b)(2)(xxi)(A), the 1-mil width wire or crack is a measurable 
criterion for a postulated crack open width dimension. Therefore, the 
1-mil width wire or crack requirement provides a minimum criteria for 
performance-based demonstrations of examination effectiveness.
    The commentors stated that the term ``enhanced VT-1'' was not 
recognized by the ASME BPV Code. The term ``enhanced VT-1'' is being 
used by knowledgeable personnel for conversational expediency. The term 
``enhanced VT-1'' is not used in the regulation. However, the use of 
the term ``enhanced magnification'' is used in the rule and may have 
been misleading. Therefore, the term ``enhanced'' will be removed from 
the regulation.

6. 10 CFR 50.55a(b)(2)(xxviii)--Evaluation Procedure and Acceptance 
Criteria for PWR Reactor Vessel Head Penetration Nozzles

    Public Comment:
    In a letter dated June 13, 2007, the ASME stated that this 
modification is being proposed because of a typographical error that 
the NRC says exists in ASME Section XI, Non-mandatory Appendix O, 
paragraph O-3220(b), equation SR, = [l--
0.82R]-\22\, where the exponent -22 should be -2.2. ASME has 
identified this error and is publishing an ERRATA in July 2007 to 
correct this error retroactively to include the 2004 Edition of Section 
XI. As such, the proposed amendment to 10 CFR 50.55a(b)(2)(xxviii) is 
unnecessary.
    NRC Response:
    The NRC finds that ASME has published an ERRATA in July 2007 to 
correct the error in the SR equation of paragraph O-3220(b) 
retroactively to include the 2004 Edition of ASME BPV Code, Section XI. 
The condition imposed in Sec.  50.55a(b)(2)(xxviii) will not be 
necessary. Therefore, the NRC is not including Sec.  
50.55a(b)(2)(xxviii) in this final rule.

7. 10 CFR 50.55a(b)(3)(v)--Subsection ISTD

    Public Comments:
    By electronic mail dated June 11, 2007, George L. Fechter of 
Southern Nuclear Operating Company stated that Article IWF-5000, 
``Inservice Inspection Requirements for Snubbers,'' was deleted from 
the 2006 Addenda of the ASME BPV Code, Section XI. With adequate 
verification of training provided to personnel performing visual exams, 
removal, testing, and reinstallation of snubbers per applicable 
Subsection ISTD, ``Inservice Testing of Dynamic Restraints (Snubbers) 
in Light-Water Reactor Power Plants,'' of the ASME OM Code and site 
licensing and maintenance criteria, it should be justifiable to allow 
performance of this type of visual examination versus a VT-3 visual 
examination. The knowledge obtained from such snubber-specific training 
and experience commonly exceeds the VT-3 visual examination criteria 
for snubbers. While IWA-2317 of the 2003 Addenda through 2004 Edition 
of the ASME BPV Code, Section XI, provides alternative VT-3 examination 
qualification requirements, the administrative burden incurred for the 
VT-3 certification may not be commensurate with any convenience 
provided by qualifying additional VT-3 personnel in this manner and, 
for reasons stated previously, does not provide higher quality 
examinations. The commenter requested that the permissive for allowing 
personnel trained specifically on snubber requirements per the 
applicable ISTD and site licensing and maintenance criteria be allowed 
to perform visual examinations for snubbers as an alternative to 
performing a VT-3 examination per the method described in IWA-2213 of 
the ASME BPV Code, Section XI.
    NRC Response:
    The commenter requested that the visual examination method required 
by Sec.  50.55a(b)(3)(v) when performing examination and testing of 
snubbers be revised. The NRC declines to adopt the commenter's 
suggestion because the proposed rule did not suggest an amendment to 
the visual examination method in Sec.  50.55a(b)(3)(v), and the NRC 
currently does not have a basis for supporting such a revision. There 
were no other public comments received on Sec.  50.55a(b)(3)(v). 
Therefore, the NRC declines to adopt the commenter's suggestion. No 
change was made to Sec.  50.55a(b)(3)(v) in the final rule as a result 
of the comment.

8. 10 CFR 50.55a(g)(6)(ii)(B)--Containment ISI Programs

    Public Comments:
    In a letter dated June 19, 2007, Duke Energy stated that when 
compliance with the requirements of the ASME BPV Code, Section XI, 
Subsections IWE and IWL was initially imposed by 10 CFR 50.55a, the 
requirements of Sec.  50.55a(g)(6)(ii)(B) did not require licensees to 
submit ISI programs that were developed to comply with the Code during 
the expedited examination period (September 9, 1996, through September 
9, 2001). However, when the initial expedited examination requirements 
were removed from Sec.  50.55a after September 9, 2001, Sec.  
50.55a(g)(6)(ii)(B) was not deleted, leaving some licensees to believe 
that the NRC wanted to retain this provision. As a result, many 
licensees continue to believe that the NRC does not want updated 
containment ISI plans to be submitted. The NRC should take action to 
clarify whether it is the intent of 10 CFR 50.55a(g)(6)(ii)(B) that 
licensees be required to submit ISI plans for Class MC and Class CC 
components for all ISI plans developed after the expedited examination 
period.
    NRC Response:
    The NRC notes that the comment was not related to the proposed rule 
but to seek clarification on Sec.  50.55a(g)(6)(ii)(B) in the current 
regulation. It is the NRC's position to retain the current Sec.  
50.55a(g)(6)(ii)(B) provision in the final rule. Sec.  
50.55a(g)(6)(ii)(B) states that

[[Page 52734]]

licensees do not have to submit to the NRC for approval of their 
containment in-service inspection (CISI) programs for Class MC and 
Class CC pressure retaining components that were developed to meet the 
requirements of the ASME BPV Code, Section XI, Subsections IWE and IWL, 
with specified modifications and limitations, under Sec.  
50.55a(g)(5)(i) and/or Sec.  50.55a(g)(4). The provision requires that 
program elements and the required documentation of the developed plan 
must be maintained on site for audit. The provision applies to the CISI 
programs developed for each operating license for the initial 120-month 
inspection interval, including the CISI program revisions made by 
licensees of operating reactors during the September 1996 to September 
2001 timeframe (i.e., expedited examination period) when the rule for 
ASME BPV Code, Section XI, compliance was initially imposed. Further, 
the provision applies to subsequent revisions to the CISI programs for 
successive 120-month inspection intervals under Sec.  50.55a(g)(4)(ii). 
Therefore, as stated in Sec.  50.55a(g)(6)(ii)(B), licensees do not 
have to submit to the NRC for approval of their CISI program that meets 
the ASME Code, Subsections IWE and IWL with specified modifications and 
limitations after the expedited examination period.
    However, the NRC would like to clarify a situation which does not 
affect 50.55a(g)(6)(ii)(B) directly but which involves the use of 
Subsections IWE and IWL. If a licensee wishes to use Subsections IWE 
and IWL of later editions and addenda (i.e., later than the code of 
record for the ISI interval in question) of the ASME Code that are 
incorporated by reference in 10 CFR 50.55a(b) to be applied to the 
specific 10-year inservice inspection interval at its nuclear plant, 
the licensee needs to submit a request for the NRC's approval to use 
the later editions and addenda of the ASME Code. As stated in Sec.  
50.55a(g)(4)(iv), licensees are required to obtain NRC approval before 
using subsequent editions and addenda (or portions thereof) of the ASME 
BPV Code, Section XI, issued after their Code of Record for any 120-
month inspection interval, if they choose to implement their ISI 
programs under Sec.  50.55a(g)(4)(iv). The regulatory issue of using 
later editions and addenda of the Code has been previously clarified in 
NRC Regulatory Issue Summary 2004-12, ``Clarification on Use of Later 
Editions and Addenda to the ASME OM Code and Section XI.'' The intent 
of the commenter is to seek a clarification rather than a suggestion. 
Therefore, no change was made to Sec.  50.55a(g)(6)(ii)(B) in the final 
rule as a result of this comment.

9. 10 CFR 50.55a(g)(6)(ii)(D)--Reactor Vessel Head Inspections

9a. Condition 10 CFR 50.55a(g)(6)(ii)(D)(1), Regarding the 
Implementation of Code Case N-729-1, as Amended, in Lieu of the First 
Revised NRC Order EA-03-009
    Some commenters requested additional information on the 
implementation of these requirements, and asked the NRC about the 
process of changing the current NRC requirements for RPV closure head 
inspection requirements from the First Revised NRC Order EA-03-009, 
issued on February 20, 2004, (Order) to the requirements provided in 
the proposed rule language for 10 CFR 50.55a(g)(6)(ii)(D). (Comment 
Numbers 14, 19 and 20)
    NRC Response:
    To allow an orderly implementation of 10 CFR 50.55a(g)(6)(ii)(D), 
the NRC finds an implementation date of no later than December 31, 
2008, for the requirements provided in this section is warranted. The 
requirements of NRC Order EA-03-009 will remain in effect until the 
provisions of 10 CFR 50.55a(g)(6)(ii)(D) are implemented. Once a 
licensee implements this requirement, the First Revised NRC Order EA-
03-009 no longer applies to that licensee and under 10 CFR 
50.55a(g)(6)(D)(1) shall be deemed to be withdrawn. All relaxations 
from the requirements of the Order will then no longer apply. If a 
licensee cannot meet the proposed requirements of 10 CFR 
50.55a(g)(6)(ii)(D), then an alternative may be requested in accordance 
with 10 CFR 50.55a(a)(3)(i) or 10 CFR 50.55a(a)(3)(ii) or 
impracticality must be shown under 10 CFR 50.55a(g)(6)(i). To 
incorporate this implementation date, section 50.55a(g)(6)(ii)(D)(1) is 
revised to incorporate this implementation date.
9b. Condition 10 CFR 50.55a(g)(6)(ii)(D)(2), Regarding the Frequency of 
Reactor Vessel Head Inspection for ``Resistant'' Materials
    Public Comment:
    Some commenters disagreed with the proposed NRC position regarding 
the frequency of inspection of Item No. B4.40 of ASME Code Case N-729-
1. The commenters made several remarks regarding previous and ongoing 
laboratory work with primary water stress corrosion cracking (PWSCC) 
``resistant'' materials. Further, they noted operational experience 
with these materials had provided a sufficient basis to allow the 
inspection interval as stated in ASME Code Case N-729-1 without the 
NRC-proposed condition, as provided in proposed 10 CFR 
50.55a(g)(6)(ii)(D)(2). One commenter, number 13, recommended extending 
the interval of inspection from every seven (7) years to every eight 
(8) years. (Comment Numbers 7, 9, 11, 13, 15, 16, 17, 19, 21, 22 and 
23)
    NRC Response:
    During the writing of the proposed rule, the NRC disagreed with the 
NDE re-inspection frequency for ``resistant'' materials, in Item B4.40 
of Table 1 of ASME Code Case N-729-1, of every ten (10) calendar years 
beyond the first 10 years. Therefore, the NRC proposed the condition 10 
CFR 50.55a(g)(6)(ii)(D)(2) to limit the inspection frequency for 
``resistant'' materials to every four refueling outages not to exceed 
seven (7) calendar years beyond the first 10 years. The proposed 
condition was based on two main factors: the availability of limited 
crack initiation and growth data on the Alloy 152/52 weld metal, and 
the accelerated susceptibility increases of replaced U.S. RPV heads 
versus the current operational experience data from international 
experience which demonstrates the resistance of Alloy 690/152/52 
materials against PWSCC.
    The available data on Alloy 152/52 weld metal resistance to PWSCC 
is an NRC concern. However, considering the comments on this issue and 
ongoing PWSCC research programs at Pacific Northwest National 
Laboratories and Argonne National Laboratory sponsored by the NRC 
Office of Nuclear Regulatory Research, NRC now finds that the current 
data is sufficient to support the re-inspection frequency of Item B4.40 
of Table 1 of ASME Code Case N-729-1. NRC research on these materials 
is scheduled to continue through CY 2010. Accordingly, there should be 
enough time to address any items of concern regarding the resistance of 
these materials to PWSCC, if and when they develop, prior to becoming a 
significant safety issue.
    The NRC acknowledges that current operating experience shows the 
resistance of Alloy 152/52 weld material to PWSCC to be superior to 
that of Alloy 82/182. However, RPV head temperatures at numerous 
international plants with replaced RPV upper heads are significantly 
less than U.S. upper-head temperatures. As PWSCC susceptibility in 
nickel based alloys like Alloy 600 has been shown to have a significant 
temperature dependence, NRC analysis of international head replacement 
data has shown that RPV heads in the U.S. will, with time, have

[[Page 52735]]

a greater susceptibility to PWSCC than a majority of the international 
plants in terms of accumulated, effective degradation years. Therefore, 
NRC has found that long-term operating experience is limited for 
components that contain Alloy 690/52/152 materials with indications and 
repairs of the scope and nature found in recently replaced U.S. RPV 
heads. Nevertheless, the NRC finds the operational experience is 
sufficient to support Code Case N-729-1 inspection frequencies while 
research on these materials continues.
    The NRC agrees with the commenters and finds that there is 
sufficient Alloy 690/152/52 laboratory data and operational experience 
to allow the inspection frequency of Item B4.40 of Table 1 of ASME Code 
Case N-729-1 for RPV upper heads containing Alloy 690/152/52 
components. Therefore, the proposed condition in 10 CFR 
50.55a(g)(6)(ii)(D)(2) of the proposed rule will not be adopted.
9c. Condition 10 CFR 50.55a(g)(6)(ii)(D)(3), Regarding RPV Head 
Inspection Requirements and Frequencies
    Public Comment:
    Some commenters disagreed with the proposed NRC condition regarding 
the implementation of Note 6 of Table 1 of ASME Code Case N-729-1, 
which is stated in the 10 CFR 50.55a proposed rule language as 10 CFR 
50.55a(g)(6)(ii)(D)(3). Several comments were concerned with the 
surface and volumetric examination coverage requirements and the 
surface examination requirement of the J-groove weld. The commenters 
requested to allow a UT ``leak-path'' examination in lieu of surface 
examination of the J-groove weld, and that a note be added to document 
that Appendix I of the Code Case may be used when approved as required 
in 10 CFR 50.55a(g)(6)(ii)(D)(6). In addition comments noted that the 
impact of Note 9 is not addressed in the elimination of the original 
Code Case N-729-1, Note 6. (Comment Numbers 7, 9, 11, 12, 13, 16, 17, 
18, 19, 20, 22 and 23)
    NRC Response:
    In development of the proposed rule, the NRC did not find 
sufficient basis to allow an inspection regime of 3.0 re-inspection 
years (RIY) as described in Code Case N-729-1. Further, the NRC noted 
that due to the lack of a non-visual leak path assessment requirement 
in Code Case N-729-1, surface examination of all J-groove welds, 
commensurate with the volumetric examination of the penetration nozzle, 
should be required. Therefore the NRC proposed the condition in 10 CFR 
50.55a(g)(6)(ii)(D)(3). The NRC found the inspection coverage as 
defined by Code Case N-729-1 using the ASME Code definition of 
``essentially 100 percent'' inspection acceptable and therefore 
retained that language in the condition. No increase in inspection 
coverage is intended in the condition.
    The NRC disagrees that the supporting probabilistic basis is 
adequate to support the 3.0 RIY option. A probabilistic fracture 
mechanics analysis was used as a basis for the 3.0 RIY inspection 
frequency option. NRC finds the supporting probabilistic model is based 
on an assumption of essentially no cracking in RPV head penetrations or 
welds with less than 4 effective years of degradation (EDY). The NRC 
considers this assumption to be non-conservative as used in the 
supporting probabilistic model. One U.S. plant at approximately 2 EDY 
identified cracking attributable to PWSCC. Many of the other near-cold-
leg temperature RPV heads (cold-head plants) with susceptible material 
will not accumulate a total of 4 EDY through the next 15 to 30 years of 
operation. Development of flaws in these heads would cause adjustment 
of the probabilistic model output for all temperature ranges of RPV 
heads. Cracking attributed to PWSCC has been identified internationally 
in head penetration nozzles and associated welds at operating 
temperatures similar to U.S. cold-head plants. In the U.S., flaws in 
other components have been attributed to PWSCC in similar cold-leg 
temperature environments. The NRC finds that relatively few more 
instances of flaws attributed to PWSCC in the cold-head sub-population 
could significantly change the probabilistic model upon which the 3.0 
RIY inspection frequency is justified. Therefore, NRC concludes that 
the supporting probabilistic model does not provide an adequate basis 
for extending the non-visual NDE inspection frequency to 3.0 RIY.
    The conditional requirement for surface examinations of all J-
groove welds is based on the need for a defense-in-depth method to 
ensure reactor coolant pressure boundary integrity through the J-groove 
weld. In Code Case N-729-1, the mechanism to identify a through-weld 
flaw in a J-groove weld is through the bare-metal visual exam using 
visual leak detection at the top of the RPV head. This method alone is 
not consistent with previous NRC inspection requirements under the 
Order which require a non-visual leak path assessment in conjunction 
with a bare-metal visual examination of the RPV head. The NRC finds 
that not performing a leak path assessment would limit the ability of 
an inspection plan to provide sufficient defense-in-depth to identify 
leakage through the J-groove weld. In the past, the NRC has accepted 
ultrasonic (UT) leak path assessments as an adequate inspection to 
provide this assurance. However, the UT leak path assessment was not 
included in Code Case N-729-1 because it had not been qualified through 
the ASME Code process. Surface examination of the J-groove weld was 
included in Code Case N-729-1, but only as an option to increase 
inspection frequency. Under the proposed condition, performance of a 
surface examination of the J-groove weld would have been the only 
option in terms of a leak path assessment.
    The commenters stated that there are current plans to demonstrate 
the effectiveness of the ultrasonic leak path assessment technique for 
use within Code Case N-729-1. As the ultrasonic leak path assessment 
was a previously acceptable alternative to surface examination of the 
J-groove weld, due to physical constraints and radiological dose 
concerns in performing a surface exam in this area, the condition 
stated in 10 CFR 50.55a(g)(6)(ii)(D)(3) has been modified in this final 
rule.
    As noted previously the Condition stated in 10 CFR 
50.55a(g)(6)(ii)(D)(2) was removed. To address stakeholder comments 
about confusion between Notes 6 and 9 of Code Case N-729-1, condition 
in 10 CFR 50.55a(g)(6)(ii)(D)(2) of the proposed rule will simply state 
in the final rule that: ``Note 9 of ASME Code Case N-729-1 shall not be 
implemented.'' Note 9 of ASME Code Case N-729-1 provides the path for 
use of the 3.0 RIY inspection frequency interval. As previously stated, 
and as directed in the change to Note 6, the 3.0 RIY inspection 
frequency will not be included in the final rule.
9d. Condition 10 CFR 50.55a(g)(6)(ii)(D)(4), Regarding Qualification 
Requirements for Volumetric Inspection of RPV Head Penetration Nozzles
    Public Comment:
    Some commenters disagreed with the NRC-proposed condition regarding 
qualification requirements for volumetric examination as stated in 
Paragraph-2500 of ASME Code Case N-729-1. This proposed condition is 
stated in 10 CFR 50.55a(g)(6)(ii)(D)(4) of the proposed rule. (Comment 
Numbers 2, 7, 9, 11, 12, 13, 17, 19 and 22).
    NRC Response:
    The NRC notes that the condition stated in 10 CFR 
50.55a(g)(6)(ii)(D)(4)

[[Page 52736]]

requires that reliable and effective ultrasonic examinations be 
performed to ensure adequate protection for public health and safety. 
Because of the emphasis placed on inspections of the penetrations, it 
is appropriate to incorporate requirements for a robust blind 
demonstration of the ability of personnel, procedures and equipment to 
reliably detect and characterize indications, consistent with the 
approach articulated in Appendix VIII of Section XI of the ASME BPV 
Code. As RPV head inspection frequencies transition to every 8 or 10 
years due to replacement heads being installed, clearly defined 
performance demonstration requirements are necessary to ensure 
effective NDE. Due to the lack of current ASME BPV Code ultrasonic 
performance demonstration qualification requirements in Section XI, 
Appendix VIII, for RPV head penetrations, the NRC is adopting the 
conditions stated in 10 CFR 50.55a(g)(6)(ii)(D)(4) in the final rule.
    With respect to the performance demonstration requirements of the 
ASME BPV Code, Section XI, Appendix VIII, have increased the 
effectiveness and reliability of ultrasonic examinations, most notably 
in the area of inspection of dissimilar metal welds. The development of 
a qualification program to meet the intermediate rigor requirements of 
ASME BPV Code, Section V, Article 14 would require an additional 
process beyond this rulemaking activity. As noted in paragraph 10 CFR 
50.55a(g)(6)(ii)(C), implementation of performance demonstration 
requirements of Appendix VIII of Section XI of the ASME BPV Code is 
currently required by 10 CFR 50.55a for Supplements 1 through 8, 10 and 
11. At this time, there is no ASME BPV Code supplement to address 
performance demonstration requirements for the qualification of 
ultrasonic inspection of Alloy 600 base material. The conditions 
identified in the paragraphs 10 CFR 50.55a(g)(6)(ii)(D)(4)(i) through 
10 CFR 50.55a(g)(6)(ii)(D)(4)(iv) of the final rule are consistent with 
the performance demonstration requirements of Appendix VIII.
    10 CFR 50.55a(g)(6)(ii)(D)(4), as stated in the proposed rule, is 
modified in the final rule to incorporate an implementation date of 
September 1, 2009, in order to address the comment which noted that 
additional time would be required to fully implement a formalized 
qualification program. The implementation date in the final rule 
addresses the time necessary for mockup production and qualification of 
sufficient numbers of NDE personnel. NRC determined that the 
implementation date of September 1, 2009, is adequate to address the 
current frequency of inspections and allow for enough qualified 
personnel resources to be available. During the interval between the 
effective date of the final rule and the implementation date, the NRC 
finds that the qualification requirements of Code Case N-729-1 will 
provide reasonable assurance of public health and safety.
    With respect to the expansion of specimen qualification set 
applicability for a range of pipe diameters and thicknesses, 10 CFR 
50.55a(g)(6)(ii)(D)(4)(i) was modified. The commenters noted that 
current demonstrations are performed on typical-sized control rod drive 
mechanism penetration nozzles. These demonstrations are used for a 
variety of similar-sized penetration nozzles (incore instrumentation, 
control rod drive and control element drive) and for smaller-size and 
thickness vent-line nozzles. The proposed draft condition specimen set 
applicability range was taken from Section XI, Appendix VIII, 
Supplement 10 requirements for dissimilar metal welds. A change to 
increase the range of applicability was made to 10 CFR 
50.55a(g)(6)(ii)(D)(4)(i) to address stakeholder comments concerning 
the number of currently available mockup assemblies and the continued 
use of them for a slightly larger range of nozzles. The commenter noted 
that a small adjustment would allow the current mockups to be 
applicable for similar sized penetration nozzles which would fall just 
outside of the range stated in the proposed draft rule language. The 
NRC has reviewed the requested increased range of applicability and 
finds that the nozzles in question have enough through-wall thickness 
to provide similar response. As the weakness of ultrasonic examination 
is near field resolution, an expanded range for pipe diameters and 
thicknesses is allowed. The NRC finds that the range now stated in 10 
CFR 50.55a(g)(6)(ii)(D)(4)(i) of the final rule is adequate to ensure 
representative specimen sets will be used in the qualification 
processes for both personnel and procedures over the entire range of 
penetration nozzles in the reactor vessel head, and address stakeholder 
concerns.
    With respect to issues that recommended an adjustment for mockup 
specimens to include a range of blind demonstration mockups previously 
manufactured, 10 CFR 50.55a(g)(6)(ii)(D)(4)(ii) was modified for 
incorporation into the final rule. Specimen set flaw location 
requirements must meet several criteria to ensure the wide range of 
possible flaws identified through operational experience are captured 
for qualification of procedures, equipment, and personnel. The NRC has 
found that the commenters' flaw location range recommendations as 
stated in public comment viii of this section satisfactorily meet the 
intent of 10 CFR 50.55a(g)(6)(ii)(D)(4)(ii), which were established to 
ensure the entire range of flaws identified through operational 
experience are represented in the mockups. The NRC accepts the comments 
and, therefore, has modified the requirements of the condition stated 
in 10 CFR 50.55a(g)(6)(ii)(D)(4)(ii) for incorporation into the final 
rule.
    With respect to asking for additional clarity when an essential 
variable may be changed outside of its demonstration range, 10 CFR 
50.55a(g)(6)(ii)(D)(4)(iii) has been revised for incorporation into the 
final rule. The identification and definition of essential variables is 
necessary to ensure proper applicability of qualification standards to 
each particular inspection. 10 CFR 50.55a(g)(6)(ii)(D)(4)(iii) has been 
revised to include specific requirements if changes to essential 
variables occur. These requirements are the same as those required in 
Section XI, Appendix VIII general requirements of Subarticle VIII-2100 
which are required for use under 10 CFR 50.55a(g)(6)(ii)(C) for 
implementation of performance demonstration requirements of Appendix 
VIII of Section XI of the ASME BPV Code.
    With respect to the objection to the proposed generic qualification 
requirements for depth and length sizing qualification, noting that the 
requirements were currently unachievable for a generic procedure and 
were not necessary from a safety standpoint, 10 CFR 
50.55a(g)(6)(ii)(D)(4)(iv) has been revised for incorporation into the 
final rule. Performance demonstration requirements provide depth sizing 
and length sizing root mean square (RMS) error tolerances to meet the 
acceptance standards of Table VIII-S10-1. The NRC reviewed the RMS 
error tolerances that the commenters recommended, and found the 
proposed RMS error tolerances of \1/8\-inch (3 mm) in depth and \3/8\-
inch (10 mm) in length were adequate to ensure the validity of 
qualification. Therefore, for qualification of procedures, equipment, 
and personnel, the acceptance standard RMS error tolerance requirements 
were updated in 10 CFR 50.55a(g)(6)(ii)(D)(4)(iv) as incorporated into 
the final rule.

[[Page 52737]]

    After review and assessment of the comments, the NRC is revising 
the proposed condition.
9e. Condition 10 CFR 50.55a(g)(6)(ii)(D)(5), Regarding Re-inspection 
Requirements Once a Plant has Identified PWSCC Flaws in Their RPV Head 
Penetration Nozzles or Associated Welds
    Public Comment:
    Some commenters disagreed with the NRC proposed condition 10 CFR 
50.55a(g)(6)(ii)(D)(5). This condition requires a volumetric and/or 
surface re-inspection each outage once a plant identifies PWSCC in its 
vessel head penetration nozzles or welds. These commenters stated that 
flaw evaluation using the crack growth rates for PWSCC should provide 
an acceptable re-inspection interval for any flaws that were accepted 
by evaluation, and an exemption should be added to exclude the 
condition of ``craze cracking'' from mandating inspections at every 
outage. (Comment Numbers 7, 9, 11, 13, 17, and 19)
    NRC Response:
    The NRC disagrees with the commenters that flaw evaluation using 
the crack growth rates for PWSCC would provide an acceptable re-
inspection interval. The proposed condition stated in 10 CFR 
50.55a(g)(6)(ii)(D)(5) is based upon operating experience, and that 
several elements of PWSCC susceptibility (e.g., cold work, specific 
material properties, etc.) are not fully included in the susceptibility 
and probabilistic models of Code Case N-729-1. At least nine plants 
have identified flaws attributable to PWSCC in the refueling outage 
immediately following an inspection which identified the degradation 
mechanism. One plant identified at least four new flaws greater than 50 
percent through-wall in one operational cycle of crack growth. The NRC 
finds that operational experience has shown that not all factors 
affecting the susceptibility of Alloy 600 materials are included within 
a standard flaw analysis model using the ASME BPV Code flaw analysis 
using the Alloy 600 crack growth rate identified in Subarticle IWB-3660 
of Section XI of the ASME BPV Code.
    The ASME BPV Code crack growth rate curve for Alloy 600 is a mean 
of the upper 50 percent of all acceptable Alloy 600 laboratory 
developed crack growth rate data points. It is not a bounding crack 
growth curve. Testing on field samples of Alloy 600 from the replaced 
RPV head of one plant by Argonne National Laboratories identified a 
crack growth rate which is at the upper bound (95th percentile) of the 
data used to develop the ASME curve. Additional factors may affect the 
initiation and growth of PWSCC in RPV upper head penetrations which 
were not fully analyzed in the laboratory tested material. These 
factors include the welding process, heats of material, and cold work 
applied in the field or during manufacturing conditions.
    If a plant is found to have a flaw attributable to PWSCC, the flaw 
may have developed due to any one or a combination of the previously 
mentioned susceptibility factors. Therefore, the plant may not be fully 
bounded by the Code Case N-729-1 PWSCC model. The model provides 
appropriate inspection frequencies to ascertain when a plant develops 
PWSCC in its RPV upper head penetrations. However, to be conservative, 
the plant should perform volumetric and/or surface examinations for 
each outage to provide reasonable assurance of the integrity of the 
reactor coolant pressure boundary and prevent leakage once conditions 
for PWSCC have been verified through inspection results. As such, the 
NRC's proposed condition is that once a plant has identified a flaw 
attributable to PWSCC in a RPV head penetration or J-groove weld, that 
plant should perform visual and volumetric and/or surface examinations 
for each outage. This is consistent with NRC Order EA-03-009. 
Therefore, the proposed provisions in 10 CFR 50.55a(g)(6)(ii)(D)(5) are 
adopted without change in the final rule.
    Indications of craze cracking have not previously been 
characterized as indications of PWSCC, and the NRC continues to find 
that indications of craze cracking are not PWSCC. Therefore, if a 
licensee determines that the indications in a vessel head penetration 
nozzle are a result of craze cracking alone, it would not be within the 
scope of proposed condition stated in 10 CFR 50.55a(g)(6)(ii)(D)(5).
9f. Condition 10 CFR 50.55a(g)(6)(ii)(D)(6), Regarding the Allowance of 
Licensee Deviation from the Requirements of ASME Code Case N-729-1 
Without NRC Review and Approval Public Comments
    Commenters disagreed with the NRC-proposed condition for use of 
Appendix I of ASME Code Case N-729-1, which is stated in 10 CFR 
50.55a(g)(6)(ii)(D)(6). The comments concerned the following items:
     It is not the place of the ASME BPV Code to require 
utilities to get NRC approval on acceptable alternatives.
     NRC review of industry implementation of Appendix I of 
Code Case N-729-1 relief from the requirements of ASME Code Case N-729-
1 is unnecessary.
     An exemption should be made for the need for NRC approval 
for use of Appendix I of Code Case N-729-1 by plants with new heads 
that use ``resistant'' material, until PWSCC is identified in those 
heads.
    (Comment Numbers 7, 12, 13, 17 and 19)
    NRC Response:
    Appendix I of Code Case N-729-1 gives an analysis procedure that 
allows licensees to demonstrate the adequacy of an NDE zone of coverage 
less than that required by Code Case N-729-1. Implementation of this 
analysis procedure does not require NRC review and approval. In 
essence, Appendix I would allow licensees to self-approve relief from 
the requirements of Code Case N-729-1, essentially usurping NRC's 
authority under 10 CFR 50.55a to evaluate alternatives. NRC experience 
in processing relaxation requests to Order requirements has shown that 
there was significant variation in technical basis approaches between 
licensees in proposing alternatives to the Order. For example, 
probabilistic analyses were used in licensee relaxation requests from 
Order requirements that the NRC found to have insufficient basis and 
therefore did not approve as a basis for relaxation. However, under 
Appendix I of Code Case N-729-1, these relaxation requests could be 
found acceptable without NRC review. While the NRC agrees that the 
methods provided in Appendix I may be used as a basis to request relief 
from the ASME Code Case requirements, NRC review and approval shall be 
required for deviations from Code Case N-729-1 examination coverage 
requirements.
    The NRC disagrees with the comment that excludes from this proposed 
condition new reactor vessel heads that use resistant material, until 
PWSCC is identified in these heads. The NRC notes that the flaw 
evaluation tools and susceptibility of new PWSCC resistant materials 
have not been established or approved by the NRC. As such, 
implementation of Appendix I of Code Case N-729-1 would be open to 
significant variation of interpretation. Therefore, the provisions in 
10 CFR 50.55a(g)(6)(ii)(D)(6) are adopted without change in the final 
rule.
9g. General Public Comments on 10 CFR 50.55a(g)(6)(ii)(D)
    Two commenters (comment numbers 8 and 11) stated that Public Law, 
PL 104-113, mandates that national consensus standards be used by 
Federal agencies where applicable. This

[[Page 52738]]

includes the use of ASME codes and standards. Because the consensus 
process used to develop the Code Case specifically considered the NRC 
comments (i.e., additional conditions being added with this rule 
change) and found them to be without technical merit, one commenter 
considered it inappropriate for NRC to impose additional conditions on 
the use of Code Case N-729-1. Therefore, the commenter requested that 
the additional conditions be removed from the rule language. 
Alternatively, if the additional conditions would not be removed from 
the rule language, the technical justifications for the need for these 
additional conditions should be included in the supplemental 
information for the final rule.
    NRC Response:
    NRC review of ASME Code Case N-729-1 concludes that its basis 
implies that leakage is acceptable as long as ejection and structural 
integrity due to wastage isn't likely to occur. All of the RPV head 
penetration and associated weld examinations required by the NRC to 
date, have been based on assuring an extremely low probability of 
leakage from these components as well as assuring their structural 
integrity. NRC's position for reactor pressure vessel upper head 
inspections is that if an active degradation mechanism is present, any 
long term inspection plan should be based on assuring an extremely low 
probability of abnormal leakage rather than allowing leakage and 
demonstrating the acceptability of its consequences. Consistent with 
this position, the NRC sets the conditions regarding the use of ASME 
Code Case N-729-1 in order to incorporate its use, by reference, into 
the Code of Federal Regulations. The technical justifications for the 
need for these conditions are included in the public comment section of 
this rulemaking activity.

10. 10 CFR 50.55a(g)(6)(ii)(E)--Reactor Coolant Pressure Boundary 
Visual Inspections

    Public Comment:
    In a letter dated June 19, 2007, Progress Energy stated that the 
ASME has not amended Section XI of the BPV Code to include Code Case N-
722. Therefore, requiring licensees to comply with a Code Case that has 
not been incorporated into the ASME Code sets a precedence of mandatory 
implementation of a Code Case which has not been subject to ASME public 
review and comment during its development.
    NRC Response:
    The NRC recognizes that the ASME has not amended Section XI of the 
ASME BPV Code to include Code Case N-722 and that during development 
code cases may be subjected to different ASME public review and comment 
than Section XI. The NRC is incorporating Code Case N-722 in the rule 
to expedite the implementation of Code Case N-722. The NRC is requiring 
expedited implementation of Code Case N-722 because the NRC concluded 
from a safety perspective that these inspections are necessary to 
ensure the integrity of the Alloy 600/82/182 components. The NRC has 
previously incorporated code cases in 10 CFR 50.55a prior to the ASME 
taking action to include the code cases in the ASME Code. The NRC 
declines to adopt commenter's suggestion. No change was made to the 
final rule as a result of this comment.
    Public Comment:
    In a letter dated June 22, 2007, Southern Nuclear Operating Company 
stated that the NRC does not reference the industry efforts, especially 
those made through the Electric Power Research Institute's Materials 
and Reliability Program (MRP) to address the issue of bare-metal visual 
examination of Alloy 600 welds. Every PWR in the United States has 
agreed to the implementation of MRP-139, which requires an augmented 
program to perform bare-metal visual examinations on the large diameter 
Alloy-600 welds on a frequency that is almost identical to the schedule 
mandated in ASME Code Case N-722. Typically, utilities are given the 
option to assess each code case and determine if that code case should 
be adopted for use. By mandating the use of Code Case N-722, the NRC 
is, in effect, writing their own code and deviating from using guidance 
from an international consensus standard body (ASME Code Committees, of 
which the NRC is a participant and voting member). The NRC and the 
industry have been working on this issue, and industry programs are in 
place to cover these examinations. Additional time should be provided 
to allow the MRP and ASME to develop the necessary enhancements.
    NRC Response:
    The MRP-139 report referenced by the commenter is an industry 
guidance document which includes guidance on bare-metal visual 
examinations of Alloy 82/182 butt welds. Because MRP-139 is written as 
inspection guidance, MRP-139 is not suitable to be incorporated by 
reference in 10 CFR 50.55a. In addition, the MRP has not issued 
inspection guidelines for partial-penetration welded components with 
Alloy 600/82/182 materials. The NRC finds Code Case N-722 with 
conditions is suitable to be incorporated by reference in the final 
rule. Given the safety significance of these inspections, the NRC 
concluded that the reactor coolant pressure boundary visual inspections 
of 10 CFR 50.55a(g)(6)(ii)(E) are necessary to ensure that the 
appropriate safety-significant visual inspections are performed.
    The NRC recognizes that the ASME is an international, consensus 
standard body, and that the ASME Code provides necessary requirements 
for the design and inspection of nuclear power plant components. 
Therefore, the NRC has incorporated by reference in 10 CFR 50.55a 
certain editions and addenda of Section III and XI of the ASME BPV 
Code. However, in certain cases, such as when an active degradation 
mechanism is affecting the integrity of pressure boundary components, 
the NRC needs to take regulatory actions to ensure safety and protect 
the public health and safety. As mandated by the Atomic Energy Act of 
1954, as amended, and the Energy Reorganization Act of 1974, the NRC 
has the statutory authority and responsibility to enact regulations 
through the rulemaking process as necessary to ensure safety.
    The NRC declines to adopt commenter's suggestion. No change was 
made to the final rule as a result of this comment.
    Public Comment:
    In a letter dated June 20, 2007, Arizona Public Service Company 
stated that 10 CFR 50.55a(g)(6)(ii)(E)(1) exempts Alloy 600/82/182 
materials that have been mitigated by weld overlay or stress 
improvement from the inspection requirements of Code Case N-722. The 
commenter recommended that nozzles and penetrations that have been 
mitigated by half-nozzle replacement or Alloy 690/52/152 weld pads 
should also be exempted from the requirements of Code Case N-722.
    NRC Response:
    Code Case N-722, as implemented by 10 CFR 50.55a(g)(6)(ii)(E), 
applies to examination of pressure retaining partial or full 
penetration welds in Class 1 components fabricated with Alloy 600/82/
182 material in PWRs. The requirements of Code Case N-722, as 
implemented by 10 CFR 50.55a(g)(6)(ii)(E), applies to nozzles and 
penetrations that have Alloy 600/82/182 materials that form the 
pressure boundary. This requirement is clear from the title and wording 
of Code Case N-722. Note the clarification in the preceding sentences 
applies even though Alloy 600/82/182 materials may not be entirely 
removed from the component, provided that pressure retaining 
penetrations and welds no

[[Page 52739]]

longer contain Alloy 600, Alloy 82, or Alloy 182 materials. In 
addition, 10 CFR 50.55a(g)(6)(ii)(E)(1) is revised in the final rule.
    Public Comment:
    In a letter dated June 20, 2007, Jack Spanner of Electric Power 
Research Institute stated that with respect to 10 CFR 
50.55a(g)(6)(ii)(E)(2), it should be sufficient to demonstrate the 
ability to characterize location, orientation and length of cracks with 
calibration blocks or mockups containing a notch in the axial and 
circumferential orientation.
    NRC Response:
    The requirements of paragraph (g)(6)(ii)(E)(2) state only that 
additional actions must be taken to characterize the location, 
orientation, and length of cracks. The comment does not provide 
sufficient information for the NRC to respond regarding the adequacy of 
calibration blocks or mockups to meet these requirements. Therefore, 
the NRC declines to adopt the commenter's suggestion. No change was 
made to the final rule as a result of this comment.
    Public Comment:
    In a letter dated June 20, 2007, Arizona Public Service Company 
recommended that the term ``Non-visual NDE'' used in paragraph 
(g)(6)(ii)(E)(3) be changed to ``surface'' or ``volumetric'' 
examination.
    NRC Response:
    The ASME Code, Section XI, paragraph IWA-2200 states that ``three 
types of examinations used during inservice inspection are defined as 
visual, surface, and volumetric.'' It is clear from this Code 
definition that non-visual examination refers to either surface or 
volumetric examination. The NRC declines to adopt the commenter's 
suggestion. No change was made to the final rule as a result of this 
comment.
    Public Comment:
    In a letter dated June 20, 2007, Arizona Public Service Company 
stated that paragraph (g)(6)(ii)(E)(4) imposes the rule of Appendix 
VIII of the ASME Code, Section XI, to components where qualification 
may not have been performed (possibly due to size and thickness). 
Therefore, the commenter recommended that because the component causing 
the implementation of this paragraph is leaking, the NDE method and 
techniques utilized to characterize the leak in paragraph 
(g)(6)(ii)(E)(2) should be sufficient qualification.
    NRC Response:
    The commentor believes that paragraph (g)(6)(ii)(E)(4) is 
unnecessary and suggests that the NDE method and techniques utilized to 
characterize the leak in (g)(6)(ii)(E)(2) be sufficient [NDE] 
qualification. The NRC disagrees with the commentor's suggestion. 
Paragraph (g)(6)(ii)(E)(2) requires that when leakage is detected in a 
component, additional action (e.g., non-visual examination) must be 
performed to characterize the location, orientation, and length of 
cracks that cause the leakage. Paragraph (g)(6)(ii)(E)(2) does not 
provide specific qualification for NDE. The intent of Paragraph 
(g)(6)(ii)(E)(2) is to provide a general requirement for non-visual 
examinations to be performed should leakage be detected. The NDE method 
and techniques utilized to characterize the leak in paragraph 
(g)(6)(ii)(E)(2) are visual examinations which cannot characterize flaw 
sizes.
    Paragraph (g)(6)(ii)(E)(4) requires that the ultrasonic examination 
be performed using the appropriate supplement of Section XI, Appendix 
VIII of the ASME Code. The intent of paragraph (g)(6)(ii)(E)(4) is to 
provide specific NDE qualification requirements for ultrasonic 
examination for Alloy 600/82/182 butt welds so that the requirements of 
paragraphs (g)(6)(ii)(E)(2) or (g)(6)(ii)(E)(3) can be satisfied.
    This position is consistent with other provisions of 10 CFR 50.55a 
in that ultrasonic examination of butt welds must be qualified in 
accordance with the appropriate supplement of Section XI, Appendix VIII 
of the ASME Code. Therefore, the NRC declines to adopt the commenter's 
suggestion. No change was made to the final rule as a result of this 
comment.
    Public Comment:
    After the public comment period closed, the NRC received an 
additional comment from Florida Power and Light Company via a phone 
call on July 8, 2008, regarding the schedule for implementing the 
initial inspections under Code Case N-722 as required by 10 CFR 
50.55a(g)(6)(ii)(E), Reactor coolant pressure boundary visual 
inspections. The commenter pointed out that Code Case N-722 specifies 
frequency of examination for each part to be examined but does not 
specify when the initial inspections shall be performed. The commenter 
recommended that the schedule for the initial inspections be specified 
in the rule.
    NRC Response:
    The NRC agrees with the commenter that the schedule for the initial 
inspections is not specified in Code Case N-722 nor is it specified in 
a NRC-proposed condition applicable to this Code Case. Code Case N-722 
contains three different inspection intervals: inspections to be 
conducted every other refueling outage, each refueling outage, and once 
per interval. The NRC has specified the following initial inspection 
requirements in a new footnote to the new paragraph.
    For inspections to be conducted every refueling outage and 
inspections conducted every other refueling outage, the initial 
inspection shall be performed at the next refueling outage after 
January 1, 2009. For inspections to be conducted once per interval, the 
inspections shall begin in the interval in effect on January 1, 2009, 
and shall be prorated over the remaining periods and refueling outages 
in this interval. For inspections to be conducted once per interval, if 
the current interval ends prior to January 1, 2009, the initial 
inspection shall be performed at the first refueling outage after 
January 1, 2009. These initial inspection schedules are believed to be 
reasonable since, in general, the inspections are straightforward to 
perform and licensees have been aware for over two years of the NRC 
intent to incorporate Code Case N-722 in the regulations during which 
to plan the inspections.

III. Section-by-Section Analysis

ASME BPV Code, Section III

10 CFR 50.55a(b)(1)
    The final rule revises Sec.  50.55a(b)(1) in the current regulation 
to incorporate by reference the 2004 Edition of Section III, Division 
1, of the ASME BPV Code into 10 CFR 50.55a. This paragraph requires new 
applicants for a nuclear power plant who submit an application for a 
construction permit under 10 CFR part 50 after the effective date of 
this rule use the 2004 Edition of Section III, Division 1 of the ASME 
BPV Code for the design and construction of the reactor coolant 
pressure boundary and Quality Group B and C components. This paragraph 
also requires that existing modifications and limitations for weld leg 
dimensions, independence of inspection and subsection NH in Sec. Sec.  
50.55a(b)(1)(ii), 50.55a(b)(1)(v), and 50.55a(b)(1)(vi), respectively, 
apply to the 2004 Edition of Section III, Division 1 of the ASME BPV 
Code. The NRC is not adopting any additional limitations with respect 
to the 2004 Edition of Section III.
10 CFR 50.55a(b)(1)(iii)--Seismic Design of Piping
    As discussed in Section II of this document, applicants or 
licensees may use Articles NB-3200, NB-3600, NC-3600, and ND-3600 for 
seismic design of piping up to and including the 1993 Addenda, subject 
to the limitation specified in paragraph (b)(1)(ii) of this section. 
Applicants or licensees may not use these Articles for seismic design 
of

[[Page 52740]]

piping in the 1994 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (b)(1) of this section. The 
final rule revises 50.55a(b)(1)(iii) in the current 10 CFR 50.55a to 
clarify the current limitation regarding seismic design. Current Sec.  
50.55a(b)(1)(iii) states that applicants or licensees may use Articles 
NB-3200, NB-3600, NC-3600, and ND-3600 for seismic design. However, the 
rules in Article NB-3200 of Section III of the ASME BPV Code contain 
criteria applicable to the seismic design of components other than 
piping systems. The NRC revises Sec.  50.55a(b)(1)(iii) to clarify that 
the limitation only applies to the seismic design of piping.

ASME BPV Code, Section XI

    The final rule revises Sec.  50.55a(b)(2) to incorporate by 
reference the 2004 Edition of the ASME BPV Code, Section XI, Division 
1, subject to the modifications and limitations discussed in the 
following paragraphs:
10 CFR 50.55a(b)(2)(xi)--Class 1 Piping
    Paragraph 50.55a(b)(2)(xi) states that ``licensees may not apply 
IWB-1220, ``Components Exempt from Examination,'' of Section XI, 1989 
Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(2) of this section, and shall apply IWB-
1220, 1989 Edition.'' Subarticle IWB-1220 of the 1989 Edition of the 
ASME BPV Code, Section XI, exempts certain components (such as small 
bore piping) from the volumetric and surface examinations. However, 
welds or portions of welds that are inaccessible due to being encased 
in concrete, buried underground, located inside a penetration, or 
encapsulated by guard pipe were included in components for exemption 
from examination and incorporated in the edition and addenda of the 
ASME BPV Code, Section XI, after the 1989 Edition. The NRC previously 
did not agree with the incorporation of these types of welds for 
exemption from examination because the NRC believed that these welds 
should be examined to monitor their structural integrity. Therefore, 
the NRC prohibited the use of 1989 addenda through the latest editions 
and addenda of the ASME BPV Code, Section XI, regarding the application 
of IWB-1220 in 10 CFR 50.55a(b)(2)(xi) (64 FR 51394; September 22, 
1999).
    The revision to the regulation removes 10 CFR 50.55a(b)(2)(xi), 
thereby permitting the use of ASME BPV Code, Section XI, IWB-1220 of 
any edition or addenda of ASME BPV Code, Section XI, incorporated by 
reference in 10 CFR 50.55a. The condition placed upon Section XI, IWB-
1220 in 10 CFR 50.55a(b)(2)(xi) is no longer necessary because of the 
following:
    1. Licensees can select an alternate weld for inspection that does 
not have limitations.
    2. Licensees have committed to perform augmented inspections of 
break exclusion zone (BEZ) welds which are located in inaccessible 
areas such as containment penetrations or encapsulated by guard pipe to 
the extent practical under the BEZ criteria.
    3. Boiling water reactor (BWR) licensees have followed the 
provisions of Generic Letter 88-01, ``NRC Position on IGSCC 
[intergranular stress corrosion cracking] in BWR Austenitic Stainless 
Steel Piping,'' and the associated NRC report, NUREG-0313, ``Technical 
Report on Material Selection and Process Guidelines for BWR Coolant 
Pressure Boundary Piping,'' and the provisions of the BEZ criteria 
(Reference: Branch Technical Position MEB 3-1 attached to Standard 
Review Plan 3.6.2) apply to the examination of the welds such as those 
that are located inside containment penetrations or encapsulated by 
guard pipe.
    4. Licensees of plants whose construction permits were issued after 
January 1, 1971, are required to have ASME Class 1 and Class 2 
components designed and provided with access to enable the performance 
of ISIs, and the removal of the limitation on the use of IWB-1220(d) 
would not permit welds to be located in reactor coolant pressure 
boundary components (including Class 1 components permitted to be 
designed to Class 2 rules) that are encased in concrete, buried 
underground, located inside a penetration, or encapsulated by guard 
pipe.
10 CFR 50.55a(b)(2)(xiii)--Mechanical Clamping Devices
    Paragraph 50.55a(b)(2)(xiii) is removed from the regulation. This 
paragraph permitted licensees to use the provisions of Code Case N-523-
1, ``Mechanical Clamping Devices for Class 2 and 3 Piping.'' Instead, 
Code Case N-523-2 provides updated requirements to those of Code Case 
N-523-1, has been accepted in Regulatory Guide (RG) 1.147, Revision 15, 
``Inservice Inspection Code Case Acceptability, ASME BPV Code, Section 
XI, Division 1,'' and Revision 15 is incorporated by reference into 10 
CFR 50.55a(g)(4)(i) and 10 CFR 50.55a(g)(4)(ii). Therefore, 10 CFR 
50.55a(b)(2)(xiii) no longer serves any useful purpose and is removed.
10 CFR 50.55a(b)(2)(xv)--Appendix VIII Specimen Set and Qualification 
Requirements
    Paragraph 50.55a(b)(2)(xv) in the current 10 CFR 50.55a regulation 
specifies provisions that may be used to modify implementation of 
Appendix VIII of Section XI, 1995 Edition through the 2001 Edition of 
the ASME BPV Code with regard to ultrasonic examinations of piping 
systems. The change specifies that licensees who have been approved by 
the NRC to use later editions and addenda than the 2001 Edition of the 
ASME BPV Code shall use the 2001 Edition of Appendix VIII. Licensees 
cannot use Appendix VIII to the editions and addenda of the ASME Code 
Section XI that are later than the Appendix VIII to 2001 Edition.
10 CFR 50.55a(b)(2)(xx)--System Leakage Tests
    10 CFR 50.55a(b)(2)(xx) in the current 50.55a regulation requires 
certain hold time when performing system leakage tests in accordance 
with IWA-5213(a) of the 1997 through 2002 addenda of the ASME Code 
Section XI. Since the publication of the current 10 CFR 50.55a, the NRC 
has noticed an NDE issue that involves the system leakage tests when 
performed in accordance with IWA-4540(a). 10 CFR 50.55a(b)(2)(xx) is 
revised to address the NDE issue. The requirements in current 10 CFR 
50.55a(b)(2)(xx) are not changed. The revised 10 CFR 50.55a(b)(2)(xx) 
provides new requirements. The revision requires, as part of repair and 
replacement activities (by welding or brazing under the 2003 Addenda 
through the latest edition and addenda incorporated by reference in 10 
CFR 50.55a(b)(2)), that NDE be performed in accordance with subarticle 
IWA-4540(a)(2) of the 2002 Addenda of the ASME BPV Code, Section XI, 
after a system leakage test is performed per subarticle IWA-4540(a)(2) 
of the 2003 Addenda through later editions and addenda of the ASME BPV 
Code, Section XI. This provision requires that after repair or 
replacement activities (1) the NDE method and acceptance criteria of 
the 1992 Edition, or later, of Section III be performed and met prior 
to returning the system to service, and that (2) a system leakage test 
be performed in accordance with IWA-5000 prior to, or as part of, 
returning the system to service.
10 CFR 50.55a(b)(2)(xxi)(A)--Table IWB-2500-1 Examination Requirements
    Paragraph 10 CFR 50.55a(b)(2)(xxi)(A) in the current 50.55a 
regulation allows the use of the visual examination with

[[Page 52741]]

enhanced magnification in lieu of an ultrasonic examination. Because of 
the latest development in visual examination requirements in the ASME 
Code, Paragraph 10 CFR 50.55a(b)(2)(xxi)(A) is revised to be consistent 
with the condition for Code Case N-648-1, ``Alternative Requirements 
for Inner Radius Examination of Class I Reactor Vessel Nozzles, Section 
XI, Division 1.'' in RG 1.147, Revision 15, which requires the 
assumption of a limiting flaw aspect ratio when using the allowable 
flaw length criteria in Table IWB-3512-1 during an enhanced visual 
examination. The revision states ``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) in 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.'' The 
limitation on the flaw aspect ratio is needed because visual 
examination cannot determine the depth of cracks. A visual examination 
requirement may be applied only when a limiting flaw aspect ratio of 
0.5 is assumed. A flaw aspect ratio of less than 0.5 would not be 
conservative. As shown in Table IWB-3512-1, there are no flaw aspect 
ratios higher than 0.5. Therefore, assuming a flaw aspect ratio of 0.5 
is appropriate.
10 CFR 50.55a(g)(6)(ii)(A)--Augmented Examination of Reactor Vessel
    Paragraph 50.55a(g)(6)(ii)(A) is removed from the regulation. This 
paragraph required a one-time, augmented ISI program for those systems 
and components the Commission determined that added assurance of 
structural reliability was necessary. Paragraph 50.55a(g)(6)(ii)(A) was 
incorporated in the regulations in 1992 to require all current 
licensees to conduct a one-time, expedited examination of reactor 
vessel shell welds. Examination requirements were specified in item 
B1.10, ``Shell Welds,'' of Examination Category B-A, ``Pressure 
Retaining Welds in Reactor Vessel,'' in Table IWB-2500-1, ``Examination 
Categories'' of the 1989 Edition of the ASME BPV Code, Section XI, 
Division 1. Because all the licensees have completed the subject 
augmented examination of the reactor vessel shell welds, the 
requirements in 10 CFR 50.55a(g)(6)(ii)(A) and associated subparagraphs 
are no longer needed. Future licensees need not conduct this augmented 
examination, because new Code provisions should adequately address the 
degradation to which the augmented examination was directed.
10 CFR 50.55a(g)(6)(ii)(D)--Reactor Vessel Head Inspections
    On September 30, 2002, the Davis-Besse Lessons Learned Task Force 
(LLTF) issued a report containing 51 recommendations for actions that 
the NRC should take to address areas that the LLTF considered 
contributors to the Davis-Besse event. On November 26, 2002, the senior 
NRC management review team endorsed all but two of the task force's 
recommendations. One endorsed high-priority recommendation was the 
following:

    The NRC should encourage American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code (ASME Code) requirement 
changes for bare metal inspections of nickel based alloy nozzles for 
which the code does not require the removal of insulation for 
inspections. The NRC should also encourage ASME Code requirement 
changes for the conduct of non-visual non-destructive examination 
(NDE) inspections of VHP [vessel head penetration] nozzles. 
Alternatively, the NRC should revise Title 10 Code of Federal 
Regulations (10 CFR) Part 50.55a to address these areas.

    Section XI of the ASME Code, which is incorporated by reference 
into NRC regulations by 10 CFR 50.55a, ``Codes and standards,'' 
currently specifies that inspections of the reactor pressure vessel 
(RPV) head need only include a visual check for leakage on the 
insulated surface or surrounding area. Experience has shown that these 
inspections may not detect small amounts of leakage from an RPV head 
penetration with cracks extending through the nozzle or the J-groove 
weld. Such leakage can create an environment that leads to 
circumferential cracks in RPV head penetration nozzles and/or corrosion 
of the RPV head.
    The NRC issued Order EA-03-009, ``Interim Inspection Requirements 
for Reactor Pressure Vessel Heads at Pressurized Water Reactors,'' 
dated February 11, 2003, which modified licensees' licenses to require 
specific inspections of the reactor pressure vessel head and associated 
penetration nozzles at pressurized water reactors. In September 2003, 
industry representatives through the Materials Reliability Program 
provided industry input to support industry alternative inspection 
programs through various public meetings and MRP-95, ``Materials 
Reliability Program: Generic Evaluation of Examination Coverage 
Requirements for the Reactor Pressure VHP Nozzles, (ML032740424).'' In 
response to internal review and stakeholder input, the NRC issued First 
Revised Order EA-03-009, February 20, 2004 (Order), which refined the 
inspection requirements of NRC Order EA-03-009 by taking into account 
lessons learned from inspections performed from February 2003 to 
January 2004.
    On July 7, 2004, after an assessment which concluded that ASME Code 
requirement revisions would not be complete in 2004, the NRC issued a 
Commission Paper (SECY-04-0115) requesting Commission approval of a 
rulemaking plan to incorporate into 10 CFR 50.55a the RPV head and 
associated head penetration inspection requirements contained in the 
Order.
    The Commission, in a Staff Requirements Memorandum, dated August 6, 
2004, approved an alternative option to evaluate the RPV inspection 
requirements of an upcoming ASME Code Case or revision of the ASME Code 
for incorporation into 10 CFR 50.55a.
    In March 2006, the ASME approved Code Case N-729-1, Alternative 
Examination Requirements for PWR Reactor Vessel Upper Heads With 
Nozzles Having Pressure-Retaining Partial-Penetration Welds, which 
provides an alternative long-term inspection program for RPV upper 
heads. The NRC participated in ASME Code development and approval of N-
729-1. The NRC has reviewed the final version of Code Case N-729-1, and 
with conditions, finds it provides reasonable assurance of public 
health and safety from failure of the reactor pressure vessel upper 
head and penetration nozzles. Therefore, the NRC is pursuing this 
rulemaking activity to incorporate by reference the inspection 
requirements of Code Case N-729-1, as conditioned, into 10 CFR 50.55a.
    The experience of the Davis-Besse RPV head degradation and the 
discovery of leaks and nozzle cracking at other plants over the past 
seven years reinforce the need for effective regulatory required 
inspections of the RPV head and penetration nozzles. The absence of an 
effective inspection regime could, over time, result in unacceptable 
circumferential cracks in RPV head penetration nozzles or in the 
degradation of the RPV head by

[[Page 52742]]

corrosion from leaks in the reactor coolant pressure boundary. These 
degradation mechanisms increase the probability of a loss of reactor 
coolant pressure boundary event through ejection of a nozzle or other 
rupture of the RPV head. The result of this rulemaking would be the 
establishment of inspection requirements that result in an extremely 
low probability of abnormal leakage, of rapidly propagating failure and 
of gross rupture of the reactor pressure vessel head and penetration 
nozzles.
    The Code Case N-729-1 inspection plan for RPV upper heads with 
Alloy 600/182/82 penetration nozzles requires periodic bare metal 
visual (BMV) examinations and periodic nonvisual examinations using 
ultrasonic testing (UT), eddy current testing (ET), or dye penetrant 
testing of the penetration nozzle base metal. BMV examinations are 
performed in order to identify primary coolant leakage based on the 
presence of boric acid deposit accumulations. Nonvisual examinations 
are performed in order to identify flaws which could lead to leakage or 
failure of the penetration nozzle.
    These same inspections are required to be performed for RPV upper 
heads with Alloy 690/152/52 penetration nozzles, but the frequency of 
inspection is greatly reduced. This reduction is due to the enhanced 
resistance these materials have demonstrated against PWSCC.
    Paragraph 50.55a(g)(6)(ii)(D) is added to the regulation to require 
licensees to comply with the reactor vessel head inspection 
requirements of ASME Code Case N-729-1, subject to conditions, by 
December 31, 2008. Compliance to Code Case N-729-1; with conditions 
regarding inspection frequency, examination coverage, qualification of 
ultrasonic examination, and re-inspection intervals; would be 
equivalent to complying with NRC Order EA-03-009, dated February 11, 
2003, and First Revised Order EA-03-009, dated February 20, 2004. Thus, 
once a licensee implements Code Case N-729-1, with conditions, the 
First Revised NRC Order EA-03-009 no longer applies to that licensee 
and is deemed to be withdrawn. This allows licensees to transfer from 
the Order requirements to the requirements of 10 CFR 
50.55a(g)(6)(ii)(D).
    Footnote 10 to 10 CFR 50.55a(b)(2) is removed because Code Case N-
729-1, as conditioned, replaces the requirements of the NRC Order EA-
03-009 cited in that footnote.
10 CFR 50.55a(g)(6)(ii)(E)--Reactor Coolant Pressure Boundary Visual 
Inspections
    A new paragraph 10 CFR 50.55a(g)(6)(ii)(E) is added to require all 
current and future licensees to apply ASME Section XI, Code Case N-722, 
with conditions. Code Case N-722 provides requirements for bare metal 
visual examination of full and partial penetration welds in Class 1 
components that are fabricated with Alloy 600/82/182 material. Surfaces 
required to be examined by the bare metal visual method have to be 
unobstructed by debris, paint, insulation or other sources of 
interference. 10 CFR 50.55a(g)(6)(ii)(E) requires the use of N-722 plus 
four additional conditions. Condition (1) requires that PWR licensees 
implement N-722 except for those welds that have been mitigated by weld 
overlay or stress improvements. Condition (2) requires that if leakage 
occurs from a component, licensees take additional actions to 
characterize the orientation of the crack that caused the leakage. 
Condition (3) requires that if the crack that leads to leakage is 
circumferentially oriented and potentially the result of primary water 
stress-corrosion cracking, licensees perform non-visual sample 
inspections of the population of the components. Condition (4) requires 
that the ultrasonic examinations of the butt welds as required by 
Condition (2) and (3) follow the appropriate supplement of Appendix 
VIII of the ASME Code, Section XI.
    The visual examinations specified in Code Case N-722 are additional 
requirements beyond the current NDE requirements of Table IWB-2500-1 in 
the ASME Code, Section XI. The application of ASME Code Case N-722 is 
necessary because current inspections are inadequate and the safety 
consequences can be significant should the components fail due to 
cracking. NRC's determination that existing inspections of the reactor 
coolant pressure boundary (RCPB) are inadequate is based upon the 
degradation of RPV head penetration nozzles at Davis-Besse and the 
discovery of leaks and cracking at other plants, such as Oconee and 
Arkansas Nuclear One Unit 1. The absence of an effective inspection 
regime could, over time, result in unacceptable circumferential 
cracking or the degradation of reactor coolant system (RCS) components 
by corrosion from leaks in the RCPB. These degradation mechanisms 
increase the probability of a loss-of-coolant accident. The inspections 
required by the 2004 Edition of the ASME BPV Code, Section XI, are 
inadequate because Examination Category B-P, ``All Pressure Retaining 
Components,'' of Table IWB-2500-1, only requires a visual examination 
of the reactor vessel with the insulation in place during a system 
leakage test each refueling outage. Visual inspections may not detect 
gradual leakage as confirmed by industry experience.
    Both the NRC and the industry took short-term actions to address 
PWSCC in the RCPB because of limitations of the ASME BPV Code 
inspection programs to address PWSCC in the RCPB. In addition to 
issuing bulletins, the NRC issued Order EA-03-009 and First Revised 
Order EA-03-009 to quickly establish interim inspection requirements 
for RPV upper heads at PWRs. However, these measures addressed the 
issue only temporarily, and for specific locations. The industry also 
responded with compensatory measures (e.g., by specifying that a one-
time, bare-metal visual inspection of all RCS nickel-based alloy 
components and weld locations be performed within two refueling 
outages). However, these were only short-term measures.
    The ASME also took actions to address PWSCC. An ASME task group 
concluded that more rigorous inspections than those currently provided 
by the ASME BPV Code were needed in the areas most susceptible to 
PWSCC. The task group developed ASME Code Case N-722 to enhance the 
current ASME BPV Code requirements for detection of leakage and 
corrosion in the components considered to be susceptible to PWSCC. The 
Code Case specifies bare-metal visual examinations for all RCS pressure 
retaining components fabricated from Alloy 600/82/182 materials. This 
Code Case was approved by ASME in July 2005 and was published in 
Supplement 6 to the 2004 Code Cases. However, the Code Case is not 
mandatory for industry to follow. The Code Case improves upon existing 
ASME BPV Code inspection requirements, because it specifies bare metal 
visual examinations.
    Beyond the bare metal visual inspection requirements and 
frequencies of inspections, ASME Code Case N-722 is relatively limited 
in scope. The NRC is requiring non-visual inspection for items where 
leakage is identified in Class 1 components. The additional non-visual 
NDE is required to determine whether circumferential cracking is 
present in the flawed material and if multiple circumferential flaws 
have initiated. Leakage detected by visual examination only identifies 
that a flaw exists, and is not able to characterize flaw orientations 
and

[[Page 52743]]

locations. The NRC is requiring NDE scope expansion once a 
circumferential flaw is identified in these components because once 
flaws are found, favorable conditions must be assumed to exist for 
additional flaws to develop in other similar components in similar 
environments. Circumferential cracking has occurred, and is a 
particularly serious safety concern because it could, if undetected by 
NDE, lead to a complete severing of the piping and a loss-of-coolant 
accident.
    Therefore, the NRC is requiring the application of Code Case N-722 
with additional conditions. The conditions require additional NDE when 
leakage is detected and expansion of the sample size if a 
circumferential PWSCC flaw is found. Operating experience has shown 
that bare metal visual inspections alone are not sufficient and that 
NDE is necessary in order to detect cracking. The requirements for the 
schedule for conducting the initial inspections are specified in a new 
footnote to the new paragraph.

ASME OM Code

    The revision to Sec.  50.55a(b)(3) incorporates by reference the 
2004 Edition of the ASME OM Code subject to no new modifications or 
limitations.
    Paragraph (b)(3)(iv)(D) is revised to be less specific with regard 
to paragraph references in subsection ISTC [Inservice testing, the Code 
for Operation and Maintenance of Nuclear Power Plants] to eliminate 
inconsistencies in paragraph numbering. This is considered to be an 
editorial change that does not affect the intent or implementation of 
the current modification regarding the discontinuance of Appendix II 
condition monitoring programs of check valves.

IV. Generic Aging Lessons Learned Report

    In September 2005, the NRC issued, ``Generic Aging Lessons Learned 
(GALL) Report,'' NUREG-1801, Volumes 1 and 2, Revision 1, 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 are adequate without change or 
augmentation for license renewal. Section XI, Division 1, of the ASME 
BPV Code is one of the existing programs in the GALL report that is 
evaluated as an aging management program (AMP) for license renewal. 
Subsections IWB, IWC, IWD, IWE, IWF, and IWL of the 2001 Edition up to 
and including the 2003 Addenda of Section XI of the ASME BPV Code for 
ISI were evaluated in the GALL report and the conclusions in the GALL 
report are valid for this edition and addenda.
    In the GALL report, Sections XI.M1, ``ASME Section XI Inservice 
Inspection, Subsections IWB, IWC, and IWD,'' XI.S1, ``ASME Section XI, 
Subsection IWE,'' XI.S2, ``ASME Section XI, Subsection IWL,'' and 
XI.S3, ``ASME Section XI, Subsection IWF,'' describe the evaluation and 
technical bases for determining the adequacy of Subsections IWB, IWC, 
IWD, IWE, IWF, and IWL, respectively. In addition, many other AMPs in 
the GALL report rely in part, but to a lesser degree, on the 
requirements in the ASME BPV Code, Section XI.
    The NRC has evaluated Subsections IWB, IWC, IWD, IWE, IWF, and IWL 
of Section XI of the ASME BPV Code, 2004 Edition as part of the Sec.  
50.55a amendment process to incorporate by reference the 2004 Edition 
of the ASME BPV Code to determine if the conclusions of the GALL report 
also apply to AMPs that rely upon the ASME BPV Code edition that is 
incorporated by reference into Sec.  50.55a by this final rule. The NRC 
finds that the 2004 Edition of Sections III and XI of the ASME BPV 
Code, as modified and limited in this final rule, are acceptable and 
the conclusions of the GALL report remain valid. Accordingly, an 
applicant may use Subsections IWB, IWC, IWD, IWE, IWF, and IWL of 
Section XI of the 2004 Edition of the ASME BPV Code, as modified and 
limited in this final rule, as acceptable alternatives to the 
requirements of the 2001 Edition up to and including the 2003 Addenda 
of the ASME BPV Code, Section XI, referenced in the GALL AMPs in its 
plant-specific license renewal application. 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 of the ASME BPV Code as acceptable alternatives to the AMPs 
described in the GALL report.
    However, a licensee must assess and follow applicable NRC 
requirements with regard to changes to its licensing basis.
    The GALL report includes AMPs that are based on the requirements in 
the 2001 Edition through the 2003 Addenda of Section XI of the ASME BPV 
Code but in which the AMPs may recommend additional augmentation of the 
Code requirements in order to achieve aging management for license 
renewal. The technical or regulatory aspects of the AMPs, for which 
augmentation is recommended, also apply when implementing the 2004 
Edition of Section XI of the ASME BPV Code. 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 aspect of its AMP.
    The NRC currently provides license renewal guidance for augmented 
inspections of PWR upper reactor vessel heads and their penetration 
nozzles in GALL AMP XI.M11A, ``Nickel-Alloy Penetration Nozzles Welded 
to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors 
(PWR Only).'' The current program elements and aging management 
recommendations in GALL AMP XI.M11A are based on the augmented 
inspection requirements in the First Revised Order EA-03-009, 
``Issuance of First Revised Order (EA-03-009) Establishing Interim 
Inspection Requirements for Reactor Pressure Vessel Heads at 
Pressurized Water Reactors.'' For licensees that have been granted a 
renewed operating license and have committed to an AMP that is based on 
both conformance with GALL AMP XI.M11A and compliance with First 
Revised Order EA-03-009, the licensees may update the program elements 
of their AMP to reflect compliance with the new requirements in 10 CFR 
50.55a(g)(6)(ii)(D) and (E) without having to identify an exception to 
GALL AMP XI.M11A. For new or current license renewal applicants, they 
may reference conformance with GALL AMP XI.M11A and compliance with the 
new augmented inspection requirements in paragraphs 10 CFR 
50.55a(g)(6)(ii)(D) and (E) without the need for taking an exception to 
the program elements in GALL AMP XI.M11A.

V. Availability of Documents

----------------------------------------------------------------------------------------------------------------
                                             Public document     Electronic
                  Document                         room         reading room               ADAMS No.
----------------------------------------------------------------------------------------------------------------
ASME BPV Code*.............................  ...............  ...............  N/A
ASME OM Code*..............................  ...............  ...............  N/A
ASME Code Case N-722.......................               X   ...............  ML070170676
ASME Code Case N-729-1.....................               X   ...............  ML070170679

[[Page 52744]]

 
Regulatory Analysis........................               X   ...............  ML081550317
EA-03-009..................................               X                X   ML030380470
First Revised NRC Order EA-03-009..........               X                X   ML040220181
GALL Report, NUREG-1801....................  ...............               X   ML012060392
                                             ...............  ...............  ML012060514
                                             ...............  ...............  ML012060521
                                             ...............  ...............  ML012060539
Staff Requirements Memorandum dated          ...............  ...............  ML003751061
 September 10, 1999.
RG 1.147, Revision 15......................               X                X   ML072070419
----------------------------------------------------------------------------------------------------------------
*Available on the ASME Web site.

VI. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113, requires agencies to use technical standards that 
are developed or adopted by voluntary consensus standards bodies unless 
the use of such a standard is inconsistent with applicable law or is 
otherwise impractical. Public Law 104-113 requires Federal agencies to 
use industry consensus standards to the extent practical; it does not 
require Federal agencies to incorporate by reference a standard into 
the regulations in its entirety. The law does not prohibit an agency 
from generally 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.
    The NRC is amending its regulations to incorporate by reference a 
more recent edition of Sections III and XI of the ASME BPV Code and 
ASME OM Code, for construction, ISI, and inservice testing of nuclear 
power plant components. ASME BPV and OM Codes are national 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. In an SRM dated September 10, 1999, 
the Commission indicated its intent that a rulemaking identify all 
parts of an adopted voluntary consensus standard that are not adopted, 
and to justify not adopting such parts. The parts of the ASME BPV Code 
and OM Code that the NRC is not adopting; or is adopting with 
conditions, modifications, or limitations under which the Codes may be 
applied; are identified in Section III of this document and in the 
regulatory analysis. If the NRC did not conditionally accept ASME Code 
Editions and Addenda, it would disapprove these items entirely. The 
effect would be that licensees would need to submit a larger number of 
relief requests which would be an unnecessary additional burden for 
both the licensee and the NRC. This situation fits the definition of 
``impractical'' under Public Law 104-113. For these reasons, the 
treatment of ASME Code Editions and Addenda, and conditions, 
modifications, or limitations placed on them in this final rule do not 
conflict with any policy on agency use of consensus standards specified 
in Office of Management and Budget Circular A-119.

VII. Finding of No Significant Environmental Impact: Environmental 
Assessment

    This action is in accordance with NRC's policy to incorporate by 
reference in 10 CFR 50.55a new editions and addenda of the ASME BPV 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.
    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. 4332(C); NEPA Sec.  102(C)).
    The Commission has determined under NEPA, as amended, and the 
Commission's regulations in subpart A of 10 CFR part 51, that this 
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 rulemaking will 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; there is no increase in 
occupational exposure; and there is no significant increase in public 
radiation exposure. Some of the changes concerning ensuring the 
integrity of the RCPB would reduce the probability of accidents and 
radiological impacts on the public. The rulemaking does not involve 
non-radiological plant effluents and has no other environmental impact. 
Therefore, no significant non-radiological impacts are associated with 
the action.
    The determination of this environmental assessment is that there 
will be no significant off-site impact to the public from this action.

VIII. Paperwork Reduction Act Statement

    This rule increases the burden on licensees to report requirements 
and maintain records for examination requirements in ASME BPV Code 
Section XI IWB-2500(b). The public burden for this information 
collection is estimated to average 3 hours every ten years per request. 
Because the burden for this information collection is insignificant, 
OMB clearance is not required. Existing requirements were approved by 
the OMB, 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.

[[Page 52745]]

IX. Regulatory Analysis

    The NRC has prepared a regulatory analysis on this final rule. The 
analysis is available for review in the NRC's PDR, located in One White 
Flint North, 11555 Rockville Pike, Rockville, Maryland. In addition, 
copies of the regulatory analysis may be obtained as indicated in 
Section V of this document.

X. Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 
605(b), the Commission certifies that this amendment will not, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This amendment affects the licensing and operation 
of nuclear power plants. The companies that own these plants do not 
fall within the scope of the definition of small entities set forth in 
the Regulatory Flexibility Act or the Small Business Size Standards set 
forth in regulations issued by the Small Business Administration at 13 
CFR part 121.

XI. Backfit Analysis

    The NRC's Backfit Rule in 10 CFR 50.109 states that the Commission 
shall require the backfitting of a facility only when it finds the 
action to be justified under specific standards stated in the rule. 
Section 50.109(a)(1) defines backfitting as the modification of or 
addition to systems, structures, components, or design of a facility; 
or the design approval or manufacturing license for a facility; or the 
procedures or organization required to design, construct or operate a 
facility; any of which may result from a new or amended provision in 
the Commission rules or the imposition of a regulatory staff position 
interpreting the Commission rules that is either new or different from 
a previously applicable NRC position after issuance of the construction 
permit or the operating license or the design approval.
    Section 50.55a requires nuclear power plant licensees to construct 
ASME BPV Code Class 1, 2, and 3 components in accordance with the rules 
provided in Section III, Division 1, of the ASME BPV Code; inspect 
Class 1, 2, 3, Class MC, and Class CC components in accordance with the 
rules provided in Section XI, Division 1, of the ASME BPV Code; and 
test Class 1, 2, and 3 pumps, valves, and dynamic restraints (snubbers) 
in accordance with the rules provided in the ASME OM Code. This rule 
incorporates by reference the 2004 Edition of Section III, Division 1, 
of the ASME BPV Code; Section XI, Division 1, of the ASME BPV Code; and 
the ASME OM Code.
    Incorporation by reference of more recent editions and addenda of 
Section III, Division 1, of the ASME BPV Code does not affect a plant 
that has received a construction permit or an operating license or a 
design that has been approved, because the edition and addenda to be 
used in constructing a plant are, by rule, determined on the basis of 
the date of the construction permit, and are not changed thereafter, 
except voluntarily by the licensee. Thus, incorporation by reference of 
a more recent edition and addenda of Section III, Division 1, does not 
constitute a ``backfitting'' as defined in Sec.  50.109(a)(1).
    Incorporation by reference of more recent editions and addenda of 
Section XI, Division 1, of the ASME BPV Code and the ASME OM Code 
affect the ISI and IST programs of operating reactors. However, the 
Backfit Rule does not apply to incorporation by reference of later 
editions and addenda of the ASME BPV Code (Section XI) and OM Code. The 
NRC's policy has been to incorporate later versions of the ASME Codes 
into its regulations. This practice is codified in Sec.  50.55a which 
requires licensees to revise their ISI and IST programs every 120 
months to the latest edition and addenda of Section XI of the ASME BPV 
Code and the ASME OM Code incorporated by reference in Sec.  50.55a 
that is in effect 12 months prior to the start of a new 120-month ISI 
and IST interval.
    Other circumstances where the NRC does not apply the Backfit Rule 
to the incorporation by reference of a later Code into the regulations 
are as follows:
    (1) When the NRC takes exception to a later ASME BPV Code or OM 
Code provision but merely retains the current existing requirement, 
prohibits the use of the later Code provision, limits the use of the 
later Code provision, or supplements the provisions in a later Code, 
the Backfit Rule does not apply because the NRC is not imposing new 
requirements. However, the NRC explains any such exceptions to the Code 
in the Statement of Considerations and regulatory analysis for the 
rule;
    (2) When an NRC exception relaxes an existing ASME BPV Code or OM 
code provision but does not prohibit a licensee from using the existing 
Code provision, the Backfit Rule does not apply because the NRC is not 
imposing new requirements and;
    (3) Modifications and limitations imposed during previous routine 
updates of Sec.  50.55a have established a precedent for determining 
which modifications or limitations are backfits or require a backfit 
analysis (e.g., final rule dated October 1, 2004 (69 FR 58804). The 
application of the backfit requirements to modifications and 
limitations in the current rule are consistent with the application of 
backfit requirements to modifications and limitations in previous 
rules.
    There are some circumstances in which the incorporation by 
reference of a later ASME BPV Code or OM Code into 10 CFR 50.55a 
introduces a backfit. In these cases, the NRC performs a backfit 
analysis or documented evaluation in accordance with Sec.  50.109. 
These include the following:
    (1) When the NRC incorporates by reference a later provision of the 
ASME BPV Code or OM Code that takes a substantially different direction 
from the existing requirements, the action is treated as a backfit, 
e.g., 61 FR 41303 (August 8, 1996).
    (2) When the NRC requires implementation of later ASME BPV Code or 
OM Code provision on an expedited basis, the action is treated as a 
backfit. This applies when implementation is required sooner than it 
would be required if the NRC simply incorporated the Code by reference 
without any expedited language, e.g., 64 FR 51370 (September 22, 1999).
    (3) When the NRC takes an exception to an ASME BPV Code or OM Code 
provision and imposes a requirement that is substantially different 
from the existing requirement as well as substantially different than 
the later Code, e.g., 67 FR 60529 (September 26, 2002).
    The backfitting discussion for the revisions to 10 CFR 50.55a is 
set forth as follows:
1. Remove 10 CFR 50.55a(b)(2)(xi) Concerning Components Exempt From 
Examination
    This change removes an existing limitation on the use of 1989 
Addenda and later editions and addenda of the ASME BPV Code, Section 
XI, regarding the use of subarticle IWB-1220 in the examinations of 
welds in the inaccessible locations. Licensees have either committed to 
perform augmented inspection or have followed the provisions of Generic 
Letter 88-01 and NUREG-0313 in examining the inaccessible welds. 
Therefore, this change is not considered as a backfit under 10 CFR 
50.109.
2. Remove 10 CFR 50.55a(b)(2)(xiii) Concerning the Provisions of Code 
Case N-523-1, ``Mechanical Clamping Devices for Class 2 and 3 Piping''
    10 CFR 50.55a(b)(2)(xiii) states that ``Licensees may use the 
provisions of

[[Page 52746]]

Code Case N-523-1, ``Mechanical Clamping Devices for Class 2 and 3 
Piping.'' 10 CFR 50.55a(b)(2)(xiii) does not require, but provides an 
option for, licensees to use Code Case N-523-1. In 2000, ASME updated 
Code Case N-523-1 to N-523-2 without changes to technical requirements. 
Code Case N-523-2, ``Mechanical Clamping Devices for Class 2 and 3 
Piping,'' has been accepted in RG 1.147, Revision 15, which is 
incorporated by reference into 10 CFR 50.55a(g)(4)(i) and 10 CFR 
50.55a(g)(4)(ii). Code Case N-523-2 may be used by licensees without 
requesting authorization. According to RG 1.147, Revision 15, Code Case 
N-523-1 has been superseded by Code Case N-523-2. It is stated in RG 
1.147, Revision 15, that ``After the ASME annuls a Code Case and the 
NRC amends 10 CFR 50.55a and this guide [RG 1.147], licensees may not 
implement that Code Case for the first time. However, a licensee who 
implemented the Code Case prior to annulment may continue to use that 
Code Case through the end of the present ISI interval. An annulled Code 
Case cannot be used in the subsequent ISI interval unless implemented 
as an approved alternative under 10 CFR 50.55a(a)(3) * * *'' The NRC 
has not annulled or prohibited the use of Code Case N-523-1 in RG 
1.147, Revision 15. Licensees who have used Code Case N-523-1 may 
continue to use it. The NRC is not imposing new requirements by 
removing 10 CFR 50.55a(b)(2)(xiii). Therefore, the removal of 10 CFR 
50.55a(b)(2)(xiii) is not a backfit.
3. Modify 10 CFR 50.55a(b)(2)(xv) To Implement Appendix VIII of Section 
XI, the 1995 Edition Through the 2004 Edition of the ASME BPV Code
    This change updates the edition of the ASME BPV Code in 10 CFR 
50.55a(b)(2)(xv). Therefore, is not considered as a backfit under 10 
CFR 50.109.
4. Add 10 CFR 50.55a(b)(2)(xx) to Require NDE Provision in IWA-
4540(a)(2) of the 2002 Addenda of Section XI When Performing System 
Leakage Tests
    Subarticle IWA-4540(a)(2) of the 2002 Addenda of the ASME BPV Code, 
Section XI, requires an NDE be performed in combination with a system 
leakage test during repair/replacement activities. Subarticle IWA-
4540(a)(2) of the 2003 Addenda through later editions and addenda of 
the ASME BPV Code, Section XI, does not specify an NDE after a system 
leakage test. The addition requires, as part of repair and replacement 
activities, that a NDE be performed per IWA-4540(a)(2) of the 2002 
Addenda of the ASME BPV Code, Section XI, after a system leakage test 
is performed per subarticle IWA-4540(a)(2) of the 2003 Addenda through 
later editions and addenda of the ASME BPV Code, Section XI.
    As stated previously, when the NRC takes exception to a later ASME 
BPV Code provision but merely retains the existing requirement, 
prohibits the use of the later Code provision, limits the use of the 
later Code provision, or supplements the provisions in a later Code, 
the Backfit Rule does not apply because the NRC is not imposing new 
requirements. The addition retains the system leakage test requirement 
in IWA-4540(a)(2) of the 2003 Addenda through the later editions and 
addenda of the ASME BPV Code, Section XI, but supplements it with the 
NDE of IWA-4540(a)(2) of the 2002 Addenda of the Code. However, the NRC 
has approved a few licensees to use IWA-4540(a) of the 2003 addenda of 
the ASME Code, Section XI without imposing the NDE requirement in 
conjunction with the system leakage tests. Therefore, some licensees 
may currently use the provisions of IWA-4540(a) in the 2003 Addenda 
without having to perform NDE. Because 10 CFR 50.55a(b)(2)(xx) imposes 
NDE requirements after these licensees are allowed not to perform the 
required NDE, the additional NDE requirements in 10 CFR 
50.55a(b)(2)(xx) may be considered backftting under 10 CFR 50.109(a)(1) 
for these licensees. However, the NRC believes that the NDE 
requirements are necessary for compliance with Commission requirements 
and/or license provisions. Therefore, a backfit analysis need not be 
prepared under the ``compliance'' exception in 10 CFR 50.109(a)(4)(i). 
The following discussion constitutes the documented evaluation to 
support the invocation of the compliance exception.
    A system leakage test does not verify fully the structural 
integrity of the repaired or replaced piping components. NDE 
examinations will most likely detect whether cracks exist and thereby 
ensure the structural integrity of the repaired or replaced components. 
The general design criteria (GDC) for nuclear power plants (Appendix A 
to 10 CFR part 50) provide the regulatory requirements for the NRC's 
assessment of the potential for, and consequences of, degradation of 
the reactor coolant pressure boundary (RCPB). The applicable GDCs 
include GDC 14 and GDC 31. GDC 14 specifies that the RCPB be designed, 
fabricated, erected, and tested so as to have an extremely low 
probability of abnormal leakage, of rapidly propagating failure, and of 
gross rupture. GDC 31 specifies that the probability of rapidly 
propagating fracture of the RCPB be minimized.
    The nuclear plants that were licensed before GDC were incorporated 
in 10 CFR Part 50 also would not be in compliance with their licensing 
basis which requires maintenance of the structural and leakage 
integrity of the RCPB.
    Cracking of primary system piping as a result of the repair or 
replacement is a non-compliance with GDC 14 because the RCPB must be 
fabricated and tested as to have an extremely low probability of 
abnormal leakage, of rapidly propagating failure and of gross rupture. 
Without an NDE, there would be no confirmation as to whether cracks 
exist in the component. The volumetric examination (NDE) will verify 
the structural integrity of the component as part of the repair or 
replacement activity. If a crack, especially a circumferential crack in 
a pipe, is not detected, it would increase the probability of rapidly 
propagating fracture of RCPB (i.e., a non-compliance with GDC 31). 
Therefore, cracking, if undetected, would be detrimental to the 
structural and leakage integrity of the RCPB. The NDE requirements in 
conjunction with system leakage testing of 50.55a(b)(2)(xx) will ensure 
the structural and leakage integrity of the RCPB, assuring an extremely 
low probability of abnormal leakage, and minimizing the probability of 
a rapidly propagating fracture of the RCPB.
    The NRC concludes that those licensees who use subsection IWA-
4540(a) of the 2003 addenda of the ASME Code, Section XI will not be in 
compliance with GDC and their licensing basis for the structural 
integrity of piping components throughout the term of their license 
(including any renewal periods) absent the imposition of NDE 
examination in conjunction with the system leakage testing. The NRC 
concludes, therefore, that 10 CFR 50.55a(b)(2)(xx) is a compliance 
backfit under 10 CFR 50.109(a)(4)(i).
5. Revise 10 CFR 50.55a(b)(2)(xxi) To Be Consistent With the NRC's 
Imposed Condition for Code Case N-648-1 in RG 1.147, Revision 15
    This change aligns the conditions imposed on visual examinations in 
10 CFR 50.55a(b)(2)(xxi) with the conditions imposed on Code Case N-
648-1 in RG 1.147, Revision 15. The imposed conditions do not represent 
a new NRC position. Therefore, this change is not considered as a 
backfit under 10 CFR 50.109.

[[Page 52747]]

6. Remove 10 CFR 50.55a(g)(6)(ii)(A) and Associated Subparagraphs on 
the Augmented Examination of the Reactor Vessel
    This change removes a one-time examination requirement which has 
been completed by all current licensees, and, therefore, is not 
considered as a backfit under 10 CFR 50.109. Future licensees will be 
subject to other Code provisions that preclude the need for this one-
time examination.
7. Add Paragraph (D) to 10 CFR 50.55a(g)(6)(ii)--Reactor Vessel Head 
Inspections
    The current regulatory requirements for RPV head inspection are set 
forth in the First Revised NRC Order EA-03-009, dated February 20, 
2004. Order EA-03-009 was issued to ensure that boric acid corrosion of 
RPV heads and PWSCC of RPV head penetration nozzles and welds, which 
could result in failure of the RPV head or head penetrations, are 
promptly identified and corrected. The NRC determined that Order EA-03-
009 constitutes backfitting as defined in 10 CFR 50.109(a)(1), but that 
the actions mandated by the Order were necessary for reasonable 
assurance of adequate protection to public health and safety. 
Therefore, a backfit analysis was not prepared for the Order in 
accordance with Sec.  50.109(a)(4)(ii). Section III of the Order also 
stated, in part, ``It is appropriate and necessary to the protection of 
public health and safety to establish a clear regulatory framework, 
pending the incorporation of revised inspection requirements into 10 
CFR 50.55a.''
    This rule revokes Order EA-03-009 as the current regulatory 
requirement for RPV head inspection, and replace it with ASME Code Case 
N-729-1, as modified in 10 CFR 50.55a per 10 CFR 
50.55a(g)(6)(ii)(D)(1). All current licensees will be required to 
implement ASME Code Case N-729-1, with the limitations and conditions 
denoted by this rule. The Code Case provisions on RPV head and head 
penetration inspections are somewhat different from those established 
in Order EA-03-009, and will require a licensee to modify its 
procedures for inspection of its RPV head and head penetrations to meet 
the requirements on the Code Case. Accordingly, NRC imposition of the 
Code Case may be deemed to be a modification of the procedures to 
operate a facility resulting from the imposition of new regulation, and 
as such, this rulemaking provision may be considered backfitting under 
10 CFR 50.109(a)(1). The NRC continues to find that RPV head 
inspections are necessary for adequate protection of public health and 
safety, and that the requirements of Code Case N-729-1, with the 
limitations and conditions denoted by this rule, represents an 
acceptable approach, developed by a voluntary consensus standards 
organization, for performing future RPV head and head penetration 
inspections. The NRC believes, in keeping with the intent of the 
National Technology Transfer and Advancement Act, that it is preferable 
to endorse a voluntary consensus standard such as Code Case N-729-1, 
with the limitations and conditions denoted by this rule, rather than 
continuing to rely upon the requirements embodied in Order EA-03-009. 
Therefore, the NRC concludes that NRC approval of Code Case N-729-1, 
with the limitations and conditions denoted by this rule, by 
incorporation by reference of that Code Case into Sec.  50.55a, 
constitutes a redefinition of the requirements necessary to provide 
reasonable assurance of adequate protection of public health and 
safety. Therefore, a backfit analysis was not prepared for this portion 
of the final rule, in accordance with Sec.  50.109(a)(4)(iii).
8. Add Paragraph (E) to 10 CFR 50.55a(g)(6)(ii)--Reactor Coolant 
Pressure Boundary Visual Inspections
    The NRC is adding 10 CFR 50.55a(g)(6)(ii)(E) to require augmented 
inspections of Class 1 components fabricated with Alloy 600/82/182 
materials. The augmented inspection will consist of the requirements in 
Code Case N-722 which specifies ISI for PWR ASME Code Class 1 
components containing materials susceptible to PWSCC and NRC imposed 
conditions to the Code Case to require additional NDE when leakage is 
detected and expansion of the inspection sample size if a 
circumferential PWSCC flaw is detected. The intent of conditioning the 
Code Case is to identify leakage of and prevent unacceptable cracks and 
corrosion in Class 1 components, which are part of RCPB. The 
requirements may be considered backfitting under 10 CFR 50.109(a)(1). 
However, the NRC believes that the requirements are necessary for 
compliance with Commission requirements and/or license provisions. 
Therefore a backfit analysis need not be prepared under the 
``compliance'' exception in 10 CFR 50.109(a)(4)(i). The following 
discussion constitutes the documented evaluation to support the 
invocation of the compliance exception.
    Failure of the RCPB could result in unacceptable challenges to 
reactor safety systems that, combined with other failures, could lead 
to the release of radioactivity to the environment. Based on PWSCC 
experience in PWRs, the NRC concludes that there is a reasonable 
likelihood that PWR licensees would not be in compliance with 
appropriate regulatory requirements and current licensing basis with 
respect to structural integrity and leak-tightness throughout the term 
of the operating license, should PWSCC occur in their plants. The 
general design criteria (GDC) for nuclear power plants (Appendix A to 
10 CFR part 50) provide the regulatory requirements for the NRC's 
assessment of the potential for, and consequences of, degradation of 
the RCPB. The applicable GDCs include GDC 14 and GDC 31. GDC 14 
specifies that the RCPB be designed, fabricated, erected, and tested so 
as to have an extremely low probability of abnormal leakage, of rapidly 
propagating failure, and of gross rupture. GDC 31 specifies that the 
probability of rapidly propagating fracture of the RCPB be minimized.
    The nuclear plants that were licensed before GDC were incorporated 
in 10 CFR Part 50 also would not be in compliance with their licensing 
basis which requires maintenance of the structural and leakage 
integrity of the RCPB.
    Leakage of primary system coolant as a result of PWSCC in Alloy 
600/82/182 material is a non-compliance with GDC 14 and 31 and 
licensing bases because there have been many cases of leakage as a 
result of PWSCC of Alloy 600/82/182 material in PWRs. Therefore, 
leakage as a result of PWSCC has not been shown to be of extremely low 
probability (i.e., a non-compliance with GDC 14). In addition, the 
operating experience has shown that the crack growth rate of PWSCC in 
Alloy 600/82/182 material can be rapid. If PWSCC is not detected and 
removed, a crack, especially a circumferential crack in a pipe, would 
increase the probability of rapidly propagating fracture of RCPB (i.e., 
a non-compliance with GDC 31). Therefore, PWSCC in Alloy 600/82/182 
material, if undetected, would be detrimental to the structural and 
leakage integrity of the RCPB. Code Case N-722 with conditions provides 
inspection requirements to detect PWSCC so that licensees can repair or 
replace the affected components, thereby maintaining the structural and 
leakage integrity of the RCPB, assuring an extremely low probability of 
abnormal leakage, and minimizing the probability of a rapidly 
propagating fracture of the RCPB.
    The NRC concludes that licensees will not be in compliance with GDC 
and their licensing basis for structural and leakage integrity of Class 
1 components

[[Page 52748]]

that were made of Alloy 600/82/182 material throughout the term of 
their license (including any renewal periods) absent the imposition of 
Code Case N-722 with conditions. The NRC concludes, therefore, that 10 
CFR 50.55a(g)(6)(ii)(E) is a compliance backfit under 10 CFR 
50.109(a)(4)(i).

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 OMB.

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.

0
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); sec. 651(e), Pub. 
L. 109-58, 119 Stat. 806-810 (42 U.S.C. 2014, 2021, 2021b, 2111).
    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(d), 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. Section 50.55a is amended by:
0
A. Revising paragraph (b) introductory text, (b)(1) introductory text, 
(b)(1)(iii), (b)(2) introductory text , (b)(2)(xv) introductory text, 
(b)(2)(xx) and (b)(2)(xxi)(A), (b)(3) introductory text, and 
(b)(3)(iv)(D);
0
B. Removing and reserving paragraphs (b)(2)(xi) and (b)(2)(xiii), and 
(g)(6)(ii)(A); and
0
C. Adding paragraphs (g)(6)(ii)(D) and (g)(6)(ii)(E), to read as 
follows:


Sec.  50.55a  Codes and standards.

* * * * *
    (b) The following 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: Sections III and XI of the ASME Boiler and 
Pressure Vessel Code and the ASME Code for Operation and Maintenance of 
Nuclear Power Plants, which are referenced in paragraphs (b)(1), 
(b)(2), and (b)(3) of this section; NRC Regulatory Guide 1.84, Revision 
34, ``Design, Fabrication, and Materials Code Case Acceptability, ASME 
Section III'' (October 2007); NRC Regulatory Guide 1.147, Revision 15, 
``Inservice Inspection Code Case Acceptability, ASME Section XI, 
Division 1'' (October 2007); and Regulatory Guide 1.192, ``Operation 
and Maintenance Code Case Acceptability, ASME OM Code'' (June 2003), 
which list ASME Code cases that the NRC has approved in accordance with 
the requirements in paragraphs (b)(4), (b)(5), and (b)(6) of this 
section; ASME Code Case N-729-1, ``Alternative Examination Requirements 
for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-
Retaining Partial-Penetration Welds, Section XI, Division 1'' (Approval 
Date: March 28, 2006), which has been approved by the NRC with 
conditions in accordance with the requirements in paragraph 
(g)(6)(ii)(D) of this section; and ASME Code Case N-722, ``Additional 
Examinations for PWR Pressure Retaining Welds in Class 1 Components 
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' 
(Approval Date: July 5, 2005), which has been approved by the NRC with 
conditions in accordance with the requirements in paragraphs 
(g)(6)(ii)(E) of this section. Copies of the ASME Boiler and Pressure 
Vessel Code, the ASME Code for Operation and Maintenance of Nuclear 
Power Plants, ASME Code Case N-729-1, and ASME Code Case N-722 may be 
purchased from the American Society of Mechanical Engineers, Three Park 
Avenue, New York, NY 10016 or through the Web http://www.asme.org/
Codes/. Single copies of NRC Regulatory Guides 1.84, Revision 34; 
1.147, Revision 15; 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 DISTRIBUTION@nrc.gov. 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/code_of_
federal_regulations/ ibr_locations.html.
    (1) As used in this section, references to Section III of the ASME 
Boiler and Pressure Vessel Code refer to Section III, and include the 
1963 Edition through 1973 Winter Addenda, and the 1974 Edition 
(Division 1) through the 2004 Edition (Division 1), subject to the 
following limitations and modifications:
* * * * *
    (iii) Seismic design of piping. Applicants and licensees may use 
Articles NB-3200, NB-3600, NC-3600, and ND-3600 for seismic design of 
piping, up to and including the 1993 Addenda, subject to the limitation 
specified in paragraph (b)(1)(ii) of this section. Applicants and 
licensees may not use these Articles for seismic design of piping in 
the 1994 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(1) of this section.
* * * * *
    (2) As used in this section, references to Section XI of the ASME 
Boiler and Pressure Vessel Code refer to Section XI, and include the 
1970 Edition through the 1976 Winter Addenda, and the 1977 Edition 
(Division 1) through the 2004 Edition (Division 1), subject to the 
following limitations and modifications:
* * * * *
    (xi) [Reserved]
* * * * *
    (xiii) [Reserved]
* * * * *
    (xv) Appendix VIII specimen set and qualification requirements. The 
following provisions may be used to modify implementation of Appendix 
VIII of Section XI, 1995 Edition through

[[Page 52749]]

the 2001 Edition. Licensees choosing to apply these provisions shall 
apply all of the following provisions under this paragraph except for 
those in Sec.  50.55a(b)(2)(xv)(F) which are optional. Licensees who 
use later editions and addenda than the 2001 Edition of the ASME Code 
shall use the 2001 Edition of Appendix VIII.
* * * * *
    (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) * * *
    (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.
* * * * *
    (3) As used in this section, references to the OM Code refer to the 
ASME Code for Operation and Maintenance of Nuclear Power Plants, and 
include the 1995 Edition through the 2004 Edition subject to the 
following limitations and modifications:
* * * * *
    (iv) * * *
    (D) The applicable provisions of subsection ISTC must be 
implemented if the Appendix II condition monitoring program is 
discontinued.
* * * * *
    (g) * * *
    (6) * * *
    (ii) * * *
    (A) [Reserved]
* * * * *
    (D) Reactor vessel head inspections.
    (1) All licensees of pressurized water reactors shall augment their 
inservice inspection program with ASME Code Case N-729-1 subject to the 
conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this 
section. Licensees of existing operating reactors as of [insert final 
date of rule] shall implement their augmented inservice inspection 
program by December 31, 2008. Once a licensee implements this 
requirement, the First Revised NRC Order EA-03-009 no longer applies to 
that licensee and shall be deemed to be withdrawn.
    (2) Note 9 of ASME Code Case N-729-1 shall not be implemented.
    (3) Instead of the specified `examination method' requirements for 
volumetric and surface examinations in Note 6 of Table 1 of Code Case 
N-729-1, the licensee shall perform volumetric and/or surface 
examination of essentially 100 percent of the required volume or 
equivalent surfaces of the nozzle tube, as identified by Figure 2 of 
ASME Code Case N-729-1. A demonstrated volumetric or surface leak path 
assessment through all J-groove welds shall be performed. If a surface 
examination is being substituted for a volumetric examination on a 
portion of a penetration nozzle that is below the toe of the J-groove 
weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface 
examination shall be of the inside and outside wetted surface of the 
penetration nozzle not examined volumetrically.
    (4) By September 1, 2009, ultrasonic examinations shall be 
performed using personnel, procedures and equipment that have been 
qualified by blind demonstration on representative mockups using a 
methodology that meets the conditions specified in 
(50.55a(g)(6)(ii)(D)(3)(i) through (50.55a(g)(6)(ii)(D)(3)(iv), instead 
of the qualification requirements of Paragraph -2500 of ASME Code Case 
N-729-1. References herein to Section XI, Appendix VIII shall be to the 
2004 Edition with no Addenda of the ASME BPV Code.
    (i) The specimen set shall have an applicable thickness 
qualification range of +25 percent to -40 percent for nominal depth 
through-wall thickness. The specimen set shall include geometric and 
material conditions that normally require discrimination from primary 
water stress corrosion cracking (PWSCC) flaws.
    (ii) The specimen set shall have a minimum of ten (10) flaws which 
provide an acoustic response similar to PWSCC indications. All flaws 
shall be greater than 10 percent of the nominal pipe wall thickness. A 
minimum of 20 percent of the total flaws shall initiate from the inside 
surface and 20 percent from the outside surface. At least 20 percent of 
the flaws shall be in the depth ranges of 10-30 percent through wall 
thickness and at least 20 percent within a depth range of 31-50 percent 
through wall thickness. At least 20 percent and no more than 40 percent 
of the flaws shall be oriented axially.
    (iii) Procedures shall identify the equipment and essential 
variables and settings used for the qualification, and are consistent 
with Subarticle VIII-2100 of Section XI, Appendix VIII. The procedure 
shall be requalified when an essential variable is changed outside the 
demonstration range as defined by Subarticle VIII-3130 of Section XI, 
Appendix VIII and as allowed by Articles VIII-4100, VIII-4200 and VIII-
4300 of Section XI, Appendix VIII. Procedure qualification shall 
include the equivalent of at least three personnel performance 
demonstration test sets. Procedure qualification requires at least one 
successful personnel performance demonstration.
    (iv) Personnel performance demonstration test acceptance criteria 
shall meet the personnel performance demonstration detection test 
acceptance criteria of Table VIII--S10-1 of Section XI, Appendix VIII, 
Supplement 10. Examination procedures, equipment, and personnel are 
qualified for depth sizing and length sizing when the RMS error, as 
defined by Subarticle VIII-3120 of Section XI, Appendix VIII, of the 
flaw depth measurements, as compared to the true flaw depths, do not 
exceed \1/8\ inch (3 mm), and the root mean square (RMS) error of the 
flaw length measurements, as compared to the true flaw lengths, do not 
exceed \3/8\ inch (10 mm), respectively.
    (5) If flaws attributed to PWSCC have been identified, whether 
acceptable or not for continued service under Paragraphs -3130 or -3140 
of ASME Code Case N-729-1, the re-inspection interval must be each 
refueling outage instead of the re-inspection intervals required by 
Table 1, Note (8) of ASME Code Case N-729-1.
    (6) Appendix I of ASME Code Case N-729-1 shall not be implemented 
without prior NRC approval.

[[Page 52750]]

    (E) Reactor coolant pressure boundary visual inspections.\1\
---------------------------------------------------------------------------

    \1\ For inspections to be conducted every refueling outage and 
inspections conducted every other refueling outage, the initial 
inspection shall be performed at the next refueling outage after 
January 1, 2009. For inspections to be conducted once per interval, 
the inspections shall begin in the interval in effect on January 1, 
2009, and shall be prorated over the remaining periods and refueling 
outages in this interval.
---------------------------------------------------------------------------

    (1) All licensees of pressurized water reactors shall augment their 
inservice inspection program by implementing ASME Code Case N-722 
subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) 
through (4) of this section. The inspection requirements of ASME Code 
Case N-722 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 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 item number. The number of components inspected 
must equal or exceed the number of components found to be leaking under 
that item number. If circumferential cracking is identified in the 
sample, non-visual NDE must be performed in the remaining components 
under that item number.
    (4) If ultrasonic examinations of butt welds are used to meet the 
NDE requirements in paragraphs (g)(6)(ii)(E)(2) or (g)(6)(ii)(E)(3) of 
this section, they must be performed using the appropriate supplement 
of Section XI, Appendix VIII of the ASME Boiler and Pressure Vessel 
Code.
* * * * *

    For the U.S. Nuclear Regulatory Commission.

    Dated at Rockville, Maryland, this 18th day of August 2008.
R.W. Borchardt,
Executive Director for Operations.
 [FR Doc. E8-20624 Filed 9-9-08; 8:45 am]
BILLING CODE 7590-01-P