Industry Codes and Standards; Amended Requirements, 52730-52750 [E8-20624]

Agencies

• NUCLEAR REGULATORY COMMISSION

[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 https://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 a