[Federal Register Volume 73, Number 176 (Wednesday, September 10, 2008)]
[Rules and Regulations]
[Pages 52730-52750]
From the Federal Register Online via the Government Publishing 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
[email protected].
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
[email protected].
FOR FURTHER INFORMATION CONTACT: L. Mark Padovan, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone 301-415-1423, e-mail [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. 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 [email protected]. Copies of the ASME Codes
and NRC Regulatory Guides incorporated by reference in this section may
be inspected at the NRC Technical Library, Two White Flint North, 11545
Rockville Pike, Rockville, MD 20852-2738 or call 301-415-5610, or at
the National Archives and Records Administration (NARA). For
information on the availability of this material at NARA, call 202-741-
6030, or go to: http://www.archives.gov/federal_register/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