[Federal Register Volume 72, Number 65 (Thursday, April 5, 2007)]
[Proposed Rules]
[Pages 16731-16741]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-6379]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AH76
Industry Codes and Standards; Amended Requirements
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
amend 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.
NRC also proposes to require the use of ASME Code Cases N-722 and N-
729-1, both with conditions, and to remove certain obsolete
requirements specified in Sec. 50.55a. This action is in accordance
with the NRC's policy to periodically update the regulations to
incorporate new editions and addenda of the ASME Codes by reference and
is intended to maintain the safety of nuclear reactors and make NRC
activities more effective and efficient.
DATES: Comments regarding the proposed amendment must be submitted by
June 19, 2007. Comments received after this date will be considered if
it is practical to do so, but the Commission is only able to ensure
consideration of comments received on or before this date.
ADDRESSES: You may submit comments by any one of the following methods.
Please include RIN 3150-AH76 in the subject line of your comments.
Comments on rulemakings submitted in writing or in electronic form will
be made available to the public in their entirety on the NRC rulemaking
Web site. Personal information will not be removed from your comments.
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
E-mail comments to: [email protected]. If you do not receive a reply e-
mail confirming that we have received your comments, contact us
directly at (301) 415-1966. You may also submit comments via the NRC's
rulemaking Web site at http://ruleforum.llnl.gov. Address questions
about our rulemaking Web site to Carol Gallagher (301) 415-5905; e-mail
[email protected].
Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland
20852, between 7:30 am and 4:15 pm Federal workdays. (Telephone (301)
415-1966).
Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at
(301) 415-1101.
Publicly available documents related to this rulemaking may be
viewed electronically on the public computers located at the NRC's
Public Document Room (PDR), O1-F21, One White Flint
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North, 11555 Rockville Pike, Rockville, Maryland. The PDR reproduction
contractor will copy documents for a fee. Selected documents, including
comments, may be viewed and downloaded electronically via the NRC
rulemaking Web site at http://ruleforum.llnl.gov.
Publicly available documents created or received at the NRC after
November 1, 1999, are available electronically at the NRC's Electronic
Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this
site, the public can gain entry into the NRC's Agencywide Document
Access and Management System (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 PDR Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected].
FOR FURTHER INFORMATION CONTACT: Lee Banic, Division of Policy and
Rulemaking, Office of Nuclear Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
2771, e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. Summary of Proposed Revisions to 10 CFR 50.55a
III. Generic Aging Lessons Learned Report
IV. Availability of Documents
V. Plain Language
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
I. Background
The NRC is proposing to amend 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.
In a separate proposed rule, published on March 13, 2006 (71 FR
12781), the Commission proposed to add language to the introductory
paragraph of Sec. 50.55a to establish the applicability of the
conditions therein to licenses and approvals issued under Part 52.
Specifically, that proposed rule would add two new sentences: ``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)'' and ``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.'' The
Commission expects that the March 13, 2006, proposed rule will become
final before the proposed rule updating Sec. 50.55a to the 2004
Edition. The net effect then is that combined licenses would be subject
to the updated requirements when the rulemaking proposed in this notice
becomes final.
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, Pub. L. 104-113, to be used by government
agencies unless the use of such a standard is inconsistent with
applicable law or is otherwise impractical. It has been the NRC's
practice to review new editions and addenda of the ASME BPV and OM
Codes and periodically update Sec. 50.55a to incorporate newer
editions and addenda by reference. 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 proposing to incorporate by reference: Section III
of the 2004 Edition of the ASME BPV Code; Section XI of the 2004
Edition of the ASME BPV Code subject to proposed modifications and
limitations; and the 2004 Edition of the ASME OM Code. The NRC is
proposing to amend its regulations as follows:
1. Remove 10 CFR 50.55a(b)(2)(xi), concerning components exempt
from examination.
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.''
3. Modify 10 CFR 50.55a(b)(2)(xv) to implement Appendix VIII of
Section XI of the 2004 Edition of the ASME BPV Code.
4. Add 10 CFR 50.55a(b)(2)(xx) to require nondestructive
examination (NDE) provision in IWA-4540(a)(2) of the 2002 Addenda of
Section XI when performing system leakage tests after repair and
replacement activities.
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 Regulatory Guide (RG) 1.147,
Revision 14.
6. Add 10 CFR 50.55a(b)(2)(xxviii) to correct a typographical error
regarding an exponent in the evaluation of pressurized water reactor
(PWR) reactor vessel head penetration nozzles.
7. Remove 10 CFR 50.55a(g)(6)(ii)(A) and associated paragraphs on
the augmented examination of the reactor vessel.
8. Add a paragraph (D) Reactor Vessel Head Inspections to 10 CFR
50.55a(g)(6)(ii) to require an inservice inspection program augmented
by the provisions of 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''
subject to conditions and remove Footnote 10.
9. Add a paragraph (E) Reactor Coolant Pressure Boundary Visual
Inspections to 10 CFR 50.55a(g)(6)(ii)--Augmented Inspection of Class 1
Components Fabricated with Alloy 600/82/182 Materials to require an
inservice inspection program augmented by the provisions of ASME Code
Case N-722, ``Additional Inspections for PWR Pressure Retaining Welds
in Class 1 Pressure Boundary Components Fabricated with Alloy 60/82/182
Materials, Section XI, Division 1'' subject to conditions.
II. Summary of Proposed Revisions to 10 CFR 50.55a
The changes to paragraphs (b) and (g) of 10 CFR 50.55a are
discussed below. Paragraphs (a), (c), (d), (e), and (f) would remain
unchanged because the requirements in these sections would not be
changed by virtue of the incorporating by reference of the 2004 Edition
of the ASME Code, Sections III and XI, and the OM Code.
Section III, ASME BPV Code
The proposed rule would revise Sec. 50.55a(b)(1) to incorporate by
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reference the 2004 Edition of Section III of the ASME BPV Code. The NRC
does not propose to adopt any limitations with respect to the 2004
Edition of Section III.
Section Xl, ASME BPV Code
The proposed rule would revise Sec. 50.55a(b)(2) to incorporate by
reference the 2004 Edition of the ASME BPV Code, Section XI, Division
1, subject to the proposed modifications and limitations discussed
below:
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 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 Section XI after the 1989 Edition. The NRC 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
Section XI regarding the application of IWB-1220 in Paragraph 10 CFR
50.55a(b)(2)(xi) (64 FR 51394).
The proposed revision would remove 10 CFR 50.55a(b)(2)(xi), thereby
permitting the use of ASME Section XI IWB-1220 of any edition or
addenda of ASME 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 (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, and (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 inservice inspections.
10 CFR 50.55a(b)(2)(xiii)--Mechanical Clamping Devices
Paragraph 50.55a(b)(2)(xiii) permits licensees to use the
provisions of Code Case N-523-1, ``Mechanical Clamping Devices for
Class 2 and 3 Piping.'' The proposed revision would remove 10 CFR
50.55a(b)(2)(xiii) because Code Case N-523-2, which provides updated
requirements to those of Code Case N-523-1, has been accepted in RG
1.147, Revision 14, ``Inservice Inspection Code Case Acceptability,
ASME Section XI, Division 1,'' which is incorporated by reference into
10 CFR 50.55a(g)(4)(I) and 10 CFR 50.55a(g)(4)(ii).
10 CFR 50.55a(b)(2)(xv)--Appendix VIII Specimen Set and Qualification
Requirements
Paragraph 50.55a(b)(2)(xv) specifies implementation of Appendix
VIII of Section XI, the 1995 Edition through the 2001 Edition of the
ASME BPV Code with regard to ultrasonic examinations of piping systems.
The proposed change would reference and allow the use of the 2004
Edition of the ASME Code.
10 CFR 50.55a(b)(2)(xx)--System Leakage Tests
Paragraph 50.55a(b)(2)(xx) would be revised to require that after
system leakage tests performed during 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), NDE must
be performed in accordance with IWA-4540(a)(2) of the 2002 Addenda of
Section XI. This provision would require that (1) the NDE method and
acceptance criteria of the 1992 edition or later of Section III be 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.
Subarticle IWA-4540(a) of the 1995 edition of ASME Section XI
requires that after welding on a pressure retaining boundary or
installing an item by welding or brazing, 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 Installation of Replacement Items by Welding
Class 1, 2, and 3, Section XI, Div. 1,'' which provides an alternative
to the hydrostatic pressure test. (NRC has accepted Code Case N-416-3
in RG 1.147, Revision 14 which has been incorporated by reference and
approved in 10 CFR 50.55a (70 FR 56809; Sept 29, 2005).
Code Case N-416-3 allows that instead of performing a hydrostatic
pressure test for welding and brazing repair/replacement activities,
performing a system leakage test if two requirements are met. The first
requirement is that a NDE be performed on welded or brazed repairs and
fabrication and installation joints in accordance with the methods and
acceptance criteria of the applicable subsection of the 1992 Edition of
Section III. Depending on the category of the weld, the NDE must
consist of, in most cases, radiography and examination by either the
liquid penetrant or magnetic particle method. The second requirement is
that prior to or immediately upon return to service, a visual
examination (VT-2) of welded or brazed repairs, fabrication, and
installation joints be performed in conjunction with a system leakage
test at nominal operating pressure and temperature in accordance with
paragraph IWA-5000 of the 1992 edition of Section XI. The technical
provisions of ASME Code Case N-416-3 were incorporated into the 2001
Edition of ASME Section XI, IWA-4540(a) and maintained, with minor
editorial changes, through the 2002 Addenda to ASME Section XI. The
2003 Addenda of the Code, IWA-4540(a) eliminated reference to the NDE
requirements of the 1992 Edition of Section III. When the ASME
developed the 2003 Addenda, the arguments in support of the Code action
state that imposing the NDE requirement in accordance with Section III
(i.e., radiography) on all repair and replacement activities is
excessively burdensome. The industry argued that the purpose of the
radiography requirements is to support the piping
[[Page 16734]]
joint efficiency factors used in the design. As such, the requirements
are appropriately imposed by the construction code or the design
specification but radiography for repair and replacement activities
would be excessive.
The industry also contended that a system leakage test compared to
a hydrostatic pressure test revealed very few cases in which leakage
occurred at the hydrostatic pressure but not at the lower pressure of
the system leakage test. Those cases involved only a small amount of
leakage and the source of the leakage would not have been detected by
additional NDE and is therefore not warranted.
NRC observes that the arguments to eliminate the NDE are from an
operational rather than a safety perspective. A safety assessment has
not been provided to demonstrate that without volumetric examination, a
system leakage pressure test alone provides a level of safety
equivalent to a hydrostatic pressure test, only that a volumetric
examination is excessively burdensome. NRC therefore concludes that to
provide reasonable assurance of adequate protection to public health
and safety, when performing a system leakage test in lieu of a
hydrostatic test after repair/replacement activities, a NDE must be
performed. It must be performed in accordance with the NDE provision in
IWA-4540(a)(2) of the 2002 Addenda of Section XI because the agency has
already accepted this provision by virtue of approving Code Case N-416-
3 in RG 1.147, Revision 14. That provision states that: (a) The NDE
method and acceptance criteria of the 1992 edition or later of Section
III shall be met prior to return to service; and (b) a system leakage
test shall be performed in accordance with IWA-5000 prior to or as part
of returning to service.
10 CFR 50.55a(b)(2)(xxi)--Table IWB-2500-1 Examination Requirements
Paragraph 10 CFR 50.55a(b)(2)(xxi)(A) would be revised to be
consistent with the condition for Code Case N-648-1, ``Alternative
Requirements for Inner Radius Examination of Class 1 Reactor Vessel
Nozzles, Section XI, Division 1,'' in RG 1.147, Revision 14, 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 proposed revision would state: ``A visual
examination with enhanced magnification that has a resolution
sensitivity to detect a 1-mil (0.001 inch) width wire or crack, using
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, where a and l are the depth and
length of the crack, respectively), may be performed instead of an
ultrasonic examination * * *''. This limitation 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.
10 CFR 50.55a(b)(2)(xxviii)--Evaluation Procedure and Acceptance
Criteria for PWR Reactor Vessel Head Penetration Nozzles
In the 2004 Edition of ASME Section XI, IWA-3660 specifies
evaluation procedure and acceptance criteria for flaws that are
detected in upper and lower reactor vessel head penetration nozzles in
PWRs. The procedure and acceptance criteria in IWB-3660 were adopted
from Code Case N-694-1, ``Evaluation Procedure and Acceptance Criteria
for PWR Reactor Vessel Head Penetration Nozzles Section XI, Division
1.'' Under IWB-3660, IWB-3662 specifies that the flaw shall be
evaluated using analytical procedures such as those described in non-
mandatory Appendix O, ``Evaluation of Flaws in PWR Reactor Vessel Upper
Head Penetration Nozzles,'' to the ASME Code, Section XI. There is a
typographical error in paragraph O-3220(b), equation SR = [
1 -0.82R] -22. The exponent should be -2.2, not -22.
Paragraph 50.55a(b)(2)(xxviii) would be added to the regulation to
ensure that the correct exponent is used. The exponent in Appendix O
was shown to be erroneous by an NRC report, NUREG/CR-6721, ``Effects of
Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue
and Stress Corrosion Cracking of Nickel Alloys and Welds,'' April 2001.
10 CFR 50.55a(g)(6)(ii)(A)--Augmented Examination of Reactor Vessel
Paragraph 50.55a(g)(6)(ii) which requires a one-time augmented
inservice inspection programs for those systems and components for
which the Commission determines that added assurance of structural
reliability is necessary would be removed. 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
implementation of the reactor vessel shell weld examinations specified
in the 1989 Edition of the ASME Code, Section XI, Division 1, in item
B1.10, ``Shell Welds,'' of Examination Category B-A, ``Pressure
Retaining Welds in Reactor Vessel,'' in Table IWB-2500-1 of the ASME
Code, Section XI. Since 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)--Augmented Inspection of PWR Reactor Vessel
Heads.
Paragraph 50.55a(g)(6)(ii)(D) of the proposed rule would be added
to require licensees to comply with the reactor vessel head inspection
requirements of ASME Code Case N-729-1, subject to conditions.
Compliance to Code Case N-729-1 with conditions 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. Footnote 10 to 10 CFR
50.55a would be removed because Code Case N-729-1, as conditioned,
would replace the requirements of the NRC Order EA-03-009 cited in that
footnote. That footnote states:
Supplemental inservice inspection requirements for reactor
vessel pressure heads have been imposed by Order EA-03-09 issued to
licensees of pressurized water reactors. The NRC expects to develop
revised supplemental inspection requirements, based in part upon a
review of the initial implementation of the order, and will
determine the need for incorporating the revised inspection
requirements into 10 CFR 50.55a by rulemaking.
Conditions are imposed on Code Case N-729-1 regarding inspection
frequency, examination coverage, qualification of ultrasonic
examination, and reinspection intervals. These conditions are being
imposed to make the requirements in N-729-1 equivalent to those of the
Order.
10 CFR 50.55a(g)(6)(ii)(E)--Augmented Inspection of Class 1 Components
Fabricated With Alloy 600/82/182 Materials
A new paragraph, 10 CFR 50.55a(g)(6)(ii)(E) Reactor Coolant
Pressure Boundary Visual Inspections would be added to require all
current and future licensees to apply ASME Code Case N-722, with
conditions.
[[Page 16735]]
The application of ASME Code Case N-722 is necessary because
current inspections are inadequate and the safety consequences can be
significant. NRC's determination that existing inspections of the
reactor coolant pressure boundary (RCPB) are inadequate are 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 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 Code, Section are inadequate because
Table IWB-2500-1, ``Examination Category B-P of Section XI'' only
requires a visual examination of the reactor vessel 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
primary water stress corrosion cracking (PWSCC) in the RCS pressure
boundary because of limitations of the ASME BPV Code inspection
programs to address PWSCC in the RCPB. In addition to issuing
bulletins, 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 measures, but these were only short term, such as 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.
ASME also took actions to address PWSCC. An ASME task group
concluded that more rigorous inspections than those currently provided
by the ASME Code are needed in the areas most susceptible to PWSCC. The
task group developed ASME Code Case N-722 to enhance the current ASME
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 Code
inspection requirements, because it specifies bare metal visual
examinations; however, such examinations are inadequate. Visual
inspections do not always detect through-wall leakage or related
corrosion until significant degradation has occurred.
Beyond the base metal visual inspection requirements and
frequencies of inspections, ASME Code Case N-722 is relatively limited
in scope. The NRC proposes to require non-visual inspection for items
where leakage is identified in Class 1 components. The additional non-
visual NDE would be 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 locations. The NRC proposes to
require 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 severance of the piping
and a loss-of-coolant-accident.
Therefore, the NRC proposes to require the application of Code Case
N-722 with additional conditions; namely, to require additional NDE
when leakage is detected and expansion of the sample size if a
circumferential PWSCC flaw is detected. Operating experience has shown
that bare metal visual inspections alone are not sufficient and that
NDE is necessary in order to detect cracking.
ASME OM Code
The proposed revision to Sec. 50.55a(b)(3) would incorporate by
reference the 2004 Edition of the ASME OM Code subject to no new
modifications or limitations.
Paragraph (b)(3)(iv)(D) would be revised to be less specific with
regard to paragraph references in subsection ISTC [In-service 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.
III. 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
in-service inspection 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 In-service
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 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 Code edition that is
proposed for incorporation by reference into Sec. 50.55a by this
proposed rule. NRC finds that the 2004 Edition of Sections III and XI
of the ASME BPV Code 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 acceptable alternatives to the requirements of the
2001 Edition up to and including the 2003 Addenda of the ASME Code,
Section XI, referenced in the GALL AMPs in its plant-specific
[[Page 16736]]
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 and the ASME Code edition and addenda used in the plant-specific
license renewal application 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 identified AMPs of the 2001 Edition through the
2003 Addenda of Section XI of the ASME Code that require augmentation
(additional requirements) for license renewal. These areas that require
augmentation also apply when implementing the 2004 edition. A license
renewal applicant may either augment its AMPs in these areas as
described in the GALL report or propose alternatives for NRC review in
its plant-specific license renewal application.
IV. Availability of Documents
The NRC is making the documents identified below available to
interested persons through one or more of the following methods as
indicated.
Public Document Room (PDR). The NRC Public Document Room is located
at 11555 Rockville Pike, Rockville, Maryland.
Rulemaking Web site (Web). The NRC's interactive rulemaking Web
site is located at http://ruleforum.llnl.gov. These documents may be
viewed and downloaded electronically via this Web site.
NRC's Electronic Reading Room. The NRC's public electronic reading
room is located at http://www.nrc.gov/reading-rm/adams.html.
NRC Staff Contact. Single copies of the Federal Register Notice
(which includes the draft Environmental Assessment) and draft
Regulatory Analysis can be obtained from Lee Banic, Division of Policy
and Rulemaking, Office of Nuclear Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001 or at (301) 415-2771,
or via e-mail at: [email protected].
----------------------------------------------------------------------------------------------------------------
Document PDR Web ADAMS No. NRC staff
----------------------------------------------------------------------------------------------------------------
ASME BPV Code*........................... ............ ............ N/A........................ X
ASME OM Code*............................ ............ ............ N/A........................ X
ASME Code Case N-722..................... X ............ ML070170676................ X
ASME Code Case N-729-1................... X ............ ML070170679................ X
Proposed Federal Register Notice......... X X ML070240552................ X
Draft Regulatory Analysis................ X X ML070290497................ X
EA-03-009................................ X X ML030380470................ X
First Revised NRC Order EA-03-009........ X X ML040220181................ X
GALL Report, NUREG-1801.................. ............ X ML012060392................ X
ML012060514................
ML012060521................
ML012060539................
Staff Requirements Memorandum (SRM) dated ............ ............ ML003751061................
September 10, 1999.
RG 1.147, Revision 14.................... X X ML052510117................ X
----------------------------------------------------------------------------------------------------------------
*Available on the ASME Web site.
V. Plain Language
The Presidential Memorandum dated June 1, 1998, entitled, ``Plain
Language in Government Writing,'' directed that the Federal
government's writing must be in plain language. This memorandum was
published on June 10, 1998 (63 FR 31883). The NRC requests comments on
this proposed rule specifically with respect to the clarity and
effectiveness of the language used. Comments should be sent to the
address listed under the ADDRESSES caption above.
VI. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995, Pub.
L. 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. Pub. L. 104-113 requires Federal agencies to use
industry consensus standards to the extent practical; it does not
require Federal agencies to endorse a standard in its entirety. 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 proposing to amend 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, in-service inspection, and in-
service 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 proposes not to
adopt, or to partially adopt, are identified in Section 2 of the
preceding section and the draft regulatory analysis. The justification
for not adopting parts of the ASME BPV Code, as set forth in these
statements of consideration and the draft regulatory analysis for this
proposed rule, satisfy the requirements of Section 12(d)(3) of Pub. L.
104-113, Office of Management and Budget (OMB) Circular A-119, and the
Commission's direction in the SRM dated September 10, 1999.
In accordance with the National Technology Transfer and Advancement
Act of 1995 and OMB Circular A-119, the NRC is requesting public
comment regarding whether other national or international consensus
standards could be endorsed as an alternative to the ASME BPV Code and
the ASME OM Code.
[[Page 16737]]
VII. Finding of No Significant Environmental Impact: Availability
This proposed 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, Pub. L. 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 (United States Code, Vol. 42, Section 4332(C) [42
U.S.C. Sec. 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, if adopted, would not be a major Federal action significantly
affecting the quality of the human environment and, therefore, an
environmental impact statement is not required.
The proposed 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 proposed changes concerning
ensuring the integrity of the RCPB would reduce the probability of
accidents and radiological impacts on the public. The proposed
rulemaking does not involve non-radiological plant effluents and has no
other environmental impact. Therefore, no significant non-radiological
impacts are associated with the proposed action.
The determination of this draft environmental assessment is that
there will be no significant off-site impact to the public from this
action. However, the NRC is seeking public comment of the draft
environmental assessment. Comments on any aspect of the environmental
assessment may be submitted to the NRC as indicated under the ADDRESSES
heading of this document.
The NRC is sending a copy of the environmental assessment and this
proposed rule to every State Liaison Officer and requesting their
comments on the environmental assessment.
VIII. Paperwork Reduction Act Statement
This proposed rule increases the burden on licensees to report
requirements and maintain records for examination requirements in ASME
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.
IX. Regulatory Analysis
The NRC has prepared a draft regulatory analysis on this proposed
rule. The draft 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 draft regulatory analysis may be
obtained as indicated in Section 4 of this document. The Commission
requests public comment on the draft regulatory analysis and comments
may be submitted to the NRC as indicated under the ADDRESSES heading.
X. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this proposed amendment will not,
if promulgated, have a significant economic impact on a substantial
number of small entities. This proposed amendment would affect 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 staff 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
proposed rule would incorporate 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
[[Page 16738]]
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 endorsement of a later Code 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 paragraph 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 proposed rule are consistent with the
application of backfit requirements to modifications and limitations in
previous rules.
There are some circumstances in which the endorsement of a later
ASME BPV Code or OM Code introduces a backfit. In these cases, the NRC
would perform a backfit analysis or documented evaluation in accordance
with paragraph 50.109. These include the following:
(1) When the NRC endorses 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, see, 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 endorsed the Code without any
expedited language, see, e.g., 64 FR 51370 (September 22, 1999).
(3) When the NRC takes an exception to a 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, see, e.g., 67 FR 60529 (September 26, 2002).
The backfitting discussion for the proposed revisions to 10 CFR
50.55a is set forth below:
1. Remove 10 CFR 50.55a(b)(2)(xi) Concerning Components Exempt From
Examination
This change would remove an existing limitation on the use of 1989
Addenda and later editions and addenda of the ASME 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.''
Paragraph 10 CFR 50.55a(b)(2)(xiii) states that ``Licensees may use
the provisions of Code Case N-523-1, ``Mechanical Clamping Devices for
Class 2 and 3 Piping.'' Paragraph 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 14, which is incorporated by reference into paragraphs 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 14, Code Case N-523-1 has been superseded by Code Case
N-523-2. It is stated in RG 1.147, Revision 14, 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 14. 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 would update the edition of the ASME 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 Code,
Section XI, requires a 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 Code, Section XI, does not specify a NDE after a system
leakage test. The proposed addition would require, as part of repair
and replacement activities, that a NDE be performed per IWA-4540(a)(2)
of the 2002 Addenda of the ASME 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 Code, Section
XI.
As it is stated above, 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 Code, Section XI, but supplements it with the NDE
of IWA-4540(a)(2) of the 2002 Addenda of the Code. The proposed
addition does not represent a new staff requirement because the NDE
requirement is specified in previous addenda of the Code. Therefore,
this change is not considered as a backfit under 10 CFR 50.109.
[[Page 16739]]
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 14
This change would align 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 14 (70 FR 5680; Sept 29, 2005).
The imposed conditions do not represent a new staff position.
Therefore, this change is not considered as a backfit under 10 CFR
50.109.
6. Add 10 CFR 50.55a(b)(2)(xxviii) To Correct a Typographical Error
Regarding an Exponent in the Evaluation of PWR Reactor Vessel Head
Penetration Nozzles
This change would correct a typographical error in an equation used
in the flaw evaluation in the ASME Section XI. Therefore, this change
is not considered as a backfit under 10 CFR 50.109.
7. Remove 10 CFR 50.55a(g)(6)(ii)(A) and Associated Subparagraphs on
the Augmented Examination of the Reactor Vessel
This change would remove 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.
8. Add Paragraph (D) to 10 CFR 50.55a(g)(6)(ii)--Augmented Inspection
of PWR Reactor Vessel Heads
The requirements in paragraph D, which impose ASME Code Case N-729-
1 with conditions, were already imposed on existing licensees under NRC
First Revised Order EA-03-009. Therefore, this requirement is not
considered a backfit under 10 CFR 50.109(a)(1).
9. Add Paragraph (E) to 10 CFR 50.55a(g)(6)(ii)--Augmented Inspection
of Class 1 Components Fabricated With Alloy 600/82/182 Materials
The NRC proposes to add 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 inservice inspection
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 proposed 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.
As discussed earlier in Section 2, ``10 CFR 50.55a(g)(6)(ii)(E)--
Augmented Inspection of Class 1 Components Fabricated with Alloy 600/
82/182 Materials,'' 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 the 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 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 the proposed 10 CFR 50.55a(g)(6)(ii)(E) is a compliance
backfit under 10 CFR 50.109(a)(4)(i).
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Incorporation by reference, Intergovernmental relations,
Nuclear power plants and reactors, Radiation protection, Reactor siting
criteria, Reporting and recordkeeping requirements.
For the reasons set forth in the preamble and under the authority
of the Atomic Energy Act of 1954, as amended, the Energy Reorganization
Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is proposing
to adopt the following amendments to 10 CFR Part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
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,
[[Page 16740]]
1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44
U.S.C. 3504 note).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). 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).
2. Section 50.55a is amended by revising the introductory text of
paragraphs (b)(1) and (b)(2), removing and reserving paragraphs
(b)(2)(xi) and (b)(2)(xiii), revising the introductory text of
paragraph (b)(2)(xv) and paragraphs (b)(2)(xx) and (b)(2)(xxi)(A),
adding paragraph (b)(2)(xxviii), revising the introductory text of
paragraph (b)(3) and paragraph (b)(3)(iv)(D), removing and reserving
paragraph (g)(6)(ii)(A), adding paragraphs (g)(6)(ii)(D) and
(g)(6)(ii)(E), and removing Footnote 10.
Sec. 50.55a Codes and standards.
* * * * *
(b) * * *
(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:
* * * * *
(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 the 2004 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.
* * * * *
(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, Item 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 enhanced
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.
* * * * *
(xxviii) Evaluation Procedure and Acceptance Criteria for PWR
Reactor Vessel Head Penetration Nozzles. When performing flaw growth
calculations in accordance with non-mandatory Appendix O of Section XI
of the ASME Code, as permitted by IWB-3660, the licensee shall use
exponent-2.2 as the exponent in the SR equation in
Subarticle O-3220.
(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 by implementing ASME Code Case N-729-1 subject to the
conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this
section.
(2) Item B4.40 of Table 1 must be inspected at least every fourth
refueling outage or at least every seven calendar years, whichever
occurs first, after the first ten-year inspection interval.
(3) Instead of fulfilling the specified `examination method'
requirements for volumetric and surface examinations of Note 6 in Table
1, the licensee shall perform a volumetric or surface examination or
both of essentially 100 percent of the required volume or equivalent
surfaces of the nozzle tube, as identified by Fig. 2 of ASME Code Case
N-729-1. A surface examination must be performed on all J-groove welds.
If a surface examination is 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 Fig. 2 of ASME Code Case N-729-1), the surface
examination must be of the inside and outside wetted surfaces of the
penetration nozzle not examined volumetrically.
(4) Ultrasonic examinations must 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 paragraphs (g)(6)(ii)(D)(4)(i) through (iv)
of this section instead of using a methodology that satisfies the
conditions specified by the qualification requirements of Paragraph-
2500 of ASME Code Case N-729-1.
(i) The diameters of pipes in the specimen set shall be within \1/
2\ in. (13 mm) of the nominal diameter of the qualification pipe size
and a thickness tolerance of 25 percent of the nominal
through-wall depth of the qualification pipe thickness. The specimen
set must contain geometric and material
[[Page 16741]]
indications that normally require discrimination from primary water
stress corrosion cracking (PWSCC) flaws.
(ii) The specimen set must have a minimum of ten (10) flaws that
provide an acoustic response similar to that of PWSCC indications. All
flaw depths in the specimen set must be greater than 10 percent of the
nominal pipe wall thickness. A minimum number of 30 percent of the
total flaws must be connected to the outside diameter and 30 percent of
the total flaws must be connected to the inside diameter. Further, at
least 30 percent of the total flaws must measure from a depth of 10 to
30 percent of the wall thickness and at least 30 percent of the total
flaws must measure from a depth of 31 to 50 percent of the wall
thickness and be connected to the inside or outside diameter, as
applicable. At least 30 percent, but no more than 60 percent, of the
flaws must be oriented axially.
(iii) The procedures must identify the equipment and essential
variable settings used to qualify the procedures. An essential variable
is defined as any variable that affects the results of the examination.
The procedure must be requalified when an essential variable is changed
to fall outside the demonstration range. A procedure must be qualified
using the equivalent of at least three test sets that are used to
demonstrate personnel performance. Procedure qualification must require
at least one successful personnel performance demonstration.
(iv) The test acceptance criteria for a personnel performance
demonstration must meet the detection test acceptance criteria for
personnel performance demonstration in Table VIII-S10-1 of Section XI,
Appendix VIII, Supplement 10. Examination procedures, equipment, and
personnel must be considered qualified for depth sizing only if the
root mean square (RMS) error of the flaw depth measurements, as
compared to the true flaw depths, does not exceed 1/32-inch (0.8 mm).
Examination procedures, equipment, and personnel must be considered
qualified for length sizing if the RMS error of the flaw length
measurements, as compared to the true flaw lengths, does not exceed 1/
16-inch (1.6 mm).
(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 reinspection interval must be each
refueling outage instead of the reinspection intervals required by
Table 1, Note (8) of ASME Code Case N-729-1.
(6) Appendix I of ASME Code Case N-729-1 must not be implemented
without prior NRC approval.
(E) Reactor Coolant Pressure Boundary Visual Inspections. (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 only apply
to components fabricated with Alloy 600/82/182 materials not 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.
* * * * *
Dated at Rockville, Maryland, this 26th day of March, 2007.
For the U.S. Nuclear Regulatory Commission.
Luis A. Reyes,
Executive Director.
[FR Doc. E7-6379 Filed 4-4-07; 8:45 am]
BILLING CODE 7590-01-P