[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

[[Page 16733]]

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