[Federal Register Volume 67, Number 187 (Thursday, September 26, 2002)]
[Rules and Regulations]
[Pages 60520-60542]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-23811]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AG61


Industry Codes and Standards; Amended Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its 
regulations to incorporate by reference a later edition and addenda of 
the American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (BPV Code) and the ASME Code for Operation and Maintenance 
of Nuclear Power Plants (OM Code) to provide updated rules for 
construction, inservice inspection (ISI), and inservice testing (IST) 
of components in light-water cooled nuclear power plants. This final 
rule incorporates by reference the latest edition and addenda of the 
ASME BPV and OM Codes that have been approved for use by the NRC 
subject to certain limitations and modifications.

EFFECTIVE DATE: October 28, 2002. The incorporation by reference of 
certain publications in this rule is approved by the Director of the 
Office of the Federal Register as of October 28, 2002

ADDRESSES: The NRC maintains an Agencywide Documents Access and 
Management System (ADAMS), which provides text and image files of NRC's 
public documents. The documents may be accessed through the NRC's 
Public Electronic Reading Room on the Internet at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC at 1-800-397-4209, (301) 415-4737, or by email to [email protected]. The 
availability of the Regulatory Analysis, Environmental Assessment, and 
Resolution of Public Comments associated with this rulemaking is 
further discussed in Section 5 below, under SUPPLEMENTARY INFORMATION.

FOR FURTHER INFORMATION CONTACT: Stephen Tingen, Division of 
Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001. Alternatively, you 
may contact Mr. Tingen at (301) 415-1280, or via e-mail at: 
[email protected].

SUPPLEMENTARY INFORMATION:

1. Background
2. Public Comments on Proposed Rule; and Final Rule
2.1 Section III
2.2 Section XI
2.2.1 Owner-Defined Requirements for Class CC and Class MC 
Components
2.2.1.1 Visual Examination Qualification Requirements (Class CC 
Components)
2.2.1.2 Visual Examination Qualification Requirements (Class MC and 
Liners of Class CC)
2.2.1.3 General and Detailed Examinations
2.2.2 Examination of Containment Bolted Connections
2.2.3 Acceptance Standard for Surfaces Requiring Augmented 
Ultrasonic Examinations
2.2.4 Containment Penetration Piping
2.2.5 Certification of Nondestructive Examination Personnel
2.2.6 Substitution of Alternative Methods
2.2.7 System Leakage Tests
2.2.8 Table IWB-2500-1 Examination Requirements
2.2.9 Supplemental Annual Training Requirements for Ultrasonic 
Examiners
2.2.10 Underwater Welding
2.3 Appendix VIII to Section XI
2.3.1 Examination Coverage for Dissimilar Metal Pipe Welds
2.3.2 Reactor Vessel Single Side Examinations
2.3.3 Qualification Test Samples
2.3.4 Implementation of Appendix VIII to Section XI
2.4 ASME OM Code
3. Section-by-Section Analysis of Substantive Changes
4. Generic Aging Lessons Learned Report
5. Availability of Documents
6. Voluntary Consensus Standards
7. Finding of No Significant Environmental Impact: Availability
8. Paperwork Reduction Act Statement
9. Regulatory Analysis
10. Regulatory Flexibility Certification
11. Backfit Analysis
12. Small Business Regulatory Enforcement Fairness Act

1. Background

    On August 3, 2001 (66 FR 40626), the NRC published a Federal 
Register notice that presented a proposed rule to amend 10 CFR part 50, 
``Domestic Licensing of Production and Utilization Facilities.'' The 
proposed rule would revise the requirements for construction, ISI, and 
IST of nuclear power plant components. For construction, the proposed 
rule would permit the use of Section III, Division 1, of the ASME BPV 
Code, 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda for 
Class 1, Class 2, and Class 3 components with no new modifications or 
limitations.
    For ISI, the proposed rule would permit the use of Section XI, 
Division 1, of the ASME BPV Code, 1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda for Class 1, Class 2, Class 3, Class MC, and 
Class CC components with new modifications and limitations.
    For IST, the proposed rule would permit the use of the ASME OM 
Code, 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda for 
Class 1, Class 2, and Class 3 pumps and valves with one new 
modification.
    In the same Federal Register notice, the Commission withdrew a 
proposed rule (64 FR 22580; April 27, 1999) that would have eliminated 
the requirement for licensees to update their ISI and IST programs 
every 120 months beyond a

[[Page 60521]]

baseline edition and addenda of the ASME Code. That withdrawal was a 
final action--not part of the proposed rule.

2.0 Public Comments on Proposed Rule; and Final Rule

    Interested parties submitted written comments on the proposed rule 
published on August 3, 2001 (66 FR 40626). Comments were received from 
17 separate sources. These sources consisted of 10 utilities, 4 service 
organizations, and 3 individuals. In response to the public comments, 
the NRC has either removed or revised some modifications and 
limitations that were proposed. A summary of the public comments 
applicable to the proposed rule and their resolution are provided in 
the following sections. Public comments on the proposed rule that are 
not addressed in the final rule are addressed in the Resolution of 
Public Comments. The availability of the Resolution of Public Comments 
is further discussed in Section 5 below.
    The NRC has considered and resolved the public comments and has 
revised the final rule accordingly. The NRC is publishing these final 
regulations in Sec.  50.55a to incorporate by reference the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of Division 1 
rules of Section III of the ASME BPV Code; the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of Division 1 rules of Section 
XI of the ASME BPV Code; and the 1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda of the ASME OM Code for construction, ISI, 
and IST of components in nuclear power plants. Section III of the ASME 
BPV Code is acceptable for use with no new limitations or 
modifications. Section XI of the ASME BPV Code is acceptable for use 
subject to limitations and modifications. The ASME OM Code is 
acceptable for use subject to one modification.
    This final rule also revises the regulations in Sec.  50.55a that 
licensees use to modify the implementation of Appendix VIII, 
``Performance Demonstration for Ultrasonic Examinations Systems,'' to 
Section XI of the ASME BPV Code. The amendment clarifies existing 
ultrasonic (UT) examination qualification requirements in Sec.  50.55a. 
The amendment also adds new requirements to clarify the coordination of 
Appendix VIII with other parts of Section XI.

2.1 Section III

    There were no public comments on the proposed rule concerning 
Section III. This final rule revises Sec.  50.55a(b)(1) to incorporate 
by reference the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 
Addenda of Section III of the ASME BPV Code; clarify that the 1963 
Edition was the initial edition of Section III incorporated by 
reference in the regulations; and extend the applicability of the 
existing regulations in Sec. Sec.  50.55a(b)(1)(ii), 50.55a(b)(1)(iii), 
and 50.55a(b)(1)(v) to the 1997 Addenda, 1998 Edition, 1999 Addenda, 
and 2000 Addenda of Section III of the ASME BPV Code.

2.2 Section XI

    Public comments on the proposed rule concerning Section XI are 
addressed in the following sections. This final rule revises Sec.  
50.55a(b)(2) to incorporate by reference the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of Section XI of the ASME BPV 
Code; clarify that the 1970 Edition was the initial edition of Section 
XI incorporated by reference in the regulations; and extend the 
applicability of the existing regulations in Sec. Sec.  
50.55a(b)(2)(viii), 50.55a(b)(2)(ix), 50.55a(b)(2)(xi), 
50.55a(b)(2)(xv), and 50.55a(b)(2)(xvii) to the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of Section XI of the ASME BPV 
Code. This final rule also deletes the existing regulations in Sec.  
50.55a(g)(6)(ii)(B)(1) through (4) because the implementation dates 
have expired and all licensees have completed the requirements or have 
been approved by an exemption for a delay. The existing requirement 
that was formerly Sec.  50.55a(g)(6)(ii)(B)(5) is redesignated as Sec.  
50.55a(g)(6)(ii)(B).
    Although Sec.  50.55a(b)(2)(vi) is not addressed in the proposed 
rule, one commenter stated that Sec.  50.55a(b)(2)(vi) should be 
revised to include references to the 1998 Edition through the 2000 
Addenda of the ASME Code for the ISI of Class MC and Class CC 
components. The commenter noted that Sec. Sec.  50.55a(b)(2)(viii) and 
(ix) in the proposed rule reference the 1998 Edition through the 2000 
Addenda, therefore, Sec.  50.55a(b)(2)(vi) should also reference the 
1998 Edition through the 2000 Addenda.
    The NRC agrees that Sec.  50.55a(b)(2)(vi) should be revised to 
clarify the applicability of the 1998 Edition through the 2000 Addenda 
to containment ISI programs but does not agree with the revision 
suggested by the commenter. The statement of considerations for the 
final rule published on September 22, 1999 (64 FR 51370), states that 
either the 1992 Edition with the 1992 Addenda, or the 1995 Edition with 
the 1996 Addenda of IWE and IWL must be used to develop and implement a 
containment ISI program within 5 years. The NRC finds that the existing 
requirements in Sec.  50.55a(b)(2)(vi) only address the applicable 
edition and addenda of IWE and IWL to be used during initial 120-month 
interval for the ISI of Class CC and Class MC components. Therefore, 
Sec.  50.55a(b)(2)(vi) is revised to clarify that the 1992 Edition with 
the 1992 Addenda, or the 1995 Edition with the 1996 Addenda of IWE and 
IWL must be used when implementing the initial 120-month interval for 
the ISI of Class MC and Class CC components, and that successive 120-
month interval updates must be implemented in accordance with Sec.  
50.55a(g)(4)(ii).
    The proposed rule would add a new Sec.  50.55a(g)(6)(ii)(B)(1) to 
clarify the start date of the first 120-month interval for the ISI of 
Class MC and Class CC components. Some commenters indicated that Sec.  
50.55a(g)(6)(ii)(B)(1) in the proposed rule did not clarify the start 
date of the first 120-month interval for the ISI of Class MC and Class 
CC components. Other commenters suggested a revised regulation that 
they thought would be more appropriate.
    The NRC finds that the proposed regulation regarding the start date 
of the first 120-month interval for the ISI of Class MC and Class CC 
components has created confusion rather than clarifying existing 
requirements as intended. The clarification in the proposed rule would 
also create a hardship for many licensees in reestablishing the start 
date of their first 120-month containment ISI interval. It was not the 
intent of the NRC to alter the 10-year examination interval in IWE or 
the 5-year examination interval in IWL already established by 
licensees. Licensees are permitted to schedule examinations of Class MC 
and Class CC components in accordance with the requirements in IWE and 
IWL. Therefore, the clarification of the first 120-month interval start 
date in Sec.  50.55a(g)(6)(ii)(B)(1) in the proposed rule is not 
adopted.
    In responding to this clarification, several commenters indicated 
that the 10-year IWE and 5-year IWL examination intervals must coincide 
with the 120-month interval update in Sec.  50.55a(b)(4)(ii). The NRC 
does not agree that the 10-year IWE and 5-year IWL examination 
intervals must coincide with the 120-month interval update in Sec.  
50.55a(b)(4)(ii). The 10-year IWE and 5-year IWL examination intervals 
are independent of the 120-month interval update in Sec.  
50.55a(g)(4)(ii). Section 50.55a(g)(4)(ii) does not prohibit licensees 
from updating to a later edition and addenda of the ASME Code midway 
through a

[[Page 60522]]

10-year IWE or 5-year IWL examination interval.
    In responding to this clarification, several commenters implied 
that the final rule dated August 8, 1996 (61 FR 41303), requiring 
licensees to develop and implement a containment ISI program for Class 
MC components in accordance with IWE of Section XI, authorized the 
extension of the first period inspection from 40 months to 60 months in 
duration. The NRC does not agree. The schedule in the containment final 
rule did not extend the duration of the 40-month inspection period 
required by IWE. This issue was addressed in the response to Question 
13 in a letter to the Nuclear Energy Institute from NRC dated May 30, 
1997.
    In responding to this clarification, several commenters indicated 
that the final rules dated August 8, 1996 (61 FR 41303), and September 
22, 1999 (64 FR 51370), create a hardship when implementing 120-month 
interval updates required by Sec.  50.55a(g)(4)(ii). The NRC agrees 
with this comment. The final rule dated August 8, 1996, required 
licensees to implement an ISI program for Class MC and Class CC 
components using the 1992 Edition with the 1992 Addenda of IWE and IWL. 
The final rule dated September 22, 1999, required licensees to 
implement Appendix VIII UT qualification requirements using the 1995 
Edition with the 1996 Addenda of Section XI. Consequently, the schedule 
for 120-month interval updates for the ISI of Class MC and Class CC 
components, Appendix VIII UT qualification requirements, and the ISI of 
Class 1, 2, and 3 components might not coincide. This creates a 
hardship for licensees because ISI programs are required to maintain up 
to 3 separate editions and addenda of Section XI--one edition and 
addenda applicable to the ISI of Class MC and Class CC components, 
another edition and addenda applicable to the ISI of Class 1, 2, and 3 
components, and a third edition and addenda applicable to Appendix VIII 
UT qualification requirements. Therefore, licensees may wish to 
synchronize 120-month interval updates such that the same edition and 
addenda of Section XI apply to the ISI of Class MC and Class CC 
components, Appendix VIII UT qualification requirements, and the ISI of 
Class 1, 2, and 3 components. Licensees wishing to synchronize their 
120-month intervals may submit a request in accordance with Sec.  
50.55a(a)(3) to obtain authorization to extend or reduce 120-month 
intervals.

2.2.1 Owner-Defined Requirements for Class CC and Class MC Components

    The proposed rule addresses NRC concerns with ``owner-defined'' 
requirements in IWE and IWL. Revisions to the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of IWE and IWL permit each 
licensee to define requirements that were previously defined in the 
ASME Code.
    A number of commenters indicated that ``owner-defined'' 
requirements are acceptable because the regulations in Appendix B of 10 
CFR 50, ``Quality Assurance Criteria For Nuclear Power Plants and Fuel 
Reprocessing Plants,'' and the Responsible Engineer/Individual 
oversight provisions (as delineated in IWE and IWL) ensure that 
requirements defined by the owner are properly implemented.
    The NRC does not agree that the quality assurance requirements in 
Appendix B of 10 CFR 50 and the oversight duties of the Responsible 
Engineer/Individual alone are adequate to ensure that owner-defined 
requirements are properly implemented. The final rule published on 
August 8, 1996 (61 FR 41303), required licensees to develop and 
implement a containment ISI program for Class MC and Class CC 
components in accordance with IWE and IWL. The final containment rule 
stated that the rule was needed because none of the existing 
requirements provide specific guidance on how to inspect containment 
surfaces. This lack of guidance resulted in a large variation with 
regard to the performance and the effectiveness of licensee containment 
inspection programs. Based on the results of inspections and audits, as 
well as plant operational experiences, it was clear to the NRC that 
without specific guidance, several licensee containment inspection 
programs were unable to detect degradation that could ultimately result 
in a compromise to the containment pressure-retaining capability. Some 
containment structures had been found to have undergone a significant 
level of degradation that was not detected by existing programs. Given 
the number and the extent of the occurrences, and the variability among 
plants with regard to the performance and the effectiveness of 
containment inspections, the NRC believed that the prudent course of 
action was to impose the more specific ISI inspection requirements in 
the 1992 Edition with the 1992 Addenda of IWE and IWL. The containment 
final rule imposed requirements that define comprehensive and 
technically sound methods that ensure uniform containment inspection 
results among all licensees.
    The NRC believes that it is inappropriate to approve Code 
provisions that do not contain specific containment inspection guidance 
when prior experience demonstrates that specific containment inspection 
guidance is necessary. The quality assurance provisions in Appendix B 
of 10 CFR 50 and the oversight duties of the Responsible Engineer/
Individual do not ensure uniform containment inspection results among 
all licensees. Furthermore, the quality assurance provisions in 
Appendix B of 10 CFR 50 did not prevent the previous problems 
associated with a lack of guidance. Reliance on Appendix B of 10 CFR 50 
resulted in a large variation in the performance and effectiveness of 
licensees' containment inspection programs that contributed to the NRC 
issuing the containment final rule.

2.2.1.1 Visual Examination Qualification Requirements (Class CC 
Components)

    Section 50.55a(b)(2)(viii)(F) in the proposed rule would require 
that personnel who conduct visual examinations of containment surfaces 
be qualified in accordance with the 1998 Edition, 1999 Addenda, and 
2000 Addenda of IWA-2300 in place of the ``owner-defined'' 
qualification provisions in the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWL-2310(d). Prior to the 1998 Edition, the NRC-approved 
provisions in IWA-2300 were used to define the qualification 
requirements for personnel who conduct visual examinations of 
containment surfaces. The qualification requirements were revised in 
IWL-2310(d), 1997 Addenda, to allow the owner to define the 
qualification requirements for personnel who perform visual 
examinations of concrete and tendon anchorage hardware, wires, or 
strands. However, the new Code provision does not provide any criteria 
that the licensee must use when developing qualification requirements. 
Therefore, the NRC proposed that licensees continue to use the 
provisions in IWA-2300 to qualify personnel who perform visual 
inspections of containment concrete surfaces and tendon anchorage 
hardware, wires, or strands.
    Several commenters recommended that the NRC specify the use of a 
more generic standard for qualification of containment examiners such 
as ANSI N45.2.6, ``Qualifications of Inspection, Examination, and 
Testing Personnel for Nuclear Power Plants,'' to define personnel 
qualification provisions in

[[Page 60523]]

place of the requirements in IWA-2300. One commenter stated that 
licensees typically commit to meet the requirements of Regulatory Guide 
1.58, ``Qualification of Nuclear Power Plant Inspection, Examination, 
and Testing Personnel (Revision 1, September 1980),'' or another NRC-
approved standard that endorses ANSI N45.2.6. Another commenter noted 
that use of the qualification standards of IWA-2300, as proposed by the 
NRC, is not appropriate because they were designed for examinations 
associated with piping systems and their supports and not containment 
examinations.
    The NRC disagrees with the comments because the use of ``owner-
defined'' qualification requirements or a generic quality assurance 
standard to qualify containment examiners does not provide adequate 
guidance to ensure that examiners are qualified to inspect containment 
surfaces. The NRC prefers instead that the ASME Code identify the 
specific elements deemed necessary to ensure containment inspection 
qualification programs are adequate, or describe specific criteria that 
licensees must use to qualify personnel performing containment 
examinations. Although the existing qualification provisions in IWA-
2300 were not developed specifically for qualifying examiners of 
concrete containment surfaces, they provide the most practical criteria 
that are presently available for qualification of personnel that 
conduct visual examinations of containment surfaces. The NRC notes that 
many of the changes in the later editions and addenda of IWE and IWL 
are more suited to containment examinations than earlier editions and 
addenda. The NRC withdrew Regulatory Guide 1.58 on July 31, 1991 (56 FR 
36175). Therefore, the NRC no longer endorses the use of ANSI 45.2.6 
for the ISI of containment surfaces in operating nuclear power plants. 
Section 50.55a(b)(2)(viii)(F) in the proposed rule is adopted without 
change.

2.2.1.2 Visual Examination Qualification Requirements (Class MC and 
Liners of Class CC)

    Section 50.55a(b)(2)(ix)(F) of the proposed rule would require that 
personnel who conduct visual examinations of containment surfaces be 
qualified in accordance with the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWA-2300 of in place of the ``owner-defined'' qualification 
provisions in the 1998 Edition, 1999 Addenda, and 2000 Addenda IWE-
2330(a). Prior to the 1998 Edition, the NRC approved provisions in IWA-
2300 were used to define the qualification requirements for personnel 
who conduct visual examinations of containment surfaces.
    There was one public comment on Sec.  50.55a(b)(2)(ix)(F), which is 
discussed in the following Section 11, Backfit Analysis. In 
consideration of the public comment, the qualification requirements for 
personnel that conduct visual inspections of containment surfaces have 
been revised to require that VT-1 and VT-3 examinations must be 
conducted in accordance with the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWA-2200. Personnel conducting examinations in accordance 
with the VT-1 or VT-3 examination method shall be qualified in 
accordance with the 1998 Edition, 1999 Addenda, and 2000 Addenda of 
IWA-2300.

2.2.1.3 General and Detailed Visual Examinations

    Section 50.55a(b)(2)(ix)(G) in the proposed rule would require that 
the general and detailed visual examinations required by the 1998 
Edition, 1999 Addenda, and 2000 Addenda of IWE-2310(b) and IWE-2310(c) 
meet the VT-1 and VT-3 examination method provisions in the 1998 
Edition, 1999 Addenda, and 2000 Addenda of IWA-2210 in place of the 
``owner-defined'' general and detailed visual examination provisions in 
the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWE-2310(a), and 
allow licensees to continue to extend Table IWA-2210-1 maximum direct 
examination distance and decrease Table IWA-2210-1 minimum illumination 
requirements as currently stated in Sec.  50.55(b)(2)(ix)(B).
    The distance and illumination requirements in Sec.  
50.55a(b)(2)(ix)(G) in the proposed rule have been removed because 
these requirements are addressed in the existing Sec.  
50.55a(b)(2)(ix)(B). There was one public comment on Sec.  
50.55a(b)(2)(ix)(G), which is discussed in the following Section 11, 
Backfit Analysis. In consideration of the public comment, Sec.  
50.55a(b)(2)(ix)(G) is revised to require that the VT-1 and VT-3 
examination methods in the 1998 Edition, 1999 Addenda, and 2000 Addenda 
of IWA-2200 be used to conduct specific visual examinations in Table 
IWE-2500-1 in place of the ``owner-defined'' visual examination methods 
in the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWE-2310(b) and 
IWE-2310(c). The VT-3 examination method must be used to conduct the 
examinations in Items E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 
examination method must be used to conduct the examination in Item 
E4.11 of Table IWE-2500-1. An examination of the pressure-retaining 
bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 
examination method must be conducted once each interval.

2.2.2 Examination of Containment Bolted Connections

    Section 50.55a(b)(2)(ix)(H) of the proposed rule would require that 
the acceptance standard in the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWC-3513 be used to evaluate flaws in pressure-retaining 
bolting greater than or equal to 51 millimeters [2.0 inches] in 
diameter which are identified during the examination of containment 
surfaces. The acceptance standard would be used in place of the 
``owner-defined'' acceptance standard in the 1998 Edition, 1999 
Addenda, and 2000 Addenda of IWE-3510.1.
    Several commenters stated that Sec.  50.55a(b)(2)(ix)(H) of the 
proposed rule is unnecessary because there are no substantial 
differences between the revised standard for bolting materials in the 
1998 Edition and the standard for bolting materials in editions and 
addenda earlier than the 1998 Edition. The NRC disagrees. The bolting 
standard for bolting materials in the editions and addenda of IWE-
3515.1 earlier than the 1998 Edition was significantly revised in the 
1998 Edition. Prior to the 1998 Edition, IWE-3515.1 stated that bolting 
material must be examined in accordance with the material specification 
for defects which may cause the bolted connection to violate either the 
containment leak-tight or structural integrity. IWE-3515.1 was revised 
and renumbered as IWE-3510.3 in the 1998 Edition to require that the 
owner define the standard for examining bolting materials. Since 
containment bolting is not unique from other bolting applications in 
Section XI, the NRC finds that the examination of containment bolting 
should be consistent with other Section XI bolting examination 
requirements.
    A number of commenters stated that IWC-3513 is not the appropriate 
standard to use to evaluate flaws in pressure-retaining bolting. One 
commenter recommended that IWB-3517.1 be used in place of IWC-3513. The 
NRC agrees and finds that the visual examination criteria for bolting 
in IWE-3517.1 is an acceptable standard because it enures that the 
integrity of reused bolting is maintained. Section 50.55a(b)(2)(ix)(H) 
is revised to require that bolting material be examined in accordance 
with the material specification or the 1997 Addenda, 1998

[[Page 60524]]

Edition, 1999 Addenda, and 2000 Addenda of IWB-3517.1.
    Section 50.55a(b)(2)(ix)(I) in the proposed rule would require 
licensees to supplement the containment bolted connection examination 
requirements in Items E1.10 and E1.11 of the 1998 Edition, 1999 
Addenda, and 2000 Addenda of Table IWE-2500-1 with additional 
requirements for examining inaccessible areas of containment bolting.
    One commenter stated that since the ASME Code requires that 
accessible areas of containment bolted connections be more frequently 
examined in the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWE 
than in the earlier editions and addenda, bolting examination 
requirements have been enhanced. The NRC disagrees. Although the 
revised provisions increase the frequency of accessible examinations of 
containment bolting, the revised provisions reduce the frequency of 
examinations of inaccessible areas of containment bolting. The 1992 
Edition with the 1992 Addenda and the 1995 Edition with the 1996 
Addenda of IWE provide acceptable provisions for conducting 
examinations of the accessible and inaccessible areas of containment 
bolted connections. Item No. E8.10 of Table IWE-2500-1 requires that a 
visual examination of the individual parts of the bolted connection 
using the VT-1 visual examination method be performed whenever a 
connection is disassembled during a scheduled ISI inspection. Item 
E8.20 of Table IWE-2500-1 requires that a bolt torque or tension test 
be performed on bolted connections that have not been disassembled 
during the inspection interval. A bolt torque or tension test provides 
an indication of the integrity of the inaccessible areas of a bolted 
connection. The requirements in Items E8.10 and E8.20 requiring that 
containment bolting either be disassembled and examined (VT-1), or 
torque tested every interval were deleted in the 1998 Edition of IWE.
    Several commenters suggest that Sec.  50.55a(b)(ix)(I) of the 
proposed rule be revised to allow the option of conducting visual 
examinations of the inaccessible areas of containment bolted 
connections during maintenance that requires a bolted connection be 
disassembled or during visual examinations that are conducted during 
scheduled ISI inspections. In consideration of the public comments, the 
modification that was formerly Sec.  50.55a(b)(ix)(I) in the proposed 
rule is revised in the final rule to allow licensees the option of 
performing visual examinations of inaccessible areas of containment 
bolted connections during maintenance evolutions or scheduled 
inspections. Any bolted connections that are disassembled during the 
scheduled performance of Item E1.11 examinations must be examined using 
the VT-3 examination method. Flaws or degradation identified during the 
performance of this VT-3 examination must be examined in accordance 
with the VT-1 examination method. The criteria in the material 
specification, or the 1998 Edition, 1999 Addenda, and 2000 Addenda of 
IWB-3517.1 must be used to evaluate bolting flaws or degradation. As an 
alternative to performing the VT-3 examination during the scheduled 
performance of Item E1.11, VT-3 examination of bolting may be conducted 
whenever containment bolting in Item E1.11 is disassembled for any 
reason. Sections 50.55a(b)(ix)(I) and 50.55a(b)(ix)(H) in the proposed 
rule have been combined as Sec.  50.55a(b)(ix)(H) in this final rule.

2.2.3 Acceptance Standard for Surfaces Requiring Augmented Ultrasonic 
Examinations

    Section 50.55a(b)(2)(ix)(J) in the proposed rule would require that 
the ultrasonic (UT) examination acceptance standard specified in the 
1998 Edition, 1999 Addenda, and 2000 Addenda of IWE-3511.3 for Class MC 
pressure-retaining components also apply to metallic liners of Class CC 
pressure-retaining components. A UT acceptance standard is needed for 
metallic liners of Class CC pressure-retaining components to evaluate 
conditions that are identified during an examination that may be 
unacceptable. Therefore, the NRC proposed to continue to use the UT 
acceptance standard in IWE-3511.3 for metallic liners of Class CC 
pressure-retaining components.
    Several commenters stated that Sec.  50.55a(b)(2)(ix)(J) of the 
proposed rule is not needed because the provisions in IWE-3122.3 
provide an appropriate standard for evaluating degradation and aging of 
metallic liners of Class CC pressure-retaining components. The NRC 
disagrees. Item E4.12 of the 1998 Edition, 1999 Addenda, and 2000 
Addenda of Table IWE-2500-1, states that IWE-3511 is the acceptance 
standard for UT examinations. IWE-3122.3 is not referenced in Table 
IWE-2500-1 as an acceptance standard. The acceptance standard in IWE-
3511 addresses Class MC pressure-retaining components and does not 
address metallic liners of Class CC pressure-retaining components. 
Prior to the 1995 Addenda to Section XI, the standard in IWE-3511 
addressed Class MC pressure-retaining components and metallic liners of 
Class CC pressure-retaining components. IWE-3511 was revised in the 
1995 Addenda to address only Class MC pressure-retaining components. 
The NRC believes that the acceptance standard in the 1995 Addenda, 1996 
Addenda, 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of 
IWE-3511 is incomplete because it does not address metallic liners of 
Class CC pressure-retaining components.
    Several commenters stated that Sec.  50.55a(b)(2)(ix)(J) of the 
proposed rule is inappropriate because a concrete metallic liner can be 
allowed to significantly degrade and still accomplish its design 
function. Therefore, imposing an acceptance limit of 10 percent of the 
nominal wall thickness is extremely conservative and unwarranted.
    The NRC disagrees and believes that the UT acceptance limit of 10 
percent of the nominal wall thickness is warranted. Concrete 
containments are constructed with metallic liners as the final leak-
tight barrier against radioactive releases to the atmosphere. By the 
virtue of being anchored to the concrete, the liner carries stresses 
and strains imparted by the concrete in addition to the loads of the 
liner itself. General or pitting corrosion occurring in a large area of 
the liner creates discontinuities in the liner behavior under accident 
pressure and earthquake loads which would result in a high stress 
concentration area in the liner. The model tests on concrete 
containments (e.g., NUREG/CR-5431, ``Round-Robin Analysis of the 
Behavior of a 1:6 Scale Reinforced Concrete Containment Model 
Pressurized to Failure: Posttest Evaluation'') have shown that once a 
liner tear occurs due to high stress concentration, the containment 
losses its ability to retain radioactive releases. Thus, the liner 
integrity must be monitored and maintained during the operating life of 
the containment. The modification in the proposed rule is identical to 
what was approved for use by the ASME Code in the 1995 Edition and 
earlier editions and addenda of the ASME Code. Section 
50.55a(b)(2)(ix)(J) of the proposed rule is presented here in the final 
rule as Sec.  50.55a(b)(2)(ix)(I). Section Sec.  50.55a(b)(2)(ix)(I) is 
otherwise adopted without change.

2.2.4 Containment Penetration Piping

    Section 50.55a(b)(2)(xii)(A) in the proposed rule would prohibit 
welds in high-energy fluid system piping that are located inside a 
containment penetration assembly or encapsulated by a guard pipe from 
being exempted from the examination provisions of

[[Page 60525]]

Subsection IWC as permitted by the 1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda of IWC-1223. The revised Code provisions 
appeared to be inconsistent with NRC's guidelines on ``break exclusion 
zone'' design and examination criteria for containment penetration 
piping. Specifically, Branch Technical Position EMEB 3-1, ``Postulated 
Rupture Locations in Fluid System Piping Inside and Outside 
Containment,'' an attachment to NRC Standard Review Plan (SRP) Section 
3.6.2, ``Determination of Rupture Locations and Dynamic Effects 
Associated with Postulated Rupture of Piping'' (NUREG-0800), allows 
that breaks and cracks in high-energy fluid piping in containment 
penetration areas need not be postulated provided that certain criteria 
are met. These criteria include a commitment that where guard pipes are 
used, the enclosed portion of fluid system piping should be seamless 
construction and without circumferential welds unless specific access 
provisions are made to permit inservice volumetric examination of the 
longitudinal and circumferential welds; and a 100 percent volumetric 
inservice examination of all pipe welds is conducted during each 
inspection interval as defined in IWA-2400 of Section XI of the ASME 
BPV Code. Licensees may have made commitments to follow the provisions 
in SRP 3.6.2 as a part of their licensing design basis.
    The commenters stated that Sec.  50.55a(b)(2)(xii)(A) of the 
proposed rule is unnecessary because the regulatory requirements 
associated with high energy line breaks are independent from the scope 
of Section XI. Commenters also noted that it is inappropriate for the 
NRC to impose limitations to maintain commitments used to license 
plants.
    The NRC agrees that the regulatory guidelines associated with high 
energy line breaks are separate from the regulatory requirements 
associated with the ISI of nuclear power plant components. The intent 
of Sec.  50.55a(b)(2)(xii)(A) in the proposed rule was to ensure that 
licensee commitments regarding high energy line breaks in Branch 
Technical Positions under SRP 3.6.2 would not be eliminated from a 
misapplication of the exemption allowed in IWC-1223. The NRC concludes 
that it is the responsibility of each licensee to ensure that changes 
to later editions and addenda of the ASME Code are not misapplied to 
licensing design bases commitments, and that it is inappropriate for 
the NRC to impose modifications or limitations in Sec.  50.55a to 
ensure that commitments, not directly related to Section XI 
requirements but part of the licensing design basis, are maintained. 
Therefore, Sec.  50.55a(b)(2)(xii)(A) in the proposed rule is not 
adopted.
    Section 50.55a(b)(2)(xii)(B) in the proposed rule would require 
that piping that penetrates the containment that is connected to a 
system not in the scope of Section XI (i.e., not safety-related) be 
pressure tested in accordance with the 1996 Addenda and earlier 
editions and addenda of IWA-5110(c).
    A number of commenters stated that Sec.  50.55a(b)(2)(xii)(B) is 
unnecessary because the Type C local leak rate test (LLRT) in Appendix 
J of 10 CFR 50, ``Primary Reactor Containment Leakage Testing for 
Water-Cooled Power Reactors,'' provides an acceptable method for 
ensuring the leak-tight integrity of the containment penetration 
piping, and that the test requirements in the editions and addenda of 
IWA-5110(c) earlier than the 1997 Addenda are redundant. The commenters 
stated that test equipment used for LLRT is capable of detecting 
extremely small leakage, and that the regulations in Appendix J of 10 
CFR 50 contain acceptance criteria for leakage identified during 
testing. Commenters also noted that Appendix J does not differentiate 
between measured leakage emanating out of the piping and out of the 
containment isolation valves. However, the commenters noted that this 
determination is unnecessary because the Appendix J maximum allowable 
leakage limit accounts for all leakage regardless of where it emanates.
    The NRC agrees that Appendix J provides an acceptable method for 
testing the leak-tightness of the containment penetration piping. 
Appendix J of 10 CFR 50 requires that piping between the containment 
isolation valves be pressurized with air during seat leak testing of 
the containment isolation valves. Any leakage emanating from the piping 
and containment isolation vales is measured and evaluated in accordance 
the criteria in Appendix J. The NRC finds that the Appendix J Type C 
LLRT provides an acceptable basis for ensuring the containment 
penetration piping integrity when the only safety function of the 
containment penetration piping is to provide containment integrity. 
Therefore, Sec.  50.55a(b)(2)(xii)(B) in the proposed rule is not 
adopted.

2.2.5 Certification of Nondestructive Examination (NDE) Personnel

    Section 50.55a(b)(2)(xviii)(A) in the proposed rule would require 
that all Level I and Level II NDE personnel be recertified on a 3-year 
interval in lieu of the 5-year interval specified in the 1997 Addenda 
and 1998 Edition of IWA-2314, and the 1999 Addenda and 2000 Addenda of 
IWA-2314(a) and IWA-2314(b). Prior to 1997, Level I and II NDE 
personnel were recertified on a 3-year interval.
    A number of commenters objected to Sec.  50.55a(b)(2)(xviii)(A) in 
the proposed rule. The commenters explained that the 1996 Addenda and 
earlier editions and addenda of IWA-2314 require that Level I and Level 
II personnel be recertified by qualification examination every 3 years, 
and that Level III personnel be recertified by qualification 
examination every 5 years. The commenters stated that the 5-year 
recertification interval should also be acceptable for Level I and 
Level II personnel because the 5-year recertification interval for 
Level III personnel has been approved by the NRC since 1989. The 
commenters also disagreed with the NRC position that available data do 
not support recertification examinations at a frequency of every 5 
years. On the contrary, the commenters stated that since the 
recertification interval was increased from 3 to 5 years in 1989 for 
Level III personnel, there is no data to support that a decrease in 
proficiency of Level III personnel has occurred. The commenters claimed 
that the improved annual practice requirements for UT examiners ensure 
the proficiency of UT examiners is maintained throughout the 5-year 
period. One commenter stated that Section XI is one of the few 
standards that require recertification by examination every 3 years, 
and that other countries recertify personnel every 5 to 10 years.
    The NRC did not approve the extension of the recertification 
frequency from 3 years to 5 years in the proposed rule because the 
proficiency of examination personnel decreases over time, and available 
data do not support recertification examinations at a frequency of 
every 5 years. Although one commenter (a licensee) stated that it has a 
100 percent recertification pass rate, the public comments did not 
provide or reference any data that NRC could review that supports 
extending the recertification frequency of Level 1 and Level 2 NDE 
personnel from 3 years to 5 years. Therefore, the NRC is not approving 
the extension of the recertification interval for Level I and Level II 
NDE personnel from 3 to 5 years at this time. Section 
50.55a(b)(2)(xviii)(A) in the proposed rule is adopted without change.
    Section 50.55a(b)(2)(xviii)(B) in the proposed rule would 
supplement the alternative qualification provisions for

[[Page 60526]]

VT-2 visual examination personnel in the 1998 Edition, 1999 Addenda, 
and 2000 Addenda of IWA-2316 with the requirements that VT-2 
examination personnel pass an initial test and then be retested every 3 
years.
    Commenters indicated that the intent of IWA-2316 is to only qualify 
personnel that observe for leakage during system leakage and 
hydrostatic tests conducted in accordance with IWA-5211(a) and (b), and 
objected to Sec.  50.55a(b)(2)(xviii)(B) in the proposed rule on the 
basis that experienced plant personnel such as system engineers, 
licensed and non-licensed operators, and maintenance staff perform the 
VT-2 examinations. The commenters argue that the basic knowledge level 
of these types of personnel is adequate to inspect plant systems during 
leakage tests. The commenters also note that the NRC has granted relief 
allowing licensees to implement the VT-2 visual examination 
qualification conditions in the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWA-2316 without requiring initial tests and periodic 
retests. The commenters also noted that the existing NRC-approved 
requirements in the 1995 Edition with the 1996 Addenda of IWA-2300 
require that personnel who conduct NDE be qualified in accordance with 
CP-189. The commenters stated that VT-2 qualification requirements are 
not in the scope of CP-189 nor are they addressed in CP-189 because 
there are no unique technical requirements associated with performing 
VT-2 examinations. VT-2 examinations are conducted to detect evidence 
of leakage from pressure-retaining components during system pressure 
tests. The use of special equipment, examination techniques, and 
evaluation of test results associated with other NDE methods such as 
volumetric and surface examinations are not applicable to VT-2 
examinations. VT-2 examinations do not include the evaluation of the 
material conditions of components, such as degraded conditions like 
loose bolting or corrosion. The commenters also stated that the 
proposed Sec.  50.55a(b)(2)(xviii)(B) is unnecessary because plant 
administrative procedures require that personnel involved in testing be 
briefed prior to the test, and special requirements for conducting the 
VT-2 examinations are covered during the pretest brief.
    The NRC agrees that there are no special or unique technical 
requirements associated with performing VT-2 examinations that require 
personnel to observe for leakage of liquids or condensation during 
system leakage or hydrostatic testing. However, VT-2 visual examiners 
also conduct other evolutions that are more complex than observing for 
leakage during a leakage or hydrostatic tests. Visual examiners that 
are VT-2 qualified also perform bubble, halogen diode leak, and mass 
spectrometer testing requiring the use of special equipment and 
examination techniques. The NRC believes that VT-2 qualification 
requirements in IWA-2316 should be limited to personnel that only 
observe for leakage of liquids or condensation during system leakage or 
hydrostatic testing. Therefore, Sec.  50.55a(b)(2)(xviii)(B) is revised 
to clarify that IWA-2316 may only be used to qualify personnel that 
observe for leakage during the performance of system leakage and that 
hydrostatic tests are to be conducted in accordance with the 1998 
Edition, 1999 Addenda, and 2000 Addenda of IWA-5211(a) and (b).
    Section 50.55a(b)(2)(xviii)(C) in the proposed rule would 
supplement the alternative qualification provisions for VT-3 visual 
examination personnel in the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWA-2317 with the requirements that VT-3 examination 
personnel pass an initial test and then retested every 3 years.
    Several commenters objected to Sec.  50.55a(b)(2)(xviii)(C) in the 
proposed rule because experienced personnel are familiar with the 
performance of VT-3 examinations, and the VT-3 examination is a 
straightforward technique. The NRC does not agree because the material 
condition of many different types of components are required to be 
evaluated during the performance of VT-3 examinations, and there are 
different technical acceptance criteria specified for the many 
different components. For example, the acceptance criteria for 
examining bolting is different from the acceptance criteria for 
examining containment metal surfaces. Furthermore, there are critical 
technical requirements associated with the minimum illumination, 
distance, and character height that must be adhered to when performing 
VT-3 examinations. There are a number of options available to the VT-3 
examiner that complicate qualification requirements. For example, 
remote visual examination can be substituted for direct visual 
examination resulting in the use of special test equipment. The NRC 
concludes that testing is required to demonstrate that VT-3 examiners 
are knowledgeable regarding the different requirements associated with 
the examination method, and that these testing requirements are 
consistent with other NDE methods in CP-189 that require testing to 
demonstrate the required knowledge. Therefore, Sec.  
50.55a(b)(2)(xviii)(C) is revised to clarify that the alternative 
qualification provisions for VT-3 visual examination personnel in the 
1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-2317 may be used 
provided that VT-3 examination personnel pass an initial test and a 
retest every 3 years.

2.2.6 Substitution of Alternative Methods

    Section 50.55a(b)(2)(xix) in the proposed rule would prohibit the 
use of the provision in IWA-2240 (1998 Edition, 1999 Addenda, and 2000 
Addenda) and IWA-4520(c) (1997 Addenda, 1998 Edition, 1999 Addenda, and 
2000 Addenda), which allows alternative examination methods, a 
combination of methods, or newly developed techniques to be substituted 
for the methods specified in the Construction Code, provided the 
Authorized Nuclear Inservice Inspector (ANII) is satisfied that the 
results are demonstrated to be equivalent or superior to those in the 
Construction Code. The revision to IWA-2240 changed the applicability 
of the paragraph from Section XI only (ISI) to both Sections III and XI 
(design/construction and ISI).
    A number of commenters stated that editions and addenda of Section 
XI approved by the NRC since the 1974 Edition of Section XI allow ANIIs 
to approve the substitution of alternative methods, a combination of 
methods, or newly developed techniques for the examinations specified 
in Section XI, Division 1. For example, the ANII can approve the 
substitution of an eddy current examination for a surface examination 
requirement in IWB and IWC of Section XI provided the ANII is satisfied 
that the results of the eddy current examination are equivalent or 
superior to those of the surface examination. Most of the commenters 
stated that the NRC should accept the revised provisions in the 1998 
Addenda, 1999 Addenda, and 2000 Addenda of IWA-2240 and the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-4520(c) 
that extend the substitution of the alternative examination provisions 
to the examinations specified in the Construction Code when performing 
repair/replacement activities. The commenters stated that ANII 
qualifications require detailed knowledge of the different examination 
methods addressed in Section XI and the Construction Code. One 
commenter stated that ASME QAI-1-1995, ``Qualifications for Authorized

[[Page 60527]]

Inspection,'' is the applicable qualification standard that must be 
used to qualify ANIIs, and that ASME QAI-1-1995 requires that ANIIs be 
certified in Section XI and Construction Code requirements. An example 
of a use of the revised provisions provided by the commenters indicated 
that in some instances it may be a hardship or impractical to perform a 
radiographic (RT) examination during a Section XI repair/replacement 
activity as specified in the Construction Code. The revised provisions 
in Section XI would allow the substitution of an alternative method 
such as an UT examination for the RT examination provided that the ANII 
is satisfied that the results of the UT examination are equivalent or 
superior to the RT examination specified in the Construction Code.
    The NRC agrees that the provisions in IWA-2240 that allow the ANII 
to approve the substitution of alternative examination methods, a 
combination of methods, or newly developed techniques for the methods 
specified in Section XI, Division 1, have been approved by the NRC 
since 1974. The NRC has reviewed the qualification standard in ASME 
QAI-1-1995, and agrees that ANIIs are required to be knowledgeable 
regarding the NDE methods, qualification requirements, and other 
requirements in Section XI and the Construction Code. However, the NRC 
believes that the substitution of alternative methods for those 
specified in the Construction Code is significantly more complex than 
what was previously approved by the NRC in editions and addenda of IWA-
2240 earlier than the 1998 Edition. For example, there are many factors 
that have to be evaluated when substituting a UT examination for an RT 
examination required by the Construction Code. Consideration needs to 
be given to the thickness of the weld, volume of the UT examination, 
appropriate UT technique, UT examination coverage criteria, UT 
examination procedure (Section V or Section XI), and performance 
demonstration methodology; calibration block material, thickness, and 
size; flaw evaluation acceptance criteria, and demonstration and 
qualification criteria for single-sided UT examinations. Weld material 
would also be a critical factor when considering the substitution of a 
UT examination for an RT examination. It may not be appropriate to 
allow the substitution of a UT examination for an RT examination for 
certain materials such as ferritic or austenitic cast products and 
corrosion resistant cladding with butt welds. Substitution of a UT 
examination for an RT examination may be acceptable for dissimilar 
metal welds but would require additional factors to be evaluated. The 
NRC finds that there is a lack of guidance in the Code to ensure proper 
consideration of factors when substituting alternative examinations for 
the examinations specified in the Construction Code. The NRC believes 
that a standardized repeatable methodology that can be consistently 
used among all licensees is needed not only to demonstrate that the 
alternative method is equivalent or superior to that specified in the 
Construction Code, but also to ensure consistent application and 
implementation of IWA-2240 and IWA-4520(c). Furthermore, the NRC notes 
that the ASME is currently developing a Code Case that will provide the 
necessary guidance to allow the substitution of a UT examination with 
an RT examination when an RT examination is required by the 
Construction Code. Therefore, Sec.  50.55a(b)(2)(xix) in the proposed 
rule is adopted without change.

2.2.7 System Leakage Tests

    Section 50.55a(b)(2)(xx) in the proposed rule would have required 
that the pressure and temperature hold time requirements in the 1995 
Edition of IWA-5213(a) be applied in place of the revised provisions of 
the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-
5213(a) when performing system leakage tests.
    Many commenters objected to this modification because pressure and 
temperature hold time requirements in the 1995 Edition of IWA-5213(a) 
imposed by the modification place what the commenters believe to be an 
undue burden on utilities. One commenter noted that the ASME is 
currently developing a new revision to IWA-5213 to clarify the pressure 
and temperature hold time requirements in IWA-5213(a). A number of 
commenters stated that the NRC is arbitrarily choosing the pressure and 
temperature hold times in the 1995 Edition, and that the NRC should 
justify the use of the pressure and temperature hold times in the 1995 
Edition.
    The NRC normally requires that the Code revision most recently 
approved by the NRC be used when it does not approve the use of a later 
Code provision. Since the NRC has not approved the elimination of the 
pressure and temperature hold times in 1995 Addenda of IWA-5213, the 
NRC proposed to require the use of the pressure and temperature 
provisions in the 1995 Edition. The NRC agrees with the commenters that 
the changes in the 1989 Addenda through the 1995 Edition in conjunction 
with the proposed modification would create unintended test conditions. 
For example, some systems are not designed to operate at test 
conditions for the period of time necessary to meet the hold time 
conditions. Also, hold times are not necessary for leakage tests of 
Class 1 components because these leakage tests are normally performed 
following each refueling outage as the reactor is heating up. The 
heatup process of the reactor is performed within the pressure-
temperature constraints of the heatup curve in the plant technical 
specifications. These constraints limit the rate of temperature and 
pressure increase resulting in a heatup period of several hours. In 
light of the substantial length of time required for the reactor heatup 
process, sufficient time is available for leakage from the Class 1 
system to collect in sufficient quantity to be detectable by visual 
examination. Holding the Class 1 components for additional time at this 
temperature and pressure is unnecessary to accomplish the purpose of 
the pressure test.
    In consideration of the public comments, the NRC has revised the 
pressure and temperature hold time requirements in Sec.  
50.55a(b)(2)(xx) to be consistent with the revisions recommended in 
several of the public comments (to use the provisions contained in the 
1989 Edition of the ASME Code). This is also consistent with the 
current ASME proposed revision to the pressure and temperature hold 
times in IWA-5213. Section 50.55a(b)(2)(xx) requires a 10-minute 
holding time after attaining test pressure for Class 2 and Class 3 
components that do not normally operate during operation, and no 
holding time is required for the remaining Class 2 and Class 3 
components provided that system has been in operation for at least 4 
hours for insulated components or 10 minutes for uninsulated 
components.

2.2.8 Table IWB-2500-1 Examination Requirements

    Section 50.55a(b)(2)(xxi)(A) in the proposed rule would require 
licensees to use the provisions in the 1998 Edition of Table IWB-2500-
1, Examination Category B-D, for Items B3.40 and B3.60 (Inspection 
Program A) and Items B3.120 and B3.140 (Inspection Program B) when 
using the 1999 Addenda and the 2000 Addenda. The 1999 Addenda 
eliminated the pressurizer and steam generator (SG) nozzle inside-
radius inspections in Table IWB-2500-1, Examination Category B-D, Items 
B3.40 and B3.60 (Inspection Program A) and Items B3.120 and B3.140 
(Inspection Program B).

[[Page 60528]]

    Several commenters summarized the results of a white paper 
developed by the ASME that provides the technical basis for eliminating 
the pressurizer and SG nozzle inside-radius UT examinations from Table 
IWB-2500-1. The commenters explained the difficulties associated with 
performing UT examinations of pressurizer and SG nozzle inside radii. 
Radiation exposure to personnel who conduct the UT examinations is a 
significant concern because the pressurizer and SG nozzles are located 
in very high radiation areas. The geometry and material of the nozzles 
significantly complicate the UT examination procedure making it 
difficult to obtain meaningful UT data. The commenters stated that the 
basis for eliminating the pressurizer and SG nozzle inner radius 
examinations is that a review of UT and visual examination data from 
pressurizer and SG nozzle inner radius examinations reveal that no 
service-induced flaws were detected in any of the examinations 
performed. Commenters claimed that pressurizer and SG nozzle cracking 
incidents have not occurred at any nuclear facilities, and that 
structural integrity evaluations of the nozzles indicate that leakage 
would occur from a through-wall flaw before any integrity problems 
would occur (ie., the nozzle would leak before it failed). In addition, 
a risk assessment indicated that the failure probability of the nozzles 
is extremely low under plant operating conditions, and shows that there 
is no change in risk if pressurizer and SG nozzle inner radius 
examinations are eliminated. Finally, the commenters stated that the 
NRC has granted relief from the pressurizer and SG inside-radius UT 
examination requirements in Table IWB-2500-1 to many licensees because 
of these concerns associated with occupational exposure and difficulty 
in obtaining meaningful UT data.
    The NRC disagrees. Operating history alone does not provide 
adequate justification to eliminate examinations of the pressurizer and 
SG nozzle inside radii because operational experience also has 
demonstrated that components degrade as they age. Although pressurizer 
and SG nozzle cracking incidents have not occurred, cracks have been 
identified in other nozzles such as the feedwater nozzles. Furthermore, 
a leak-before-break evaluation is not adequate justification to 
eliminate the examination of the pressurizer and SG nozzle inside radii 
because the primary purpose of the ISI requirements in Section XI is to 
identify and correct component degradation before it becomes 
significant. Leakage from any pressurizer or SG nozzle would be 
significant because such leakage would represent an unisolable breach 
of the reactor coolant pressure boundary.
    The NRC agrees that a number of licensees have requested relief 
from the UT examination requirements for SG nozzle inner radii and 
pressurizer nozzle inner radii. In these cases, the NRC has authorized, 
as an alternative to UT examination, the performance of a visual 
examination which utilizes equipment with enhanced magnification that 
has a resolution sensitivity to detect a 1-mil width wire or crack. The 
flaw length acceptance criteria specified for the UT examination in 
Table IWB-3512-1 is applicable to the visual examination. The primary 
degradation mode for these nozzles is fatigue which produces hairline 
surface indications that network along the circumference of the nozzle 
at the inner radius section. Ultrasonic examination of the inner radii 
from the outside surface should detect these indications. However, even 
with the use of improved technology from the outside surface, the 
complex geometry of these nozzle inner radius sections prevents 
complete coverage. Visual examination for some of these nozzles from 
the inside surface is easier and less costly to accomplish, and 
coverage is more complete. The examinations can be performed when the 
pressurizer and SG are opened for other maintenance or inspection 
activities. Use of video equipment with enhanced magnification that has 
a resolution sensitivity to detect a 1-mil width wire or crack is 
similar to UT examination regarding the capability of detecting 
fatigue-type cracks on nozzle inside radii before they become 
detrimental to structural integrity.
    Therefore, Sec.  50.55a(b)(2)(xxi)(A) is revised to allow the 
option of performing 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 of Table IWB-3512-1 
in place of a UT examination. Section 50.55a(b)(2)(xxi)(A) requires 
that licensees use the provisions of Table IWB-2500-1, Examination 
Category B-D, Items B3.40 and B3.60 (Inspection Program A) and Items 
B3.120 and B3.140 (Inspection Program B) of the 1998 Edition when using 
the 1999 Addenda and the 2000 Addenda. 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 the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of 
Table IWB-3512-1 may be performed in place of a UT examination.
    Section 50.55a(b)(2)(xxi)(B) in the proposed rule would require 
that licensees apply the provisions in the 1995 Edition of Table IWB-
2500-1, Examination Category B-G-2, Item B7.80 when using the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda. The 1995 Edition 
and earlier editions and addenda of Section XI require a visual 
examination of control rod drive (CRD) housing bolting using the VT-1 
visual examination method whenever the CRD housing is disassembled. The 
requirement to examine CRD bolting whenever the CRD housing is 
disassembled was deleted in the 1995 Addenda.
    Several commenters stated that Sec.  50.55a(b)(2)(xxi)(B) should be 
deleted because the skill of the craft and maintenance practices are 
sufficient to ensure that bolting is not damaged during maintenance 
activities. The NRC agrees that the scope of Section XI does not 
normally include examinations that are conducted during routine 
maintenance activities, but notes there may be maintenance-related 
activities associated with ISI. The ISI of components to verify that 
service-related degradation is not occurring is within the scope of 
Section XI.
    The majority of the commenters stated that no degradation of CRD 
bolting has occurred in 30 years of experience, and hence the 
requirement to examine the CRD bolting should be eliminated. The NRC 
disagrees. Operating history alone does not provide adequate 
justification to eliminate examinations of CRD bolting because 
operational experience also has demonstrated that components degrade as 
they age. Furthermore, the NRC is aware of an example where CRD bolting 
was replaced in two units because examination of CRD bolting identified 
cracks.
    Several commenters stated that the NRC is misinterpreting the ASME 
Code because Item B7.80 of Table IWB-B7.80 does not require that the 
CRD housing be disassembled to perform the examination of CRD bolting. 
The NRC notes that although the Code does not require disassembly of 
the CRD housing to examine the bolting, Item B7.80 of Table IWB-2500-1 
in the 1995 Edition and earlier editions and addenda of Section XI 
states that the extent and frequency of the examination is to include 
bolts, studs, and nuts in CRD housings when disassembled. The NRC finds 
that the 1995 Edition and earlier editions and addenda of Section XI 
only require that CRD bolting be examined when the CRD housing is 
disassembled

[[Page 60529]]

such as during a repair or maintenance activity.
    Several other commenters stated that since CRD mechanisms are 
usually contaminated and in high radiation areas, elimination of the 
bolting examinations reduces radiation exposure to personnel. The NRC 
notes that CRD bolting is normally relocated to a storage area after 
disassembly of the CRD housing. Therefore, VT-1 examination personnel 
typically examine the bolting when it is removed and remotely located 
from the CRD mechanism, reducing the exposure to individuals.
    One commenter requested that the NRC revise Sec.  
50.55a(b)(2)(xxi)(B) to include a statement that only CRD bolting that 
is reused is required to be examined. It was the NRC's intent to 
require examination of the CRD bolting material only if it was to be 
reused. Therefore, Sec.  50.55a(b)(2)(xxi)(B) is revised to clarify 
that only CRD bolting that is reused must be re-examined.
    Section 50.55a(b)(2)(xxi)(C) in the proposed rule would require 
that licensees use the provisions in the 1995 Addenda of Table IWB-
2500-1, Examination Category B-K, Item B10.10, when using the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda for the 
examination of welded attachments to pressure vessels. The 1997 Addenda 
permits performance of a single-side surface examination in place of a 
surface examination from both sides of the weld, whereas the 1995 
Addenda requires the performance of a single-side volumetric 
examination of the weld in place of surface examination of the 
inaccessible surface if surface examination from both sides of the weld 
is not performed.
    Several commenters noted that volumetric examination of reactor 
pressure vessel (RPV) skirt welds is not practical because UT 
calibration blocks were typically not supplied for RPV skirt welds and 
the UT performance demonstration requirements of Appendix VIII do not 
address RPV support attachment welds. If a licensee wanted to perform a 
volumetric examination in place of surface examination of both 
surfaces, it would have to fabricate its own calibration blocks and 
sample specimens, develop its own procedures, and set up its own 
demonstration program.
    The NRC recognizes that UT examination of RPV skirt welds is not 
addressed in Appendix VIII at this time. However, the applicable 
examination requirements are addressed in Article I-2000 of Section XI 
which in turn references Section V of the ASME BPV Code. Furthermore, 
Section V of the ASME BPV Code addresses the qualification and use of 
suitable alternative calibration blocks.
    Commenters stated that access under the RPV bottom head for 
performing a visual examination is a confined space that is also a high 
radiation area. The inside surface geometry is such that preparation 
for a surface examination is difficult, thus extending the time spent 
in the high radiation area. The commenters conclude that the radiation 
exposure to personnel who examine the inside surface of the RPV skirt 
weld is not justified. The NRC agrees that access to such confined 
spaces is very difficult. However, the NRC also believes that the 1995 
Addenda of the ASME Code, which already provides for an alternative UT 
examination in place of a surface examination of the inaccessible 
surface, appropriately accommodates the commenters concerns. These UT 
examinations are performed on the accessible surface of the RPV skirt 
welds. Therefore, personnel are not required to enter the confined 
space area under the RPV bottom head.
    Commenters also stated that RPV skirt weld materials are very flaw-
tolerant, with slow flaw-propagation rates. Flaws originating on the 
inside surface would grow through-wall long before their length would 
threaten the structural integrity/function of the weld. The NRC notes 
that the assumption that flaws will be detected before affecting 
structural integrity is an assumption based on limited surface 
examination experience and is not supported by rigorous study. The 
commenters have not presented any analyses or studies which support 
such an assumption.
    Commenters stated that RPV skirt welds are similar to non-pressure 
boundary core shroud circumferential welds in boiling water reactors. 
The commenters also stated that safety analyses performed by the 
Boiling Water Reactor Vessel & Internals Program found that core shroud 
circumferential welds could be cracked through-wall for 360[deg] and 
still perform their function. The NRC considers the inference that the 
structural performance, response, and safety implications of operating 
with a significantly cracked RPV skirt weld is no different than 
operating with significantly cracked core shroud circumferential welds 
to be inappropriate. Operation with cracked core shroud welds has been 
extensively evaluated for all operating and accident loading 
conditions. The core shroud is contained within the confines of the 
reactor pressure vessel with positive restraints holding it in place to 
assure integrity and adequate coolant flow through the core. However, 
operation with a significantly cracked RPV skirt weld has not been 
evaluated. Therefore, the NRC has no basis to conclude that operation 
under such conditions is acceptable. Commenters also claim that the 
excellent service history of RPV skirt welds demonstrates that inside 
surface examinations of welds is not warranted. The NRC considers that 
operating history alone does not provide adequate justification to 
eliminate examinations of components because operational experience has 
also demonstrated that components degrade as they age. Therefore, Sec.  
50.55a(b)(2)(xxi)(C) in the proposed rule is adopted without change.

2.2.9 Supplemental Annual Training Requirements for Ultrasonic 
Examiners

    Section 50.55a(b)(2)(xxii) in the proposed rule would require 
licensees to apply the UT examiner supplemental annual training 
provisions in the 1998 Edition of Paragraph VII-4240 of Appendix VII, 
in place of the revised provisions in the 1999 Addenda and 2000 Addenda 
of VII-4240.
    Several commenters stated that the NRC position on training 
requirements for UT examiners in Sec.  50.55a(b)(2)(xxii) of the 
proposed rule is inconsistent with the NRC position on training 
requirements for UT examiners in final rule 64 FR 51370 (September 22, 
1999). The commenters noted that the final rule imposed Sec.  
50.55a(b)(2)(xiv) because the 10-hour classroom training requirement in 
VII-4240 was inadequate. The commenters stated that Code Case N-583, 
``Annual Training Alternative,'' was developed by the ASME to 
specifically address the NRC concern with the 10-hour classroom 
training requirement in the 1995 Edition and 1996 Addenda of VII-4240. 
Code Case N-583 was incorporated into the 1999 Addenda of VII-4240, 
replacing the 10-hour classroom training requirement with an 8-hour 
training requirement to analyze data from material or welds containing 
flaws similar to those that may be encountered during UT examinations. 
The commenters stated that the revised training requirements in the 
1999 Addenda of VII-4240 are an improvement over the training 
requirements in the 1998 Edition and earlier editions and addenda of 
VII-4240. The revised training requirements provide specific criteria 
that result in uniform training programs among all licensees.
    The commenters have clarified to the NRC that the training 
requirements in the 1999 Addenda and 2000 Addenda of VII-4240 specify 
hands on training in

[[Page 60530]]

place of classroom training. Therefore, Sec.  50.55a(b)(2)(xxii) in the 
proposed rule is not adopted because after further clarification, the 
NRC finds that the training requirements in 1999 Addenda and 2000 
Addenda of VII-4240 are consistent with the NRC position on training 
requirements for UT examiners in final rule 64 FR 51370 (September 22, 
1999).
    Commenters requested that licensees be allowed to substitute the 
supplemental practice in the 1999 Addenda and 2000 Addenda of VII-4240 
for the existing hands-on training requirement in Sec.  
50.55a(b)(2)(xiv). The NRC finds that the supplemental practice as 
described in VII-4240 of Supplement VII of Section XI, 1999 Addenda and 
2000 Addenda, is an acceptable alternative to the existing hands-on 
training requirement in Sec.  50.55a(b)(2)(xiv) provided that the 
supplemental practice is performed on material or welds that contain 
cracks, or by analyzing prerecorded data from material or welds that 
contain cracks. Therefore, Sec.  50.55a(b)(2)(xiv) is revised to allow 
the option of performing the supplemental practice as described in VII-
4240 of Supplement VII of Section XI, 1999 Addenda and 2000 Addenda, or 
the existing hands-on training requirement.

2.2.10 Underwater Welding

    Section 50.55a(b)(2)(xxiii) in the proposed rule would require 
licensees to demonstrate the acceptability of the underwater welding 
method through the use of a mockup using material with similar neutron 
fluence levels, when welding irradiated material underwater in 
accordance with the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWA-4660.
    Several commenters stated that the use of a mockup to demonstrate 
the acceptability of an underwater welding method is impractical due to 
unavailability of materials with similar neutron fluence levels, 
personnel exposure, high-cost of mockups, and handling and disposal 
requirements. The commenters also stated that the industry is currently 
developing an acceptable underwater welding technique for irradiated 
materials in conjunction with the Boiling Water Reactor Vessel & 
Internals Project that will be submitted to the NRC for approval.
    The NRC proposed the use of a mockup because underwater weld 
repairs using conventional welding techniques on in-vessel components 
exposed to high neutron fluences may be unsuccessful due to helium-
induced cracking and radiation damage, unless special welding 
techniques are used. The NRC has revised the proposed underwater 
welding mockup requirement because of the impracticality of developing 
and using a mockup with similar neutron fluence levels. Section 
50.55a(b)(2)(xxiii) is revised to prohibit the use of the 1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-4660 to weld 
irradiated material underwater. Licensees must obtain NRC approval in 
accordance with Sec.  50.55a(a)(3) of the technique used to weld 
irradiated material underwater. Section 50.55a(b)(2)(xxiii) of the 
proposed rule is presented here in the final rule as Sec.  
50.55a(b)(2)(xii).

2.3 Appendix VIII to Section XI

    This final rule extends the applicability of the existing 
regulations in Sec.  50.55a(b)(2)(xv) to the 1997 Addenda, the 1998 
Edition, 1999 Addenda, and 2000 Addenda of Appendix VIII of Section XI 
of the ASME BPV Code.

2.3.1 Examination Coverage for Dissimilar Metal Pipe Welds

    The existing requirements in Sec.  50.55a(g)(6)(ii)(C)(1) state 
that Supplement 10, ``Qualification Requirements for Dissimilar Metal 
Piping Welds,'' of Appendix VIII to Section XI must be implemented by 
November 22, 2002. Therefore, the proposed rule would have updated 
Sec.  50.55a(b)(2)(xv)(A) to reference Supplement 10. Specifically, the 
proposed rule would revise Sec. Sec.  50.55a(b)(2)(xv)(A)(1) and (A)(2) 
to provide UT examination coverage criteria for dissimilar metal piping 
welds. Examination coverage criteria for dissimilar metal piping welds 
are specified in the 1989 Edition and earlier editions and addenda of 
Appendix III of Section XI. Appendix VIII was added in the 1989 Addenda 
of Section XI, and Section XI would require that the UT examination 
criteria for piping welds in Appendix VIII supercede the examination 
criteria in Appendix III. Although Appendix VIII addresses 
qualification of personnel, procedures, and equipment used to conduct 
UT examinations of dissimilar metal piping welds, Appendix VIII (unlike 
Appendix III) does not address UT examination coverage criteria for 
dissimilar metal piping welds.
    The commenters agreed that Sec. Sec.  50.55a(b)(2)(xv)(A), (A)(1) 
and (A)(2) should be revised to provide UT examination coverage 
criteria for dissimilar metal piping welds. However, the commenters did 
not agree with the examination coverage criteria in Sec.  
50.55a(b)(2)(xv)(A)(2) of the proposed rule requiring that dissimilar 
metal welds be examined from the austenitic side of the weld when 
examination from both sides is not possible. The commenters stated that 
Sec.  50.55a(b)(2)(xv)(A)(2) should be revised to allow coverage from 
either the austenitic or ferritic side of the weld when UT examination 
from both sides is not possible because the composition of the base 
material is of minor consequence when compared to the effects of the 
austenitic weld material. Furthermore, the commenters argued that the 
examination should be conducted from the side of the weld that is most 
accessible.
    The NRC does not agree that the composition of the base material is 
of minor consequence when compared to the effects of austenitic weld 
material. There is a higher probability and reliability of identifying 
flaws in dissimilar metal welds when using a UT procedure qualified to 
perform examinations from the austenitic side than when using a UT 
procedure qualified to perform examinations from the ferritic side. 
Therefore, coverage from the austenitic side of the weld is preferred 
when UT examination from both sides is not possible.
    Sections 50.55a(b)(2)(xv)(A) and (A)(1) in the proposed rule are 
adopted without change. Section 50.55a(b)(2)(xv)(A)(2) is revised to 
clarify that dissimilar metal weld qualifications must be demonstrated 
from the austenitic side of the weld, and that the examination from the 
austenitic side of the weld may be used to perform examinations from 
either side of the weld.

2.3.2 Reactor Vessel Single Side Examinations

    The proposed rule would remove the existing Sec.  
50.55a(b)(2)(xv)(G)(4) because the examination criteria are redundant 
with the examination criteria contained in Sec.  50.55a(b)(2)(xv)(G)(3) 
and, therefore unnecessary. Both Sec. Sec.  50.55a(b)(2)(xv)(G)(3) and 
(4) allow credit for the full volume when the examination volume is 
covered from a perpendicular and parallel direction. There were no 
public comments on the proposed revision; therefore, Sec.  
50.55a(b)(2)(xv)(G)(4) is removed.

2.3.3 Qualification Test Samples

    The revision to Sec.  50.55a(b)(2)(xv)(K)(1)(i) in the proposed 
rule would resolve a discrepancy between the existing Sec. Sec.  
50.55a(b)(2)(xv)(K)(1)(i) and 50.55a(b)(2)(xv)(K)(4). Currently, Sec.  
50.55a(b)(2)(xv)(K)(1)(i) states that

[[Page 60531]]

flaws which are perpendicular to the weld are not required to be 
included in the qualification test sample. This requirement conflicts 
with a provision in Sec.  50.55a(b)(2)(xv)(K)(4), which states that 
test samples must contain flaws that are perpendicular to the weld in 
the inner 15 percent of the weld, but that these same flaws are not 
required to be located in the outer 85 percent of the weld. There were 
no public comments on the proposed revision; therefore, the revision to 
Sec.  50.55a(b)(2)(xv)(K)(1)(i) is adopted without change.

2.3.4 Implementation of Appendix VIII to Section XI

    Section 50.55a(b)(2)(xv)(M) in the proposed rule would clarify that 
only those provisions in Supplement 12 to Appendix VIII that relate to 
the coordinated implementation of Supplement 3 to Supplement 2 
performance demonstrations must be implemented. Supplement 12 provides 
coordinated implementation provisions for the performance 
demonstrations in Supplements 2, 3, 10, and 11 of Appendix VIII; 
however, with the exception of the coordinated implementation of 
Supplement 3 to Supplement 2 performance demonstration, the other 
coordinated implementation provisions in Supplement 12 are incomplete. 
Supplement 12 does not provide provisions for implementing single-side 
examinations as part of the coordinating process, or provide provisions 
for the coordinated implementation of Supplement 2 or Supplement 11 
performance demonstrations to Supplements 3 and 10. There were no 
public comments on the proposed Sec.  50.55a(b)(2)(xv)(M); therefore, 
Sec.  50.55a(b)(2)(xv)(M) is adopted without change.
    Section 50.55a(g)(6)(ii)(C)(1) in the proposed rule would clarify 
that Appendix VIII to Section XI, 1995 Edition with the 1996 Addenda, 
as well as its supplements, are mandatory and must be implemented. 
Although the final rule that implemented Appendix VIII (64 FR 51370; 
September 22, 1999) requires a phased implementation of Appendix VIII 
over a 3-year period, the final rule addressed the implementation of 
the Appendix VIII supplements only and failed to mention the 
implementation of Appendix VIII itself. The failure to address the 
implementation of Appendix VIII was an oversight. Section 
50.55a(g)(6)(ii)(C)(1) in the proposed rule would also eliminate 
Supplements 12 and 13 of Appendix VIII from the implementation schedule 
that is currently in Sec.  50.55a(g)(6)(ii)(C)(1). Supplements 12 and 
13 coordinate the implementation of selected aspects of Supplements 2, 
3, 4, 5, 6, 7, 10, and 11 of Appendix VIII. Since the implementation 
schedule for Supplements 2, 3, 4, 5, 6, 7, 10, and 11 of Appendix VIII 
is addressed in Sec.  50.55a(g)(6)(ii)(C)(1), the imposition of a 
mandatory implementation date for Supplements 12 and 13 is redundant. 
There were no public comments on either of the proposed revisions; 
therefore, the revisions to Sec.  50.55a(g)(6)(ii)(C)(1) are adopted 
without change.
    Section 50.55a(g)(6)(ii)(C)(2) in the proposed rule would clarify 
that the requirements of Appendix VIII and the supplements to Appendix 
VIII to Section XI, of the 1995 Edition with the 1996 Addenda are 
mandatory when implementing the 1989 Edition and earlier editions and 
addenda of IWA-2232 of Section XI. Paragraph IWA-2232 provides rules 
for conducting UT examinations. Appendix VIII was introduced into 
Section XI in the 1989 Addenda. Before that time, Appendix VIII did not 
exist in Section XI. Therefore, the 1989 Edition and earlier editions 
and addenda of IWA-2232 do not reference Appendix VIII. It is not clear 
to some licensees that they are required to perform UT examinations 
using personnel, procedures, and equipment qualified in accordance with 
Appendix VIII. The NRC believes that the final rule dated September 22, 
1999 (64 FR 51370), by imposing an expedited implementation of the 
supplements to Appendix VIII to Section XI, 1995 Edition with the 1996 
Addenda, makes it clear that all licensees are required to implement 
the provisions of Appendix VIII, including those licensees implementing 
the 1989 Edition or earlier editions and addenda of IWA-2232.
    A commenter pointed out that Sec.  50.55a(g)(6)(ii)(C)(2) in the 
proposed rule is inconsistent with the statement of considerations for 
the proposed rule. The NRC agrees. The purpose of Sec.  
50.55a(g)(6)(ii)(C)(2) in the proposed rule was to clarify the 
relationship between the 1989 Edition and earlier editions and addenda 
of IWA-2232 of Section XI, and Appendix VIII of Section XI. However, in 
making this clarification, the NRC inadvertently worded Sec.  
50.55a(g)(6)(ii)(C)(2) such that licensees would be required to update 
their Appendix VIII program to the latest edition and addenda of 
Section XI incorporated by reference in Sec.  50.55a(b)(2) following 
every update. It was not the intent of the NRC to revise the existing 
120-month inspection interval update requirement. Therefore, Sec.  
50.55a(g)(6)(ii)(C)(2) is revised to clarify that licensees 
implementing the 1989 Edition and earlier editions and addenda of IWA-
2232 of Section XI must implement the 1995 Edition with the 1996 
Addenda of Appendix VIII of Section XI.

2.4 ASME OM Code

    The final rule revises Sec.  50.55a(b)(3) to incorporate by 
reference the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 
Addenda of the ASME OM Code, and extends the applicability of the 
existing regulations in Sec. Sec.  50.55a(b)(3)(ii), 50.55a(b)(3)(iii), 
50.55a(b)(3)(iv), and 50.55a(b)(3)(v) to the 1997 Addenda, 1998 
Edition, 1999 Addenda, and the 2000 Addenda of the ASME OM Code. 
Subsections of the ASME OM Code were renumbered in the 1998 Edition; 
therefore, Sec. Sec.  50.55a(b)(3)(ii), 50.55a(b)(3)(iii), and 
50.55a(b)(3)(iv) are revised and Sec.  50.55a(b)(3)(iv)(D) is added to 
account for the renumbering.
    Although the technical requirements in Sec.  50.55a(b)(3)(ii) were 
not revised in the proposed rule, several commenters stated that the 
reference to motor-operated valve (MOV) stroke-time testing in the 
existing Sec.  50.55a(b)(3)(ii) is confusing because there are other 
MOV test requirements in the ASME OM Code (such as position indication 
and seat leakage testing) that are applicable in addition to stroke-
time testing. The commenters suggested that a licensee might 
incorrectly interpret Sec.  50.55a(b)(3)(ii) as requiring that only MOV 
stroke-time testing be performed in accordance with the OM Code. The 
NRC believes the current regulation clearly states that licensees must 
meet all of the ASME Code provisions for testing MOVs. The NRC is not 
aware of any misunderstanding among licensees regarding the intent of 
the regulatory requirement for MOVs. However, to avoid any potential 
confusion in the future, Sec.  50.55a(b)(3)(ii) is revised to clarify 
that licensees must comply with the provisions of the ASME OM ISTC Code 
for testing MOVs.
    Section 50.55a(b)(3)(vi) in the proposed rule would require an 
exercise interval of 2 years for manual valves within the scope of the 
ASME OM Code rather than the exercise interval of 5 years specified in 
the 1999 Addenda and the 2000 Addenda of the ASME OM Code. The 1998 
Edition of the ASME OM Code specified an exercise interval of 3 months 
for manual valves within the scope of the Code. The 1999 Addenda to the 
ASME OM Code revised ISTC-3540 to extend the exercise frequency for 
manual valves to 5 years.

[[Page 60532]]

    A number of commenters stated that Sec.  50.55a(b)(3)(vi) in the 
proposed rule should be withdrawn because sufficient justification 
exists to allow the extension of the exercise interval for manual 
valves to 5 years. The justification for the 5-year frequency is the 
simplicity of manual valves (limited number of failure causes) and that 
the ASME OM Code allows other valves (safety and relief valves) to be 
tested on a 5-year or longer frequencies.
    The NRC does not agree that there is sufficient justification to 
extend the exercise interval for manual valves to 5 years. The NRC 
review of licensee IST programs indicate that manual valves are 
exercised every 3 months except in instances where it is impractical to 
operate valves during unit operation. Valves are then exercised when 
the unit is in a cold shutdown condition, and the exercise frequency 
cannot exceed 2 years. Therefore, a 2-year interval for exercising 
manual valves is justified because the available manual valve exercise 
data supports the 2-year interval. The NRC has approved longer test 
intervals for other types of valves in the ASME OM Code but the longer 
test intervals include additional means to determine component 
degradation. For example, although the ASME OM Code test strategy for 
Class 2 and 3 relief valves has a testing interval of 10 years, Class 2 
and 3 relief valves are subject to grouping and sample expansion if 
there is a test failure. Manual valves that are required to be 
exercised are not subject to grouping and sample expansion. 
Furthermore, obstruction from silting or blockage, or corrosion of 
valve internals are possible failure modes for safety-related manual 
valves that are not applicable to other types of valves with longer 
test intervals. Exercising manual valves minimizes both of these 
failure modes and also allows for more immediate detection if an 
obstruction or corrosion induced failure occurs. Section 
50.55a(b)(3)(vi) is revised to clarify that the interval for exercising 
manual valves may not exceed 2 years when using the 1999 Addenda and 
2000 Addenda of ISTC-3540. Licensees are not prohibited from exercising 
manual valves more frequently than every 2 years.

3. Section-by-Section Analysis of Substantive Changes

    Paragraph (b)(1). This paragraph incorporates by reference the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of Section III, 
Division 1, of the ASME BPV Code. New applicants for a nuclear power 
plant submitting an application for a construction permit under 10 CFR 
50 or design certification under 10 CFR 52 are required to use the 1998 
Edition up to and including the 2000 Addenda for the design and 
construction of the reactor coolant pressure boundary and Quality Group 
B and C components.
    Paragraph (b)(1)(ii). This paragraph extends the applicability of 
the existing regulation on weld leg dimension requirements to the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of Section III of 
the ASME BPV Code. Applicants and licensees using these Edition and 
Addenda are not allowed to apply paragraph NB-3683.4(c)(1), Footnote 11 
to Figure NC-3673.2(b)-1, and Figure ND-3673.2(b)-1.
    Paragraph (b)(1)(iii). This paragraph extends the applicability of 
the existing regulation on seismic design requirements to the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of Section III of 
the ASME BPV Code. Applicants and licensees using these edition and 
addenda are not allowed to use Articles NB-3200, NB-3600, NC-3600, and 
ND-3600.
    Paragraph (b)(1)(v). This paragraph extends the applicability of 
the existing regulation on independence of inspection requirements to 
the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of 
Section III of the ASME BPV Code. Applicants and licensees using these 
edition and addenda are not allowed to apply Sub-subparagraph NCA-
4134.10(a).
    Paragraph (b)(2). This paragraph incorporates by reference the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of Section XI, 
Division 1, of the ASME BPV Code. Licensees of nuclear power plants are 
required to use the 1998 Edition up to and including the 2000 Addenda 
when updating their ISI programs in their subsequent 120-month interval 
under Sec.  50.55a(g)(4)(ii).
    Paragraph (b)(2)(vi). This paragraph clarifies that either the 1992 
Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda 
of Subsection IWE and Subsection IWL as modified and supplemented by 
the requirements in Sec.  50.55a(b)(2)(viii) and Sec.  50.55a(b)(2)(ix) 
must be used when implementing the initial 120-month inspection 
interval for the containment inservice inspection requirements. 
Successive 120-month interval updates must be implemented in accordance 
with Sec.  50.55a(g)(4)(ii).
    Paragraph (b)(2)(viii). This paragraph extends the applicability of 
the existing regulation in paragraph (b)(2)(viii)(E) on concrete 
containment examination requirements to the 1998 Edition, 1999 Addenda, 
and 2000 Addenda of IWL, and clarifies that the new modification in 
paragraph (b)(2)(viii)(F) applies only to the 1998 Edition, 1999 
Addenda, and 2000 Addenda of IWL.
    Paragraph (b)(2)(viii)(F). This paragraph requires that personnel 
who perform visual inspections of containment surfaces and tendon 
anchorage hardware, wires, or strands be qualified in accordance with 
IWA-2300 in place of the ``owner-defined'' personnel qualification 
provision in the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWL-
2310(d).
    Paragraph (b)(2)(ix). This paragraph clarifies that the existing 
modifications in paragraphs (b)(2)(ix)(A) through (E) of this section 
on examination of metal containments and liners of Class CC components 
apply to the 1992 Edition with the 1992 Addenda or the 1995 Edition 
with the 1996 Addenda of IWE. It also extends the applicability of the 
regulations in paragraphs (b)(2)(ix)(A) and (b)(2)(ix)(B) to the 1998 
Edition, 1999 Addenda, and 2000 Addenda of IWE, and clarifies that the 
new modifications in paragraphs (b)(2)(ix)(F) through (I) apply only to 
the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWE.
    Paragraph (b)(2)(ix)(F). This paragraph requires that VT-1 and VT-3 
examinations of containment surfaces be conducted in accordance with 
IWA-2200, and that personnel who perform visual inspections of 
containment surfaces be qualified in accordance with IWA-2300 in place 
of the ``owner-defined'' examination and personnel qualification 
provisions in the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWE.
    Paragraph (b)(2)(ix)(G). This paragraph requires that the VT-3 
examination method be used to conduct the examinations in Items E1.12 
and E1.20 in the 1998 Edition, 1999 Addenda, and 2000 Addenda of Table 
IWE-2500-1 in place of the ``owner-defined'' general visual examination 
provisions; the VT-1 examination method be used to conduct the 
examination in Item E4.11 of Table IWE-2500-1 in place of ``owner-
defined'' detailed visual examinations; and an examination of the 
pressure-retaining bolted connections in Item E1.11 of Table IWE-2500-1 
using the VT-3 examination method must be conducted once each interval.
    Paragraph (b)(2)(ix)(H). This paragraph supplements the examination 
requirements for containment bolted connections that are in Item E1.11 
of the 1998 Edition, 1999 Addenda, and 2000 Addenda of Table IWE-2500-
1. Containment bolted connections that are disassembled during the 
scheduled

[[Page 60533]]

performance of the examinations in Item E1.11 of Table IWE-2500-1 must 
be examined using the VT-3 examination method. Flaws or degradation 
identified during the performance of a VT-3 examination must be 
examined in accordance with the VT-1 examination method. The criteria 
in the material specification or IWB-3517.1 must be used to evaluate 
containment bolting flaws or degradation. As an alternative to 
performing VT-3 examinations of containment bolted connections that are 
disassembled during the scheduled performance of Item E1.11, VT-3 
examinations of containment bolted connections may be conducted 
whenever containment bolted connections are disassembled for any 
reason.
    Paragraph (b)(2)(ix)(I). This paragraph requires that the UT 
examination acceptance standard specified in the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of IWE-3511.3 for Class MC 
pressure-retaining components also apply to metallic liners of Class CC 
pressure-retaining components.
    Paragraph (b)(2)(xi). This paragraph extends the applicability of 
the existing regulation on the use of IWB-1220 to the 1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda of Section XI of the ASME 
BPV Code. Licensees using editions and addenda later than the 1989 
Addenda of Section XI are prohibited from exempting components from 
volumetric and surface examination as allowed by IWB-1220.
    Paragraph (b)(2)(xii). This paragraph prohibits the use of the 
irradiated material underwater weld provisions in the 1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-4660. Licensees 
must obtain NRC authorization in accordance with Sec.  50.55a(a)(3) of 
the method used to weld irradiated material underwater.
    Paragraph (b)(2)(xiv). This paragraph allows 8 hours of annual 
practice as described in VII-4240 of Supplement VII of Section XI, 1999 
Addenda and 2000 Addenda, to be performed in place of the existing 
hands-on training requirement in paragraph (b)(2)(xiv), provided that 
the supplemental practice is performed on material or welds that 
contain cracks, or by analyzing prerecorded data from material or welds 
that contain cracks. In either case, training must be completed no 
earlier than 6 months prior to performing ultrasonic examinations at a 
licensee's facility.
    Paragraph (b)(2)(xv). This paragraph extends the applicability of 
the existing regulations on Appendix VIII specimen set and 
qualification requirements to the 1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda of Section XI of the ASME BPV Code. Licensees 
choosing to use these modifications are required to apply all the 
modifications under paragraph (b)(2)(xv) except for those in 
(b)(2)(xv)(F) which are optional.
    Paragraphs (b)(2)(xv)(A), (A)(1), and (A)(2). These paragraphs 
update the UT examination coverage criteria to include examination 
coverage criteria for dissimilar metal piping welds when using 
personnel, procedures and equipment that are qualified in accordance 
with Supplement 10 of Appendix VII to Section XI. Dissimilar metal 
welds must be examined axially and circumferentially. Where examination 
from both sides is not possible on dissimilar metal welds, full 
coverage credit from a single side may be claimed only after completing 
a successful single-sided Appendix VIII demonstration using flaws on 
the opposite side of the weld. Dissimilar metal weld qualifications 
must be demonstrated from the austenitic side of the weld and may be 
used to perform examinations from either side of the weld.
    Paragraph (b)(2)(xv)(G)(4). Paragraph (b)(2)(xv)(G)(4) is removed. 
This requirement is redundant given the requirement in paragraph 
(b)(2)(xv)(G)(3) and is unnecessary. As a result, this revision 
involves no substantive change.
    Paragraph (b)(2)(xv)(K)(1)(i). This paragraph clarifies that flaws 
perpendicular to the weld located in the outer eighty-five (85) percent 
of the weld are not required to be included in the qualification test 
sample. The revision neither increases nor decreases current 
requirements, but clarifies conflicting requirements that currently 
exist.
    Paragraph (b)(2)(xv)(M). This paragraph clarifies that only the 
provisions in Supplement 12 to Appendix VIII that are related to the 
coordinated implementation of Supplement 3 to Supplement 2 performance 
demonstrations are required to be implemented.
    Paragraph (b)(2)(xvii). This paragraph extends the applicability of 
the existing regulation on reconciliation of quality requirements to 
the 1997 Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of 
Section XI of the ASME BPV Code. Licensees using IWA-4200 of this 
edition and these addenda are required to procure replacement and 
repair items under its approved quality assurance program required by 
Appendix B of 10 CFR 50. The limitation does not permit licensees to 
use IWA-4200 to procure repair and replacement items to be used in ASME 
Code safety-related applications that are manufactured under a non-
nuclear code or non-nuclear standard without an approved quality 
assurance program.
    Paragraph (b)(2)(xviii)(A). This paragraph requires that Level I 
and II NDE personnel be recertified on a 3-year interval in lieu of the 
5-year interval specified in IWA-2314.
    Paragraph (b)(2)(xviii)(B). This paragraph requires that IWA-2316 
may only be used to qualify personnel that observe for leakage during 
system leakage and hydrostatic tests conducted in accordance with IWA-
5211(a) and (b).
    Paragraph (b)(2)(xviii)(C). This paragraph requires that when 
qualifying VT-3 examination personnel in accordance with IWA-2317, the 
proficiency of the training must be demonstrated by administering an 
initial qualification examination and administering subsequent 
examinations on a 3-year interval.
    Paragraph (b)(2)(xix). This paragraph prohibits the use of the 
provisions in IWA-2240 and IWA-4520(c) which would allow alternative 
examination methods, a combination of methods, or newly developed 
techniques to be substituted for the methods specified in the 
Construction Code during repair and replacement activities.
    Paragraph (b)(2)(xx). This paragraph supplements the 1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-5213(a) to require 
a 10-minute hold time after attaining test pressure for Class 2 and 
Class 3 components that are not in use during normal operating 
conditions, and no hold time for the remaining Class 2 and Class 3 
components provided that system has been in operation for at least 4 
hours for insulated components or 10 minutes for uninsulated 
components.
    Paragraph (b)(2)(xxi)(A). This paragraph requires that licensees 
perform pressurizer and steam generator nozzle inside-radius 
inspections of Table IWB-2500-1, Examination Category B-D, Items B3.40 
and B3.60 (Inspection Program A) and Items B3.120 and B3.140 
(Inspection Program B) of the 1998 Edition. The 1999 Addenda and the 
2000 Addenda of Section XI are not permitted to be used. A visual 
examination with enhanced magnification that has a resolution 
sensitivity to detect a 1-mil width wire or crack, using the allowable 
flaw length criteria in Table IWB-3512-1, may be performed in place of 
a UT examination.
    Paragraph (b)(2)(xxi)(B). This paragraph requires that the CRD 
bolting examinations of Table IWB-2500-1, Examination Category B-G-2, 
Item

[[Page 60534]]

B7.80, of the 1995 Addenda of Section XI be retained only for used CRD 
bolting in ISI programs when using the 1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda of Section XI.
    Paragraph (b)(2)(xxi)(C). This paragraph requires that the 
attachment weld single-side volumetric examination of Table IWB-2500-1, 
Examination Category B-K, Item B10.10, of the 1995 Addenda of Section 
XI be retained in ISI programs when using the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of Section XI.
    Paragraph (b)(3). This paragraph incorporates by reference the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of the ASME OM 
Code. Licensees of nuclear power plants are required to use the 1998 
Edition up to and including the 2000 Addenda when updating their 
inservice testing programs in their subsequent 120-month inspection 
interval under Sec.  50.55a(f)(4)(ii).
    Paragraph (b)(3)(ii). This paragraph extends the applicability of 
the existing regulations on MOV test requirements to the 1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda of the ASME OM Code. 
Licensees using this edition and these addenda are required to 
establish a program to ensure that MOVs continue to be capable of 
performing their design basis safety functions. This paragraph 
clarifies that licensees must comply with the provisions of the ASME OM 
ISTC Code for testing MOVs, and reconciles the different subsection and 
paragraph numbers of the ASME OM Code that were renumbered in the 1998 
Edition and subsequent editions and addenda.
    Paragraph (b)(3)(iii). This paragraph extends the applicability of 
the existing regulation that permits the use of Code Case OMN-1 in 
place of stroke time test requirements to the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of the ASME OM Code, and 
reconciles those subsections of the ASME OM Code that were renumbered 
in the 1998 Edition. The modification continues to allow, as a 
voluntary alternative, the use of Code Case OMN-1 in place of the 
stroke-time testing requirements of paragraph (b)(3)(ii) when using 
this edition and these addenda.
    Paragraph (b)(3)(iv). This paragraph extends the applicability of 
the existing regulations in paragraphs (b)(3)(iv)(A), (B), and (C) on 
check valve condition monitoring requirements to the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of the ASME OM Code. There are 
no substantive changes in the requirements. This paragraph also 
reconciles the different subsection and paragraph numbers of the ASME 
OM Code that were renumbered in the 1998 Edition and subsequent 
editions and addenda.
    Paragraph (b)(3)(iv)(D). There are no substantive changes to the 
check valve condition monitoring requirements in ASME OM Code in this 
paragraph. This paragraph reconciles the different subsection and 
paragraph numbers of that were renumbered in the 1998 Edition and 
subsequent editions and addenda.
    Paragraph (b)(3)(v). This paragraph extends the applicability of 
the existing snubber ISI requirements to the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda of the ASME OM Code.
    Paragraph (b)(3)(vi). This paragraph requires that manual valves 
within the scope of the ASME OM Code be exercised on a 2-year interval 
rather than the 5-year interval specified in the 1999 Addenda and 2000 
Addenda of the ASME OM Code, provided that adverse conditions do not 
require more frequent testing. Paragraph ISTC-3540 of the ASME OM Code 
describes adverse conditions as harsh service environment, lubricant 
hardening, corrosive or sediment-laden process fluid, or degraded valve 
components.
    Paragraph (g)(6)(ii)(B). The paragraph removes the containment 
examination requirements in Sec. Sec.  50.55a(g)(6)(ii)(B)(1) through 
(4) because the implementation dates have expired and all licensees 
have completed the requirements (or a delay has been approved by an 
exemption); and redesignates the existing Sec.  50.55a(g)(6)(ii)(B)(5) 
as Sec.  50.55a(g)(6)(ii)(B). Licensees do not have to submit to the 
NRC staff for approval of their containment inservice inspection 
programs which were developed to satisfy the requirements of Subsection 
IWE and Subsection IWL with specified modifications and limitations. 
The program elements and the required documentation must be maintained 
on site for audit.
    Paragraph (g)(6)(ii)(C)(1). This paragraph clarifies that Appendix 
VIII to Section XI, 1995 Edition with the 1996 Addenda, as well as its 
supplements, must be implemented. Supplements 12 and 13 of Appendix 
VIII are eliminated from the implementation schedule.
    Paragraph (g)(6)(ii)(C)(2). This paragraph clarifies the 
requirements of Appendix VIII and the supplements to Appendix VIII to 
Section XI. Licensees implementing the 1989 Edition and earlier 
editions and addenda of IWA-22323 of Section XI must implement the 1995 
Edition with the 1996 Addenda of Appendix VIII of Section XI.

4. Generic Aging Lessons Learned Report

    In July 2001, the NRC issued, ``Generic Aging Lessons Learned 
(GALL) Report,'' NUREG-1801, Volumes 1 and 2, for use by applicants in 
preparing their license renewal applications. The GALL report evaluates 
existing generic programs, documents the basis for determining when 
generic existing programs are adequate without change, and documents 
when generic existing programs should be augmented for license renewal. 
Section XI, Division 1, of the ASME BPV Code is one of the generic 
existing programs in the GALL report that is evaluated as an aging 
management program (AMP) for license renewal. Subsections IWB, IWC, 
IWD, IWF, IWE, and IWL of the 1995 Edition up to and including the 1996 
Addenda of Section XI of the ASME BPV for ISI were evaluated in the 
GALL report and the conclusions in the GALL report are valid for these 
edition and addenda.
    In the GALL report Sections XI.M1, ``ASME Section XI Inservice 
Inspection, Subsections IWB, IWC, and IWD,'' XI.S1, ``ASME Section XI, 
Subsection IWE,'' XI.S2, ``ASME Section XI, Subsection IWL,'' and 
XI.S3, ``ASME Section XI, Subsection IWF,'' describe the evaluation and 
technical basis for determining the adequacy of Subsections IWB, IWC, 
IWD, IWE, IWL, and IWF, 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 (i.e., XI.M3, XI.M4, XI.M5, 
XI.M6, XI.M7, XI.M8, XI.M9, XI.M11, XI.M12, XI.M13, XI.M14, XI.M15, 
XI.M16, XI.M18. XI.M24, XI.M25, and XI.M32). These AMPs were evaluated 
for 10 specific elements with such attributes as scope of program, 
preventive actions, parameters monitored/inspected, detection of aging 
effects, monitoring and trending, acceptance criteria, corrective 
actions, confirmation process, administrative controls, and operating 
experience. If an applicant takes credit for a program in GALL, it is 
incumbent on the applicant to ensure that the plant program contains 
all the elements of the referenced GALL program. The GALL report 
contains one acceptable way to manage aging effects for license 
renewal. An applicant may propose alternatives for NRC review in its 
plant-specific license renewal application.
    The NRC has completed an evaluation of Subsections IWB, IWC, IWD, 
IWE, IWF, and IWL of Section XI of the ASME BPV Code (1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda) as part of the Sec.  
50.55a

[[Page 60535]]

amendment process to ensure that the conclusions of the GALL report 
remain valid. Although some of the revisions in Section XI of the ASME 
BPV Code relax the provisions of the 1995 Edition with the 1996 
Addenda, the revisions are acceptable (except as discussed below) 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 ASME BPV Code (1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda) as acceptable alternatives to the 
requirements of the 1995 Edition up to and including the 1996 Addenda 
of the ASME Code, Section XI, referenced in the GALL AMPs without the 
need to submit these alternatives for NRC review in its plant-specific 
license renewal application. Similarly, a licensee approved for license 
renewal that relied on the GALL AMPs may use Subsections IWB, IWC, IWD, 
IWE, IWF, and IWL of Section XI of the ASME BPV Code (1997 Addenda, 
1998 Edition, 1999 Addenda, and 2000 Addenda) as acceptable 
alternatives to the AMPs described in the GALL report.
    Several of the revisions to Subsections IWB, IWE, and IWL that are 
discussed in the preceding Section 2, Public Comments on Proposed Rule; 
and Final Rule, might affect the validity of the conclusions in the 
GALL report because provisions in the 1995 Edition up to and including 
the 1996 Addenda that address examination requirements and acceptance 
standards have been relaxed or eliminated in the 1997 Addenda, 1998 
Edition, 1999 Addenda, and 2000 Addenda. The new limitations and 
modifications in Sec.  50.55a(b) require that the revised provisions be 
supplemented with additional inspection requirements as a condition for 
their use. The conclusions of the GALL report remain valid for the 1997 
Addenda, 1998 Edition, 1999 Addenda, and 2000 Addenda of Section XI of 
the ASME BPV Code with the use of these new limitations and 
modifications as discussed in this final rulemaking. However, it should 
be noted that the NRC is imposing these limitations and modifications 
to ensure consistency and an acceptable level of safety in the 
examination requirements and acceptance standards, and not solely to 
validate the conclusions in the GALL report.
    The GALL report identified areas of the 1995 Edition with the 1996 
Addenda of Section XI of the ASME Code that require augmentation for 
license renewal. A license renewal applicant may either augment their 
AMPS in these areas as described in the GALL report, or propose 
alternatives for NRC review in its plant-specific license renewal 
application. The GALL report's conclusions with respect to augmentation 
in connection with a license renewal application also apply when 
implementing the 1998 Edition, 1999 Addenda, and 2000 Addenda of 
Section XI of the ASME Code.

5. 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 Website (Web). The NRC's interactive rulemaking Website 
is located at http://ruleforum.llnl.gov. These documents may be viewed 
and downloaded electronically via this Website.
    NRC's Public Electronic Reading Room (PERR). 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, 
Regulatory Analysis, Environmental Assessment, and Resolution of Public 
Comments be obtained from Stephen Tingen, Division of Engineering, 
Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001. Alternatively, you may contact 
Mr. Tingen at (301) 415-1280, or via e-mail at: [email protected].

----------------------------------------------------------------------------------------------------------------
                   Document                          PDR              Web             PERR          NRC staff
----------------------------------------------------------------------------------------------------------------
Federal Register Notice......................               X                X   ..............               X
Regulatory Analysis..........................               X                X     ML 022130308               X
Environmental Assessment.....................               X                X     ML 022130316               X
Resolution of Public Comments................               X                X     ML 022130320               X
Public Comments..............................               X                X     ML 021480072               X
----------------------------------------------------------------------------------------------------------------

6. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, Pub. 
L. 104-113, requires that Federal agencies 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. In this final rule, the NRC is amending its 
regulations to incorporate by reference a later edition and addenda of 
Sections III and XI of the ASME BPV Code and the ASME OM Code, for 
construction, ISI, and IST of nuclear power plant components, as 
identified in the preceding Section 2, Public Comments on Proposed 
Rule; and Final Rule.
    A number of commenters stated that the NRC approval of the ASME 
Code with exceptions (i.e., modifications and limitations) does not 
meet the spirit of Pub. L. 104-113. The NRC disagrees because although 
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, nor does it forbid Federal 
agencies from endorsing industry consensus standards with limitations 
or modifications. 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. Moreover, 
there is no legislative history suggesting that Congress intended 
agencies to take an ``all or nothing'' approach to endorsement of 
voluntary consensus standards under the Act, and the OMB guidance 
implementing Pub. L. 104-113 does not address the matter. The 
discussion in the statement of considerations of the limitations and 
modifications is sufficient to satisfy the requirements of Section 
12(d)(3) of Pub. L. 104-113, and the relevant requirements of OMB 
Circular A-119 (1998). In light of these factors, the NRC concludes 
that the explanations for the modifications and limitations to the ASME 
BPV and OM Codes, as set forth

[[Page 60536]]

in the statement of considerations for this final rule, satisfy the 
requirements of Section 12(d)(3) of Pub. L. 104-113, and OMB Circular 
A-119.

7. Finding of No Significant Environmental Impact: Availability

    The Commission has determined under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 
Subpart A of 10 CFR Part 51, that this rule is not a major Federal 
action significantly affecting the quality of the human environment, 
and, therefore, an environmental impact statement is not required.
    This rulemaking will not significantly increase the probability or 
consequences of accidents; no changes are being made in the types of 
any effluents that may be released off-site; the environmental 
assessment for this rule demonstrates that there is a small decrease in 
occupational exposure; and there is no significant increase in public 
radiation exposure. Therefore, there are no significant radiological 
impacts associated with the action. The rulemaking does not involve 
non-radiological plant effluents and has no other environmental impact. 
Therefore, no significant non-radiological impacts are associated with 
the action.
    The determination for this rule is that there will be no 
significant off-site impact to the public from this action. The NRC has 
prepared an environmental assessment on this final rule. The 
environmental assessment is available as indicated in Section 5, 
Availability of Documents, under the Supplementary Information heading.
    The NRC requested the views of the States on the environmental 
assessment for the rule and did not receive any comments from the 
States.

8. Paperwork Reduction Act Statement

    This final rule amends information collection requirements that are 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). These requirements were approved by the Office of Management and 
Budget, approval number 3150-0011.
    Because the rule will reduce existing information collection 
requirements, the public burden for these information collections is 
expected to be decreased by 14 hours per licensee. This reduction 
includes the time required for reviewing instructions, searching 
existing data sources, gathering and maintaining the data needed, and 
completing and reviewing the information collection. Send comments on 
any aspect of these information collections, including suggestions for 
further reducing the burden, to the Records Management Branch (T-6 E6), 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by 
Internet electronic mail to [email protected]; and to the Desk 
Officer, Office of Information and Regulatory Affairs, NEOB-10202 
(3150-0011), Office of Management and Budget, Washington, DC 20503.

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

9. Regulatory Analysis

    The NRC has prepared a regulatory analysis on this final rule. The 
analysis examines the costs and benefits of the action considered by 
the Commission. The regulatory analysis is available as indicated in 
Section 5, Availability of Documents, under the Supplementary 
Information heading.
    One commenter stated that the regulatory analysis for the proposed 
amendment failed to address the values and impacts associated with a 
number of the modifications and limitations in the proposed rule. The 
NRC notes that the purpose of the regulatory analysis is to identify 
any significant values and impact associated with updating from the 
1995 Edition with the 1996 Addenda to the 1997 Addenda, 1998 Edition, 
1999 Addenda, and 2000 Addenda of the ASME Code. Therefore, 
modifications and limitations that require licensees to use the 
existing Code provisions in the 1995 Edition with the 1996 Addenda of 
the ASME Code are not addressed in the regulatory analysis.

10. Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 
605(b), the Commission certifies that this rule will not have a 
significant economic impact on a substantial number of small entities. 
This final rule affects only 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 size standards established by the NRC 
(10 CFR 2.810).

11. 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 Boiler and Pressure Vessel Code (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 
and valves in accordance with the rules provided in the ASME Code for 
Operation and Maintenance of Nuclear Power Plants (OM Code). This final 
rule incorporates by reference the 1997 Addenda, 1998 Edition, 1999 
Addenda, and 2000 Addenda 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 later editions and addenda of Section 
III, Division 1, of the ASME BPV Code is prospective in nature. The 
later editions and addenda do 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 later 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 later 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 
generally does not apply to incorporation by reference of later 
editions and addenda of the ASME BPV (Section XI) and OM Codes for the 
following reasons--
    (1) The NRC's longstanding policy has been to incorporate later 
versions of the

[[Page 60537]]

ASME Codes into its regulations. This is codified in Sec.  50.55a which 
requires licensees to revise their ISI and IST programs every 120 
months to the latest edition and addenda of Section XI of the ASME BPV 
Code and the ASME OM Code incorporated by reference into Sec.  50.55a 
that is in effect 12 months prior to the start of a new 120-month ISI 
and IST interval. Thus, when the NRC endorses a later version of the 
Code, it is implementing this longstanding policy and requirement.
    (2) 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 utilities) participate. This 
consideration is consistent with both the intent and spirit of the 
Backfit Rule (i.e., the NRC provides for the protection of the public 
health and safety, and does not unilaterally imposed undue burden on 
applicants or licensees).
    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 or OM Code 
provision, but merely retains the current existing requirement, 
prohibits the use of the use of the later Code provision, or limits the 
use of the later Code provision, 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 for the rule. Sections 50.55a(b)(2)(viii)(F), 
(b)(2)(ix)(F), (b)(2)(ix)(G), (b)(2)(ix)(H), (b)(2)(xviii)(A), (B) and 
(C), (b)(2)(xix), (b)(2)(xxi)(A) and (C) in this final rule either 
retain current existing requirements, prohibit the use of the later 
Code provision, or limit the use of the later Code provision.
    (2) When an NRC exception relaxes an existing ASME BPV or OM Code 
provision but does not prohibit a licensee from using the existing Code 
provision. Section 50.55a(b)(3)(vi) in this final rule relaxes the use 
of an existing Code provision but does not prohibit a licensee from 
using the existing Code provision.
    There are some circumstances where the NRC considers it appropriate 
to treat as a backfit the endorsement of a later ASME BPV or OM Code--
    (1) When the NRC endorses a later provision of the ASME BPV or OM 
Code that takes a substantially different direction from the currently 
existing requirements, the action is treated as a backfit. An example 
was the NRC's initial endorsement of Subsections IWE and IWL of Section 
XI, which imposed containment inspection requirements on operating 
reactors for the first time. The final rule dated August 8, 1996 (61 FR 
41303), incorporated by reference in Sec.  50.55a the 1992 Edition with 
the 1992 Addenda of IWE and IWL of Section XI to require that 
containments be routinely inspected to detect defects that could 
compromise a containment's structural integrity. This action expanded 
the scope of Sec.  50.55a to include components that were not 
considered by the existing regulations to be within the scope of ISI. 
Since those requirements involved a substantially different direction, 
they were treated as backfits, and justified in accordance with the 
standards of 10 CFR 50.109. There are no provisions similar to this in 
the final rule.
    (2) When the NRC requires implementation of later ASME BPV 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. An example was the final rule dated September 22, 
1999 (64 FR 51370), which incorporated by reference the 1989 Addenda 
through the 1996 Addenda of Section III and Section XI of the ASME BPV 
Code, and the 1995 Edition with the 1996 Addenda of the ASME OM Code. 
The final rule expedited the implementation of the 1995 Edition with 
the 1996 Addenda of Appendix VIII of Section XI of the ASME BPV Code 
for qualification of personnel and procedures for performing UT 
examinations. The expedited implementation of Appendix VIII was 
considered a backfit because licensees were required to implement the 
new requirements in Appendix VIII prior to the next 120-month ISI 
program inspection interval update. Another example was the final rule 
dated August 6, 1992 (57 FR 34666), which incorporated by reference in 
Sec.  50.55a the 1986 Addenda through the 1989 Edition of Section III 
and Section XI of the ASME BPV Code. The final rule added a requirement 
to expedite the implementation of the revised reactor vessel shell weld 
examinations in the 1989 Edition of Section XI. Imposing these 
examinations was considered a backfit because licensees were required 
to implement the examinations prior to the next 120-month ISI program 
inspection interval update. There are no provisions similar to this in 
the final rule.
    (3) When the NRC takes an exception to a ASME BPV or OM Code 
provision and imposes a requirement that is substantially different 
from the current existing requirement as well as substantially 
different than the later Code.
    In Sec. Sec.  50.55a(b)(2)(xv)(A), (A)(1) and (A)(2) that are 
discussed in the preceding Section 2, Final Rule and Comments on 
Proposed Rule, the NRC is adopting dissimilar metal piping weld 
ultrasonic (UT) examination coverage requirements. The NRC concludes 
that the addition of dissimilar metal piping weld UT examination 
coverage requirements to the regulation is necessary to correct the 
omission by the ASME BPV Code to ensure adequate protection of public 
health and safety. This backfit falls into the ``adequate protection'' 
exception under 10 CFR 50.109(a)(4)(ii), and the documented evaluation 
required by 10 CFR 50.109(a)(6) is below. Therefore, a backfit analysis 
under 10 CFR 50.109(a)(3) is not required.

Documented Evaluation

    Dissimilar metal piping weld examination coverage requirements, 
although contained in the 1989 Edition, and earlier editions and 
addenda of Appendix III of Section XI of the ASME BPV Code, are not 
addressed in later editions and addenda of Section XI. Appendix VIII 
was added in the 1989 Addenda of Section XI, and the UT examination 
criteria for piping welds in Appendix VIII supercede the examination 
criteria for piping welds in Appendix III. Although Appendix VIII 
addresses qualification of personnel, procedures, and equipment used to 
conduct UT examinations of dissimilar metal piping welds, Appendix VIII 
(unlike Appendix III) does not define UT examination coverage criteria 
for dissimilar metal piping welds. Therefore, the addition of 
dissimilar metal piping weld examination coverage requirements to the 
regulation is necessary to correct the omission by the ASME BPV Code.
    The purpose of ISI is to monitor for degrading conditions and 
ensure that any flaws which develop during service can be detected, 
sized, and evaluated, and that components with unacceptable flaws are 
repaired or replaced to adequately maintain the integrity of the 
pressure boundary. Another purpose of ISI is to identify any possible 
generic-type defects that were unforeseen during the design stage so 
that corrective actions can be taken prior to a breach of the pressure 
boundary. Although plants have generally been designed with sufficient 
margin so that important components will not crack or undergo excessive 
degradation, uncertainties in the definition of

[[Page 60538]]

service-induced loads and operating environments may have led to a less 
than optimum choice of materials, and may have permitted degradation 
mechanisms to progress more rapidly, or allowed different mechanisms to 
be active during plant operation, than were foreseen in the design.
    Section XI defines inspection criteria for ISI and indicates 
allowable flaw sizes (with margin) based on fracture mechanics for 
various locations within reactor coolant pressure boundary (RCPB) 
components. If a flaw is found that exceeds the allowable size, (1) the 
component must be repaired, or (2) a safety analysis must be conducted, 
using fracture mechanics, to show that the flaw will not grow to an 
extent that could impair the integrity of the component. To conduct 
reliable and credible safety evaluations using fracture mechanics, 
information from the UT examination is required regarding the flaw 
size, shape, orientation, and location within the component. 
Consequently, examination information is key to detecting flaws and 
assessing the continued reliability and safety of flawed RCPB 
components.
    Dissimilar metal welds are used to connect RCPB components. 
Operating history shows serious degradation of RCPB dissimilar metal 
welds have occurred at several nuclear power plants in the United 
States and at one foreign nuclear power plant. The NRC believes that 
additional occurrences are possible. Therefore, comprehensive and 
technically sound UT examination coverage criteria for dissimilar metal 
piping welds are needed to ensure that each facility provides adequate 
protection to the health and safety of the public. Sections 
50.55a(b)(2)(xv)(A), (A)(1) and (A)(2) impose requirements that define 
comprehensive and technically sound UT examination coverage criteria 
for dissimilar metal piping welds that ensure uniform examination 
results among all licensees. These UT examination coverage requirements 
are necessary to detect flaws in dissimilar metal welds in RCPB 
components, thereby maintaining an extremely low probability of 
abnormal leakage or rapidly propagating failure, and gross rupture.
    The remaining portion of this section addresses public comments 
related to backfitting or backfit issues on the proposed rule.
    A number of commenters raised a generic concern with regard to the 
NRC's position on imposing exceptions (i.e., modification or 
limitation) to consensus standards that are incorporated by reference 
in the Code of Federal Regulations. The commenters believe that, 
contrary to the NRC's determination, imposing any modification or 
limitation to the ASME Code constitutes a backfit for which a backfit 
analysis is required. Commenters stated that NRC is required to 
demonstrate that modifications and limitations result in an increase in 
quality or safety.
    The NRC has reviewed the comments and has concluded that the 
commenters do not raise concerns which would alter the previous 
conclusion that the Backfit Rule does not require a backfit analysis of 
the modifications and limitations imposed by the NRC in the final rule. 
Furthermore, many of the modifications and limitations imposed during 
previous routine updates of Sec.  50.55a have established a precedence 
for determining which modifications or limitations are backfits or 
require a backfit analysis (final rules dated August 6, 1992 (57 FR 
34666), August 8, 1996 (61 FR 41303), and September 22, 1999 (64 FR 
51370)). The NRC finds that the application of the backfit requirements 
to modifications and limitations in the current rule are consistent 
with the application of backfit requirements to modifications and 
limitations in previous rules. Since the modification and limitations 
in the current rule are not considered backfits or do not require 
backfit analyses, the NRC is not required to demonstrate that the new 
modifications and limitations result in an increase in quality or 
safety.
    Section 50.55a(b)(2)(ix)(F) of the proposed rule would require that 
personnel who conduct visual examinations of containment surfaces be 
qualified in accordance with the 1998 Edition, 1999 Addenda, and 2000 
Addenda of IWA-2300 in place of the ``owner-defined'' qualification 
provisions in the 1998 Edition, 1999 Addenda, and 2000 Addenda IWE-
2330(a). One commenter stated that the NRC is imposing additional 
qualification requirements for personnel that conduct general visual 
examinations in accordance with the 1998 Edition, 1999 Addenda, and 
2000 Addenda of IWE that were not imposed on general visual 
examinations conducted in accordance with earlier editions and addenda 
of IWE.
    The NRC agrees with the commenter. The NRC proposed additional 
qualification requirements for personnel that conduct general visual 
examinations. Editions and addenda of IWE earlier than the 1998 Edition 
required the use of the VT-1 visual inspection method, the VT-3 visual 
inspection method, and a general visual inspection. The provisions in 
IWA-2300 were used to define the qualification requirements for 
personnel that conduct VT-1 and VT-3 visual examinations; however, 
detailed qualification requirements were not provided in the ASME Code 
for personnel that conduct general visual examinations. There are 
significant changes in the visual examination requirements in the 1998 
Edition of IWE. Paragraph IWE-2330(a) requires that the licensee define 
the qualification requirements for personnel that conduct all visual 
examinations of containment surfaces, and a number of visual 
examinations are recategorized as general visual examinations that were 
formerly categorized as VT-1 or VT-3 in earlier editions and addenda of 
IWE. The intent of Sec.  50.55a(b)(2)(ix)(F) in the proposed rule was 
not to allow licensees to use ``owner-defined'' qualification 
requirements to qualify personnel that conduct examinations that were 
formerly categorized as VT-1 or VT-3. However, the NRC inadvertently 
worded the modification such that additional qualification requirements 
would also be imposed on personnel that conduct general visual 
examinations. Therefore, the qualification requirements for personnel 
that conduct visual inspections of containment surfaces are revised in 
the final rule to require that personnel who conduct VT-1 and VT-3 
visual examinations of containment surfaces be qualified in accordance 
with the 1998 Edition, 1999 Addenda, and 2000 Addenda of IWA-2300.
    Section 50.55a(b)(2)(ix)(G) in the proposed rule would require that 
the general visual examinations required by the 1998 Edition, 1999 
Addenda, and 2000 Addenda of IWE-2310(b) and IWE-2310(c) meet VT-3 
examination method provisions in the 1998 Edition, 1999 Addenda, and 
2000 Addenda of IWA-2210 in place of the ``owner-defined'' general and 
detailed visual examination provisions in the 1998 Edition, 1999 
Addenda, and 2000 Addenda of IWE-2310(a).
    One commenter stated that it is inappropriate for the NRC to impose 
Sec.  50.55a(b)(2)(ix)(G) without performing a backfit analysis because 
the modification increases the frequency of VT-3 examinations of 
containment surfaces beyond that which was previously required in the 
editions and addenda of IWE earlier than the 1998 Edition. The 
commenter is correct. It was not the intent of the NRC to increase the 
frequency of VT-3 visual examinations of containment surfaces. The NRC 
inadvertently worded the modification such that the frequency of VT-3 
examinations of containment areas was increased. Therefore, Sec.  
50.55a(b)(2)(ix)(G) is revised in the

[[Page 60539]]

final rule to require that VT-3 visual examinations for certain 
containment areas be performed once during each 10-year inspection 
interval which is consistent with the provisions in the editions and 
addenda of IWE earlier than the 1998 Edition.
    Sections 50.55a(b)(2)(ix)(J), (b)(2)(xx), and (b)(2)(xxi)(B) in the 
proposed rule involve provisions in Section XI that were deleted in the 
1995 Addenda that the NRC is reinstating in the final rule (Sec.  
50.55a(b)(2)(ix)(J) of the proposed rule is renumbered as Sec.  
50.55a(b)(2)(ix)(I) in the final rule). Section 50.55a(b)(2)(xxiii) of 
the proposed rule involves underwater welding provisions in Section XI 
that were added in the 1996 Addenda that the NRC is prohibiting the use 
of in the final rule (Sec.  50.55a(b)(2)(xxiii) of the proposed rule is 
renumbered as Sec.  50.55a(b)(2)(xii) in the final rule).
    Several commenters stated that it is inappropriate for the NRC to 
reinstate or prohibit the use of these Code provisions because the 
elimination or addition of these Code provisions was previously 
accepted by the NRC the final rule dated September 22, 1999 (64 FR 
51370). The NRC disagrees. These modifications were not included in the 
final rule that incorporated by reference the 1995 Addenda and 1996 
Addenda of Section XI in 10 CFR 50.55a (64 FR 51370) due to an 
oversight by the NRC. The NRC did not identify that these Code 
provisions were eliminated or added when it reviewed the 1995 Addenda 
and 1996 Addenda of Section XI. The NRC has determined that these 
modifications should only apply to those licensees who implement the 
1997 Addenda and later editions and addenda of Section XI, and should 
not be backfit to those licensees who update their ISI programs to the 
1995 Edition with the 1996 Addenda in accordance with 10 CFR 
50.55a(g)(4)(ii). The NRC has determined it is acceptable not to 
backfit the licensees who update their ISI programs to the 1995 Edition 
with the 1996 Addenda, because those licensees will be required at the 
next 10-year interval to update their ISI programs to include or 
prohibit the relevant Code provisions. Thus, any problems would be 
caught during the next 10-year interval. The reinstatement or 
prohibition of the relevant Code provisions are not considered 
backfits, because they are imposed only as part of the routine updating 
required as part of the 120-month updating, and do not constitute a 
significant change to, or fundamental modification of the existing ISI 
program.
    Section 50.55a(b)(3)(vi) in the proposed rule would prohibit the 
extension of the exercise interval for manual valves from 3 months to 5 
years when using the 1999 Addenda and 2000 Addenda of ISTC-3540. One 
commenter stated that the NRC should delete Sec.  50.55a(b)(3)(vi) or 
conduct a backfitting analysis justifying the imposition of the 
proposed modification.
    The NRC disagrees that a backfit analysis is required for Sec.  
50.55a(b)(3)(vi). The intent of the ASME consensus process was to 
extend the exercise interval for manual valves, and in this case, the 
NRC is accommodating the ASME consensus process to the extent that the 
NRC believes the extended exercise interval can be justified (i.e., 2 
years). In this case the NRC is allowing a relaxation from the current 
requirements, but not as much of a relaxation as the later Code would 
allow. Licensees are free to continue to implement the existing 
requirement (e.g., testing every three months).
    The proposed rule would add a new Sec.  50.55a(g)(6)(ii)(B)(1) to 
clarify the start date of the first 120-month interval for the ISI of 
Class MC and Class CC components. One commenter noted that since 
licensees have already established the start date of the first 120-
month interval for the ISI of Class MC and Class CC components, it is a 
backfit for the NRC to now impose a different start date than that 
already established by licensees. The NRC agrees with this comment. 
Therefore, Sec.  50.55a(g)(6)(ii)(B)(1) in the proposed rule is not 
adopted.

12. Small Business Regulatory Enforcement Fairness Act

    In accordance with the Small Business Regulatory Enforcement 
Fairness Act of 1996, the NRC has determined that this action is not a 
major rule and has verified this determination with the Office of 
Information and Regulatory Affairs of OMB.

List of Subjects in 10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
protection, Incorporation by reference, Intergovernmental relations, 
Nuclear power plants and reactors, Radiation protection, Reactor siting 
criteria, Reporting and recordkeeping requirements.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended, the Energy Reorganization 
Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting 
the following amendments to 10 CFR Part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    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, 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, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, 
as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 as amended by Pub L. 102-486, sec. 2902, 106 Stat. 3123 (42 
U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68 
Stat. 936, 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(dd), 
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 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:
    (a) removing paragraphs (b)(2)(xv)(G)(4), (g)(6)(ii)(B)(1) through 
(g)(6)(ii)(B)(4);
    (b) redesignating and revising paragraph (g)(6)(ii)(B)(5) as 
(g)(6)(ii)(B);
    (c) revising the introductory text of paragraph (b)(1), paragraphs 
(b)(1)(ii), (b)(1)(iii) and (b)(1)(v), the introductory text of 
paragraph (b)(2), paragraph (b)(2)(vi), the introductory text of 
paragraphs (b)(2)(viii) and (b)(2)(ix), paragraphs (b)(2)(xi), and 
(b)(2)(xiv), the introductory text of paragraph (b)(2)(xv), paragraphs 
(b)(2)(xv)(A), (b)(2)(xv)(K)(1)(i) and (b)(2)(xvii), the introductory 
text of paragraph (b)(3), paragraph (b)(3)(ii), the introductory text 
of paragraphs (b)(3)(iii) and (b)(3)(iv), and paragraphs (b)(3)(v) and 
(g)(6)(ii)(C)(1); and
    (d) adding paragraphs (b)(2)(viii)(F), (b)(2)(ix)(F) through 
(b)(2)(ix)(I), (b)(2)(xii), (b)(2)(xv)(M), (b)(2)(xviii) through 
(b)(2)(xxi), (b)(3)(iv)(D), (b)(3)(vi), and (g)(6)(ii)(C)(2).
    The amended text is set forth to read as follows:


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

[[Page 60540]]

1973 Winter Addenda, and the 1974 Edition (Division 1) through the 2000 
Addenda (Division 1), subject to the following limitations and 
modifications:
* * * * *
    (ii) Weld leg dimensions. When applying the 1989 Addenda through 
the latest edition and addenda incorporated by reference in paragraph 
(b)(1) of this section, licensees may not apply paragraph NB-
3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-
3673.2(b)-1.
    (iii) Seismic design. Licensees may use Articles NB-3200, NB-3600, 
NC-3600, and ND-3600 up to and including the 1993 Addenda, subject to 
the limitation specified in paragraph (b)(1)(ii) of this section. 
Licensees may not use these Articles in the 1994 Addenda through the 
latest edition and addenda incorporated by reference in paragraph 
(b)(1) of this section.
* * * * *
    (v) Independence of inspection. Licensees may not apply NCA-
4134.10(a) of Section III, 1995 Edition through the latest edition and 
addenda incorporated by reference in paragraph (b)(1) of this section.
    (2) As used in this section, references to Section XI of the ASME 
Boiler and Pressure Vessel Code refer to Section XI, and include the 
1970 Edition through the 1976 Winter Addenda, and the 1977 Edition 
(Division 1) through the 2000 Addenda (Division 1), subject to the 
following limitations and modifications:
* * * * *
    (vi) Effective edition and addenda of Subsection IWE and Subsection 
IWL, Section XI. Licensees may use either the 1992 Edition with the 
1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection 
IWE and Subsection IWL as modified and supplemented by the requirements 
in paragraphs (b)(2)(viii) and (b)(2)(ix) of this section when 
implementing the initial 120-month inspection interval for the 
containment inservice inspection requirements of this section. 
Successive 120-month interval updates must be implemented in accordance 
with paragraph (g)(4)(ii) of this section.
* * * * *
    (viii) Examination of concrete containments. Licensees applying 
Subsection IWL, 1992 Edition with the 1992 Addenda, shall apply 
paragraphs (b)(2)(viii)(A) through (b)(2)(viii)(E) of this section. 
Licensees applying the 1995 Edition with the 1996 Addenda shall apply 
paragraphs (b)(2)(viii)(A), (b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of 
this section. Licensees applying the 1998 Edition with the 1999 and 
2000 Addenda shall apply paragraphs (b)(2)(viii)(E) and (b)(2)(viii)(F) 
of this section.
* * * * *
    (F) Personnel that examine containment concrete surfaces and tendon 
hardware, wires, or strands must meet the qualification provisions in 
IWA-2300. The ``owner-defined'' personnel qualification provisions in 
IWL-2310(d) are not approved for use.
    (ix) Examination of metal containments and the liners of concrete 
containments. Licensees applying Subsection IWE, 1992 Edition with the 
1992 Addenda, or the 1995 Edition with the 1996 Addenda, shall satisfy 
the requirements of paragraphs (b)(2)(ix)(A) through (b)(2)(ix)(E) of 
this section. Licensees applying the 1998 Edition with the 1999 Addenda 
and 2000 Addenda shall satisfy the requirements of paragraphs 
(b)(2)(ix)(A), (b)(2)(ix)(B), and (b)(2)(ix)(F) through (b)(2)(ix)(I) 
of this section.
* * * * *
    (F) VT-1 and VT-3 examinations must be conducted in accordance with 
IWA-2200. Personnel conducting examinations in accordance with the VT-1 
or VT-3 examination method shall be qualified in accordance with IWA-
2300. The ``owner-defined'' personnel qualification provisions in IWE-
2330(a) for personnel that conduct VT-1 and VT-3 examinations are not 
approved for use.
    (G) The VT-3 examination method must be used to conduct the 
examinations in Items E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 
examination method must be used to conduct the examination in Item 
E4.11 of Table IWE-2500-1. An examination of the pressure-retaining 
bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 
examination method must be conducted once each interval. The ``owner-
defined'' visual examination provisions in IWE-2310(a) are not approved 
for use for VT-1 and VT-3 examinations.
    (H) Containment bolted connections that are disassembled during the 
scheduled performance of the examinations in Item E1.11 of Table IWE-
2500-1 must be examined using the VT-3 examination method. Flaws or 
degradation identified during the performance of a VT-3 examination 
must be examined in accordance with the VT-1 examination method. The 
criteria in the material specification or IWB-3517.1 must be used to 
evaluate containment bolting flaws or degradation. As an alternative to 
performing VT-3 examinations of containment bolted connections that are 
disassembled during the scheduled performance of Item E1.11, VT-3 
examinations of containment bolted connections may be conducted 
whenever containment bolted connections are disassembled for any 
reason.
    (I) The ultrasonic examination acceptance standard specified in 
IWE-3511.3 for Class MC pressure-retaining components must also be 
applied to metallic liners of Class CC pressure-retaining components.
* * * * *
    (xi) Class 1 piping. 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.
    (xii) Underwater Welding. The provisions in IWA-4660, ``Underwater 
Welding,'' of Section XI, 1997 Addenda through the latest edition and 
addenda incorporated by reference in paragraph (b)(2) of this section, 
are not approved for use on irradiated material.
* * * * *
    (xiv) Appendix VIII personnel qualification. All personnel 
qualified for performing ultrasonic examinations in accordance with 
Appendix VIII shall receive 8 hours of annual hands-on training on 
specimens that contain cracks. Licensees applying the 1999 Addenda 
through the latest edition and addenda incorporated by reference in 
paragraph (b)(2) of this section may use the annual practice 
requirements in VII-4240 of Supplement VII of Section XI in place of 
the 8 hours of annual hands-on training provided that the supplemental 
practice is performed on material or welds that contain cracks, or by 
analyzing prerecorded data from material or welds that contain cracks. 
In either case, training must be completed no earlier than 6 months 
prior to performing ultrasonic examinations at a licensee's facility.
    (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 latest edition and addenda 
incorporated by reference in paragraph (b)(2) of this section. 
Licensees choosing to apply these provisions shall apply all of the 
following provisions under paragraph (b)(2)(xv) except for those in 
paragraph (b)(2)(xv)(F) which are optional.
    (A) When applying Supplements 2, 3, and 10 to Appendix VIII, the 
following examination coverage criteria requirements must be used:

[[Page 60541]]

    (1) Piping must be examined in two axial directions, and when 
examination in the circumferential direction is required, the 
circumferential examination must be performed in two directions, 
provided access is available. Dissimilar metal welds must be examined 
axially and circumferentially.
    (2) Where examination from both sides is not possible, full 
coverage credit may be claimed from a single side for ferritic welds. 
Where examination from both sides is not possible on austenitic welds 
or dissimilar metal welds, full coverage credit from a single side may 
be claimed only after completing a successful single-sided Appendix 
VIII demonstration using flaws on the opposite side of the weld. 
Dissimilar metal weld qualifications must be demonstrated from the 
austenitic side of the weld and may be used to perform examinations 
from either side of the weld.
* * * * *
    (K) * * *
    (1) * * *
    (i) For detection, a minimum of four flaws in one or more full-
scale nozzle mock-ups must be added to the test set. The specimens must 
comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for 
flaw locations specified in Table VIII S6-1. Flaws may be either 
notches, fabrication flaws or cracks. Seventy-five (75) percent of the 
flaws must be cracks or fabrication flaws. Flaw locations and 
orientations must be selected from the choices shown in paragraph 
(b)(2)(xv)(K)(4) of this section, Table VIII-S7-1--Modified, with the 
exception that flaws in the outer eighty-five (85) percent of the weld 
need not be perpendicular to the weld. There may be no more than two 
flaws from each category, and at least one subsurface flaw must be 
included.
* * * * *
    (M) When implementing Supplement 12 to Appendix VIII, only the 
provisions related to the coordinated implementation of Supplement 3 to 
Supplement 2 performance demonstrations are to be applied.
* * * * *
    (xvii) Reconciliation of Quality Requirements. When purchasing 
replacement items, in addition to the reconciliation provisions of IWA-
4200, 1995 Edition through the latest edition and addenda incorporated 
by reference in paragraph (b)(2) of this section, the replacement items 
must be purchased, to the extent necessary, in accordance with the 
licensee's quality assurance program description required by 10 CFR 
50.34(b)(6)(ii).
    (xviii) Certification of NDE personnel. (A) Level I and II 
nondestructive examination personnel shall be recertified on a 3-year 
interval in lieu of the 5-year interval specified in the 1997 Addenda 
and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 
1999 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(2) of this section.
    (B) Paragraph IWA-2316 of the 1998 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section, may only be used to qualify personnel that observe for 
leakage during system leakage and hydrostatic tests conducted in 
accordance with IWA-5211(a) and (b), 1998 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(2) of 
this section.
    (C) When qualifying visual examination personnel for VT-3 visual 
examinations under paragraph IWA-2317 of the 1998 Edition through the 
latest edition and addenda incorporated by reference in paragraph 
(b)(2) of this section, the proficiency of the training must be 
demonstrated by administering an initial qualification examination and 
administering subsequent examinations on a 3-year interval.
    (xix) Substitution of alternative methods. The provisions for the 
substitution of alternative examination methods, a combination of 
methods, or newly developed techniques in the 1997 Addenda of IWA-2240 
must be applied. The provisions in IWA-2240, 1998 Edition through the 
latest edition and addenda incorporated by reference in paragraph 
(b)(2) of this section, are not approved for use. The provisions in 
IWA-4520(c), 1997 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (b)(2) of this section, allowing 
the substitution of alternative examination methods, a combination of 
methods, or newly developed techniques for the methods specified in the 
Construction Code are not approved for use.
    (xx) System leakage tests. When performing system leakage tests in 
accordance IWA-5213(a), 1997 Addenda through the latest edition and 
addenda incorporated by reference in paragraph (b)(2) of this section, 
a 10-minute hold time after attaining test pressure is required for 
Class 2 and Class 3 components that are not in use during normal 
operating conditions, and 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.
    (xxi) Table IWB-2500-1 examination requirements. (A) The provisions 
of Table IWB-2500-1, Examination Category B-D, Full Penetration Welded 
Nozzles in Vessels, Items B3.40 and B3.60 (Inspection Program A) and 
Items B3.120 and B3.140 (Inspection Program B) 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, may be performed in place of an 
ultrasonic examination.
    (B) The provisions of Table IWB-2500-1, Examination Category B-G-2, 
Item B7.80, that are in the 1995 Edition are applicable only to reused 
bolting when using the 1997 Addenda through the latest edition and 
addenda incorporated by reference in paragraph (b)(2) of this section.
    (C) The provisions of Table IWB-2500-1, Examination Category B-K, 
Item B10.10, of the 1995 Addenda must be applied when using the 1997 
Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(2) of this section.
    (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 2000 Addenda subject to the 
following limitations and modifications:
* * * * *
    (ii) Motor-Operated Valve testing. Licensees shall comply with the 
provisions for testing motor-operated valves in OM Code ISTC 4.2, 1995 
Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (b)(3) of this section, and shall establish a program to 
ensure that motor-operated valves continue to be capable of performing 
their design basis safety functions.
    (iii) Code Case OMN-1. As an alternative to paragraph (b)(3)(ii) of 
this section, licensees may use Code Case OMN-1, ``Alternative Rules 
for Preservice and Inservice Testing of Certain Electric Motor-Operated 
Valve Assemblies in Light Water Reactor Power Plants,'' Revision 0, in

[[Page 60542]]

conjunction with ISTC 4.3, 1995 Edition with the 1996 and 1997 Addenda, 
or ISTC-3600, 1998 Edition through the latest edition and addenda 
incorporated by reference in paragraph (b)(3) of this section. 
Licensees choosing to apply the Code Case shall apply all of its 
provisions.
* * * * *
    (iv) Appendix II. Licensees applying Appendix II, ``Check Valve 
Condition Monitoring Program,'' of the OM Code, 1995 Edition with the 
1996 and 1997 Addenda, shall satisfy the requirements of paragraphs 
(b)(3)(iv)(A), (b)(3)(iv)(B), and (b)(3)(iv)(C) of this section. 
Licensees applying Appendix II, 1998 Edition through the latest edition 
and addenda incorporated by reference in paragraph (b)(3) of this 
section, shall satisfy the requirements of paragraphs (b)(3)(iv)(A), 
(b)(3)(iv)(B), and (b)(3)(iv)(D) of this section.
* * * * *
    (D) The provisions of ISTC-3510, ISTC-3520, and ISTC-3540 in 
addition to ISTC-5221 must be implemented if the Appendix II condition 
monitoring program is discontinued.
    (v) Subsection ISTD. Article IWF-5000, ``Inservice Inspection 
Requirements for Snubbers,'' of the ASME BPV Code, Section XI, provides 
inservice inspection requirements for examinations and tests of 
snubbers at nuclear power plants. Licensees may use Subsection ISTD, 
``Inservice Testing of Dynamic Restraints (Snubbers) in Light-Water 
Reactor Power Plants,'' ASME OM Code, 1995 Edition through the latest 
edition and addenda incorporated by reference in paragraph (b)(3) of 
this section, in place of the requirements for snubbers in Section XI, 
IWF-5200(a) and (b) and IWF-5300(a) and (b), by making appropriate 
changes to their technical specifications or licensee-controlled 
documents. Preservice and inservice examinations must be performed 
using the VT-3 visual examination method described in IWA-2213.
    (vi) Exercise interval for manual valves. Manual valves must be 
exercised on a 2-year interval rather that the 5-year interval 
specified in paragraph ISTC-3540 of the 1999 Addenda through the latest 
edition and addenda incorporated by reference in paragraph (b)(3) of 
this section, provided that adverse conditions do not require more 
frequent testing.
* * * * *
    (g) * * *
    (6) * * *
    (ii) * * *
    (B) Licensees do not have to submit to the NRC staff for approval 
of their containment inservice inspection programs which were developed 
to satisfy the requirements of Subsection IWE and Subsection IWL with 
specified modifications and limitations. The program elements and the 
required documentation must be maintained on site for audit.
    (C) * * *
    (1) Appendix VIII and the supplements to Appendix VIII to Section 
XI, Division 1, 1995 Edition with the 1996 Addenda of the ASME Boiler 
and Pressure Vessel Code must be implemented in accordance with the 
following schedule: Appendix VIII and Supplements 1, 2, 3, and 8--May 
22, 2000; Supplements 4 and 6--November 22, 2000; Supplement 11--
November 22, 2001; and Supplements 5, 7, and 10--November 22, 2002.
    (2) Licensees implementing the 1989 Edition and earlier editions 
and addenda of IWA-2232 of Section XI, Division 1, of the ASME Boiler 
and Pressure Vessel Code must implement the 1995 Edition with the 1996 
Addenda of Appendix VIII and the supplements to Appendix VIII of 
Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code.
* * * * *

    Dated at Rockville, Maryland this 9th day of September 2002.

    For the U.S. Nuclear Regulatory Commission.
William D. Travers,
Executive Director For Operations.
[FR Doc. 02-23811 Filed 9-25-02; 8:45 am]
BILLING CODE 7590-01-P