[Federal Register Volume 64, Number 183 (Wednesday, September 22, 1999)]
[Rules and Regulations]
[Pages 51370-51400]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-24256]



[[Page 51369]]

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Part II





Nuclear Regulatory Commission





_______________________________________________________________________



10 CFR Part 50



Industry Codes and Standards; Amended Requirements; Final Rule

  Federal Register / Vol. 64, No. 183 / Wednesday, September 22, 1999 / 
Rules and Regulations  

[[Page 51370]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AE26


Industry Codes and Standards; Amended Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

-----------------------------------------------------------------------

SUMMARY: The Nuclear Regulatory Commission is amending its regulations 
to incorporate by reference more recent editions and addenda of the 
ASME Boiler and Pressure Vessel Code and the ASME Code for Operation 
and Maintenance of Nuclear Power Plants for construction, inservice 
inspection, and inservice testing. These provisions provide updated 
rules for the construction of components of light-water-cooled nuclear 
power plants, and for the inservice inspection and inservice testing of 
those components. This final rule permits the use of improved methods 
for construction, inservice inspection, and inservice testing of 
nuclear power plant components.

DATES: Effective November 22, 1999. The incorporation by reference of 
certain publications listed in the regulations is approved by the 
Director of the Federal Register as of November 22, 1999.

FOR FURTHER INFORMATION CONTACT: Thomas G. Scarbrough, Division of 
Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-
2794, or Robert A. Hermann, Division of Engineering, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, Telephone: 301-415-2768.

SUPPLEMENTARY INFORMATION:

1.  Background
2.  Summary of Comments
2.1  List of Each Revision, Implementation Schedule, and Backfit 
Status
2.2  Discussion
2.3  120-Month Update
2.3.1  Section XI
2.3.1.1  Class 1, 2, and 3 Components, Including Supports
2.3.1.2  Limitations:
2.3.1.2.1  Engineering Judgment (Deleted)
2.3.1.2.2  Quality Assurance
2.3.1.2.3  Class 1 Piping
2.3.1.2.4  Class 2 Piping (Deleted)
2.3.1.2.5  Reconciliation of Quality Requirements
2.3.2  OM Code (120-Month Update)
2.3.2.1  Class 1, 2, and 3 Pumps and Valves
2.3.2.2  Background--OM Code
2.3.2.2.1  Comments on the OM Code
2.3.2.3  Clarification of Scope of Safety-Related Valves Subject to 
IST
2.3.2.4  Limitation:
2.3.2.4.1  Quality Assurance
2.3.2.5  Modification:
2.3.2.5.1  Motor-Operated Valve Stroke-Time Testing
2.4  Expedited Implementation
2.4.1  Appendix VIII
2.4.1.1  Modifications:
2.4.1.1.1  Appendix VIII Personnel Qualification
2.4.1.1.2  Appendix VIII Specimen Set and Qualification Requirements
2.4.1.1.3  Appendix VIII Single Side Ferritic Vessel and Piping and 
Stainless Steel Piping Examination
2.4.2  Generic Letter on Appendix VIII
2.4.3  Class 1 Piping Volumetric Examination (Deferred)
2.5  Voluntary Implementation
2.5.1  Section III
2.5.1.1  Limitations:
2.5.1.1.1  Engineering Judgement (Deleted)
2.5.1.1.2  Section III Materials
2.5.1.1.3  Weld Leg Dimensions
2.5.1.1.4  Seismic Design
2.5.1.1.5  Quality Assurance
2.5.1.1.6  Independence of Inspection
2.5.1.2  Modification:
2.5.1.2.1  Applicable Code Version for New Construction
2.5.2  Section XI (Voluntary Implementation)
2.5.2.1  Subsection IWE and Subsection IWL
2.5.2.2  Flaws in Class 3 Piping; Mechanical Clamping Devices
2.5.2.3  Application of Subparagraph IWB-3740, Appendix L
2.5.3  OM Code (Voluntary Implementation)
2.5.3.1  Code Case OMN-1
2.5.3.2  Appendix II
2.5.3.3  Subsection ISTD
2.5.3.4  Containment Isolation Valves
2.6  ASME Code Interpretations
2.7  Direction Setting Issue 13
2.8  Steam Generators
2.9  Future Revisions of Regulatory Guides Endorsing Code Cases
3.  Voluntary Consensus Standards
4.  Finding of No Significant Environmental Impact
5.  Paperwork Reduction Act Statement
6.  Regulatory Analysis
7.  Regulatory Flexibility Certification
8.  Backfit Analysis
9.  Small Business Regulatory Enforcement Fairness Act

1. Background

    The Nuclear Regulatory Commission (NRC) is amending its regulations 
to incorporate by reference the 1989 Addenda, 1990 Addenda, 1991 
Addenda, 1992 Edition, 1992 Addenda, 1993 Addenda, 1994 Addenda, 1995 
Edition, 1995 Addenda, and 1996 Addenda of Section III, Division 1, of 
the American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (BPV Code) with five limitations; the 1989 Addenda, 1990 
Addenda, 1991 Addenda, 1992 Edition, 1992 Addenda, 1993 Addenda, 1994 
Addenda, 1995 Edition, 1995 Addenda, and 1996 Addenda of Section XI, 
Division 1, of the ASME BPV Code with three limitations; and the 1995 
Edition and 1996 Addenda of the ASME Code for Operation and Maintenance 
of Nuclear Power Plants (OM Code) with one limitation and one 
modification. The final rule imposes an expedited implementation of 
performance demonstration methods for ultrasonic examination systems. 
The final rule permits the optional implementation of the ASME Code, 
Section XI, provisions for surface examinations of High Pressure Safety 
Injection Class 1 piping welds. The final rule also permits the use of 
evaluation criteria for temporary acceptance of flaws in ASME Code 
Class 3 piping (Code Case N-523-1); mechanical clamping devices for 
ASME Code Class 2 and 3 piping (Code Case N-513); the 1992 Edition 
including the 1992 Addenda of Subsections IWE and IWL in lieu of 
updating to the 1995 Edition and 1996 Addenda; alternative rules for 
preservice and inservice testing of certain motor-operated valve 
assemblies (OMN-1) in lieu of stroke-time testing; a check valve 
monitoring program in lieu of certain requirements in Subsection ISTC 
of the ASME OM Code (Appendix II to the OM Code); and guidance in 
Subsection ISTD of the OM Code as part of meeting the ISI requirements 
of Section XI for snubbers. This final rule deletes a previous 
modification for inservice testing of containment isolation valves.
    On December 3, 1997 (62 FR 63892), the NRC published a proposed 
rule in the Federal Register that presented an amendment to 10 CFR part 
50, ``Domestic Licensing of Production and Utilization Facilities,'' 
that would revise the requirements for construction, inservice 
inspection (ISI), and inservice testing (IST) of nuclear power plant 
components. For construction, the proposed amendment would have 
permitted the use of Section III, Division 1, of the ASME BPV Code, 
1989 Addenda through the 1996 Addenda, for Class 1, Class 2, and Class 
3 components with six proposed limitations and a modification.
    For ISI, the proposed amendment would have required licensees to 
implement Section XI, Division 1, of the ASME BPV Code, 1995 Edition up 
to and including the 1996 Addenda for Class 1, Class 2, and Class 3 
components with five proposed limitations. The proposed amendment 
included permission for licensees to implement Code Cases N-513, 
``Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 
Piping,'' and N-523, ``Mechanical Clamping Devices for Class 2 and 3 
Piping.'' The proposed

[[Page 51371]]

amendment also would allow licensees to use the 1992 Edition including 
the 1992 Addenda of Subsections IWE and IWL in lieu of updating to the 
1995 Edition and the 1996 Addenda. The proposed rule included expedited 
implementation of Appendix VIII, ``Performance Demonstration for 
Ultrasonic Examination Systems,'' to Section XI, Division 1, with three 
proposed modifications. An expedited examination schedule would also 
have been required for a proposed modification to Section XI which 
addresses volumetric examination of Class 1 high pressure safety 
injection (HPSI) piping systems in pressurized water reactors (PWRs).
    For IST, the proposed amendment would have required licensees to 
implement the 1995 Edition up to and including the 1996 Addenda of the 
ASME OM Code for Class 1, Class 2, and Class 3 pumps and valves with 
one limitation and one modification. The proposed amendment included 
permission for licensees to implement Code Case OMN-1 in lieu of 
stroke-time testing for motor-operated valves; Appendix II which 
provides a check valve condition monitoring program as an alternative 
to certain check valve testing requirements in Subsection ISTC of the 
OM Code; and Subsection ISTD of the OM Code as part of meeting the ISI 
requirements in Section XI for snubbers. Finally, the proposed rule 
would delete the modification presently in Sec. 50.55a(b) for IST of 
containment isolation valves.
    The NRC regulations currently require licensees to update their ISI 
and IST programs every 120 months to meet the version of Section XI 
incorporated by reference into 10 CFR 50.55a and in effect 12 months 
prior to the start of a new 120-month interval. The NRC published a 
supplement to the proposed rule on April 27, 1999 (64 FR 22580), that 
would eliminate the requirement for licensees to update their ISI and 
IST programs beyond a baseline edition and addenda of the ASME BPV 
Code. Under that proposed rule, licensees would continue to be allowed 
to update their ISI and IST programs on a voluntary basis to more 
recent editions and addenda of the ASME Code incorporated by reference 
in the regulations. Upon further review, the Commission decided to 
issue this final rule to incorporate by reference the 1995 Edition with 
the 1996 Addenda of the ASME BPV Code and the ASME OM Code with 
appropriate limitations and modifications. The Commission also decided 
to consider the proposal to eliminate the requirement to update ISI and 
IST programs every 120 months as a separate rulemaking effort. 
Following consideration of the public comments on the April 27, 1999, 
proposed rule, the NRC may prepare a final rule addressing the 
continued need for the requirement to update periodically ISI and IST 
programs and, if necessary, establishing an appropriate baseline 
edition of the ASME Code.

2. Summary of Comments

    Interested parties were invited to submit written comments for 
consideration on the proposed rule published on December 3, 1997. 
Comments were received from 65 separate sources on the proposed rule. 
These sources consisted of 27 utilities and service organizations, the 
Nuclear Energy Institute (NEI), the Nuclear Utility Backfitting and 
Reform Group (NUBARG) represented by the firm of Winston & Strawn, the 
ASME Board on Nuclear Codes and Standards, the Electric Power Research 
Institute (EPRI), the Performance Demonstration Initiative (PDI), the 
Nuclear Industry Check Valve Group, the State of Illinois Department of 
Nuclear Safety, Oak Ridge National Laboratory, the Southwest Research 
Institute, three consulting firms (one firm submitted three separate 
letters), and 24 individuals. The commenters' concerns related 
principally to one or more of the proposed limitations and 
modifications included in the proposed rule. Many of these limitations 
and modifications have been renumbered in the final rule because some 
limitations and modifications that were contained in the proposed rule 
were deleted.
    The proposed rule divided the proposed revisions to 10 CFR 50.55a 
into three groups based on the implementation schedule (i.e., 120-month 
update, expedited, and voluntary). These groupings have been retained 
in the discussion of the final rule. For each of these groups, it is 
indicated below in parentheses whether or not particular items are 
considered a backfit under 10 CFR 50.109 as discussed in Section 8, 
Backfit Analysis. This section provides a list of each revision and its 
implementation schedule, followed by a brief summary of the comments 
and their resolution. The summary and resolution of public comments and 
all of the verbatim comments which were received (grouped by subject 
area) are contained in Resolution of Public Comments. This document is 
available for inspection and copying for a fee in the NRC Public 
Document Room, 2120 L Street NW (Lower Level), Washington, DC.
2.1  List of Each Revision, Implementation Schedule, and Backfit 
Status.
  120-Month Update [in accordance with Secs. 50.55a(f)(4)(i) 
and 50.55a(g)(4)(i)]
  Section XI (Not A Backfit)
2.3.1.1  Class 1, 2, and 3 Components, Including Supports
2.3.1.2.1  Engineering Judgement (Deleted)
2.3.1.2.2  Quality Assurance
2.3.1.2.3  Class 1 Piping
2.3.1.2.4  Class 2 Piping (Deleted)
2.3.1.2.5  Reconciliation of Quality Requirements
  OM Code (Not A Backfit)
2.3.2.1  Class 1, 2, and 3 Pumps and Valves
2.3.2.3  Clarification of Scope of Safety-Related Valves Subject to IST
2.3.2.4.2  Quality Assurance
2.3.2.5.1  Motor-Operated Valve Stroke-Time Testing
 Expedited Implementation [after 6 months from the date of the 
final rule--Backfit]
2.4.1  Appendix VIII
2.4.1.1.1  Appendix VIII Personnel Qualification
2.4.1.1.2  Appendix VIII Specimen Set and Qualification Requirements
2.4.1.1.3  Appendix VIII Single Side Ferritic Vessel and Piping and 
Stainless Steel Piping Examination
2.4.3  Class 1 Piping Volumetric Examination (Deferred)
 Voluntary Implementation [may be used when final rule 
published--Not A Backfit]
 Section III
2.5.1.1.1  Engineering Judgement (Deleted)
2.5.1.1.2  Section III Materials
2.5.1.1.3  Weld Leg Dimensions
2.5.1.1.4  Seismic Design
2.5.1.1.5  Quality Assurance
2.5.1.1.6  Independence of Inspection
2.5.1.2.1  Applicable Code Version for New Construction
 Section XI
2.5.2.1  Subsection IWE and Subsection IWL
2.5.2.2  Flaws in Class 3 Piping; Mechanical Clamping Devices
2.5.2.3  Application of Subparagraph IWB-3740, Appendix L
 OM Code
2.5.3.1  Code Case OMN-1
2.5.3.2  Appendix II
2.5.3.3  Subsection ISTD
2.5.3.4  Containment Isolation Valves
2.2  Discussion
2.3  120-Month Update
2.3.1  Section XI
2.3.1.1  Class 1, 2, and 3 Components, Including Supports

    Section 50.55a(b)(2) endorses the 1995 Edition with the 1996 
Addenda of

[[Page 51372]]

Section XI, Division 1, for Class 1, Class 2, and Class 3 components 
and their supports. The proposed rule contained five limitations to 
address NRC positions on the use of Section XI: engineering judgment, 
quality assurance, Class 1 piping, Class 2 piping, and reconciliation 
of quality requirements. As a result of public comment, the NRC has 
reconsidered its positions on the use of engineering judgment and Class 
2 piping. These two limitations have been eliminated from the final 
rule. In addition, the NRC has modified the scope of the limitation 
related to reconciliation of quality requirements. A discussion of each 
of the five proposed limitations and their comment resolution follows.
2.3.1.  Limitations.
2.3.1.2.1  Engineering Judgement.
    The first proposed limitation to the implementation of Section XI 
(Sec. 50.55a(b)(2)(xi) in the proposed rule) addressed an NRC position 
with regard to the Foreword in the 1992 Addenda through the 1996 
Addenda of the BPV Code. That Foreword addresses the use of 
``engineering judgement'' for ISI activities not specifically 
considered by the Code. The December 3, 1997, proposed rule contained a 
limitation which would have specified that licensees receive NRC 
approval for those activities prior to implementation.
    Twenty-three commenters provided 30 separate comments on the 
proposed limitation to the use of engineering judgment with regard to 
Section XI activities. After reviewing the comments, it is apparent 
that the proposed rule did not accurately communicate the NRC's 
concerns with regard to the use of engineering judgment for Section XI 
activities. All of the commenters construed the limitation to prohibit 
the use of engineering judgment for all activities. The NRC understands 
that the use of engineering judgement is routinely exercised on a daily 
basis at each plant. It was not the NRC's intent to interject itself in 
this process by requiring prior approval as suggested by most 
commenters. The limitation was added to the proposed rule to address 
specific situations where engineering judgment was used and a 
regulatory requirement was not observed. Upon reconsideration of this 
issue and after reviewing all of the comments, the NRC has deleted this 
limitation from the final rule. The summary and the detailed 
discussions provided in the responses to the public comments should 
adequately address NRC concerns with regard to past applications of 
engineering judgment.
    The NRC acknowledges that the use of engineering judgment is a 
valid and necessary part of engineering activities. However, in 
applying such judgment, licensees must remain cognizant of the need to 
assure continued compliance with regulatory requirements. Specific 
examples of cases where application of engineering judgment resulted in 
failure to satisfy regulatory requirements are discussed in detail in 
the Response to Public Comments, Section 2.3.1.2.1, Engineering 
Judgment, and Section 2.6, ASME Code Interpretations. Questions were 
raised by the industry regarding Interpretations, the use of 
engineering judgment, and related enforcement actions. At NEI's 
request, the NRC staff met with NEI on January 11, 1995, to discuss the 
use of engineering judgment and Code interpretations. On November 12, 
1996, a meeting was held between representatives from the NRC and the 
ASME to discuss the same issues as well as the related enforcement 
actions. NRC Inspection Manual Part 9900, ``Technical Guidance,'' which 
had been developed in response to industry questions was also 
discussed. The ASME representatives agreed that the NRC guidance with 
respect to engineering judgment was consistent with their understanding 
of the relationship between the ASME Code and federal regulations. The 
ASME stated that the NRC should not establish a formal method for 
reviewing ASME Code interpretations. This position was based primarily 
on the understanding that it would be tantamount to NRC becoming the 
interpreter of the Code.
    It is apparent from the comments received on the proposed 
limitation that there is continuing confusion regarding the 
relationship between ASME Code requirements and NRC regulations. The 
NRC incorporates the ASME Code by reference into 10 CFR 50.55a. Upon 
adoption, the Code provisions become a part of NRC regulations as 
modified by other provisions in the regulations. Several commenters 
argued that a modification or limitation in the regulations cannot 
replace or overrule a Code provision or Interpretation. They also 
argued that, because the NRC did not accept all ASME Interpretations, 
the NRC was reinterpreting the Code. The NRC recognizes that the ASME 
is the official interpreter of the Code. However, only the NRC can 
determine whether the ASME Interpretation is acceptable such that it 
constitutes compliance with the NRC's regulations and does not 
adversely affect safety. The NRC cannot a priori approve Code 
Interpretations. While it is true that the ASME is the official 
interpreter of the Code, if the ASME interprets the Code in a manner 
which the NRC finds unacceptable (e.g., results in non-compliance with 
NRC regulatory requirements, a license condition, or technical 
specifications), the NRC can take exception to the Interpretation and 
is not bound by the ASME Interpretation. To put it another way, only 
the ASME can provide an Interpretation of the Code, but the NRC may 
make the determination whether that Interpretation constitutes 
compliance with NRC regulations. Hence, licensees need to consider the 
guidance on the use of Interpretations contained in the NRC Inspection 
Manual Part 9900, ``Technical Guidance.''
2.3.1.2.2  Quality Assurance.
    The second proposed limitation to the implementation of Section XI 
[Sec. 50.55a(b)(2)(xii) in the proposed rule] pertained to the use of 
ASME Standard NQA-1, ``Quality Assurance Requirements for Nuclear 
Facilities,'' with Section XI. Six comments were received and all were 
considered in arriving at the NRC's decision to retain the limitation 
as contained in the proposed rule. This limitation has been renumbered 
as Sec. 50.55a(b)(2)(x) in the final rule.
    As part of the licensing basis for nuclear power plants, NRC 
licensees have committed to certain quality assurance program 
provisions that are identified in both their Technical Specifications 
and Quality Assurance Programs. These provisions, as explained below, 
are taken from several sources (e.g., ASME, ANSI) and together, they 
constitute an acceptable Quality Assurance Program. The licensee 
quality assurance program commitments describe how the requirements of 
Appendix B, ``Quality Assurance Criteria for Nuclear Power Plants and 
Fuel Processing Plants,'' to 10 CFR part 50 will be satisfied by 
referencing applicable industry standards and the NRC Regulatory Guides 
(RGs) that endorsed the industry standards (e.g., the ANSI N45 series 
standards and applicable regulatory guides or NQA-1-1983 as endorsed by 
RG 1.28 (Revision 3), ``Quality Assurance Program Requirements (Design 
and Construction),'' and by prescriptive text contained in the program. 
Further, owners of operating nuclear power plants have committed to the 
additional operational phase quality assurance and administrative 
provisions contained in ANSI N18.7 as endorsed by RG 1.33, ``Quality 
Assurance Program Requirements (Operations).''

[[Page 51373]]

    Section XI references the use of either NQA-1 or the owner's 
Appendix B Quality Assurance Program (10 CFR part 50, Appendix B) as 
part of its individual provisions for a QA program. However, NQA-1 (any 
version) does not contain some of the quality assurance provisions and 
administrative controls governing operational phase activities that are 
contained in the ANSI standards as well as other documents which, as a 
group, constitute an acceptable program. When the NRC originally 
endorsed NQA-1, it did so with the knowledge that NQA-1 was not 
entirely adequate and must be supplemented by other commitments such as 
the ANSI standards. The later versions of NQA-1 also, by themselves, 
would not constitute an acceptable Quality Assurance Program. Hence, 
NQA-1 is not acceptable for use without the other quality assurance 
program provisions identified in Technical Specifications and licensee 
Quality Assurance Programs. The NRC staff has received questions 
regarding the relationship between commitments made relative to the 
Appendix B QA Program and Section XI as endorsed by 10 CFR 50.55a. It 
is apparent from public comments that there is confusion with regard to 
Section XI permitting the use of either NQA-1 or the owner's QA 
Program. The proposed limitation clarified that, when performing 
Section XI activities, licensees must meet other applicable NRC 
regulations. The limitation has been retained in the final rule to 
provide emphasis that licensees must comply with other applicable NRC 
regulations in addition to the quality assurance provisions contained 
in Section XI. As further clarification, the following discussion is 
provided.
    Although not discussed in the proposed amendment to 10 CFR 50.55a, 
the requirements of Secs. 50.34(b)(6)(ii) and 50.54(a) for establishing 
and revising QA Program descriptions during the operational phase are 
required to be followed and are not superseded or usurped by any of the 
requirements presently contained in 10 CFR 50.55a. Therefore, even 
though the present text of 10 CFR 50.55a does not take exception to 
applying the quality assurance provisions of NQA-1-1979 to ASME Section 
XI work activities, licensees of commercial nuclear power plants are 
required to comply not only with the QA provisions included in the 
Codes referenced in 10 CFR 50.55a, but also the quality assurance 
program developed to satisfy the requirements contained in 
Sec. 50.34(b)(6)(ii). This means that, regardless of the specific 
quality assurance controls delineated in Section XI as referenced in 10 
CFR 50.55a, licensees must meet the additional quality assurance 
provisions of their NRC approved quality assurance program description 
and other administrative controls governing operational phase 
activities.
2.3.1.2.3  Class 1 Piping.
    The third proposed limitation to the implementation of Section XI 
[Sec. 50.55a(b)(2)(xiii) in the proposed rule] pertained to the use of 
Section XI, IWB-1220, ``Components Exempt from Examination,'' that are 
contained in the 1989 Edition in lieu of the rules in the 1989 Addenda 
through the 1996 Addenda. Subparagraph IWB-1220 in these later Code 
addenda contain provisions from three Codes Cases: N-198-1, ``Exemption 
from Examination for ASME Class 1 and Class 2 Piping Located at 
Containment Penetrations;'' N-322, ``Examination Requirements for 
Integrally Welded or Forged Attachments to Class 1 Piping at 
Containment Penetrations;'' and N-334, ``Examination Requirements for 
Integrally Welded or Forged Attachments to Class 2 Piping at 
Containment Penetrations,'' which the NRC found to be unacceptable. The 
provisions of Code Case N-198-1 were determined by the NRC to be 
unacceptable because industry experience has shown that welds in 
service-sensitive boiling water reactor (BWR) stainless steel piping, 
many of which are located in containment penetrations, are subjected to 
an aggressive environment (BWR water at reactor operating temperatures) 
and will experience Intergranular Stress Corrosion Cracking. Exempting 
these welds from examination could result in conditions which reduce 
the required margins to failure to unacceptable levels. The provisions 
of Code Cases N-322 and N-334 were determined to be unacceptable 
because some important piping in PWRs and BWRs was exempted from 
inspection. Access difficulty was the basis in the Code cases for 
exempting these areas from examination. However, the NRC developed the 
break exclusion zone design and examination criteria utilized for most 
containment penetration piping expecting not only that Section XI 
inspections would be performed but that augmented inspections would be 
performed. These design and examination criteria are contained in 
Branch Technical Position MEB 3-1, an attachment of NRC Standard Review 
Plan 3.6.2, ``Determination of Rupture Locations and Dynamic Effects 
Associated with the Postulated Rupture of Piping.''
    Twenty-one comments were received on this limitation. Some 
commenters understood the bases for the limitation and did not believe 
that significant hardship would result. Many of the commenters argued 
that the Code cases were developed because these configurations are 
generally inaccessible and cannot be examined. Some argued that the 
piping in question is not safety significant and, thus, the 
examinations are unwarranted and the repairs which will be required are 
unnecessary.
    The NRC disagrees with these comments. The provisions of 
Sec. 50.55a(g)(2) require that facilities who received their 
construction permit on or after January 1, 1971, for Class 1 and 2 
systems be designed with provisions for access for preservice 
inspections and inservice inspections. Several early plants with 
limited access have been granted plant specific relief for certain 
configurations. These exemptions were granted on the basis that the 
examinations were impractical because these plants were not designed 
with access to these areas. Modifications to the plant would have been 
required at great expense to permit examination. Therefore, narrow 
exceptions were granted to these early plants. For later plants, 
however, Sec. 50.55a(g)(2) required that plants be constructed to 
provide access. The rationale for granting exemptions to early plants 
is not applicable to these later plants. In addition, there have been 
improvements in technology for the performance of examination using 
remote automated equipment. In designs where these welds are truly 
inaccessible, relief will continue to be granted when appropriate bases 
are provided by the licensee per Sec. 50.55a(g)(5). With regard to the 
safety significance of this piping, failure of Class 1 piping within a 
containment penetration may lead to loss of containment integrity and 
an unisolable pipe break. These areas were considered break exclusion 
zones as part of their initial design, in part, due to the augmented 
examinations performed on this portion of the piping system. Further, 
this issue could affect the large early release frequency (LERF). For 
these reasons, the limitation has been retained in the final rule 
(Sec. 50.55a(b)(2)(xi)) to require licensees to use the rules for IWB-
1220 that are contained in the 1989 Edition in lieu of the rules in the 
1989 Addenda through the 1996 Addenda.
2.3.1.2.4  Class 2 Piping.
    The fourth proposed limitation to the implementation of Section XI 
(Sec. 50.55a(b)(2)(xiv) in the proposed rule) would have confined 
implementation of

[[Page 51374]]

Section XI, IWC-1220, ``Components Exempt from Examination;'' IWC-1221, 
``Components Within RHR (Residual Heat Removal), ECC (Emergency Cool 
Cooling), and CHR (Containment Heat Removal) Systems or Portions of 
Systems;'' and IWC-1222, ``Components Within Systems or Portions of 
Systems Other Than RHR, ECC, and CHR Systems,'' to the 1989 Edition 
(i.e., it was determined that the 1989 Addenda through the 1996 Addenda 
were unacceptable). The provisions of Code Case N-408-3, ``Alternative 
Rules for Examination of Class 2 Piping,'' were incorporated into 
Subsection IWC in the 1989 Addenda. These provisions contain rules for 
determining which Class 2 components are subject to volumetric and 
surface examination. The NRC limitation on the use of the Code case and 
its revisions has consistently been that an ``applicant for an 
operating license should define the Class 2 piping subject to 
volumetric and surface examination in the Preservice Inspection for 
determination of acceptability by the NRC staff.'' Approval was 
required to ensure that safety significant components in the Residual 
Heat Removal, Emergency Core Cooling, and Containment Heat Removal 
systems are not exempted from appropriate examination requirements. The 
limitation in the proposed rule would have extended the approval 
required for preservice examination to inservice examination. Twenty 
comments were received, all disagreeing with the need for this 
limitation. Commenters pointed out that the information of interest is 
contained in the ISI program plan which is required by the Code to be 
submitted to the NRC. In addition, the intent of the limitation is 
current practice, and suitable controls are presently in place to 
ensure that adequate inspections of this piping are being performed. 
The NRC has reconsidered its bases for this limitation and agrees with 
the comments. Hence, the limitation has been eliminated from the final 
rule.
2.3.1.2.5  Reconciliation of Quality Requirements.
    The fifth proposed limitation to the implementation of Section XI 
(Sec. 50.55a(b)(2)(xx) in the proposed rule) addressed reconciliation 
of quality requirements when implementing Section XI, IWA-4200, 1995 
Addenda through the 1996 Addenda. Specifically, there were two 
provisions addressing the reconciliation of replacement items 
(Sec. 50.55a(b)(2)(xx)(A)) and the definition of Construction Code 
(Sec. 50.55a(b)(2)(xx)(B)). The limitation was included in the proposed 
rule to address the concern that, due to changes made to IWA-4200, 
``Items for Repair/Replacement Activities,'' in the 1995 Addenda, and 
IWA-9000, ``Glossary,'' definition of Construction Code in the 1993 
Addenda, a Section III component could be replaced with a non-Section 
III component, or that Construction Codes earlier than the Code of 
record might be used to procure components.
    Twelve comments were received on the limitation. Most of the 
commenters stated that the limitation was too extensive; i.e., rather 
than taking exception to Subparagraph IWA-4200, the limitation should 
specifically address Subparagraph IWA-4222, ``Reconciliation of Code 
and Owner's Requirements.'' Several comments suggested that the 
limitation be simplified to require only that ``Code items shall be 
procured with Appendix B requirements.'' Additional comments were 
provided relating to the need to remove the limitation on the 
definition of Construction Code, the use of the quality provisions 
contained in the Construction Code, and the historical provisions 
contained in Section XI for reconciling of technical requirements.
    The NRC has carefully reviewed the comments and agrees with the 
conclusions that: (1) A non-Section III item cannot be used to replace 
a Section III item; (2) only the same or later editions of the same 
Construction Code, or one that is higher in the evolutionary scale of 
the Code may be used; and (3) when using an earlier Construction Code, 
licensees must remain within the same Construction Code. The limitation 
has been revised in the final rule to address the reconciliation 
requirements contained in IWA-4222. However, changes to IWA-4222 in the 
1995 Addenda specifically exempt quality assurance requirements from 
the reconciliation process. The various changes implemented in the 1995 
Addenda, including the new definition of Construction Code, the 
identification of new Construction Codes, and the specific exemption to 
reconcile quality assurance requirements, could result in codes and 
standards being utilized which do not contain any quality assurance 
requirements, or contain quality assurance requirements which do not 
fully comply with Appendix B to 10 CFR part 50. Thus, the NRC has 
adopted the commenters' suggestion to clarify that Code items shall be 
procured in accordance with Appendix B requirements. Hence, when 
implementing the 1995 Addenda through the 1996 Addenda, the limitation 
(Sec. 50.55a(b)(2)(xvii) in the final rule) will require, in addition 
to the reconciliation provisions of IWA-4200, that the replacement 
items be purchased to the extent necessary to comply with the owner's 
quality assurance program description required by 10 CFR 
50.34(b)(6)(ii). The rewording of the limitation addresses the NRC's 
concerns with regard to definitions. That portion of the proposed 
limitation has been eliminated from the final rule.
2.3.2  OM Code (120-Month Update).
2.3.2.1  Class 1, 2, and 3 Pumps and Valves.
    This rule incorporates by reference for the first time into 10 CFR 
50.55a the ASME Code for Operation and Maintenance of Nuclear Power 
Plants (OM Code).
2.3.2.2  Background--OM Code.
    Until 1990, the ASME Code requirements addressing IST of pumps and 
valves were contained in Section XI, Subsections IWP (pumps) and IWV 
(valves). The provisions of Subsections IWP and IWV were last 
incorporated by reference into 10 CFR 50.55a in a final rulemaking 
published on August 6, 1992 (57 FR 34666). In 1990, the ASME published 
the initial edition of the OM Code which provides rules for IST of 
pumps and valves. The requirements contained in the 1990 Edition are 
identical to the requirements contained in the 1989 Edition of Section 
XI, Subsections IWP (pumps) and IWV (valves). Subsequent to the 
publication of the 1990 OM Code, the ASME Board on Nuclear Codes and 
Standards (BNCS) transferred responsibility for maintenance of these 
rules on IST from Section XI to the OM Committee. As such, the Section 
XI rules for inservice testing of pumps and valves that are presently 
incorporated by reference into NRC regulations are no longer being 
updated by Section XI.
    The 1990 Edition of the ASME OM Code consists of one section 
(Section IST) entitled ``Rules for Inservice Testing of Light-Water 
Reactor Power Plants.'' This section is divided into four subsections: 
ISTA, ``General Requirements,'' ISTB, ``Inservice Testing of Pumps in 
Light-Water Reactor Power Plants,'' ISTC, ``Inservice Testing of Valves 
in Light-Water Reactor Power Plants,'' and ISTD, ``Examination and 
Performance Testing of Nuclear Power Plant Dynamic Restraints 
(Snubbers).'' The testing of snubbers is governed by the ISI 
requirements of Section XI of the ASME BPV Code. Therefore, the rule 
only requires implementation of Subsections ISTA, ISTB, and ISTC. 
Because this final rule for the first time incorporates by reference 
the OM Code, the NRC has determined that the latest

[[Page 51375]]

endorsed Edition and Addenda of the OM Code (i.e., 1995 Edition up to 
and including the 1996 Addenda) should be used. Therefore, there is no 
need to incorporate by reference earlier Editions and Addenda of the OM 
Code (e.g., 1990 Edition or 1992 Edition).
2.3.2.2.1  Comments on the OM Code.
    There were four commenters addressing the proposed endorsement of 
the OM Code. The ASME BNCS (commenter one) agreed that the action was 
appropriate based on the ASME moving the responsibility for developing 
and maintaining IST program requirements from Section XI to the OM 
Code. A utility (commenter two) requested clarification as to when 
licensees would be required to begin using the 1995 Edition with the 
1996 Addenda for the OM Code. Licensees are presently required by 
Section XI to perform IST of pumps and valves. The regulations in 10 
CFR 50.55a currently require licensees to update their IST (and ISI) 
programs to the latest Code incorporated by reference in Sec. 50.55a(b) 
every 120 months. Hence, there is not a need to accelerate the 
transition to the OM Code.
    A utility (commenter three) stated that changes to the OM Code that 
appear in the 1995 Edition with the 1996 Addenda would require their 
facilities to modify the test loop piping for demonstrating pump design 
flow rate. The NRC is aware that some licensees may have difficulty 
fully implementing these tests and in certain cases, due to the 
impracticality of implementation, a request for relief under 
Sec. 50.55a(f)(5) would be appropriate. However, the OM committees 
developed these provisions in an effort to improve functional testing 
of pumps because present pump testing programs may not be capable of 
fully demonstrating that pumps are performing as designed. Some 
licensees have preoperational test loops which may be used to 
demonstrate full flow for this testing. Hence, the NRC has concluded 
that current regulatory requirements address this issue and a 
modification to the final rule in response to this comment is not 
required.
    The fourth commenter (an individual) stated that the NRC was 
primarily responsible for the changes in the 1994 Addenda (referred to 
as the Comprehensive Pump Test) which will result in additional pump 
testing. Further, the commenter believes that the changes were more the 
result of pressure by the NRC than actions determined prudent by the OM 
committees. Hence, the conclusion is drawn that, because the changes 
were not instituted exclusively by the OM committees, a backfit 
analysis is appropriate. With respect to the addition of the 
Comprehensive Pump Test, the OM Code committees had decided to pursue 
new approaches to pump testing for a long time before its actual 
development. In some cases, the changes resulted in less stringent 
requirements or in the deletion of certain requirements. The NRC staff 
raised concerns with certain changes and discussed these concerns with 
the ASME/OM representatives in ASME/OM committee meetings. As a result, 
the ASME/OM decided to develop an approach to pump testing that would 
include a nominal ``bump'' test (i.e., a more frequent, but less 
rigorous test) complemented by a biennial ``comprehensive'' test (i.e., 
a less frequent, but more rigorous test). Subsequent changes to the 
1990 OM Code were developed and adopted through a consensus process in 
which members of the nuclear industry are the primary participants. The 
NRC's position on the backfit issue is discussed in Section 8, Backfit 
Analysis, of the final rule, and in the response to public comments on 
the proposed rule. The NRC does not regard the development of the 
Comprehensive Pump Test to be an example of ``coercion'' by the NRC; 
rather it is an example of a properly functioning consensus process.
2.3.2.3  Clarification of Scope of Safety-Related Valves Subject to 
IST.
    The previous language in Sec. 50.55a(f)(1) had been interpreted by 
some licensees as a requirement to include all safety-related pumps and 
valves regardless of ASME Code Class (or equivalent) in the IST program 
of plants whose construction permits were issued before January 1, 
1971. The NRC proposed to revise this paragraph in the draft rule 
amendment to clarify which safety-related pumps and valves are 
addressed by 10 CFR 50.55a. The intent of the revision was to ensure 
that the IST scope of pumps and valves for these earlier-licensed 
plants was similar to the scope for plants licensed after January 1, 
1971. A corresponding revision was also proposed for Sec. 50.55a(g)(1) 
for ISI requirements.
    Fifteen separate commenters responded to the proposed clarification 
to Sec. 50.55a(f)(1). During consideration of their comments, it became 
apparent that the proposed language in Sec. 50.55a(f)(1) for IST did 
not fully accomplish its intended purpose. Instead of narrowing the IST 
scope of earlier-licensed plants to be consistent with the scope of 
later plants as intended, the proposed language inadvertently expanded 
the scope to include all pumps and valves in safety-related steam, 
water, air, and liquid-radioactive waste systems. The scope of pumps 
and valves to be included in IST should be dependent on the safety-
related function of the component rather than the function of the 
system. That is, a safety-related system might include many pumps and 
valves. However, not all of the pumps and valves might have a safety-
related function. For example, some valves in a safety-related system 
might be used for maintenance purposes only although they might be 
classified as safety-related because they are part of the safety-
related system pressure boundary. Accordingly, these valves would not 
need to be tested under the IST program, but the welds connecting the 
valve to the piping might be required to be examined under the ISI 
program. For this reason, the NRC further concluded that, unlike the 
scope issue that arose in Sec. 50.55a(f)(1) for IST, the scope issue 
did not apply to ISI, and a modification to the language of 
Sec. 50.55a(g)(1) pertaining to ISI is not appropriate. Therefore, the 
existing language of Sec. 50.55a(g)(1) will remain unchanged.
    However, the need to modify the language for IST requirements 
exists. The final rule revises Sec. 50.55a(f)(1) to ensure that the 
scope of inservice testing of pumps and valves in earlier plants is 
consistent with the scope applicable to later plants. This was 
accomplished by making the language of Sec. 50.55a(f)(1) consistent 
with the scope of Paragraph 1.1 in Subsections ISTB and ISTC of the OM 
Code. Hence, Sec. 50.55a(f)(1) in the final rule specifies that those 
pumps and valves that perform a specific function to shut down the 
reactor or maintain the reactor in a safe shutdown condition, mitigate 
the consequences of an accident, or provide overpressure protection for 
safety-related systems must meet the test requirements applicable to 
components which are classified as ASME Code Class 2 and Class 3 to the 
extent practical. The new language establishes the scope of pumps and 
valves that are to be included in an IST program based on the safety-
related function of the pump or valve. The requirements for pumps and 
valves that are part of the reactor coolant pressure boundary have not 
been changed. This change in the regulation will clarify the scope of 
IST for earlier-licensed plants resulting in a more consistent scope in 
pump and valve IST programs for all nuclear power plants.

[[Page 51376]]

2.3.2.4  Limitation.
2.3.2.4.1  Quality Assurance.
    The proposed rule contained one limitation (Sec. 50.55a(b)(3)(i)) 
to implementation of the OM Code addressing quality assurance (QA). 
This limitation pertained to the use of ASME Standard NQA-1, ``Quality 
Assurance Requirements for Nuclear Facilities,'' with the OM Code. 
Three comments were received and all were considered in arriving at the 
NRC's decision to retain the limitation as contained in the proposed 
rule.
    As part of the licensing basis for nuclear power plants, NRC 
licensees have committed to certain quality assurance program 
provisions which are identified in both their Technical Specifications 
and Quality Assurance Programs. These provisions are taken from several 
sources (e.g., ASME, ANSI) and together, they constitute an acceptable 
Quality Assurance Program. The licensee quality assurance program 
commitments describe how the requirements of appendix B to 10 CFR part 
50 will be satisfied by referencing applicable industry standards and 
the NRC Regulatory Guides (RGs) which endorsed the industry standards 
(e.g., the ANSI N45 series standards and applicable regulatory guides 
or NQA-1-1983 as endorsed by RG 1.28, Revision 3) and by prescriptive 
text contained in the program. Further, owners operating nuclear power 
plants have committed to the additional operational phase quality 
assurance and administrative provisions contained in ANSI N18.7 as 
endorsed by RG 1.33.
    The OM Code references the use of either NQA-1 or the owner's 
Appendix B Quality Assurance Program (10 CFR part 50, appendix B) as 
part of its individual provisions for a QA program. However, NQA-1 (any 
version) does not contain some of the quality assurance provisions and 
administrative controls governing operational phase activities which 
would be required in order to use NQA-1 in lieu of an owner's Appendix 
B QA Program Description. When the NRC originally endorsed NQA-1, it 
did so with the knowledge that NQA-1 was not entirely adequate and must 
be supplemented by other commitments such as the ANSI standards. The 
later versions of NQA-1 also, by themselves, would not constitute an 
acceptable Quality Assurance Program. Hence, NQA-1 is not acceptable 
for use without the other quality assurance program provisions 
identified in Technical Specifications and licensee Quality Assurance 
Programs. The NRC staff has received questions regarding the 
relationship between commitments made relative to the Appendix B QA 
Program and the proposed endorsement of the OM Code by 10 CFR 50.55a. 
It is apparent from the public comments that there is confusion with 
regard to the OM Code permitting the use of either NQA-1 or the owner's 
QA Program. The proposed limitation clarified that, when performing 
Section XI activities, licensees must meet other applicable NRC 
regulations. The limitation (Sec. 50.55a(b)(3)(i)) is retained in the 
final rule to provide emphasis that owners must comply with other 
applicable NRC regulations in addition to the quality provisions 
contained in the OM Code. The following discussion provides further 
clarification.
    Although not discussed in the proposed amendment to 10 CFR 50.55a, 
the requirements of Secs. 50.34(b)(6)(ii) and 50.54(a) for establishing 
and revising QA Program descriptions during the operational phase are 
required to be followed and are not superseded or usurped by any of the 
requirements presently contained in 10 CFR 50.55a. Therefore, even 
though the present text of 10 CFR 50.55a does not take exception to 
applying the quality provisions of NQA-1-1979 to ASME OM Code work 
activities, owners of commercial nuclear power plants are required to 
comply not only with the QA provisions included in the Codes referenced 
in 10 CFR 50.55a, but also the quality assurance program developed to 
satisfy the requirements contained in Sec. 50.34(b)(6)(ii). This means 
that, regardless of the specific quality assurance controls delineated 
in the OM Code as referenced in 10 CFR 50.55a, owners must meet the 
additional quality assurance provisions of their NRC approved quality 
assurance program description and other administrative controls 
governing operational phase activities.
2.3.2.5  Modification.
2.3.2.5.1  Motor-Operated Valve Stroke-Time Testing.
    The proposed rule contained a modification (Sec. 50.55a(b)(3)(ii)) 
pertaining to supplementing the stroke-time testing requirement of 
Subsection ISTC of the OM Code applicable for motor-operated valves 
(MOVs) with programs that licensees have previously committed to 
perform, prior to issuance of this amendment to 10 CFR 50.55a, for 
demonstrating the design-basis capability of MOVs. Stroke-time testing 
of MOVs is also specified in ASME Section XI. Seven commenters 
responded to the proposed change. The primary concern raised was that 
licensees would be required to comply with the provisions on stroke-
time testing in the OM Code as well as the programs developed under 
their licensing commitments for demonstrating MOV design-basis 
capability. This might result in a duplication of activities associated 
with inservice testing of safety-related MOVs and the periodic 
verification of the design-basis capability of safety-related MOVs at 
nuclear power plants.
    Since 1989, it has been recognized that the quarterly stroke-time 
testing requirements for MOVs in the Code are not sufficient to provide 
assurance of MOV operability under design-basis conditions. For 
example, in Generic Letter (GL) 89-10, ``Safety-Related Motor-Operated 
Valve Testing and Surveillance,'' the NRC stated that ASME Section XI 
testing alone is not sufficient to provide assurance of MOV operability 
under design-basis conditions. Therefore, in GL 89-10, the NRC staff 
requested licensees to verify the design-basis capability of their 
safety-related MOVs and to establish long-term MOV programs. The NRC 
subsequently issued GL 96-05, ``Periodic Verification of Design-Basis 
Capability of Safety-Related Motor-Operated Valves,'' to provide 
updated guidance for establishing long-term MOV programs. Licensees 
have made licensing commitments pursuant to GL 96-05 that are being 
reviewed by the NRC staff. Most licensees have voluntarily committed to 
participate in an industry-wide Joint Owners Group (JOG) Program on MOV 
Periodic Verification. This program will help provide consistency among 
the individual plant long-term MOV programs.
    At this time, the OM Code committees are working to update the Code 
with respect to its provisions for quarterly MOV stroke-time testing. 
For example, the ASME is considering incorporating Code Case OMN-1, 
``Alternative Rules for Preservice and Inservice Testing of Certain 
Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power 
Plants,'' into the OM Code. These provisions would allow users to 
replace quarterly MOV stroke-time testing with a combination of MOV 
exercising at least every refueling outage and MOV diagnostic testing 
on a longer interval. (The NRC has determined that, for MOVs, Code Case 
OMN-1 is acceptable in lieu of Subsection ISTC, with a modification. 
See Section 2.5.3.1 for further information.)
    In light of the present weakness in the information provided by 
quarterly MOV stroke-time testing, this modification has been retained 
in the final rule. However, the NRC agrees with the

[[Page 51377]]

public comment that the language in the proposed rule referring to 
licensing commitments was cumbersome and the language has been 
clarified. The final rule supplements the Code requirements for MOV 
stroke-time testing with a provision that licensees periodically verify 
MOV design-basis capability. The changes to Sec. 50.55a(b)(3)(ii) do 
not alter expectations regarding existing licensee commitments relating 
to MOV design-basis capability. Without being overly prescriptive, the 
final rule allows licensees to implement the regulatory requirements in 
a manner that best suits their particular application. The rulemaking 
does not require licensees to implement the JOG program on MOV periodic 
verification. The final rule in Sec. 50.55a(b)(3)(iii) allows licensees 
the option of using ASME Code Case OMN-1 to meet the requirements of 
Sec. 50.55a(b)(3)(ii).
2.4  Expedited Implementation.
2.4.1  Appendix VIII.
    The proposed rule contained a requirement 
(Sec. 50.55a(g)(6)(ii)(C)) that licensees expedite implementation of 
mandatory Appendix VIII, ``Performance Demonstration for Ultrasonic 
Examination Systems,'' to Section XI, 1995 Edition with the 1996 
Addenda. Three proposed modifications were included to address NRC 
positions on the use of Appendix VIII. The proposed rule would have 
required licensees to implement Appendix VIII for all examinations of 
the pressure vessel, piping, nozzles, and bolts and studs which occur 
after 6 months from the date of the final rule. The proposed rule would 
not have required any change to a licensee's ISI schedule for 
examination of these components, but would have required that the 
provisions of Appendix VIII be used for all examinations after that 
date.
    The 1989 Addenda to Section XI added mandatory Appendix VIII to 
enhance the requirements for performance demonstration for ultrasonic 
examination (UT) procedures. In 1991, the Performance Demonstration 
Initiative (PDI) was organized and funded. PDI is an organization of 
all U. S. nuclear utilities formed for the express purpose of 
developing efficient, cost-effective, and technically sound 
implementation of the performance demonstration requirements described 
in the ASME Code Section XI, Appendix VIII. The EPRI NDE Center 
provides technical support and administration for this program on 
behalf of the utilities. The PDI program has been evolving. Changes to 
the program were being made as difficulties in implementing some Code 
provisions were discovered. Other changes resulted when agreements were 
reached on issues such as training. Finally, the program has evolved as 
programs were developed for each Appendix VIII supplement.
    Sixty comments were received related to the proposed expedited 
implementation of Appendix VIII to Section XI. The issues raised by the 
commenters were generally uniform and narrow in scope; i.e., in 
agreement with the principles behind the development of Appendix VIII, 
but opposed to the manner in which the proposed rule would implement 
performance demonstration. In addition, commenters argued that 
implementation of Appendix VIII within 6 months from the date of the 
final rule was not possible because:
    (1) Some Appendix VIII supplements have not yet been implemented by 
PDI;
    (2) The number of qualified individuals is not yet sufficient;
    (3) The rule would require UT personnel to requalify; and
    (4) PDI's implementation of Appendix VIII differs from the Code.
    The NRC staff met four times with representatives from PDI, EPRI, 
and NEI between the dates of May 12, 1998, and November 19, 1998, to 
discuss items such as the current status of the PDI program, and 
Appendix VIII of Section XI as modified by PDI during the development 
of the program. Piping, bolting, and RPV samples, for the initial phase 
of the program, were completed in 1994. Procedure and personnel 
demonstrations were initiated in April of 1994. Since that time, a 
large number of personnel and procedures have been qualified. However, 
additional time and effort will be required to complete the industry 
qualification process for the remaining supplements of Appendix VIII.
    Subsequent to these meetings and consideration of the public 
comments, the NRC has reviewed the latest version of the PDI program 
for examination of vessels, piping, and bolting. The NRC agrees that 
this version will provide reasonable assurance of detecting the flaws 
of concern in ferritic vessels and piping. In addition, adoption in the 
final rule of Appendix VIII as modified by PDI during the development 
of the program means that the present test specimens are acceptable. 
The PDI program requires scanning the examination volume from both 
sides of the same surface of piping welds when it is accessible. 
Examinations performed from one side of a pipe weld may be conducted 
with procedures and personnel demonstrated at PDI; i.e., confirmed 
proficiency with single sided examinations. For the vessel weld, the 
volume must be examined in 4 directions from the clad-to-basemetal 
interface to a depth of 15 percent through-wall. Examinations performed 
from one side of a vessel weld may be conducted on the remaining 
portion of the weld volume provided the procedure shows the ability to 
detect flaws at angles up to 45 degrees from normal. In addition, to 
demonstrate equivalency to two sided examinations, the NRC staff and 
PDI agree that the demonstration be performed with specimens containing 
flaws with non-optimum sound energy reflecting characteristics or flaws 
similar to those in the vessel or pipe being examined. Because Appendix 
VIII supplements were designed for two-sided examinations, given the 
uniqueness in some instances of single side examinations, 
requalification may be necessary to demonstrate proficiency for these 
special cases. Single side examinations are not permitted for 15 
percent of the vessel volume adjacent to the cladding, and thus cannot 
be used for Supplement 4 performance demonstration.
    Evidence indicates that there are shortcomings in the 
qualifications of personnel and procedures in ensuring the reliability 
of nondestructive examination of the reactor vessel and other 
components of the reactor coolant system, the emergency core cooling 
systems, and portions of the steam and feedwater systems. Imposition of 
performance demonstration will greatly enhance the overall level of 
assurance of the reliability of ultrasonic examination techniques in 
detecting and sizing flaws. Hence, the final rule will expedite the 
implementation of these safety significant performance demonstration 
programs. The final rule will permit licensees to implement either 
Appendix VIII, ``Performance Demonstration for Ultrasonic Examination 
Systems,'' to Section XI, Division 1, 1995 Edition with the 1996 
Addenda, or Appendix VIII as executed by PDI. Because PDI is not a 
consensus standards body, its program document cannot be referenced in 
the final rule. Thus, the PDI requirements are directly contained in 
the final rule in Sec. 50.55a(b)(2)(xv).
    In Sec. 50.55a(g)(6)(ii)(C), the final rule incorporates a phased 
implementation of Appendix VIII over a three-year period. Licensees are 
required to implement the supplements to Appendix VIII according to the 
following schedule:
    (1) Six months after the effective date of the final rule: 
Supplement 1,

[[Page 51378]]

``Evaluating Electronic Characteristics of Ultrasonic Systems,'' 
Supplement 2, ``Qualification Requirements for Wrought Austenitic 
Piping Welds,'' Supplement 3, ``Qualification Requirements for Ferritic 
Piping Welds,'' and Supplement 8, ``Qualification Requirements for 
Bolts and Studs;''
    (2) One year after the effective date of the final rule: Supplement 
4, ``Qualification Requirements for the Clad/Base Metal Interface of 
Reactor Vessel,'' and Supplement 6, ``Qualification Requirements for 
Reactor Vessel Welds Other Than Clad/Base Metal Interface;''
    (3) Two years after the effective date of the final rule: 
Supplement 11, ``Qualification Requirements for Full Structural 
Overlaid Wrought Austenitic Piping Welds;'' and
    (4) Three years after the effective date of the final rule: 
Supplement 5, ``Qualification Requirements for Nozzle Inside Radius 
Section,'' Supplement 7, ``Qualification Requirements for Nozzle-to-
Vessel Weld,'' Supplement 10, ``Qualification Requirements for 
Dissimilar Metal Piping Welds,'' Supplement 12, ``Requirements for 
Coordinated Implementation of Selected Aspects of Supplements 2, 3, 10, 
and 11,'' and Supplement 13, ``Requirements for Coordinated 
Implementation of Selected Aspects of Supplements 4, 5, 6, and 7.''
    Performance demonstration requirements for Supplement 9, 
``Qualification Requirements for Cast Austenitic Piping Welds,'' have 
not yet been initiated pending completion of the other supplements. 
Hence, the final rule does not address Supplement 9.
    The final rule has been structured so that the equipment and 
procedures previously qualified under the PDI program are acceptable. 
Personnel previously qualified by PDI will remain qualified with the 
exception of a small population of individuals qualified for 
Supplements 4 and 6.
2.4.1.1  Modifications.
2.4.1.1.1  Appendix VIII Personnel Qualification.
    The first proposed modification of Appendix VIII 
(Sec. 50.55a(b)(2)(xvii) in the proposed rule) related to its 
requirement that ultrasonic examination personnel meet the requirements 
of Appendix VII, ``Qualification of Nondestructive Examination 
Personnel for Ultrasonic Examination,'' to Section XI. Appendix VII-
4240 contains a requirement for personnel to receive a minimum of 10 
hours of training on an annual basis. The NRC had determined that this 
requirement was inadequate for two reasons. The first reason was that 
the training does not require laboratory work and examination of flawed 
specimens. Signals can be difficult to interpret and, as detailed in 
the regulatory analysis for this rulemaking, experience and studies 
indicate that the examiner must practice on a frequent basis to 
maintain the capability for proper interpretation. The second reason is 
related to the length of training and its frequency. Studies have shown 
that an examiner's capability begins to diminish within approximately 6 
months if skills are not maintained. Thus, the NRC had determined that 
10 hours of annual training is not sufficient practice to maintain 
skills, and that an examiner must practice on a more frequent basis to 
maintain proper skill level. The modification in the proposed rule 
would have required 40 hours of annual training including laboratory 
work and examination of flawed specimens.
    Thirty-five comments were received on this proposed modification to 
Appendix VIII. Many of the commenters stated that 40 hours of required 
training were excessive because:
    (1) The EPRI NDE Center did not have the facilities which would be 
required to satisfy this requirement;
    (2) An ample supply of training specimens would cost each site 
$75,000; and
    (3) The requirement would result in administrative as well as cost 
burdens for both the utility and the vendor.
    Based on the public comments and the meetings with PDI and EPRI, 
the NRC has reconsidered its position. The PDI program has adopted a 
requirement for 8 hours of training, but it is required to be hands-on 
practice. In addition, the training must be taken no earlier than 6 
months prior to performing examinations at a licensee's facility. PDI 
believes that 8 hours will be acceptable relative to an examiner's 
abilities in this highly specialized skill area because personnel can 
gain knowledge of new developments, material failure modes, and other 
pertinent technical topics through other means. Thus, the NRC has 
decided to adopt in the final rule the PDI position on this matter. 
These changes are reflected in Sec. 50.55a(b)(2)(xiv) of the final 
rule.
2.4.1.1.2  Appendix VIII Specimen Set and Qualification Requirements.
    The second proposed modification of Appendix VIII 
(Sec. 50.55a(b)(2)(xviii) in the proposed rule) would have required 
that all flaws in the specimen sets used for performance demonstration 
for piping, vessels, and nozzles be cracks. For piping, Appendix VIII 
requires that all of the flaws in a specimen set be cracks. However, 
for vessels and nozzles, Appendix VIII would allow as many as 50 
percent of the flaws to be notches. The NRC had previously believed 
that, for the purpose of demonstrating nondestructive examination (NDE) 
capabilities, notches are not realistic representations of service 
induced cracks. The flaws in the specimen sets utilized for piping by 
EPRI for the PDI are all cracks.
    Thirty-two comments were received on this proposed modification to 
Appendix VIII. A majority of the commenters stated that this 
modification should be deleted from the rule because it would require 
the manufacture of new specimens and that the majority of procedure and 
examiner qualifications performed to date would be nullified. Many 
commenters argued that notches are realistic representations of cracks. 
Another comment was that fabrication defects should be permitted in 
order to test an examiner's ability to discriminate between real flaws 
and innocuous reflectors.
    The NRC believes that flaws in test specimens used for UT should be 
representative of the flaws normally found or expected to be found in 
operating plants. Based on the public comments, the final rule in 
Sec. 50.55a(b)(2)(xv) permits a population of notches and fabrication 
flaws on a limited basis for vessel and nozzle test specimen sets 
(Supplements 4, 5, 6, and 7). For these components, the NRC has 
concluded that a mix of cracks and notches is acceptable as long as 
they provide a similar detection and sizing challenge to that seen in 
actual service induced degradation. These types of notches will ensure 
that the qualification demonstration tests the ability of an examiner 
to discriminate between real flaws and innocuous reflectors. In 
addition, a mix of cracks and notches means that the present specimens 
can continue to be used for qualification. For wrought austenitic, 
ferritic, and dissimilar metal welds, however, these flaws can best be 
represented with cracks. Cracks span the ultrasonic spectra of flaw 
surface conditions from rough to smooth, jagged to straight, single to 
multiple tip, and tight to wide tip. Notches generally have smooth 
surfaces that reflect a narrow ultrasonic spectrum that represents a 
small population of flaws contained in components. Some variations in 
UT examination techniques may be more challenged with a notch located 
in specific locations, whereas other variations in UT examination 
techniques may not. With respect to

[[Page 51379]]

bolting, the NRC believed it would be clear that bolting was not 
addressed by the proposed modification. The NRC does not consider it 
necessary to use cracks for performance qualification for Supplement 8 
as notches are appropriate reflectors in the specimen test sets.
2.4.1.1.3  Appendix VIII Single Side Ferritic Vessel and Piping and 
Stainless Steel Piping Examination.
    The third proposed modification of Appendix VIII 
(Sec. 50.55a(b)(2)(xix) in the proposed rule) would have required that 
all specimens for single-side tests contain microstructures like the 
components to be inspected and flaws with non-optimum characteristics 
consistent with field experience that provide realistic challenges to 
the UT technique. The industry would have been required to develop 
specimen sets that contain microstructures similar to the types found 
in the components to be inspected and flaws with non-optimum 
characteristics (such as skew, tilt, and roughness) consistent with 
field experience that provide realistic challenges for single-sided 
performance demonstration. Appendix VIII does not distinguish specimens 
for two-sided examinations from those used for single-sided examination 
since Appendix VIII was originally developed using UT lessons learned 
from two-sided examinations of welds.
    Thirty comments were received on this proposed modification to 
Appendix VIII. Many commenters stated that the NRC should delete this 
modification because it would invalidate the current PDI test specimens 
and the procedures and examiners already qualified. Another prevalent 
comment was that the flaws being used by PDI in vessel and piping 
specimens represent the microstructure and flaw orientation of 
postulated in-service flaws in vessel welds and, therefore, ferritic 
vessels should be exempted from the proposed requirement.
    Based on the consideration of public comments, the final rule 
permits either Appendix VIII, as contained in the 1995 Edition with the 
1996 Addenda, or Appendix VIII, as modified by PDI during development 
of the program, to be implemented. The PDI program requirements are 
contained in Sec. 50.55a(b)(2)(xv). The NRC agrees that the latest 
version of the PDI program will provide reasonable assurance of 
detecting the flaws of concern in ferritic vessels and piping. In 
addition, adoption in the final rule of Appendix VIII as modified by 
PDI during the development of the PDI program means that the present 
test specimens are acceptable. The PDI program requires scanning the 
examination volume from both sides of the piping weld on the same 
surface when it is accessible. Examinations performed from one side of 
a vessel weld may be conducted with procedures and personnel 
demonstrated at PDI; i.e., confirmed proficiency with single sided 
examinations by a procedure that shows the ability to detect flaws at 
angles up to 45 degrees from the normal. The equipment, procedures, and 
personnel must demonstrate proficiency with single side examination. In 
addition, to demonstrate equivalency to two sided examinations, PDI 
requires that the demonstration be performed with specimens containing 
flaws with non-optimum sound energy reflecting characteristics or flaws 
similar to those in the ferritic vessel or pipe being examined. Because 
Appendix VIII supplements were designed for two-sided examinations, 
given the uniqueness in some instances of single side examinations, 
requalification may be necessary to demonstrate proficiency for these 
special cases. Single side examinations are not permitted for 15 
percent of the vessel volume adjacent to the cladding, and thus cannot 
be used for Supplement 4 performance demonstration.
    The final rule recognizes the difficulties of performance 
demonstration for two sided examination of austenitic stainless steel. 
However, PDI does not endorse single side inspection of austenitic 
welds because current technology cannot consistently satisfy Appendix 
VIII criteria. Thus, for certain situations, the final rule in 
Sec. 50.55a(b)(2)(xvi) contains criteria for demonstrating equivalency 
to two sided examinations.
    Single side examination of wrought-to-cast stainless steel is 
outside the scope of the current qualification program for austenitic 
piping. Current technology is not reliable for detecting flaws on the 
opposite side of wrought-to-cast stainless steel welds. Given these 
shortcomings, single side examination of stainless steel piping is 
considered ``best effort.'' The results of best-effort examination on 
the cast side of these welds is, in the NRC's view, marginal at best.
2.4.2  Generic Letter on Appendix VIII.
    The proposed rule contained a summary of a draft generic letter 
published in the Federal Register for public comment on December 31, 
1996 (61 FR 69120). The purpose of the generic letter was to alert the 
industry to the importance of using equipment, procedures, and 
examiners capable of reliably detecting and sizing flaws in the 
performance of comprehensive examinations of reactor vessels and 
piping. The NRC received 16 comment letters on the generic letter.
    Eighteen comments were received on the summary. Many of the 
comments reiterated comments submitted on Appendix VIII (i.e., Section 
2.4.1). Some commenters stated that the summary in the proposed rule 
inappropriately categorized and consolidated comments providing 
generalized responses to the industry's detailed comments. One 
commenter stated that an alternative to the proposed rule would be to 
mandate the use of PDI through a generic letter.
    The NRC disagrees with the characterization of its consideration of 
the comments submitted on the generic letter. The NRC thoroughly 
considered each comment. Commenters generally were not in agreement 
with the proposed NRC action and a determination was made to withdraw 
the generic letter pending rulemaking. Thus, the NRC's action to 
withdraw the generic letter was consistent with the commenters' 
recommendations. The summary of the comments in the Statement of 
Considerations for the proposed rule was not intended to provide a 
detailed response to every comment received on the generic letter. The 
purpose of the summary was to provide some history and background 
related to the proposed Appendix VIII action and to alert the industry 
that it was the NRC's intent to withdraw the generic letter. 
Implementation of Appendix VIII was included in the proposed and final 
rules partly as a result of public comment that a generic letter should 
not be used to mandate new examination requirements.
2.4.3  Class 1 Piping Volumetric Examination (Deferred).
    A proposed modification of Section XI (Sec. 50.55a(b)(2)(xv) in the 
proposed rule) would have required licensees of pressurized water 
reactor (PWR) plants to supplement the surface examination of Class 1 
High Pressure Safety Injection (HPSI) system piping as required by 
Examination Category B-J of Table IWB-2500-1 for nominal pipe sizes 
(NPS) between 4 (inches) and 1+ (inches), with a volumetric 
(ultrasonic) examination. This requirement was proposed because:
    (1) Inside diameter cracking of HPSI piping in the subject size 
range has been previously discovered (as detailed in NRC Generic Letter 
85-20, ``High Pressure Injection/Make-Up Nozzle Cracking in Babcock and 
Wilcox Plants,'' and in NRC Information Notice

[[Page 51380]]

97-46, ``Unisolable Crack in High-Pressure Injection Piping'');
    (2) Failure of this line could result in a small break loss of 
coolant accident while directly affecting the system designed to 
mitigate such an event;
    (3) Volumetric examinations are already required by the Code for 
Class 2 portions of this system (Table IWC-2500-1, Examination Category 
C-F-1) within the same NPS range; and
    (4) Surface examinations are not highly effective in identifying 
cracks and flaws in piping as evidenced by events at nuclear power 
plants and comparisons to other examination techniques.
    Implementation of this requirement was proposed to be performed 
during any ISI program inspection of the HPSI system performed after 6 
months from the date of the final rule. Using a licensee's existing ISI 
schedules would result in the volumetric examinations being implemented 
in a reasonable period of time while not impacting lengths of outages 
or requiring facility shutdown solely for performance of these 
examinations. In light of recent industry initiatives to address Class 
1 piping volumetric examination, the NRC is deferring rulemaking in 
this area at this time.
    Fifteen comments were received on this modification to Section XI. 
Several concerns were raised in the comments.
    (1) Volumetric examination of piping components in this size range 
is not very effective.
    (2) Given the general ineffectiveness of volumetric examination for 
this piping, the occupational exposure which would be incurred 
outweighs the perceived need.
    (3) The expedited implementation does not allow sufficient time to 
prepare specimen sets to comply with Appendix VIII.
    (4) There was no evidence that this problem would occur in all PWRs 
(i.e., the concern should be limited to Babcock & Wilcox (B&W) plants 
which have already addressed this problem).
    (5) The ASME Section XI Subcommittee on Inservice Inspection has 
initiated an action to address Class 1 piping.
    These five concerns are addressed in order below.
    As detailed in the regulatory analysis for the proposed rule, the 
initiation and propagation of pipe cracks at several plants have shown 
that surface examinations alone are not sufficient to detect the types 
of cracks which have occurred. It is agreed that these examinations for 
certain configurations may be difficult. The basic thermohydraulic 
phenomenon which caused the thermal fatigue cracking in the piping is 
well understood. However, current modeling limitations make it 
difficult to predict when this phenomenon will occur and at what 
locations. At this time, the most reliable means of detection is 
volumetric examination of the entire system in accordance with Section 
XI provisions for other Class 1 piping systems. In addition, experience 
has shown that, after initially discovering a section of degraded HPSI 
piping via leakage detection at one unit, it was possible to 
successfully identify similar degradation in the HPSI lines at sister 
units during subsequent ultrasonic examinations (in locations 
considered difficult to inspect). Therefore, it is the NRC's view that 
the usefulness of ultrasonic examinations in discovering thermal 
fatigue cracking in these lines has already been demonstrated in 
practice. Additionally, it is not clear to the NRC that the integrity 
of this piping can be assured in the presence of a through-wall flaw 
under all normal, emergency, upset, and faulted operating conditions 
for all PWR facilities. In short, the NRC does not believe that visual 
walkdowns should be the principal means of detecting leakage from pipes 
in these safety systems.
    The NRC is aware that the imposition of any additional inspections 
of the reactor coolant pressure boundary may result in additional cost 
and/or additional worker radiation exposure depending on the plant. 
Some units have already implemented these examinations in response to 
occurrences of thermal fatigue cracking at that unit. Given the safety 
significance of the HPSI system (i.e., failure of this line could 
result in a small break loss of coolant accident while directly 
affecting the system designed to mitigate such an event) and the number 
of failures reported to date (failures have occurred in the U.S. and 
several foreign countries), the NRC concludes that the burden 
associated with such examinations is minimal.
    The provisions of Appendix VIII are applicable to these 
examinations. The NRC staff has had several meetings with 
representatives from the industry's Performance Demonstration 
Initiative (PDI) group to discuss the status of the performance 
demonstration program. It is the NRC's understanding that the PDI 
program for piping is complete and can be implemented as soon as the 
administrative procedures have been developed.
    The NRC does not concur that the absence of piping failures for 
certain portions of the HPSI system in other reactor designs precludes 
the need for attention to this issue in those systems at those 
facilities. Thermal fatigue damage attributed to diverse initiating 
phenomena has been reported at several facilities in the U.S. and in 
Europe. As discussed, it is difficult to predict when and where this 
phenomenon might occur. Until data consistent with the failures that 
occurred are determined, and the thermohydraulic phenomenon which 
caused the failures is reproducible by analytical means, there is 
limited assurance that a given analytical method will provide a 
reliable assessment under all potential cyclic stratification 
circumstances, except in special cases where the technique is obviously 
conservative with respect to known data. At this time, the most 
reliable means of detection is volumetric examination.
    General Design Criterion (GDC) 14, ``Reactor coolant pressure 
boundary,'' of 10 CFR part 50, appendix A, or similar provisions in the 
licensing basis, requires that the reactor coolant pressure boundary 
(of which the unisolable portions of the HPSI system are a part) be 
tested so as to have an extremely low probability of abnormal leakage, 
of propagating failure, and of gross rupture. The ASME Section XI 
Subcommittee on Inservice Inspection is considering the need for 
volumetric examination of Class 1 HPSI systems. Further, the nuclear 
industry has initiated a voluntary effort being coordinated by the 
Nuclear Energy Institute to address the issue of thermal fatigue of 
nuclear power plant piping. The NRC has decided to defer regulatory 
action on the volumetric examination of Class 1 HPSI system piping 
while evaluating the industry initiative and determining the need for 
interim action during performance of the initiative. The NRC does not 
believe that deferral of regulatory action in this rulemaking while 
evaluating the need for interim action for HPSI Class 1 weld 
examinations will significantly affect plant safety, because staff 
evaluations indicate that a minimal increase in core damage frequency 
would result from potentially undiscovered flaws in HPSI Class 1 piping 
welds over this short time period. In light of the limited benefit of 
surface examinations of Class 1 HPSI system piping and concerns 
regarding occupational radiation exposure in the performance of those 
examinations, this rule in Sec. 50.55a(g)(4)(iii) endorses but does not 
mandate the provision in the ASME Code for surface weld examinations of 
Class 1 HPSI system piping.

[[Page 51381]]

2.5  Voluntary Implementation.
2.5.1  Section III.
    The proposed rule stated that the NRC had reviewed the 1989 
Addenda, 1990 Addenda, 1991 Addenda, 1992 Edition, 1992 Addenda, 1993 
Addenda, 1994 Addenda, 1995 Edition, 1995 Addenda, and 1996 Addenda of 
Section III, Division 1, for Class 1, Class 2, and Class 3 components, 
and had determined that they were acceptable for voluntary use with six 
proposed limitations. The final rule contains five limitations to the 
implementation of Section III. The proposed limitation on the use of 
engineering judgment during Section III activities has been deleted 
from the rule. In addition, the proposed rule stated that 10 CFR 50.55a 
would be modified to ensure consistency between 10 CFR 50.55a and NCA-
1140. The ASME initiated an action to address this issue and requested 
that the NRC delete this modification from the final rule. The NRC 
agrees in principle with the ASME action and has deleted the 
modification.
    The version of Section III utilized by applicants and licensees is 
established prior to construction as required by Sec. 50.55a(b), (c), 
and (d). For operating plants, Sec. 50.55a permits licensees to use the 
original construction code during the operational phase or voluntarily 
update to a later version which has been endorsed by 10 CFR 50.55a. 
Accordingly, the limitations to Section III apply to design and 
construction of new nuclear plants and become applicable to operating 
plants only if a licensee voluntarily updates to a later version.
2.5.1.1  Limitations.
2.5.1.1.1  Engineering Judgment (Deleted).
    The first proposed limitation to the implementation of Section III 
(Sec. 50.55a(b)(1)(i) in the proposed rule) addressed an NRC position 
with regard to the Foreword in the 1992 Addenda through the 1996 
Addenda of the ASME BPV Code. That Foreword addresses the use of 
``engineering judgement'' for ISI activities not specifically 
considered by the Code. The proposed rule would have required licensees 
to receive NRC approval for those activities prior to implementation.
    Twenty-three commenters provided 26 separate comments on the 
proposed limitation to the use of engineering judgment with regard to 
Section III activities. This proposed limitation has been dealt with in 
the same manner as the proposed limitation on the use of engineering 
judgment for Section XI activities. The NRC has deleted this limitation 
from the final rule as discussed in Section 2.3.1.2.1. The response to 
public comments in Section 2.3.1.2.1 addresses all of the comments 
which were received and provides specific examples of cases where 
application of engineering judgment resulted in failure to satisfy 
regulatory requirements.
2.5.1.1.2  Section III Materials.
    The second proposed limitation to the implementation of Section III 
(Sec. 50.55a(b)(1)(ii) in the proposed rule) pertained to a reference 
to Part D, ``Properties,'' of Section II, ``Materials.'' Section II, 
Part D, contained many printing errors in the 1992 Edition. These 
errors were corrected in the 1992 Addenda. The limitation would require 
that Section II, 1992 Addenda, be applied when using the 1992 Edition 
of Section III to ensure that the design stresses intended by the ASME 
Code are used.
    Four comments were received on the proposed limitation. One 
commenter agreed with the proposed action. The second commenter 
disagreed with the severity of the errors but had no objection to the 
proposed action. The third commenter stated that alerting users of the 
Code to such errors in a rulemaking was inappropriate. The fourth 
commenter argued that every version of Section II contains errors and 
that the NRC should recommend the use of the latest version because it 
contains the fewest number of errors. The limitation was not included 
in the proposed rule to initiate a debate over how conservative the 
errors were or whether the errors could cause faulty designs. There 
were over 160 Errata in the 1992 Edition (as identified in the 1992 
Addenda) apparently because of a printing error. By comparison, there 
were only 16 Errata in the 1993 Addenda. The NRC was simply attempting 
to alert users of the Code to that fact. This limitation has been 
retained in the final rule to ensure that these particular design 
stress tables will not be used. This limitation is contained in 
Sec. 50.55a(b)(1)(i) in the final rule.
2.5.1.1.3  Weld Leg Dimensions.
    The third proposed limitation to the implementation of Section III 
[Sec. 50.55a(b)(1)(iii) in the proposed rule] would correct a conflict 
in the design and construction requirements in Subsection NB (Class 1), 
Subsection NC (Class 2), and Subsection ND (Class 3) of Section III, 
1989 Addenda through the 1996 Addenda of the BPV Code. Two equations in 
NB-3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-
3673.2(b)-1 were modified in the 1989 Addenda and are no longer in 
agreement with Figures NB-4427-1, NC-4427-1, and ND-4427-1. This change 
results in a different weld leg dimension depending on whether the 
dimension is derived from the text or calculated from the figures. 
Thus, the proposed limitation was included to ensure consistency by 
specifying use of the 1989 Edition for the above referenced paragraphs 
and figures in lieu of the 1989 Addenda through the 1996 Addenda.
    Four comments were received on this proposed limitation. One 
commenter believed that the limitation was necessary. A second 
commenter believed that it was inappropriate to address Code errors in 
a rulemaking and this action should be accomplished through an 
information notice. The third commenter agreed that there appears to be 
a conflict, but they did not believe that the conflict would result in 
designs which do not satisfy the requirements and recommended deletion 
of the limitation. The fourth commenter stated that a conflict did not 
exist as a result of the changes made in the 1989 Addenda; i.e., the 
changes were deliberate to permit the designer an option on determining 
the proper weld size. However, this commenter did state that a printing 
error had been made in another change to the 1994 Addenda which has 
been corrected in the 1998 Edition.
    The NRC disagrees that the limitation should be deleted from the 
final rule. The weld size requirements that were used in the majority 
of U.S. operating nuclear power plant piping systems were provided by 
ANSI B31.7, Nuclear Power Piping Code, ANSI B31.1, Power Piping Code, 
and early editions of the ASME Code, Section III. Specifically, these 
standards required that the minimum socket weld size equal 1.25 t but 
not less than \1/8\ inch, where t is the nominal pipe wall thickness. 
The same weld size requirements as those specified in the above listed 
codes are also required by other nationally recognized codes and 
standards such as ANSI B31.3, Petroleum Refinery Piping Code. Those 
sizes were established as a result of many years of experience 
associated with the design and construction of piping systems, piping 
equipment, and components. In 1981, Code Case N-316, ``Alternative 
Rules for Fillet Weld Dimensions for Socket Welded Fittings,'' was 
published permitting a reduction in socket weld sizes to 1.09 t. In 
essence, the Code case was developed to provide relief for certain 
utilities having difficulty complying with the minimum socket weld size 
requirement of 1.25 t. The

[[Page 51382]]

provisions contained in the Code case were incorporated into the 1989 
Edition of the ASME Code. The NRC accepted this reduction because the 
new weld size was still greater than the pipe. In the 1989 Addenda of 
Section III of the ASME Code, the requirements for the size of socket 
welds were further reduced to 0.75 t which would permit welds smaller 
than the thickness of the pipe. The NRC is concerned with the 
structural integrity of a joint with a weld size which is less than the 
pipe wall thickness. The reduction to 0.75 t was not supported with 
test results or operating experience. Thus, a good technical basis has 
not been provided for reducing minimum socket weld sizes in nuclear 
power plant piping. It should be noted that the petrochemical industry 
has not made a corresponding change to the standards governing weld 
sizes in refinery piping. Hence, this limitation has been retained in 
Sec. 50.55a(b)(1)(ii).
2.5.1.1.4  Seismic Design.
    The fourth proposed limitation to the implementation of Section III 
(Sec. 50.55a(b)(1)(iv) in the proposed rule) pertained to new 
requirements for piping design evaluation contained in the 1994 Addenda 
through the 1996 Addenda of the ASME BPV Code. The NRC had determined 
that changes to articles NB-3200, ``Design by Analysis,'' NB-3600, 
``Piping Design,'' NC-3600, ``Piping Design,'' and ND-3600, ``Piping 
Design,'' of Section III for Class 1, 2, and 3 piping design evaluation 
for reversing dynamic loads (e.g., earthquake and other similar type 
dynamic loads which cycle about a mean value) were unacceptable. The 
new requirements are based, in part, on industry evaluations of the 
test data performed under sponsorship of the EPRI and the NRC. NRC 
evaluations of the data do not support the changes and indicate lower 
margins than those estimated in earlier evaluations. The ASME has 
established a special working group to reevaluate the bases for the 
seismic design for piping.
    Six comments were received on this proposed limitation to Section 
III. None of the commenters agreed with the proposed limitation and 
recommended its deletion from the final rule. The primary argument was 
that present seismic design of safety related piping is ``overly 
conservative both as it relates to the seismic capacity of structures 
which house or support such piping as well as the potential for a 
reduction in overall piping safety and reliability.'' Several 
commenters stated that, while it is true that there is an ongoing 
review within the ASME concerning the revised criteria, the data 
support the revised rules.
    An extensive discussion of this issue is provided in both the 
regulatory analysis and the response to public comments. In summary, in 
1993 prior to publication of the new ASME Code rules, the NRC initiated 
a research program at the U.S. Department of Energy (DOE) Energy 
Technology Engineering Center (ETEC) to evaluate the technical basis 
for the Code changes, and to assess the impact of the Code changes. In 
December 1994, the NRC informed the ASME that there were technical 
concerns regarding the new criteria, and the NRC would not endorse the 
criteria changes in the 1994 Addenda pending the results from the 
research program. By letter dated May 24, 1995, the NRC restated its 
technical concerns, and transmitted preliminary findings from those 
ETEC studies which had been completed to date along with the peer 
review comments. After receiving comments and input from other members 
of the ASME BPV Code as well as representatives from other countries, 
the ASME established a Special Working Group--Seismic Rule (SWG-SR) in 
September 1995 to assess the concerns identified by the NRC and others 
regarding the new piping design rules, and provide a proposed 
resolution to address these concerns.
    The ETEC efforts are now complete, and the results of the research 
indicate that the technical bases for the new piping design rules as 
published in the 1994 Addenda were incomplete. The results of the 
research are contained in NUREG/CR-5361, ``Seismic Analysis of 
Piping,'' which was published in May 1998. The SWG-SR is considering 
ETEC's recommendations and is conducting some additional studies.
    The NRC has concluded that additional technical bases need to be 
developed before the new rules could be found to be acceptable and will 
continue to interact via normal NRC staff participation with the Code 
committees. Thus, this limitation has been retained in 
Sec. 50.55a(b)(1)(iii). Licensees will be permitted to use articles NB-
3200, NB-3600, NC-3600, and ND-3600, in the 1989 Addenda through the 
1993 Addenda, but are prohibited from using these articles as contained 
in the 1994 Addenda through the 1996 Addenda.
2.5.1.1.5  Quality Assurance.
    The fifth proposed limitation to the implementation of Section III 
[Sec. 50.55a(b)(1)(v) in the proposed rule] pertained to the use of 
ASME Standard NQA-1, ``Quality Assurance Requirements for Nuclear 
Facilities.'' Section III references NQA-1 as part of its individual 
requirements for a QA program by integrating portions of NQA-1 into the 
QA program defined in NCA-4000, ``Quality Assurance,'' rather than 
permitting NQA-1 as a stand alone document similar to Section XI and 
the OM Code. Hence, even though NQA-1 by itself does not adequately 
describe how to satisfy the requirements of 10 CFR part 50, appendix B, 
the same concern does not exist regarding Section III and the use of 
NQA-1 as exists with Section XI. However, the limitation has been 
included in the final rule to provide consistency between the 
requirements of Section III, Section XI, and the OM Code, and to 
eliminate any possible confusion which could be created by not 
addressing the use of NQA-1 under each circumstance. The NRC had 
reviewed the requirements of NQA-1, 1986 Addenda through the 1992 
Addenda, that are part of the incorporation by reference of Section 
III, and had determined that the provisions of NQA-1 are acceptable for 
use in the context of Section III activities. Portions of NQA-1 are 
integrated into Section III administrative, quality, and technical 
provisions which provide a complete QA program for design and 
construction. The additional criteria contained in Section III, such as 
nuclear accreditation, audits, and third party inspection, establishes 
a complete program and satisfies the requirements of 10 CFR part 50, 
appendix B (i.e., the provisions of Section III integrated with NQA-1). 
Licensees may voluntarily choose to apply later provisions of Section 
III. Hence, a limitation was included in the proposed rule which would 
require that the edition and addenda of NQA-1 specified by NCA-4000 of 
Section III be used in conjunction with the administrative, quality, 
and technical provisions contained in the edition of Section III being 
utilized.
    Five comments were received on this proposed limitation. One 
commenter stated that the limitation was reasonable. The other 
commenters found the limitation confusing given that the NRC had 
determined that the provisions of NQA-1 were acceptable.
    Section III is a design and construction code used by the 
manufacturers and suppliers of new Code items. However, Section III is 
also used for controlling the construction of replacement Code items 
during the operational phase at nuclear power plants. The basis for the 
limitation in the proposed rule was that the quality provisions 
contained in NQA-1 (any version) are not adequate to describe how to 
satisfy the applicable 10 CFR

[[Page 51383]]

requirements for these activities. The NRC has not taken any exceptions 
to the quality or administrative provisions contained in Section III. 
However, in the proposed limitation for Section III, the NRC emphasized 
that the quality provisions of NQA-1 are acceptable for use in the 
context of Section III activities for the construction of new and 
replacement Code items. Therefore, the NRC has concluded that the 
quality provisions contained in Section III are acceptable for the 
construction of new and replacement items; i.e., NQA-1 is not adequate 
by itself. Thus, the limitation has been retained in 
Sec. 50.55a(b)(1)(iv).
2.5.1.1.6  Independence of Inspection.
    The sixth proposed limitation to the implementation of Section III 
[Sec. 50.55a(b)(1)(vi) in the proposed rule] related to prohibiting 
licensees from using subparagraph NCA-4134.10(a), ``Inspection,'' in 
the 1995 Edition through the 1996 Addenda. Before this edition and 
addenda, inspection personnel were prohibited from reporting directly 
to the immediate supervisors responsible for performing the work being 
inspected. However, in the 1995 Edition, NCA-4134.10(a) was modified so 
that independence of inspection was no longer required. This could 
result in noncompliance with Criterion I, ``Organization,'' of 10 CFR 
part 50, appendix B. This criterion requires that persons performing QA 
functions report to a management level such that authority and 
organizational freedom, including sufficient independence from cost and 
schedule when opposed to safety considerations, are provided.
    Four comments were received on this limitation. One commenter 
stated that the proposed limitation was reasonable. The second 
commenter stated that this position is consistent with NRC's previous 
positions. The third commenter stated the change in the Code provisions 
had been made because the previous Code requirements exceeded the 
requirements of appendix B. The fourth commenter stated that there has 
never been a provision in appendix B that prohibited inspectors from 
reporting to the supervisor responsible for the work being inspected.
    The NRC disagrees with both the third and fourth commenters. 
Criterion I, ``Organization,'' of 10 CFR part 50, appendix B requires 
the establishment and execution of a quality assurance program which 
includes establishing and delineating in writing the authority and 
duties of persons and organizations performing activities affecting the 
safety-related functions of structures, systems, and components. In 
particular, Criterion I states: ``These activities include both the 
performing functions of attaining quality objectives and the quality 
assurance functions. The quality assurance functions are those of (a) 
assuring that an appropriate quality assurance program is established 
and effectively executed and (b) verifying, such as by checking, 
auditing, and inspection, that activities affecting safety-related 
functions have been correctly performed.'' Criterion I continues by 
stating that ``[t]he persons and organizations performing quality 
assurance functions shall have sufficient authority and organizational 
freedom to identify quality problems; to initiate, recommend, or 
provide solutions; and to verify implementation of solutions. Such 
persons and organizations performing quality assurance functions shall 
report to a management level such that this required authority and 
organizational freedom, including sufficient independence from cost and 
schedule when opposed to safety considerations, are provided.'' 
Criterion X, ``Inspection,'' of Appendix B requires ``[s]uch inspection 
shall be performed by individuals other than those who performed the 
activity being inspected.''
    The requirements of 10 CFR part 50, appendix B could not be met for 
persons performing the quality function of inspection if those persons 
were reporting to the individual directly responsible for meeting cost, 
schedule, etc. (e.g., the requirement that personnel performing quality 
functions, such as inspection and auditing, shall have sufficient 
authority and organizational freedom to identify quality problems; to 
initiate, recommend, or provide solutions; and to verify implementation 
of solutions).
    As discussed in the first paragraph in this section, earlier 
versions of Section III contained a requirement for reporting 
independence. The requirement was contained in Supplement 10S-1, 
``Supplementary Requirements for Inspection.'' Supplement 10S-1, 
paragraph 2.1 states that, ``Inspection personnel shall not report 
directly to the immediate supervisors who are responsible for 
performing the work being inspected.'' The Code change substitutes the 
more general wording in Basic Requirement 1 that applies to the overall 
organization. Applying this general requirement for the more specific 
requirements applied to independence of inspectors could promote 
noncompliance with established licensee QA program commitments in the 
absence of compensating measures. Thus, the limitation has been 
retained in Sec. 50.55a(b)(1)(v). Licensees will be permitted to use 
the provisions contained in NCA-4134.10(a) in the 1989 Addenda through 
the 1994 Addenda, but will be prohibited from using these provisions as 
contained in the 1995 Edition through the 1996 Addenda.
2.5.1.2  Modification.
2.5.1.2.1  Applicable Code Version for New Construction.
    The modification of Section III contained in the proposed rule 
addressed a possible conflict between NCA-1140, ``Use of Code Editions, 
Addenda, and Cases,'' and 10 CFR 50.55a for new construction. NCA-1140 
of Section III requires that the length of time between the date of the 
edition and addenda used for new construction and the docket date of 
the construction permit application for a nuclear power plant be no 
greater than three years. Section 50.55a(b)(1) requires that the 
edition and addenda utilized be incorporated by reference into the 
regulations. The possibility exists that the edition and addenda 
required by the ASME Code to be used for new construction would not be 
incorporated by reference into 10 CFR 50.55a. In order to resolve this 
possible discrepancy, the NRC proposed to modify existing 
Secs. 50.55a(c)(3)(i), 50.55a(d)(2)(i), and 50.55a(e)(2)(i), to permit 
an applicant for a construction permit to use the latest edition and 
addenda which has been incorporated by reference into Sec. 50.55a(b)(1) 
if the requirements of the ASME Code and the regulations cannot 
simultaneously be satisfied.
    Three comments were received regarding this proposed modification 
to Section III. The ASME Board on Nuclear Codes and Standards (BNCS) 
agreed that there would be a conflict for new construction, but stated 
that the modification would preclude a Section III requirement for 
stamping. The BNCS recommendation was to delete this modification. The 
ASME is considering a Code case to resolve this by providing an 
alternative to NCA-1140(a)(2) which would allow an exception to this 
requirement when permitted by the enforcement authority. The NRC agrees 
with the suggested comment. The NRC, through its normal participation 
in the ASME committee process, will work with the appropriate ASME 
committees to provide an alternative when the requirements of the ASME 
Code and the regulations cannot simultaneously be satisfied. Hence, the 
proposed

[[Page 51384]]

modification has been deleted from the final rule.
2.5.2  Section XI (Voluntary Implementation).
    The proposed rule contained provisions intended to permit licensees 
to voluntarily implement specific portions of the Code. One provision 
related to Subsection IWE and Subsection IWL of the 1995 Edition with 
the 1996 Addenda. Another provision related to Code Case N-513, 
``Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 
Piping,'' and Code Case N-523-1, ``Mechanical Clamping Devices for 
Class 2 and 3 Piping.''
2.5.2.1  Subsection IWE and Subsection IWL.
    A final rule was published on August 8, 1996 (61 FR 41303), which 
incorporated by reference for the first time the 1992 Edition with the 
1992 Addenda of Subsection IWE, ``Requirements for Class MC and 
Metallic Liners of Class CC Components of Light-Water Cooled Power 
Plants,'' and Subsection IWL, ``Requirements for Class CC Concrete 
Components of Light-Water Cooled Power Plants.'' The final containment 
rule contained a requirement for licensees to develop and implement a 
containment ISI program within 5 years. Some licensees have begun the 
development of this program. However, other licensees have expressed an 
interest in using later versions of the Code for this program. During 
review of the 1995 Edition with the 1996 Addenda, the NRC determined 
that the provisions contained in Subsection IWE and Subsection IWL 
would be acceptable when used in conjunction with the modifications 
contained in the final rule published on August 8, 1996 (61 FR 41303). 
Thus, the proposed rule contained a provision [Sec. 50.55a(b)(2)(vi)] 
to permit licensees to implement either the presently required 1992 
Edition with the 1992 Addenda, or the 1995 Edition with the 1996 
Addenda.
    Twenty comments were received related to this provision. One 
commenter agreed with the action as proposed, and another did not 
object to the action but expressed a preference for the 1998 Edition. 
Three commenters stated that the NRC should give consideration to 
deferring action on this proposed amendment so that the 1998 Edition 
for containment ISI can be incorporated into this rulemaking. There are 
several provisions in Subsections IWE and IWL, 1992 Edition with the 
1992 Addenda, that licensees are finding cumbersome to implement. The 
commenters indicated that relief requests relative to these provisions 
will be submitted. Because these implementation difficulties have been 
addressed in the 1998 Edition, incorporation of the 1998 Edition would 
preclude the need to seek relief. Five commenters believe that the NRC 
did not perform the mandatory backfit analysis for the August 8, 1996 
(61 FR 41303), final rule; and, therefore, did not adequately justify 
its implementation. Further, the commenters believe that the NRC 
responses to the public comments were inadequately substantiated. Based 
on this, the comments argued that the proposed rule should be revised 
to make these subsections voluntary. Finally, one commenter believes 
that these subsections should be used on a trial basis before they are 
mandated.
    The NRC has made a determination to go forward with the final rule. 
Given the high priority of some of the items contained in the rule, 
deferral of the final rule to consider the 1998 Edition for containment 
ISI would result in an unacceptable delay. Approval of the 1998 Edition 
for containment ISI would involve not only review of Subsections IWE 
and IWL but review of the related Code requirements such as Subsection 
IWA, ``General Requirements,'' Section V, ``Nondestructive 
Examination,'' and Section IX, ``Welding and Brazing Qualifications.'' 
In addition, incorporation by reference of these additional Code 
requirements would result in the renoticing of the rule in the Federal 
Register for public comment. The NRC staff has met with NEI, EPRI, and 
utility representatives to discuss several industry concerns with 
regard to implementation of a containment ISI program. It is the NRC's 
understanding that these concerns can be addressed through the use of 
alternative examination requirements provided by an ASME Code case or 
the submittal of a relief request (e.g., some containment designs 
cannot meet Code access for examination requirements).
    The NRC performed the mandatory backfit analysis for the August 8, 
1996, rulemaking. Twelve commenters including NUBARG submitted comments 
on the documented evaluation which was performed in accordance with 
Sec. 50.109(a)(4). The industry developed examination rules for 
containments in response to a perceived need. The reported occurrences 
of containment degradation and the potential for additional serious 
occurrences was well documented in the final rule. No technical basis 
has been provided for the comment that this rule should be used to 
revise the implementation status of Subsections IWE and IWL from 
mandatory to voluntary. Therefore, the provision has not been changed 
in the final rule. However, the proposed provision 
(Sec. 50.55a(b)(2)(ix) in the proposed rule) containing supplemental 
requirements for the examination of concrete containments has been 
renumbered as Sec. 50.55a(b)(2)(viii) in the final rule. The proposed 
provision (Sec. 50.55a(b)(2)(x) in the proposed rule) containing 
supplemental requirements for the examination of metal containments and 
liners of concrete containments has been renumbered as 
Sec. 50.55a(b)(2)(ix) in the final rule.
    As licensees have begun developing their containment ISI programs, 
the NRC has received requests to clarify the implementation schedule 
for ISI of concrete containments and their post-tensioning systems. The 
current wording of Sec. 50.55a(g)(6)(ii)(B)(2) requiring licensees to 
implement ``the inservice examinations which correspond to the number 
of years of operation which are specified in Subsection IWL'' has 
created confusion regarding whether the first examination of concrete 
is required to meet the examination schedule in Section XI, Subsection 
IWL, IWL-2410, which is based on the date of the Structural Integrity 
Test (SIT), or may be performed at any time between September 9, 1996, 
and September 9, 2001. In addition, the examination schedule for post-
tensioning systems relative to the examination schedule for concrete 
was not clear. According to Sec. 50.55a(g)(6)(ii)(B)(2) of the final 
rulemaking of August 8, 1996, the first examination of concrete may be 
performed at any time between September 9, 1996, and September 9, 2001. 
The intent of the rule was that, for operating plants, the date of the 
first examination of concrete not be linked to the date of the SIT. The 
first examination of concrete will set the schedule for subsequent 
concrete examinations. With regard to examination of the post-
tensioning system, operating plants are to maintain their present 5-
year schedule as they transition to Subsection IWL. For operating 
reactors, there is no need to repeat the 1, 3, 5-year implementation 
cycle.
    Section 50.55a(g)(6)(ii)(B)(2) also stated that the first 
examination performed shall serve the same purpose for operating plants 
as the preservice examination specified for plants not yet in 
operation. The affected plants are presently operating, but they will 
be performing the examination of concrete under Subsection IWL for the 
first time.

[[Page 51385]]

Because the plants are operating, a Section XI preservice examination 
cannot be performed. Therefore, the first concrete examination is to be 
an inservice examination which will serve as the baseline (the same 
purpose for operating plants as the preservice examination specified 
for plants not yet in operation). With completion of this first 
examination of concrete, the second 5-year ISI interval would begin. 
Likewise, examinations of the post-tensioning system at the nth year 
(e.g., the 15th year post-tensioning system examination), if performed 
to the requirements of Subsection IWL, are to be performed to the ISI 
requirements, not the preservice requirements.
    The NRC has also been requested to clarify the schedule for future 
examinations of concrete and their post-tensioning systems at both 
operating and new plants. There is no requirement in Subsection IWL to 
perform the examination of the concrete and the examination of the 
post-tensioning system at the same time. The examination of the 
concrete under Subsection IWL and the examination of the liner plates 
of concrete containments under Subsection IWE may be performed at any 
time during the 5-year expedited implementation. This examination of 
the concrete and liner plate provides the baseline for comparison with 
future containment ISI. Coordination of these schedules in future 
examinations is left to each licensee. New plants would be required to 
follow all of the provisions contained in Subsection IWL, i.e., satisfy 
the preservice examination requirements and adopt the 1, 3, 5-year 
examination schedule linked to the Structural Integrity Test. The final 
rule has been clarified in Sec. 50.55a(g)(6)(ii)(B)(2) with respect to 
the examination schedules.
    The NRC has also received a request to clarify 
Sec. 50.55a(g)(4)(v)(C) regarding the replacement requirements of 
Subsection IWL-7000 for concrete and the post-tensioning systems. 
Section 50.55a(g)(4)(v)(A) and (B) each state the inservice inspection, 
repair, and replacement requirements must be met for metal containments 
and metallic shell and penetration liners, respectively. However, 
Sec. 50.55a(g)(4)(v)(C) states only that the inservice inspection and 
repair requirements applicable to concrete and the post-tensioning 
systems be met. This raised a question regarding whether the omission 
of the word ``replacement'' was intentional.
    The intent of the rule was to require implementation of all the 
Articles of Subsection IWL. The failure to include ``replacements'' was 
an oversight. Section 50.55a(g)(4) requires that ``* * * components 
which are classified as Class CC pressure retaining components and 
their integral attachments must meet the requirements, except for 
design and access provisions and preservice examination requirements, 
set forth in Section XI of the ASME Boiler and Pressure Vessel Code and 
Addenda that are incorporated by reference in paragraph (b).'' Section 
50.55a(g)(4)(v)(C) has been clarified in this final rule by including 
``replacement'' in order to eliminate any further confusion.
2.5.2.2  Flaws in Class 3 Piping.
    Section 50.55a(b)(2)(xvi) in the proposed rule pertained to use of 
ASME Code Case N-513, ``Evaluation Criteria for Temporary Acceptance of 
Flaws in Class 3 Piping,'' and Code Case N-523-1, ``Mechanical Clamping 
Devices for Class 2 and 3 Piping.'' These Code cases were developed to 
address criteria for temporary acceptance of flaws (including through-
wall leaking) of moderate energy Class 3 piping where a Section XI Code 
repair may be impractical for a flaw detected during plant operation 
(i.e., a plant shutdown would be required to perform the Code repair). 
In the past, licensees had to request NRC staff approval to defer 
Section XI Code repair for these Class 3 moderate energy (200  deg.F, 
275 psig) piping systems. The NRC had determined that Code Case N-513 
is acceptable except for the scope and Section 4.0. Code Case N-523-1 
is acceptable without limitation. When using Code Case N-523-1, it 
should be noted that the Code case erroneously references Table NC-
3321-2, rather than Table NC-3321-1 for pressure-retaining clamping 
devices designed by stress analysis. The use of Code Case N-513, with 
the limitations, and Code Case N-523-1 will obviate the need for 
licensees to request approval for deferring repairs; thus saving NRC 
and licensee resources.
    Section 1.0(a) of the Scope to Code Case N-513 limits the use of 
the requirements to Class 3 piping. However, Section 1.0(c) would allow 
the flaw evaluation criteria to be applied to all sizes of ferritic 
steel and austenitic stainless steel pipe and tube. Without some 
limitation on the scope of the Code case, the flaw evaluation criteria 
could be applied to components such as pumps and valves, and pressure 
boundary leakage; applications for which the criteria should not be 
utilized. Thus, paragraph (B) of the proposed provision limited the use 
of Code Case N-513 to those applications for which it was developed.
    The first paragraph of Section 4.0 of Code Case N-513 contains the 
flaw acceptance criteria. The criteria provide a safety margin based on 
service loading conditions. The second paragraph of Section 4.0, 
however, would permit a reduction of the safety factors based on a 
detailed engineering evaluation. Criteria and guidance are not provided 
for justifying a reduction, or limiting the amount of reduction. The 
NRC had determined that this provision was unacceptable because the 
second paragraph could permit available margins to become unacceptably 
low. Hence, Sec. 50.55a(b)(2)(xvi)(A) of the proposed provision 
required that, when implementing Code Case N-513, the specific safety 
factors in the first paragraph of Section 4.0 must be satisfied.
    There were seven commenters on the proposed use of these Code 
cases. One commenter agreed with the proposed action. Five commenters 
believed that the endorsement of these Code cases in a rulemaking is 
not appropriate. Five commenters disagreed with the limitations to Code 
Case N-513.
    The reason for incorporating the Code cases in the proposed rule 
was that Sec. 50.55a(g)(4) requires the application of Section XI 
during all phases of plant operation. Under Section XI structural and 
operability requirements, piping containing indications greater than 75 
percent of the pipe thickness are deemed unsatisfactory for continued 
service. A limitation must be included in the rulemaking to modify the 
above mentioned Section XI regulatory requirements. Because regulatory 
guides are not mandatory, inclusion of the Code cases in Regulatory 
Guide 1.147 would not modify the Section XI repair requirements. In 
addition, the preparation of these relief requests consumes 
considerable industry resources, and the review and issuance consume 
considerable NRC staff resources. Therefore, the NRC is implementing 
this limited use of these Code cases through the final rule.
    With regard to the limitations on the use of Code Case N-513, some 
commenters questioned the restrictions and believe that the Code case 
should be permitted in other applications such as socket welded 
connections. The Code case has been approved for use on moderate energy 
Class 3 piping and tubing (which is the ASME scope of the Code case). 
The NRC does not believe that the criteria are applicable to socket 
welds because NDE methods are not available for adequate flaw 
characterization. In addition, the NRC

[[Page 51386]]

does not agree that the level of reduction of safety margins which 
would be permitted by the Code case is appropriate. The margins 
available in an unflawed component are expected to be higher than for a 
degraded component. Margins less than the minimums specified for Level 
A, B, C, and D loading conditions are not acceptable. Hence, these 
restrictions have been maintained in the final rule except for the 
limitation related to original construction. The NRC agrees with 
commenters that any defects remaining from construction that have been 
determined by evaluation to be permissible are acceptable and has 
removed this limitation from the final rule. Code Cases N-513 and N-
523-1 are addressed in Sec. 50.55a(b)(2)(xiii) of the final rule.
2.5.2.3  Application of Subparagraph IWB-3740, Appendix L.
    Appendix L of Subparagraph IWB-3740 permits a licensee to 
demonstrate that a component is acceptable with regard to cumulative 
fatigue effects by performing a flaw tolerance evaluation of the 
component as an alternative to meeting the fatigue requirements of 
Section III. The NRC has reviewed Appendix L and determined that its 
use is generally acceptable. However, licensees should be aware of the 
following two items, which have been under consideration by certain 
ASME committees and may affect future revisions of Appendix L. The 
first item is that the assumption of a postulated flaw with a fixed 
aspect ratio of 6 may not be conservative depending on the extent of 
cumulative usage factor (CUF) criteria exceedance along the surface of 
the component. The assumption of a fixed aspect ratio can have an 
impact on crack growth rates and projected remaining fatigue life in a 
component. The second item pertains to the influence of environmental 
effects on both fatigue usage and crack growth evaluations in Appendix 
L. Environmental crack growth data from laboratory studies indicate the 
potential for a growth rate which is different from that currently 
reflected in a draft Section XI Code case which has been under ASME 
consideration. In addition, some environmental effects data on fatigue 
usage are available that may be considered for a revision to Section 
III.
2.5.3  OM Code (Voluntary Implementation).
    The proposed rule contained three provisions 
[Secs. 50.55a(b)(3)(iii), 50.55a(b)(3)(iv), and 50.55a(b)(3)(v)] 
pertaining to voluntary implementation of alternatives to specific OM 
Code requirements. The first provision involved implementation of ASME 
Code Case OMN-1, ``Alternative Rules for Preservice and Inservice 
Testing of Certain Electric Motor-Operated Valve Assemblies in Light-
Water Reactor Power Plants,'' in lieu of stroke time testing as 
required in Subsection ISTC, with a modification. The second provision 
involved implementation of a check valve condition monitoring program 
under Appendix II as an alternative to the testing or examination 
provisions contained in Subsection ISTC, with three modifications. The 
third provision involved use of Subsection ISTD to satisfy certain ISI 
requirements for snubbers provided in ASME BPV Code, Section XI. Each 
of these provisions is discussed separately below.
2.5.3.1  Code Case OMN-1.
    Section 50.55a(b)(3)(iii) of the proposed rule addressed the 
voluntary implementation of Code Case OMN-1 in lieu of stroke time 
testing as required for motor-operated valves (MOVs) in Subsection 
ISTC. In particular, Code Case OMN-1 permits licensees to replace 
quarterly stroke-time testing of MOVs with a program of exercising on 
intervals of one year or one refueling outage (whichever is longer) and 
diagnostic testing on longer intervals. As indicated in Attachment 1 to 
GL 96-05, the Code case meets the intent of the generic letter, but 
with certain limitations which were discussed in the generic letter. 
For MOVs, Code Case OMN-1 is acceptable in lieu of Subsection ISTC, 
except for leakage rate testing (ISTC 4.3) which must continue to be 
performed. In addition, OMN-1 contains a maximum MOV test interval of 
10 years, which the NRC supports. However, the NRC believed it prudent 
to include the modification requiring licensees to evaluate the 
information obtained for each MOV, during the first 5 years or three 
refueling outages (whichever is longer) of use of the Code case, to 
validate assumptions made in justifying a longer test interval. These 
conditions on the use of OMN-1 were included in the rule as a 
modification [Sec. 50.55a(b)(3)(iii)(A) in the final rule].
    Paragraph 3.7 of OMN-1 discusses the use of risk insights in 
implementing the provisions of the Code case such as those involving 
MOV grouping, acceptance criteria, exercising requirements, and testing 
frequency. For example, Paragraph 3.6.2 of OMN-1 states that exercising 
more frequently than once per refueling cycle shall be considered for 
MOVs with high risk significance. Since the proposed rule was issued, 
the NRC has reviewed plant-specific requests to use OMN-1 and has 
determined that a clarification of the rule is appropriate regarding 
the provision in the Code case for the consideration of risk insights 
if extending the exercising frequencies for MOVs with high risk 
significance beyond the quarterly frequency specified in the ASME Code. 
In particular, licensees should ensure that increases in core damage 
frequency and/or risk associated with the increased exercise interval 
for high-risk MOVs are small and consistent with the intent of the 
Commission's Safety Goal Policy Statement (51 FR 30028; August 21, 
1986). The NRC also considers it important for licensees to have 
sufficient information from the specific MOV, or similar MOVs, to 
demonstrate that exercising on a refueling outage frequency does not 
significantly affect component performance. The information may be 
obtained by grouping similar MOVs and staggering the exercising of MOVs 
in the group equally over the refueling interval. This clarification is 
provided in Sec. 50.55a(b)(3)(iii)(B) of the final rule.
    Thus, Code Case OMN-1 is acceptable as an optional alternative to 
MOV stroke-time test requirements with
    (1) The modification that, at 5 years or three refueling outages 
(whichever is longer) from initial implementation of Code Case OMN-1, 
the adequacy of the test interval for each MOV must be evaluated and 
adjusted as necessary; and
    (2) The clarification of the provision in OMN-1 for the 
establishment of exercise intervals for high risk MOVs in that the 
licensee will be expected to ensure that the potential increase in core 
damage frequency and risk associated with extending exercise intervals 
beyond a quarterly frequency is small and consistent with the intent of 
the Commission's Safety Goal Policy Statement.
    In addition, as noted in GL 96-05, licensees are cautioned that, 
when implementing Code Case OMN-1, the benefits of performing a 
particular test should be balanced against the potential adverse 
effects placed on the valves or systems caused by this testing. Code 
Case OMN-1 specifies that an IST program should consist of a mixture of 
static and dynamic testing. While there may be benefits to performing 
dynamic testing, there are also potential detriments to its use (i.e., 
valve damage). Licensees should be cognizant of this for each MOV when 
selecting the appropriate method or combination of methods for the IST 
program.
    Seven commenters responded to the proposed voluntary use of Code 
Case

[[Page 51387]]

OMN-1. All of the commenters agreed with the action to permit use of 
the Code case. However, four of the commenters did not believe that it 
was appropriate to do so in a rulemaking. Two commenters believe that 
the rule codifies individual licensee responses to Generic Letters 89-
10 and 96-05 which is unnecessary. Two commenters did not believe that 
the NRC had adequately justified limits on the test intervals.
    The proposed rule referenced Code Case OMN-1 as one method for 
developing a long-term MOV program that satisfies the recommendations 
of GL 96-05. This issue is closely related to Section 2.3.2.5.1. The 
amendment does not require the use of Code Case OMN-1. Licensees will 
be allowed the option of using the Code case as an alternative to the 
Code-required provisions for MOV stroke-time testing with the specified 
limitation and clarification. The voluntary use of Code Case OMN-1 by a 
licensee (in accordance with the rule and GL 96-05) would resolve 
weaknesses in the Code requirements for quarterly MOV stroke-time 
testing, and would also address the need to establish a long-term MOV 
program in response to GL 96-05.
    With regard to the concerns that the rule would require licensees 
to comply with the provisions on stroke-time testing in the OM Code and 
also with the programs developed under their licensing commitments for 
demonstrating MOV design-basis capability, it has been recognized since 
1989 that the quarterly stroke-time testing requirements for MOVs in 
the ASME Code are not sufficient to provide assurance of MOV 
operability under design-basis conditions. For example, in GL 89-10, 
the NRC stated that ASME BPV Code, Section XI, testing alone is not 
sufficient to provide assurance of MOV operability under design-basis 
conditions. Therefore, in GL 89-10, the NRC requested licensees to 
verify the design-basis capability of their safety-related MOVs and to 
establish long-term MOV programs. The NRC subsequently issued GL 96-05 
to provide updated guidance for establishing long-term MOV programs. 
However, the NRC agrees with the public comment that the language in 
the proposed rulemaking referring to licensing commitments is 
cumbersome. The paragraph has been revised in the final rule to be 
performance-based to focus on maintaining MOV design-basis capability.
    With regard to the question of limits on test intervals, the 
amendment does not limit the diagnostic test interval in Code Case OMN-
1 for MOVs to 5 years or three refueling outages. In endorsing the 
allowable use of Code Case OMN-1, the amendment states that the 
adequacy of the test interval for each MOV shall be evaluated and 
adjusted as necessary but not later than 5 years or three refueling 
outages (whichever is longer) from initial implementation of Code Case 
OMN-1. In other words, the amendment requires when applying Code Case 
OMN-1, prior to extending diagnostic test intervals for a specific MOV 
beyond 5 years (or three refueling outages), that the licensee evaluate 
test information on similar MOVs to ensure that the aging mechanisms 
are sufficiently understood such that the MOV will remain capable of 
performing its safety function over the entire diagnostic test 
interval. After evaluating the test information on similar MOVs, a 
licensee can extend the diagnostic test interval on other MOVs beyond 5 
years or three refueling outages up to 10-year limit specified in Code 
Case OMN-1.
2.5.3.2  Appendix II.
    Paragraph ISTC 4.5.5 of Subsection ISTC permits the owner to use 
Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM 
Code as an alternative to the testing or examination provisions of ISTC 
4.5.1 through ISTC 4.5.4. If an owner elects to use Appendix II, the 
provisions of Appendix II become mandatory per OM Code requirements. 
However, upon reviewing the appendix, the NRC determined that the 
requirements in Appendix II must be supplemented in three areas. The 
first area is testing or examination of the check valve obturator 
movement to both the open and closed positions to assess its condition 
and confirm acceptable valve performance. Bi-directional testing of 
check valves was approved by the ASME OM Main Committee for inclusion 
in the 1996 Addenda to the Code. The NRC agrees with the need for a 
required demonstration of bi-directional exercising movement of the 
check valve disc. Single direction flow testing of check valves, as an 
interpreted requirement, will not always detect degradation of the 
valve. The classic example of this faulty testing strategy is that the 
departure of the disc would not be detected during forward flow tests. 
The departed disc could be lying in the valve bottom or another part of 
the system, and could move to block flow or disable another valve. 
Although the ASME's Working Group on Check Valves (OM Part 22) is 
considering Code rules for bi-directional testing of check valves, 
Appendix II does not presently require it. Hence, the modification in 
Sec. 50.55a(b)(3)(iv)(A) was included so that an Appendix II condition 
monitoring program includes bi-directional testing of check valves to 
assess their condition and confirm acceptable valve performance (as is 
presently required by the OM Code).
    The second area needing supplementation is the length of test 
interval. Appendix II would permit a licensee to extend check valve 
test intervals without limit. Under the current check valve IST 
program, most valves are tested quarterly during plant operation. The 
interval for certain valves has been extended to refueling outages. The 
NRC has concluded that operating experience exists at this time to 
support longer test intervals for the condition monitoring concept. A 
policy of prudent and safe interval extension dictates that any 
additional interval extension must be limited to one fuel cycle, and 
this extension must be based on sufficient experience to justify the 
additional time. Condition monitoring and current experience may 
qualify some valves for an initial extension to every other fuel cycle, 
while trending and evaluation of the data may dictate that the testing 
interval for some valves be reduced. Extensions of IST intervals must 
consider plant safety and be supported by trending and evaluating both 
generic and plant-specific performance data to ensure the component is 
capable of performing its intended function over the entire IST 
interval. Thus, the modification (Sec. 50.55a(b)(3)(iv)(B)) limits the 
time between the initial test or examination and second test or 
examination to two fuel cycles or three years (whichever is longer), 
with additional extensions limited to one fuel cycle. The total 
interval is limited to a maximum of 10 years. An extension or reduction 
in the interval between tests or examinations would have to be 
supported by trending and evaluation of performance data.
    The third area in Appendix II which the NRC determined should be 
supplemented is the requirement applicable to a licensee who 
discontinues a condition monitoring program. A licensee who 
discontinues use of Appendix II, under Subsection ISTC 4.5.5, is 
required to return to the requirements of Subsection ISTC 4.5.4. 
However, the NRC has concluded that the requirements of ISTC 4.5.1 
through ISTC 4.5.4 must be also met. Hence, if the monitoring program 
is discontinued, the modification [Sec. 50.55a(b)(3)(iv)(C)] specifies 
that licensees implement the provisions of ISTC 4.5.1 through ISTC 
4.5.4.
    Thirty-four comments were received relative to the proposed 
voluntary implementation of Appendix II. There were seven comments 
supporting the

[[Page 51388]]

option to utilize the requirements of Appendix II. Most of the 
commenters did not agree with the limitations on the use of Appendix 
II. However, during its June 1997 meeting, the ASME's Working Group on 
Check Valves (OM Part 22) identified the following issues related to 
Condition Monitoring (as reported in the December 1, 1997, meeting 
minutes) that still needed to be resolved: consideration of safety 
significance; trending; interval limits; step-wise interval limits; and 
bi-directional testing. The proposed modifications addressed these 
issues. Based on its interaction with OM-22, the NRC believes the ASME 
will address these issues in future updates of the Code.
    Condition Monitoring, as described in Appendix II, is a program 
consisting of a general process without specified requirements, 
interval extension limits, and criteria. Condition Monitoring is a new 
Code approach with a promise of better detection of check valve 
degradation, improved valve performance, and maintaining reliable 
component capability over extended intervals, while adjusting test and 
examination intervals. The Condition Monitoring approach has not yet 
been implemented. Therefore, the nuclear industry lacks sufficient 
experience upon which to provide confidence of a uniform industry 
application of the process, or that equivalent requirements and 
interval extension limits will be applied, or assurance that components 
are capable of maintaining safe and reliable performance over extended 
intervals. Failure to ensure proper implementation of the process 
without specified requirements, interval extension limits, and criteria 
could result in inadvertent degradation in safety. Ensuring proper 
implementation could present an unwieldy compliance and inspection 
process for the NRC and licensees. The modifications to Appendix II 
contained in the rule provide for a safe and prudent progression of 
extending test and examination intervals consistent with historical 
experience and performance expectations. In addition, the modifications 
allow the licensee to conduct self-compliance inspections and minimize 
the expenditure of owner and NRC resources. Hence, the NRC has 
concluded that the modifications are justified and they have been 
retained in the final rule.
    The NRC considers the Condition Monitoring approach of Appendix II 
for check valves to be a significant improvement over present Code 
requirements, and encourages licensees to implement Appendix II. Where 
a licensee's Code of record is an earlier edition or addenda of the 
ASME Code, the regulations in Sec. 50.55a(f)(4)(iv) allow the licensee 
to implement portions of subsequent Code editions and addenda that are 
incorporated by reference in the regulations subject to the limitations 
and modifications listed in the rule, and subject to Commission 
approval. The NRC staff will favorably consider a request by a licensee 
under Sec. 50.55a(f)(4)(iv) to apply Appendix II, in advance of 
incorporating the 1995 Edition with the 1996 Addenda of the ASME OM 
Code as its Code of record, if the licensee justifies the following in 
its submitted request:
    (1) The modifications to Appendix II contained in the rule have 
been satisfied; and
    (2) All portions of the 1995 Edition with the 1996 Addenda of the 
OM Code that apply to check valves are implemented for the remaining 
check valves not included in the Appendix II program.
2.5.3.3  Subsection ISTD.
    Article IWF-5000, ``Inservice Inspection Requirements for 
Snubbers,'' of the ASME BPV Code, Section XI, 1996 Addenda, requires 
examinations and tests of snubbers at nuclear power plants as part of 
the licensee's ISI program in accordance with ASME/ANSI OM, Part 4. 
Some licensees control testing of snubbers through plant technical 
specifications. Although the ASME BPV Code, Section XI, establishes ISI 
requirements for examination and tests of snubbers, the ASME OM Code 
also provides guidance on snubber examination and testing in Subsection 
ISTD, ``Inservice Testing of Dynamic Restraints (Snubbers) in Light-
Water Reactor Power Plants.'' The proposed rule (Sec. 50.55a(b)(3)(v)) 
stated that licensees may use the guidance in Subsection ISTD, OM Code, 
1995 Edition with the 1996 Addenda, for testing snubbers. The final 
rule (Sec. 50.55a(b)(3)(v)) clarifies that Subsection ISTD, OM Code, 
1995 Edition, up to and including the 1996 Addenda may be used to meet 
certain ISI requirements for snubbers provided in IWF-5000 of the ASME 
BPV Code, Section XI. The licensee must still meet those requirements 
of IWF-5000, Section XI, not included in or addressed by Subsection 
ISTD. Consistent with IWF-5000, the rule specifies that preservice and 
inservice examinations must be performed using the VT-3 visual 
examination method in IWA-2213.
    Eleven comments were received on the endorsement of Subsection ISTD 
of the ASME OM Code. Seven commenters indicated that some owners have 
modified their Technical Specifications Snubber Surveillance 
Requirements to follow the provisions of GL 90-09, ``Alternative 
Requirements for Snubber Visual Inspection Intervals and Corrective 
Actions,'' to move the specific visual inspection and functional 
testing requirements to a Technical Requirements Manual. The NRC has 
addressed these comments in the final rule by referencing technical 
specifications or licensee-controlled documents for snubber test or 
examination requirements.
    One commenter noted that Article IWF-5000, Section XI, requires 
examination of snubbers be performed in accordance with ASME OM-1987, 
Part 4. Licensees of plants with a large number of snubbers have found 
the required visual inspection schedule in Part 4 to be excessively 
restrictive. As a result, some licensees have expended a significant 
amount of resources and have subjected plant personnel to unnecessary 
radiological exposure to comply with the visual examination 
requirements. Many licensees have been granted relief based on 
application of the snubber visual inspection intervals contained in GL 
90-09. The final rule allows licensees to use the snubber visual 
inspection interval contained in Table ISTD 6.5.2-1, ``Refueling 
Outage-Based Visual Examination Table,'' Subsection ISTD, OM Code, as 
an alternative to the Table in OM-1987, Part 4. Table ISTD 6.5.2-1 is 
substantially similar to the guidance provided in GL 90-09 for snubber 
visual inspection intervals. The final rule should help resolve the 
concerns regarding the visual inspection schedule in OM-1987, Part 4.
    Some commenters proposed Subsection ISTD as an acceptable 
alternative to the preservice and inservice examination requirements in 
IWF-5000, Section XI. The NRC has not accepted this suggestion because 
some preservice and inservice examinations for snubbers are not 
included in the OM Code. For example, Subsection ISTD does not address 
inspection of integral and non-integral attachments, such as lugs, 
bolting, pins, and clamps. Further, Subsection ISTD does not address 
snubbers in systems required to maintain the integrity of reactor 
coolant pressure boundary.
    Section 2.5.3.3, ``Subsection ISTD,'' of the Statement of 
Considerations for the proposed rule (62 FR 63903; December 3, 1997) 
stated that inservice testing of dynamic restraints or snubbers is 
governed by plant technical specifications and, thus, has never been 
included in 10 CFR 50.55a. It was apparent from comments received on

[[Page 51389]]

this section that this statement was confusing and needed to be 
clarified. First, it is true that 10 CFR 50.55a never directly required 
inservice testing of snubbers although the language in the current rule 
would appear to indicate otherwise. The language in the current rule 
states in Sec. 50.55a(f)(4), ``Throughout the service life of a boiling 
or pressurized water-cooled nuclear power facility, components 
(including supports) which are classified as ASME Code Class 1, Class 
2, and Class 3 must meet the requirements * * * set forth in section XI 
of editions of the ASME Boiler and Pressure Vessel Code and Addenda * * 
*'' (emphasis added). Although the language clearly states that 
``components (including supports)'' are within the scope of inservice 
testing, and it appears that inservice testing of snubbers is included 
under this statement, this statement was an editorial error. In the 
1992 final rule amending 10 CFR 50.55a to more clearly distinguish the 
requirements for inservice testing from those for inservice inspection 
(57 FR 34666; August 6, 1992), paragraph (g) was split into two 
separate paragraphs--paragraph (f) for inservice testing and paragraph 
(g) was retained for inservice inspection. In the 1992 final rule, 
similar requirements that applied to both inservice inspection and 
inservice testing were carried over from paragraph (f) to paragraph 
(g). The terminology, ``components (including supports),'' which 
existed in paragraph (g) was changed in paragraph (f) to read, ``pumps 
and valves,'' except in this one instance. Therefore, the Commission 
views this error as an editorial oversight. In the final rule, the 
language in paragraph (f)(4) has been corrected to read, ``pumps and 
valves,'' instead of ``components (including supports).''
    Based on this discussion, Sec. 50.55a never directly required 
inservice testing of snubbers. However, confusion resulted because some 
licensees interpreted this to mean that the NRC was implying that 
inservice testing of snubbers was never a regulatory requirement. 
Inservice testing of snubbers is a regulatory requirement and has been 
for many years. Section 50.55a(g)(4) requires that ASME Code Class 1, 
2, and 3 components (including supports) must meet the inservice 
inspection requirements of ASME Code, Section XI. Article IWF-5000 of 
Section XI, ``Inservice Inspection Requirements for Snubbers,'' 
provides requirements for the examination and testing of snubbers in 
nuclear power plants. Therefore, inservice testing of snubbers is 
required by 10 CFR 50.55a because it incorporates by reference Section 
XI requirements including Article IWF-5000. Inservice testing of 
snubbers has been a requirement in IWF-5000 since Subsection IWF was 
first issued in the Winter 1978 Addenda of the ASME Code, Section XI.
2.5.3.4  Containment Isolation Valves.
    The proposed rule contained a provision to delete the existing 
modification in Sec. 50.55a(b)(2)(vii) for IST of containment isolation 
valves (CIVs), which was added to the regulations in a rulemaking 
published on August 6, 1992 (57 FR 34666). That rulemaking incorporated 
by reference, among other things, the 1989 Edition of ASME Section XI, 
Subsection IWV that endorsed part 10 of ASME/ANSI OMa-1988 for valve 
inservice testing. A modification to the testing requirements of part 
10 related to CIVs was included in the rulemaking indicating that 
paragraphs 4.2.2.3(e) and 4.2.2.3(f) of part 10 were to be applied to 
CIVs. Since that time, the ASME OM Committee has performed a 
comprehensive review of OM Part 10 CIV testing requirements and 
acceptance standards, and has developed a basis document supporting 
removal of the requirements for analysis of leakage rates and 
corrective actions in Part 10 for those CIVs that do not provide a 
reactor coolant system pressure isolation function. The NRC reviewed 
this OM Committee basis document and determined that the modification 
addressing CIVs could be removed from the regulation. The requirements 
of 10 CFR part 50, Appendix J, ensure adequate identification analysis, 
and corrective actions for leakage monitoring of CIVs. There were four 
separate commenters on the proposed deletion of this modification and 
all were in agreement with the action. The final rule deletes this 
requirement.
2.6  ASME Code Interpretations.
    The ASME issues ``Interpretations'' to clarify provisions of the 
ASME BPV and OM Codes. Requests for interpretation are submitted by 
users and, after appropriate committee deliberations and balloting, 
responses are issued by the ASME. Generally, the NRC agrees with these 
interpretations. However, in a few cases interpretations have been 
issued which conflicted with or were inconsistent with NRC 
requirements. Following the guidance in these interpretations resulted 
in noncompliance with the regulations. Some cases were discussed 
earlier on engineering judgment. Additional discussion is provided on 
the use of interpretations in the Response to Public Comments. The 
proposed rule contained a discussion of NRC concerns related to ASME 
Code Interpretations, and referenced part 9900, Technical Guidance, of 
the NRC Inspection Manual. Part 9900 provides that licensees should 
exercise caution when applying Interpretations as they are not 
specifically part of the incorporation by reference into 10 CFR 50.55a 
and have not received NRC approval.
    Twenty-two comments were submitted by 21 separate commenters. 
Interpretations were also discussed in Sections 2.3.1.2.1 and 2.5.1.1.1 
as the use of engineering judgment and interpretations is intrinsically 
linked. Many of the commenters believe that the NRC position on ASME 
Code Interpretations is inconsistent. The NRC recognizes that the ASME 
is the official interpreter of the Code, but the NRC will not accept 
ASME interpretations that, in NRC's opinion, are contrary to NRC 
requirements or may adversely impact facility operations. It should be 
noted that, considering the large number of Code interpretations that 
are issued, there have been very few cases where the NRC has taken 
exception to an ASME interpretation. Interpretations have been of great 
benefit in clarifying the Code. The NRC is not restricting the use of 
ASME Code interpretations. A proposed limitation on their use was not 
placed in 10 CFR 50.55a; the discussion being limited to the Statement 
of Considerations. The purpose of the discussion was to merely alert 
Code users to be prudent when applying interpretations.
    As discussed in Section 2.3.1.2.1, a meeting was held on November 
12, 1996, between representatives from the ASME and the NRC (in part 
because of the continuing questions from the industry regarding ASME 
interpretations). The guidance given in NRC Inspection Manual, Part 
9900, regarding ASME Code interpretations was discussed. ASME 
representatives stated that the guidance is consistent with the ASME's 
understanding of the relationship between the ASME Code and NRC 
regulations. There were discussions regarding the mechanism for the NRC 
to inform the ASME of Code interpretations to which the NRC takes 
exception. It was agreed that the NRC should not establish a formal 
method for reviewing ASME Code interpretations for acceptance. This 
conclusion was based primarily on the understanding that it would be 
tantamount to the NRC becoming the interpreter of the Code. It was 
agreed that any concerns the NRC has regarding specific ASME Code 
interpretations would be brought to the ASME's attention through the 
NRC

[[Page 51390]]

staff's normal interaction with the Code. This has been routine 
practice for many years.
    Many commenters suggested that the NRC should adopt all 
interpretations because the ASME is the official interpreter of the 
Code. The NRC cannot a priori approve interpretations as suggested. 
This would delegate the NRC's statutory oversight responsibility to the 
ASME. In addition, the NRC cannot accept an interpretation when it 
conflicts with regulatory requirements. Finally, an interpretation may 
not be accepted that changes the requirements of the Code subsequent to 
the NRC endorsement of a particular edition or addenda in 10 CFR 
50.55a. Several commenters stated that the NRC should accept 
interpretations because, interpretations do not change the Code, they 
clarify it. As discussed in the responses to the public comments, there 
is evidence in a few cases to the contrary.
2.7  Direction Setting Issue 13.
    The proposed rule contained a discussion of issues under 
consideration relative to the Commission's endorsement of ASME Codes. 
The first item discussed was an October 21, 1993, Cost Beneficial 
Licensing Action (CBLA) submittal from Entergy Operations, Inc., 
requesting relief from the requirement to update ISI and IST programs 
to the latest ASME Code edition and addenda incorporated by reference 
into 10 CFR 50.55a. The underlying premise of the request was that a 
licensee should not be required to upgrade its ISI and IST programs 
without considering whether the costs of the upgrade are warranted in 
light of the increased safety afforded by the updated Code edition and 
addenda. The second item discussed was the National Technology Transfer 
and Advancement Act of 1995, Public Law 104-113. The Act directs 
Federal agencies to achieve greater reliance on technical standards 
developed by voluntary consensus standards development organizations. 
The third item was Direction Setting Issue (DSI) 13, which is part of 
an NRC Commission Strategic Assessment and Rebaselining Initiative. The 
Commission has directed the NRC staff to address how industry 
initiatives should be evaluated, and to evaluate several issues related 
to NRC endorsement of industry codes and standards. As part of this 
evaluation, the NRC staff is addressing issues relevant to the NRC's 
endorsement of the ASME Code, including periodic updating, the impact 
of 10 CFR 50.109 (the Backfit Rule), and streamlining the process for 
NRC review and endorsement of the ASME Code.
    Thirty-five comments were received from 21 commenters. Eight of the 
commenters supported NRC endorsement of the ASME Code, but submitted 
comments encouraging more timely endorsement. The Nuclear Energy 
Institute (NEI), the ASME Board on Nuclear Codes and Standards, and one 
utility requested that the NRC hold public meetings regarding the 
proposed rule. The reasons cited were: (1) Difficulties in implementing 
Appendix VIII as modified by the NRC; (2) concerns with the number of 
modifications and limitations and their content; and (3) licensee use 
of ASME Code editions later than 1989 should be voluntary and NRC staff 
endorsement need not be reflected in revisions to 10 CFR 50.55a.
    With regard to the comments related to difficulties in implementing 
Appendix VIII as modified by the NRC, as discussed under Section 2.4.1, 
the NRC staff met with representatives from PDI, EPRI, and NEI on May 
12, 1998, and again on June 18, 1998, to discuss items such as the 
current status of the PDI program, and Appendix VIII as modified during 
the development of the PDI program. The final rule endorses the latest 
version of Appendix VIII as modified by PDI during the development of 
the PDI program which, the NRC believes, satisfies the industry's 
concerns relative to this issue.
    Nine commenters stated that the modifications and limitations in 
the proposed rule violate or are contrary to the spirit of the National 
Technology Transfer and Advancement Act of 1995, Pub. L. 104-113, which 
codified OMB Circular A-119. However, the NRC disagrees that Pub. L. 
104-113 requires, without exception, the use of industry consensus 
standards. Section 12(d)(3) clearly allows agencies to decline to adopt 
voluntary consensus standards if they are inconsistent with applicable 
law or otherwise impractical. Furthermore, the Commission believes that 
it is in keeping with the intent of the Act if industry consensus 
standards are endorsed with limitations, rather than failing to endorse 
them in their entirety because of a few objectionable provisions. Ten 
commenters suggested that the modifications and limitations, in effect, 
reject the ASME consensus process. Some further suggested that many of 
the issues had not previously been brought to the ASME's attention. The 
NRC disagrees that the limitations and modifications exemplify NRC's 
failure to accept the consensus process of standards development. There 
are several examples, such as the new Section III piping seismic design 
criteria, which illustrate that the consensus process failed to 
consider the NRC representatives' comments that the bases for some of 
the criteria were flawed. This has been conclusively confirmed through 
additional testing performed by ETEC. Nearly all of the issues had 
previously been brought to the attention of committee members directly 
or as a result of public issuances such as NUREGs and generic 
communications.
    On April 27, 1999 (64 FR 22580), the NRC published a supplement to 
the proposed rule dated December 3, 1997 (63 FR 63892), that would 
eliminate the requirement for licensees to update their ISI and IST 
programs beyond a baseline edition and addenda of the ASME BPV Code. 
Under the proposed rule, licensees would continue to be allowed to 
update their ISI and IST programs to more recent editions and addenda 
of the ASME Code incorporated by reference in the regulations. In a 
Staff Requirements Memorandum dated June 24, 1999, the Commission 
directed the NRC staff to complete expeditiously the issuance of the 
final rule to incorporate by reference the 1995 Edition with the 1996 
Addenda of the ASME BPV Code and the ASME OM Code with appropriate 
limitations and modifications, and to consider the elimination of the 
requirement to update ISI and IST programs every 120 months as a 
separate rulemaking effort. The NRC is currently reviewing the public 
comments received on the proposed rule dated April 27, 1999. The NRC 
will indicate the decision regarding the need for periodic updating of 
ISI and IST programs and, if necessary, an appropriate baseline edition 
of the ASME Code following the review of public comments.
2.8  Steam Generators.
    ASME Code requirements for repair of heat exchanger tubes by 
sleeving were added to Section XI in the 1989 Addenda. This portion of 
the Code contains requirements for sleeving of heat exchanger tubes by 
several methods (e.g., explosion welding, fusion welding, expansion, 
etc.). The NRC has reviewed the Code requirements for sleeving and 
determined that they are acceptable. However, it should be recognized 
that, typically, there are other relevant requirements that need to be 
addressed for the application of sleeving to steam generator tubing. 
Some of the other requirements are as follows: periodic inservice 
inspections, repair of sleeves containing flaws exceeding the plugging 
limit (i.e., tube repair criteria), structural design and operational 
leakage limits. All of these sleeving requirements (ASME Code and

[[Page 51391]]

otherwise) would need to be addressed in the technical specifications 
sleeving license amendment request. Thus, the NRC determination that 
the ASME Code sleeving requirements are acceptable should be kept in 
perspective.
2.9  Future Revisions of Regulatory Guides Endorsing Code Cases.
    Section 50.55a indicates the ASME Code edition and addenda which 
have been approved for use by the NRC. In addition, Footnote 6 to 10 
CFR 50.55a references NRC Regulatory Guide 1.84, ``Design and Code Case 
Acceptability--ASME Section III Division 1,'' NRC Regulatory Guide 
1.85, ``Materials Code Case Acceptability--ASME Section III Division 
1,'' and NRC Regulatory Guide 1.147, ``Inservice Inspection Code Case 
Acceptability--ASME Section XI Division 1,'' which list the ASME Code 
cases that have been determined suitable by the NRC for use and may be 
applied to: (1) The design and construction of a particular component; 
or (2) the performance of inservice examination of systems and 
components. A determination has been made that the regulatory guide 
process must change in order to assure that the Code cases endorsed in 
the Regulatory Guides are incorporated by reference into the 
regulations and constitute legally-binding alternatives to the existing 
requirements in Sec. 50.55a. Draft Revision 31 to Regulatory Guide 
1.84, draft Revision 31 to Regulatory Guide 1.85, and draft Revision 12 
to Regulatory Guide 1.147 were published for public comment in May 
1997. The final regulatory guides were published in May 1999, in 
accordance with the present process. Future revisions to these 
regulatory guides, however, will be accompanied by rulemaking which 
will change the footnote reference to indicate the acceptable 
regulatory guide revisions, and to reflect approval for incorporation 
by reference of the endorsed Code cases by the Office of the Federal 
Register.

3. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, Pub. 
L. 104-113, requires that 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 otherwise 
impractical. In this final rule, the NRC is amending its regulations to 
incorporate by reference more recent editions and addenda of the ASME 
Boiler and Pressure Vessel Code and the ASME Code for Operation and 
Maintenance of Nuclear Power Plants for construction, inservice 
inspection, and inservice testing as identified in the SUPPLEMENTARY 
INFORMATION of this document.

4. Finding of No Significant Environmental Impact

    Based upon an environmental assessment, 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 will not have a significant effect on the quality of 
the human environment and therefore an environmental impact statement 
is not required.
    The final rule is one part of a regulatory framework directed to 
ensuring pressure boundary integrity and the operational readiness of 
pumps and valves. The final rule incorporates provisions contained in 
the ASME BPV Code and the OM Code for the construction, inservice 
inspection, and inservice testing of components used in nuclear power 
plants. These provisions have been updated to incorporate improved 
technology and methodology. Therefore, in the general sense, the final 
rule would have a positive impact on the environment.
    The final rule endorses ASME BPV Code, Section XI, 1995 Edition 
with the 1996 Addenda. As most of the technical changes to this 
edition/addenda merely incorporate improved technology and methodology, 
imposition of these requirements is not expected to either increase or 
decrease occupational exposure. However, imposition of paragraphs IWF-
2510, Table IWF-2500-1, Examination Category F-A, and IWF-2430, will 
result in fewer supports being examined which will decrease the 
occupational exposure compared to present support inspection plans. It 
is estimated that an examiner receives approximately 100 millirems for 
every 25 supports examined. Adoption of the new provisions is expected 
to decrease the total number of supports to be examined by 
approximately 115 per unit per interval. Thus, the reduction in 
occupational exposure is estimated to be 460 millirems per unit each 
inspection interval or 50.14 rems for 109 units.
    The final rule endorses the 1995 Edition with the 1996 Addenda of 
the ASME OM Code. The provisions of the OM Code are not expected to 
either increase or decrease occupational exposure. The types of testing 
associated with the 1995 Edition with the 1996 Addenda of the OM Code 
are essentially the same as the OM standards contained in the 1989 
Edition of Section XI referenced in a final rule published on August 6, 
1992 (57 FR 34666).
    Actions by applicants and licensees in response to the final rule 
are of the same nature as those applicants and licensees have been 
performing for many years. Therefore, this action should not increase 
the potential for a negative environmental impact.
    The Commission has determined, in accordance with the National 
Environmental Policy Act of 1969, as amended and the Commission's 
regulations in subpart A of 10 CFR part 51, that this rulemaking is not 
a major action significantly affecting the quality of the human 
environment, and, therefore, an environmental impact statement is not 
required. This final rule amends the NRC regulations pertaining to ISI 
and IST requirements for nuclear power plant components. The current 
regulations in 10 CFR 50.55a incorporates by reference the 1989 Edition 
of the ASME BPV Code, Section III, Division 1; the 1989 Edition of the 
ASME BPV Code, Section XI, Division 1, for Class 1, Class 2, and Class 
3 components; the 1992 Edition with the 1992 Addenda of the ASME BPV 
Code, Section XI, Division 1, for Class MC and Class CC components; and 
the 1989 Edition of the ASME BPV Code, Section XI, Division 1, for 
Class 1, Class 2, and Class 3 pumps and valves. The Commission is 
amending its regulations to incorporate by reference the 1989 Addenda, 
1990 Addenda, 1991 Addenda, 1992 Edition, 1992 Addenda, 1993 Addenda, 
1994 Addenda, 1995 Edition, 1995 Addenda, and 1996 Addenda of Section 
III, Division 1, of the ASME BPV Code with five limitations; the 1989 
Addenda, 1990 Addenda, 1991 Addenda, 1992 Edition, 1992 Addenda, 1993 
Addenda, 1994 Addenda, 1995 Edition, 1995 Addenda, and 1996 Addenda of 
Section XI, Division 1, of the ASME BPV Code with three limitations; 
and the 1995 Edition and 1996 Addenda of the ASME OM Code with one 
limitation and one modification. The final rule imposes an expedited 
implementation of performance demonstration methods for ultrasonic 
examination systems. The final rule permits the optional implementation 
of the ASME Code, Section XI, provisions for surface examinations of 
High Pressure Safety Injection Class 1 piping welds. The final rule 
also permits the use of evaluation criteria for temporary acceptance of 
flaws in ASME Code Class 3 piping (Code Case N-523-1); mechanical 
clamping devices for ASME Code Class 2 and 3 piping (Code Case N-513); 
the 1992 Edition including the 1992 Addenda of Subsections IWE and IWL

[[Page 51392]]

in lieu of updating to the 1995 Edition and 1996 Addenda; alternative 
rules for preservice and inservice testing of certain motor-operated 
valve assemblies (OMN-1) in lieu of stroke-time testing; a check valve 
monitoring program in lieu of certain requirements in Subsection ISTC 
of the ASME OM Code (Appendix II to the OM Code); and guidance in 
Subsection ISTD of the OM Code as part of meeting the ISI requirements 
of Section XI for snubbers. This final rule deletes a previous 
modification for inservice testing of containment isolation valves. The 
editions and addenda of the ASME BPV Code and OM Code incorporated by 
reference provide updated rules for the construction of components of 
light-water-cooled nuclear power plants, and for the inservice 
inspection and inservice testing of those components. This final rule 
permits the use of improved methods for construction, inservice 
inspection, and inservice testing of nuclear power plant components. 
For these reasons, the Commission concludes that this rule should have 
no significant adverse impact on the operation of any licensed facility 
or the environment surrounding these facilities.
    The conclusion of this environmental assessment is that there will 
be no significant offsite impact to the general public from this 
action. However, the general public should note that the NRC has also 
committed to comply with Executive Order (EO) 12898, ``Federal Actions 
to Address Environmental Justice in Minority Populations and Low-Income 
Populations,'' dated February 11, 1994, in all its actions. Therefore, 
the NRC has also determined that there is no disproportionately high 
adverse impacts on minority and low-income populations. In the letter 
and spirit of EO 12898, the NRC is requesting public comment on any 
environmental justice considerations or questions that the public 
thinks may be related to this final rule. The NRC uses the following 
working definition of ``environmental justice': the fair treatment and 
meaningful involvement of all people, regardless of race, ethnicity, 
culture, income, or education level with respect to the development, 
implementation, and enforcement of environmental laws, regulations, and 
policies. Comments on any aspect of the environmental assessment, 
including environmental justice may be submitted to the NRC.
    The NRC will send a copy of this final rule including the foregoing 
Environmental Assessment to every State Liaison Officer.
    The environmental assessment is available for inspection at the NRC 
Public Document Room, 2120 L Street NW (Lower Level), Washington, DC. 
Single copies of the environmental assessment are available from Thomas 
G. Scarbrough, Division of Engineering, Office of Nuclear Reactor 
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, Telephone: 301-415-2794, or Robert A. Hermann, Division of 
Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-
2768.

5. 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.
    The public reporting burden for this information collection is 
estimated to average 85 person-hours per response, including the time 
for reviewing instructions, searching existing data sources, gathering 
and maintaining the data needed, and completing and reviewing the 
collection of information.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a collection of information unless it displays a currently 
valid OMB control number.

6. Regulatory Analysis

    The Commission has prepared a regulatory analysis on this final 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by the Commission. The analysis is available 
for inspection in the NRC Public Document Room, 2120 L Street NW (Lower 
Level), Washington DC. Single copies of the analysis may be obtained 
from Thomas G. Scarbrough, Division of Engineering, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, Telephone: 301-415-2794, or Robert A. Hermann, Division of 
Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-
2768.

7. 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 involves the licensing and operation of nuclear power 
plants. The companies that own these plants do not fall within the 
scope of the definition of ``small entities'' set forth in the 
Regulatory Flexibility Act or the Small Business Size Standards set out 
in regulations issued by the Small Business Administration at 13 CFR 
part 121. Public comment received on this section suggested that the 
implementation of Appendix VIII of ASME BPV Code, Section XI, on 
performance qualification for ultrasonic testing might negatively 
impact small entities that contract their examination personnel to 
nuclear power plants. However, the final rule permits licensees to 
implement either Appendix VIII as contained in the 1995 Edition with 
the 1996 Addenda of the ASME Code, or Appendix VIII as implemented by 
the industry's PDI program. As a result, the NRC is unaware of any 
small entities in this area of expertise that are adversely affected 
such that they cannot satisfy either Appendix VIII as written or as 
implemented by PDI and endorsed in the rule.

8. Backfit Analysis

    The NRC regulations in 10 CFR 50.55a require that nuclear power 
plant owners--
    (1) Construct Class 1, Class 2, and Class 3 components in 
accordance with the rules provided in Section III, Division 1, 
``Requirements for Construction of Nuclear Power Plant Components,'' of 
the ASME BPV Code;
    (2) Inspect Class 1, Class 2, Class 3, Class MC (metal containment) 
and Class CC (concrete containment) components in accordance with the 
rules provided in Section XI, Division 1, ``Requirements for Inservice 
Inspection of Nuclear Power Plant Components,'' of the BPV Code; and
    (3) Test Class 1, Class 2, and Class 3 pumps and valves in 
accordance with the rules provided in Section XI, Division 1.
    The amendment to 10 CFR 50.55a endorses the 1995 Edition with the 
1996 Addenda of Section XI, Division 1, of the ASME BPV Code for ISI of 
Class 1, Class 2, Class 3, Class MC, and Class CC components; and the 
1995 Edition with the 1996 Addenda of the ASME OM Code for IST of Class 
1, Class 2, and Class 3 pumps and valves. The final rule requires 
licensees to implement Appendix VIII, ``Performance Demonstration for 
Ultrasonic Examination Systems,'' to Section XI, Division 1, as 
contained in the 1995 Edition with the 1996 Addenda of the ASME BPV 
Code, or Appendix VIII as

[[Page 51393]]

modified during the development of the PDI program.
    Under Sec. 50.55a(a)(3), licensees may voluntarily update to the 
1989 Addenda through the 1996 Addenda of Section III of the BPV Code, 
with limitations. In addition, the modification for containment 
isolation valve inservice testing that applied to the 1989 Edition of 
the BPV Code has been deleted.
    The NRC regulations currently require licensees to update their ISI 
and IST programs every 120 months to the version of Section XI 
incorporated by reference into 10 CFR 50.55a 12 months prior to the 
start of a new 10-year interval. In the past, the NRC position has 
consistently been that 10 CFR 50.109 does not ordinarily require a 
backfit analysis of the routine 120-month update to 10 CFR 50.55a. The 
basis for the NRC position is that
    (1) Section III, Division 1, update applies only to new 
construction (i.e., the edition and addenda to be used in the 
construction of a plant are selected based upon the date of the 
construction permit and are not changed thereafter, except voluntarily 
by the licensee);
    (2) Licensees understand that 10 CFR 50.55a requires that they 
update their ISI and IST programs every 10 years to the latest edition 
and addenda of the ASME Code that were incorporated by reference in 10 
CFR 50.55a and in effect 12 months before the start of the next 
inspection interval; and
    (3) The ASME Code is a national consensus standard developed by 
participants with broad and varied interests where all interested 
parties (including the NRC and utilities) participate; the consensus 
process includes an examination of the cost and benefits of proposed 
Code revisions.
    This consideration is consistent with both the intent and spirit of 
the backfit rule (i.e., NRC provides for the protection of the public 
health and safety, and does not unilaterally impose undue burden on 
applicants or licensees). Finally, to ensure that any interested member 
of the public that may not have had an opportunity to participate in 
the national consensus standard process is able to communicate with the 
NRC, proposed rules are published in the Federal Register. However, it 
should be noted that the Commission's initial endorsement of new 
subsections or appendices which would expand the scope of 10 CFR 50.55a 
to, e.g., include components that are not presently considered by the 
regulation (e.g., containment structures under Subsection IWE and 
Subsection IWL) would be subject to the Backfit Rule, unless one or 
more of the exceptions to 10 CFR 50.109(a)(4) apply.
    The Nuclear Utility Backfitting and Reform Group (NUBARG) and the 
Nuclear Energy Institute (NEI) each raised a concern with regard to the 
NRC's position on routine updates to 10 CFR 50.55a. Both NUBARG and NEI 
believe that, contrary to the NRC's determination, the routine updating 
of 10 CFR 50.55a to incorporate by reference new ASME Code provisions 
for ISI and IST constitutes a backfit for which a backfit analysis is 
required. The NRC has reviewed all of NUBARG's and NEI's comments in 
detail and has concluded that neither NUBARG nor NEI raise legal 
concerns which would alter the previous legal conclusion that the 
Backfit Rule does not require a backfit analysis of routine updates to 
10 CFR 50.55a to incorporate new ASME Code ISI and IST requirements. 
Based on the historical evolution of the ISI requirements in 10 CFR 
50.55a, the NRC believes it manifest that the ``automatic update'' of 
ISI programs under Sec. 50.55a(g) exists in tandem with the periodic 
updating and endorsement of new Code editions and addenda for ISI under 
Sec. 50.55a(b), and that the Commission intended that they be treated 
as an integrated regulatory structure for ISI which should not be 
subject to the Backfit Rule except in limited circumstances as 
discussed above. However, even though the NRC has determined that 
updating and endorsement of new Code editions and addenda are not 
subject to the Backfit Rule, the NRC is still considering these issues 
in the context of DSI 13. In particular, on April 27, 1999 (64 FR 
22580), the NRC published a supplement to the proposed rule dated 
December 3, 1997 (62 FR 63892), to eliminate the requirement for 
licensees to update their ISI and IST programs beyond a baseline 
edition and addenda of the ASME BPV Code. Under that proposed rule, 
licensees would continue to be allowed to update their ISI and IST 
programs to more recent editions and addenda of the ASME Code 
incorporated by reference in the regulations. Upon further review, the 
Commission decided to complete the issuance of this final rule 
endorsing the 1995 Edition with the 1996 Addenda of the ASME BPV Code 
and the ASME OM Code with appropriate limitations and modifications and 
to consider the elimination of the requirement to update ISI and IST 
programs every 120 months as a separate rulemaking effort. Following 
consideration of the public comments on the April 27, 1999, proposed 
rule, the NRC may prepare a final rule addressing the continued need 
for the requirement to update periodically ISI and IST programs and, if 
necessary, establishing an appropriate baseline edition of the ASME 
Code.
    The provisions for IST of pumps and valves were originally 
contained in Section XI Subsections IWP and IWV of the ASME BPV Code, 
but have now been moved by ASME to a new OM Code. Section XI, 1989 
Edition was incorporated by reference in the August 6, 1992, rulemaking 
(57 FR 34666). The 1990 OM Code standards, Parts 1, 6, and 10 of ASME/
ANSI-OM-1987, are identical to Section XI, 1989 Edition. This amendment 
is an administrative change simply referencing the 1995 Edition with 
the 1996 Addenda of the OM Code. Therefore, imposition of the 1995 
Edition with the 1996 Addenda of the OM Code is not a backfit.
    Appendix VIII to ASME BPV Code, Section XI, or Appendix VIII as 
modified during the development of the PDI program will be used to 
demonstrate the qualification of personnel and procedures for 
performing nondestructive examination of welds in components of systems 
that include the reactor coolant system and the emergency core cooling 
systems in nuclear power facilities. These performance demonstration 
programs will greatly increase the reliability of detection and sizing 
of cracks and flaws. Current requirements have been demonstrated not to 
be able to consistently and accurately identify and size cracks and 
flaws and thus are not effective. The Appendix delineates a method for 
qualification of the personnel and procedures. Appendix VIII changes 
the Code rules from a prescriptive set of requirements to a performance 
based approach that allows for implementation of improved technology 
without changes to the regulations. Performance demonstration would 
normally be imposed by the 120-month update requirement but, because of 
its importance, implementation of Appendix VIII is being expedited by 
the rulemaking. Because of the fundamental change in the nature of the 
qualification requirements, Appendix VIII is being considered a 
backfit. The proposed rule would have required licensees to implement 
Appendix VIII, including the modifications, for all examinations of the 
pressure vessel, piping, nozzles, and bolts and studs which occur after 
6 months from the date of the final rule. However, based on public 
comment, the final rule adopts a phased implementation approach for 
Appendix VIII, ranging from 6 months to 3 years, depending on the 
supplement. The final rule will not require any change to a licensee's 
ISI schedule for examination of these components, but will require

[[Page 51394]]

that the provisions of Appendix VIII as contained in the 1995 Edition 
with the 1996 Addenda (as supplemented by the final rule) or Appendix 
VIII as modified during the development of the PDI program (as 
supplemented by the final rule) be used for all examinations after that 
date rather than the UT procedures and personnel requirements presently 
being utilized by licensees.
    On the basis of the documented evaluation required by 
Sec. 50.109(a)(4), the NRC has concluded that imposition of Appendix 
VIII is necessary to bring the facilities described into compliance 
with GDC 14, 10 CFR Part 50, Appendix A, or similar provisions in the 
licensing basis for these facilities, and Criterion II, ``Quality 
Assurance Program,'' and Criterion XVI, ``Corrective Actions,'' of 
appendix B to 10 CFR part 50. Criterion II requires, in part, that a QA 
program shall take into account the need for special controls, 
processes, test equipment, tools, and skills to attain the required 
quality and the need for verification of quality by inspection and 
test. Evidence indicates that there are shortcomings in the 
qualifications of personnel and procedures in ensuring the reliability 
of the examinations. These safety significant revisions to the Code 
include specific requirements for UT performance demonstration, with 
statistically based acceptance criteria for blind testing of UT systems 
(procedures, equipment, and personnel) used to detect and size flaws. 
Criterion XVI requires that measures shall be established to assure 
that conditions adverse to quality, such as failures, malfunctions, 
deficiencies, deviations, defective material and equipment, and 
nonconformances, are promptly identified and corrected. Because of the 
serious degradation which has occurred, and the belief that additional 
occurrences of noncompliance with GDC 14, and Criteria II and XVI will 
occur, the NRC has determined that imposition of Appendix VIII 
beginning 6 months after the final rule has been published under the 
compliance exception to Sec. 50.109(a)(4)(i) is appropriate. Therefore, 
a backfit analysis is not required and the cost-benefit standards of 
Sec. 50.109(a)(3) do not apply. A complete discussion is contained in 
the documented evaluation.
    The rationale for application of the backfit rule and the backfit 
justification for the various items contained in this final rule are 
contained in the regulatory analysis and documented evaluation. The 
regulatory analysis and documented evaluation are available for 
inspection at the NRC Public Document Room, 2120 L Street NW (Lower 
Level), Washington, DC. Single copies of the regulatory analysis and 
documented evaluation are available from Thomas G. Scarbrough, Division 
of Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-
2794, or Robert A. Hermann, Division of Engineering, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, Telephone: 301-415-2768.

9. 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: Sections 102, 103, 104, 105, 161, 182, 183, 186, 189, 
68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 
234, 83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 
2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 
206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 
5846).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 
185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. 
L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(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 sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued 
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
    2. Section 50.55a is amended as follows:
    a. By removing paragraph (b)(2)(vii);
    b. By redesignating and revising paragraphs (b)(2)(viii), 
(b)(2)(ix), and (b)(2)(x) as (b)(2)(vii), (b)(2)(viii), and (b)(2)(ix), 
respectively;
    c. By adding paragraphs (b)(1)(i) through (b)(1)(v), (b)(2)(x) 
through (b)(2)(xvii), (b)(3), (g)(4)(iii), and (g)(6)(ii)(C); and
    d. By revising the introductory paragraph, the introductory text of 
paragraph (b), paragraph (b)(1), the introductory text of paragraph 
(b)(2), paragraph (b)(2)(vi), the introductory text of paragraph (f), 
paragraphs (f)(1), the introductory text of paragraph (f)(3), 
paragraphs (f)(3)(iii), (f)(3)(iv), the introductory text of paragraph 
(f)(4), paragraph (g)(1), the introductory text of paragraph (g)(3), 
paragraph (g)(3)(i), the introductory paragraph of (g)(4), and 
paragraphs (g)(4)(v)(C), (g)(6)(ii)(B)(1), and (g)(6)(ii)(B)(2), to 
read as follows:


Sec. 50.55a  Codes and standards.

    Each operating license for a boiling or pressurized water-cooled 
nuclear power facility is subject to the conditions in paragraphs (f) 
and (g) of this section and each construction permit for a utilization 
facility is subject to the following conditions in addition to those 
specified in Sec. 50.55.
* * * * *
    (b) The ASME Boiler and Pressure Vessel Code, and the ASME Code for 
Operation and Maintenance of Nuclear Power Plants, which are referenced 
in the following paragraphs, were approved for incorporation by 
reference by the Director of the Federal Register. A notice of any 
changes made to the material incorporated by reference will be 
published in the Federal Register. Copies of the ASME Boiler and 
Pressure Vessel Code and the ASME Code for Operation and Maintenance of 
Nuclear Power Plants may be purchased from the American Society of 
Mechanical Engineers, Three Park Avenue, New York, NY 10016. They are 
also available for inspection at the NRC Library, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland 20852-2738.

[[Page 51395]]

Copies are also available at the Office of the Federal Register, 800 N. 
Capitol Street, Suite 700, Washington, DC.
    (1) As used in this section, references to Section III of the ASME 
Boiler and Pressure Vessel Code refer to Section III, Division 1, and 
include editions through the 1995 Edition and addenda through the 1996 
Addenda, subject to the following limitations and modifications:
    (i) Section III Materials. When applying the 1992 Edition of 
Section III, licensees must apply the 1992 Edition with the 1992 
Addenda of Section II of the ASME Boiler and Pressure Vessel Code.
    (ii) Weld leg dimensions. When applying the 1989 Addenda through 
the 1996 Addenda of Section III, 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 shall not use these Articles in the 1994 Addenda through the 
1996 Addenda.
    (iv) Quality assurance. When applying editions and addenda later 
than the 1989 Edition of Section III, the requirements of NQA-1, 
``Quality Assurance Requirements for Nuclear Facilities,'' 1986 Edition 
through the 1992 Edition, are acceptable for use provided that the 
edition and addenda of NQA-1 specified in NCA-4000 is used in 
conjunction with the administrative, quality, and technical provisions 
contained in the edition and addenda of Section III being used.
    (v) Independence of inspection. Licensees may not apply NCA-
4134.10(a) of Section III, 1995 Edition with the 1996 Addenda.
    (2) As used in this section, references to Section XI of the ASME 
Boiler and Pressure Vessel Code refer to Section XI, Division 1, and 
include editions through the 1995 Edition and addenda through the 1996 
Addenda, 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 Sec. 50.55a(b)(2)(viii) and Sec. 50.55a(b)(2)(ix) when implementing 
the containment inservice inspection requirements of this section.
    (vii) Section XI References to OM Part 4, OM Part 6 and OM Part 10 
(Table IWA-1600-1). When using Table IWA-1600-1, ``Referenced Standards 
and Specifications,'' in the Section XI, Division 1, 1987 Addenda, 1988 
Addenda, or 1989 Edition, the specified ``Revision Date or Indicator'' 
for ASME/ANSI OM Part 4, ASME/ANSI Part 6, and ASME/ANSI Part 10 must 
be the OMa-1988 Addenda to the OM-1987 Edition. These requirements have 
been incorporated into the OM Code which is incorporated by reference 
in paragraph (b)(3) of this section.
    (viii) Examination of concrete containments. Licensees applying 
Subsection IWL, 1992 Edition with the 1992 Addenda, shall apply all of 
the modifications in this paragraph. Licensees choosing to apply the 
1995 Edition with the 1996 Addenda shall apply paragraphs 
(b)(2)(viii)(A), (viii)(D)(3), and (viii)(E) of this section.
    (A) Grease caps that are accessible must be visually examined to 
detect grease leakage or grease cap deformations. Grease caps must be 
removed for this examination when there is evidence of grease cap 
deformation that indicates deterioration of anchorage hardware.
    (B) When evaluation of consecutive surveillances of prestressing 
forces for the same tendon or tendons in a group indicates a trend of 
prestress loss such that the tendon force(s) would be less than the 
minimum design prestress requirements before the next inspection 
interval, an evaluation must be performed and reported in the 
Engineering Evaluation Report as prescribed in IWL-3300.
    (C) When the elongation corresponding to a specific load (adjusted 
for effective wires or strands) during retensioning of tendons differs 
by more than 10 percent from that recorded during the last measurement, 
an evaluation must be performed to determine whether the difference is 
related to wire failures or slip of wires in anchorage. A difference of 
more than 10 percent must be identified in the ISI Summary Report 
required by IWA-6000.
    (D) The licensee shall report the following conditions, if they 
occur, in the ISI Summary Report required by IWA-6000:
    (1) The sampled sheathing filler grease contains chemically 
combined water exceeding 10 percent by weight or the presence of free 
water;
    (2) The absolute difference between the amount removed and the 
amount replaced exceeds 10 percent of the tendon net duct volume;
    (3) Grease leakage is detected during general visual examination of 
the containment surface.
    (E) For Class CC applications, the licensee shall evaluate the 
acceptability of inaccessible areas when conditions exist in accessible 
areas that could indicate the presence of or result in degradation to 
such inaccessible areas. For each inaccessible area identified, the 
licensee shall provide the following in the ISI Summary Report required 
by IWA-6000:
    (1) A description of the type and estimated extent of degradation, 
and the conditions that led to the degradation;
    (2) An evaluation of each area, and the result of the evaluation, 
and;
    (3) A description of necessary corrective actions.
    (ix) Examination of metal containments and the liners of concrete 
containments.
    (A) For Class MC applications, the licensee shall evaluate the 
acceptability of inaccessible areas when conditions exist in accessible 
areas that could indicate the presence of or result in degradation to 
such inaccessible areas. For each inaccessible area identified, the 
licensee shall provide the following in the ISI Summary Report as 
required by IWA-6000:
    (1) A description of the type and estimated extent of degradation, 
and the conditions that led to the degradation;
    (2) An evaluation of each area, and the result of the evaluation, 
and;
    (3) A description of necessary corrective actions.
    (B) When performing remotely the visual examinations required by 
Subsection IWE, the maximum direct examination distance specified in 
Table IWA-2210-1 may be extended and the minimum illumination 
requirements specified in Table IWA-2210-1 may be decreased provided 
that the conditions or indications for which the visual examination is 
performed can be detected at the chosen distance and illumination.
    (C) The examinations specified in Examination Category E-B, 
Pressure Retaining Welds, and Examination Category E-F, Pressure 
Retaining Dissimilar Metal Welds, are optional.
    (D) Section 50.55a(b)(2)(ix)(D) may be used as an alternative to 
the requirements of IWE-2430.
    (1) If the examinations reveal flaws or areas of degradation 
exceeding the acceptance standards of Table IWE-3410-1, an evaluation 
must be performed to determine whether additional component 
examinations are required. For each flaw or area of degradation 
identified which exceeds acceptance standards, the licensee shall

[[Page 51396]]

provide the following in the ISI Summary Report required by IWA-6000:
    (i) A description of each flaw or area, including the extent of 
degradation, and the conditions that led to the degradation;
    (ii) The acceptability of each flaw or area, and the need for 
additional examinations to verify that similar degradation does not 
exist in similar components, and;
    (iii) A description of necessary corrective actions.
    (2) The number and type of additional examinations to ensure 
detection of similar degradation in similar components.
    (E) A general visual examination as required by Subsection IWE must 
be performed once each period.
    (x) Quality Assurance. When applying Section XI editions and 
addenda later than the 1989 Edition, the requirements of NQA-1, 
``Quality Assurance Requirements for Nuclear Facilities,'' 1979 Addenda 
through the 1989 Edition, are acceptable as permitted by IWA-1400 of 
Section XI, if the licensee uses its 10 CFR Part 50, Appendix B, 
quality assurance program, in conjunction with Section XI requirements. 
Commitments contained in the licensee's quality assurance program 
description that are more stringent than those contained in NQA-1 must 
govern Section XI activities. Further, where NQA-1 and Section XI do 
not address the commitments contained in the licensee's Appendix B 
quality assurance program description, the commitments must be applied 
to Section XI activities.
    (xi) Class 1 piping. Licensees may not apply IWB-1220, ``Components 
Exempt from Examination,'' of Section XI, 1989 Addenda through the 1996 
Addenda, and shall apply IWB-1220, 1989 Edition.
    (xii) Reserved.
    (xiii) Flaws in Class 3 Piping. Licensees may use the provisions of 
Code Case N-513, ``Evaluation Criteria for Temporary Acceptance of 
Flaws in Class 3 Piping,'' Revision 0, and Code Case N-523-1, 
``Mechanical Clamping Devices for Class 2 and 3 Piping.'' Licensees 
choosing to apply Code Case N-523-1 shall apply all of its provisions. 
Licensees choosing to apply Code Case N-513 shall apply all of its 
provisions subject to the following:
    (A) When implementing Code Case N-513, the specific safety factors 
in paragraph 4.0 must be satisfied.
    (B) Code Case N-513 may not be applied to:
    (1) Components other than pipe and tube, such as pumps, valves, 
expansion joints, and heat exchangers;
    (2) Leakage through a flange gasket;
    (3) Threaded connections employing nonstructural seal welds for 
leakage prevention (through seal weld leakage is not a structural flaw, 
thread integrity must be maintained); and
    (4) Degraded socket welds.
    (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. This 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 with the 1996 Addenda. Licensees 
choosing to apply these provisions shall apply all of the provisions 
except for those in Sec. 50.55a(b)(2)(xv)(F) which are optional.
    (A) When applying Supplements 2 and 3 to Appendix VIII, the 
following examination coverage criteria requirements must be used:
    (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.
    (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, 
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.
    (B) The following provisions must be used in addition to the 
requirements of Supplement 4 to Appendix VIII:
    (1) Paragraph 3.1, Detection acceptance criteria--Personnel are 
qualified for detection if the results of the performance demonstration 
satisfy the detection requirements of ASME Section XI, Appendix VIII, 
Table VIII-S4-1 and no flaw greater than 0.25 inch through wall 
dimension is missed.
    (2) Paragraph 1.1(c), Detection test matrix--Flaws smaller than the 
50 percent of allowable flaw size, as defined in IWB-3500, need not be 
included as detection flaws. For procedures applied from the inside 
surface, use the minimum thickness specified in the scope of the 
procedure to calculate a/t. For procedures applied from the outside 
surface, the actual thickness of the test specimen is to be used to 
calculate a/t.
    (C) When applying Supplement 4 to Appendix VIII, the following 
provisions must be used:
    (1) A depth sizing requirement of 0.15 inch RMS shall be used in 
lieu of the requirements in Subparagraphs 3.2(a) and 3.2(b).
    (2) In lieu of the location acceptance criteria requirements of 
Subparagraph 2.1(b), a flaw will be considered detected when reported 
within 1.0 inch or 10 percent of the metal path to the flaw, whichever 
is greater, of its true location in the X and Y directions.
    (3) In lieu of the flaw type requirements of Subparagraph 
1.1(e)(1), a minimum of 70 percent of the flaws in the detection and 
sizing tests shall be cracks. Notches, if used, must be limited by the 
following:
    (i) Notches must be limited to the case where examinations are 
performed from the clad surface.
    (ii) Notches must be semielliptical with a tip width of less than 
or equal to 0.010 inches.
    (iii) Notches must be perpendicular to the surface within 
 2 degrees.
    (4) In lieu of the detection test matrix requirements in paragraphs 
1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain 
a representative distribution of flaw orientations, sizes, and 
locations.
    (D) The following provisions must be used in addition to the 
requirements of Supplement 6 to Appendix VIII:
    (1) Paragraph 3.1, Detection Acceptance Criteria--Personnel are 
qualified for detection if:
    (i) No surface connected flaw greater than 0.25 inch through wall 
has been missed.
    (ii) No embedded flaw greater than 0.50 inch through wall has been 
missed.
    (2) Paragraph 3.1, Detection Acceptance Criteria--For procedure 
qualification, all flaws within the scope of the procedure are 
detected.
    (3) Paragraph 1.1(b) for detection and sizing test flaws and 
locations--Flaws smaller than the 50 percent of allowable flaw size, as 
defined in IWB-3500, need not be included as detection flaws. Flaws 
which are less than the allowable flaw size, as defined in IWB-3500, 
may be used as detection and sizing flaws.
    (4) Notches are not permitted.
    (E) When applying Supplement 6 to Appendix VIII, the following 
provisions must be used:
    (1) A depth sizing requirement of 0.25 inch RMS must be used in 
lieu of the requirements of subparagraphs 3.2(a), 3.2(c)(2), and 
3.2(c)(3).
    (2) In lieu of the location acceptance criteria requirements in 
Subparagraph

[[Page 51397]]

2.1(b), a flaw will be considered detected when reported within 1.0 
inch or 10 percent of the metal path to the flaw, whichever is greater, 
of its true location in the X and Y directions.
    (3) In lieu of the length sizing criteria requirements of 
Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch 
RMS must be used.
    (4) In lieu of the detection specimen requirements in Subparagraph 
1.1(e)(1), a minimum of 55 percent of the flaws must be cracks. The 
remaining flaws may be cracks or fabrication type flaws, such as slag 
and lack of fusion. The use of notches is not allowed.
    (5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test 
matrix, personnel demonstration test sets must contain a representative 
distribution of flaw orientations, sizes, and locations.
    (F) The following provisions may be used for personnel 
qualification for combined Supplement 4 to Appendix VIII and Supplement 
6 to Appendix VIII qualification. Licensees choosing to apply this 
combined qualification shall apply all of the provisions of Supplements 
4 and 6 including the following provisions:
    (1) For detection and sizing, the total number of flaws must be at 
least 10. A minimum of 5 flaws shall be from Supplement 4, and a 
minimum of 50 percent of the flaws must be from Supplement 6. At least 
50 percent of the flaws in any sizing must be cracks. Notches are not 
acceptable for Supplement 6.
    (2) Examination personnel are qualified for detection and length 
sizing when the results of any combined performance demonstration 
satisfy the acceptance criteria of Supplement 4 to Appendix VIII.
    (3) Examination personnel are qualified for depth sizing when 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws 
are sized within the respective acceptance criteria of those 
supplements.
    (G) When applying Supplement 4 to Appendix VIII, Supplement 6 to 
Appendix VIII, or combined Supplement 4 and Supplement 6 qualification, 
the following additional provisions must be used, and examination 
coverage must include:
    (1) The clad to base metal interface, including a minimum of 15 
percent T (measured from the clad to base metal interface), shall be 
examined from four orthogonal directions using procedures and personnel 
qualified in accordance with Supplement 4 to Appendix VIII.
    (2) If the clad-to-base-metal-interface procedure demonstrates 
detectability of flaws with a tilt angle relative to the weld 
centerline of at least 45 degrees, the remainder of the examination 
volume is considered fully examined if coverage is obtained in one 
parallel and one perpendicular direction. This must be accomplished 
using a procedure and personnel qualified for single-side examination 
in accordance with Supplement 6. Subsequent examinations of this volume 
may be performed using examination techniques qualified for a tilt 
angle of at least 10 degrees.
    (3) The examination volume not addressed by 
Sec. 50.55a(b)(2)(xv)(G)(1) is considered fully examined if coverage is 
obtained in one parallel and one perpendicular direction, using a 
procedure and personnel qualified for single sided examination when the 
provisions of Sec. 50.55a(b)(2)(xv)(G)(2) are met.
    (4) Where applications are limited by design to single side access, 
credit may be taken for the full volume provided the examination volume 
is covered from a single direction perpendicular to the weld and the 
weld volume is examined from at least one direction parallel to the 
weld.
    (H) When applying Supplement 5 to Appendix VIII, at least 50 
percent of the flaws in the demonstration test set must be cracks and 
the maximum misorientation shall be demonstrated with cracks. Flaws in 
nozzles with bore diameters equal to or less than 4 inches may be 
notches.
    (I) When applying Supplement 5, Paragraph (a), to Appendix VIII, 
the following provision must be used in calculating the number of 
permissible false calls:
    (1) The number of false calls allowed must be D/10, with a maximum 
of 3, where D is the diameter of the nozzle.
    (J) When applying the requirements of Supplement 5 to Appendix 
VIII, qualifications for the nozzle inside radius performed from the 
outside surface may be performed in accordance with Code Case N-552, 
``Qualification for Nozzle Inside Radius Section from the Outside 
Surface,'' provided that 10 CFR 50.55a(b)(2)(xv)(I)(1) is also 
satisfied.
    (K) When performing nozzle-to-vessel weld examinations, the 
following provisions must be used when the requirements contained in 
Supplement 7 to Appendix VIII are applied for nozzle-to-vessel welds in 
conjunction with Supplement 4 to Appendix VIII, Supplement 6 to 
Appendix VIII, or combined Supplement 4 and Supplement 6 qualification.
    (1) For examination of nozzle-to-vessel welds conducted from the 
bore, the following provisions are required to qualify the procedures, 
equipment, and personnel:
    (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 percent of the flaws 
must be cracks or fabrication flaws. Flaw locations and orientations 
must be selected from the choices shown in Sec. 50.55a(b)(2)(xv)(K)(4), 
Table VIII-S7-1--Modified, except flaws perpendicular to the weld are 
not required. There may be no more than two flaws from each category, 
and at least one subsurface flaw must be included.
    (ii) For length sizing, a minimum of four flaws as in 
Sec. 50.55a(b)(2)(xv)(K)(1)(i) must be included in the test set. The 
length sizing results must be added to the results of combined 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The 
combined results must meet the acceptance standards contained in 
Sec. 50.55a(b)(2)(xv)(E)(3
    (iii) For depth sizing, a minimum of four flaws as in 
Sec. 50.55a(b)(2)(xv)(K)(1)(i) must be included in the test set. Their 
depths must be distributed over the ranges of Supplement 4, Paragraph 
1.1, to Appendix VIII, for the inner 15 percent of the wall thickness 
and Supplement 6, Paragraph 1.1, to Appendix VIII, for the remainder of 
the wall thickness. The depth sizing results must be combined with the 
sizing results from Supplement 4 to Appendix VIII for the inner 15 
percent and to Supplement 6 to Appendix VIII for the remainder of the 
wall thickness. The combined results must meet the depth sizing 
acceptance criteria contained in Secs. 50.55a(b)(2)(xv)(C)(1), 
50.55a(b)(2)(xv)(E)(1), and 50.55a(b)(2)(xv)(F)(3).
    (2) For examination of reactor pressure vessel nozzle-to-vessel 
welds conducted from the inside of the vessel,
    (i) The clad to base metal interface and the adjacent examination 
volume to a minimum depth of 15 percent T (measured from the clad to 
base metal interface) must be examined from four orthogonal directions 
using a procedure and personnel qualified in accordance with Supplement 
4 to Appendix VIII as modified by Secs. 50.55a(b)(2)(xv)(B) and 
50.55a(b)(2)(xv)(C).
    (ii) When the examination volume defined in 
Sec. 50.55a(b)(2)(xv)(K)(2)(i) cannot be effectively examined in all 
four directions, the examination must be

[[Page 51398]]

augmented by examination from the nozzle bore using a procedure and 
personnel qualified in accordance with Sec. 50.55a(b)(2)(xv)(K)(1).
    (iii) The remainder of the examination volume not covered by 
Sec. 50.55a(b)(2)(xv)(K)(2)(ii) or a combination of 
Sec. 50.55a(b)(2)(xv)(K)(2)(i) and Sec. 50.55a(b)(2)(xv)(K)(2)(ii), 
must be examined from the nozzle bore using a procedure and personnel 
qualified in accordance with Sec. 50.55a(b)(2)(xv)(K)(1), or from the 
vessel shell using a procedure and personnel qualified for single sided 
examination in accordance with Supplement 6 to Appendix VIII, as 
modified by Secs. 50.55a(b)(2)(xv)(D), 50.55a(b)(2)(xv)(E), 
50.55a(b)(2)(xv)(F), and 50.55a(b)(2)(xv)(G).
    (3) For examination of reactor pressure vessel nozzle-to-shell 
welds conducted from the outside of the vessel,
    (i) The clad to base metal interface and the adjacent metal to a 
depth of 15 percent T, (measured from the clad to base metal interface) 
must be examined from one radial and two opposing circumferential 
directions using a procedure and personnel qualified in accordance with 
Supplement 4 to Appendix VIII, as modified by Secs. 50.55a(b)(2)(xv)(B) 
and 50.55a(b)(2)(xv)(C), for examinations performed in the radial 
direction, and Supplement 5 to Appendix VIII, as modified by 
Sec. 50.55a(b)(2)(xv)(J), for examinations performed in the 
circumferential direction.
    (ii) The examination volume not addressed by 
Sec. 50.55a(b)(2)(xv)(K)(3)(i) must be examined in a minimum of one 
radial direction using a procedure and personnel qualified for single 
sided examination in accordance with Supplement 6 to Appendix VIII, as 
modified by Secs. 50.55a(b)(2)(xv)(D), 50.55a(b)(2)(xv)(E), 
50.55a(b)(2)(xv)(F), and 50.55a(b)(2)(xv)(G).
    (4) Table VIII-S7-1, ``Flaw Locations and Orientations,'' 
Supplement 7 to Appendix VIII, is modified as follows:

                        Table VIII-S7-1--Modified
------------------------------------------------------------------------
                     Flaw Locations and Orientations
-------------------------------------------------------------------------
                                                Parallel   Perpendicular
                                                 to weld      to weld
------------------------------------------------------------------------
Inner 15 percent.............................          X             X
OD Surface...................................          X   .............
Subsurface...................................          X   .............
------------------------------------------------------------------------

    (L) As a modification to the requirements of Supplement 8, 
Subparagraph 1.1(c), to Appendix VIII, notches may be located within 
one diameter of each end of the bolt or stud.
    (xvi) Appendix VIII single side ferritic vessel and piping and 
stainless steel piping examination.
    (A) Examinations performed from one side of a ferritic vessel weld 
must be conducted with equipment, procedures, and personnel that have 
demonstrated proficiency with single side examinations. To demonstrate 
equivalency to two sided examinations, the demonstration must be 
performed to the requirements of Appendix VIII as modified by this 
paragraph and Secs. 50.55a(b)(2)(xv) (B) through (G), on specimens 
containing flaws with non-optimum sound energy reflecting 
characteristics or flaws similar to those in the vessel being examined.
    (B) Examinations performed from one side of a ferritic or stainless 
steel pipe weld must be conducted with equipment, procedures, and 
personnel that have demonstrated proficiency with single side 
examinations. To demonstrate equivalency to two sided examinations, the 
demonstration must be performed to the requirements of Appendix VIII as 
modified by this paragraph and Sec. 50.55a(b)(2)(xv)(A).
    (xvii) Reconciliation of Quality Requirements. When purchasing 
replacement items, in addition to the reconciliation provisions of IWA-
4200, 1995 Edition with the 1996 Addenda, the replacement items must be 
purchased, to the extent necessary, in accordance with the owner's 
quality assurance program description required by 10 CFR 
50.34(b)(6)(ii).
    (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 and the 1996 Addenda subject to the following 
limitations and modifications:
    (i) Quality Assurance. When applying editions and addenda of the OM 
Code, the requirements of NQA-1, ``Quality Assurance Requirements for 
Nuclear Facilities,'' 1979 Addenda, are acceptable as permitted by ISTA 
1.4 of the OM Code, provided the licensee uses its 10 CFR part 50, 
Appendix B, quality assurance program in conjunction with the OM Code 
requirements. Commitments contained in the licensee's quality assurance 
program description that are more stringent than those contained in 
NQA-1 govern OM Code activities. If NQA-1 and the OM Code do not 
address the commitments contained in the licensee's Appendix B quality 
assurance program description, the commitments must be applied to OM 
Code activities.
    (ii) Motor-Operated Valve stroke-time testing. Licensees shall 
comply with the provisions on stroke time testing in OM Code ISTC 4.2, 
1995 Edition with the 1996 Addenda, 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 Sec. 50.55a(b)(3)(ii), 
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, 1995 
Edition with the 1996 Addenda, in conjunction with ISTC 4.3, 1995 
Edition with the 1996 Addenda. Licensees choosing to apply the Code 
case shall apply all of its provisions.
    (A) The adequacy of the diagnostic test interval for each valve 
must be evaluated and adjusted as necessary but not later than 5 years 
or three refueling outages (whichever is longer) from initial 
implementation of ASME Code Case OMN-1.
    (B) When extending exercise test intervals for high risk motor-
operated valves beyond a quarterly frequency, licensees shall ensure 
that the potential increase in core damage frequency and risk 
associated with the extension is small and consistent with the intent 
of the Commission's Safety Goal Policy Statement.
    (iv) Appendix II. The following modifications apply when 
implementing Appendix II, ``Check Valve Condition Monitoring Program,'' 
of the OM Code, 1995 Edition with the 1996 Addenda:
    (A) Valve opening and closing functions must be demonstrated when 
flow testing or examination methods (nonintrusive, or disassembly and 
inspection) are used;
    (B) The initial interval for tests and associated examinations may 
not exceed two fuel cycles or 3 years, whichever is longer; any 
extension of this interval may not exceed one fuel cycle per extension 
with the maximum interval not to exceed 10 years; trending and 
evaluation of existing data must be used to reduce or extend the time 
interval between tests.
    (C) If the Appendix II condition monitoring program is 
discontinued, then the requirements of ISTC 4.5.1 through 4.5.4 must be 
implemented.
    (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

[[Page 51399]]

use Subsection ISTD, ``Inservice Testing of Dynamic Restraints 
(Snubbers) in Light-Water Reactor Power Plants,'' ASME OM Code, 1995 
Edition up to and including the 1996 Addenda, in lieu 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 shall be performed using the VT-3 visual 
examination method described in IWA-2213.
* * * * *
    (f) Inservice testing requirements. Requirements for inservice 
inspection of Class 1, Class 2, Class 3, Class MC, and Class CC 
components (including their supports) are located in Sec. 50.55a(g).
    (1) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit was issued prior to January 1, 1971, 
pumps and valves must meet the test requirements of paragraphs (f)(4) 
and (f)(5) of this section to the extent practical. Pumps and valves 
which are part of the reactor coolant pressure boundary must meet the 
requirements applicable to components which are classified as ASME Code 
Class 1. Other pumps and valves that perform a function to shut down 
the reactor or maintain the reactor in a safe shutdown condition, 
mitigate the consequences of an accident, or provide overpressure 
protection for safety-related systems (in meeting the requirements of 
the 1986 Edition, or later, of the Boiler and Pressure Vessel or OM 
Code) must meet the test requirements applicable to components which 
are classified as ASME Code Class 2 or Class 3.
* * * * *
    (3) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit was issued on or after July 1, 1974:
* * * * *
    (iii)(A) Pumps and valves, in facilities whose construction permit 
was issued before November 22, 1999, which are classified as ASME Code 
Class 1 must be designed and be provided with access to enable the 
performance of inservice testing of the pumps and valves for assessing 
operational readiness set forth in Section XI of editions of the ASME 
Boiler and Pressure Vessel Code and Addenda \6\ applied to the 
construction of the particular pump or valve or the Summer 1973 
Addenda, whichever is later.
    (B) Pumps and valves, in facilities whose construction permit is 
issued on or after November 22, 1999, which are classified as ASME Code 
Class 1 must be designed and be provided with access to enable the 
performance of inservice testing of the pumps and valves for assessing 
operational readiness set forth in editions and addenda of the ASME OM 
Code referenced in paragraph (b)(3) of this section at the time the 
construction permit is issued.
    (iv)(A) Pumps and valves, in facilities whose construction permit 
was issued before November 22, 1999, which are classified as ASME Code 
Class 2 and Class 3 must be designed and be provided with access to 
enable the performance of inservice testing of the pumps and valves for 
assessing operational readiness set forth in Section XI of editions of 
the ASME Boiler and Pressure Vessel Code and Addenda 6 
applied to the construction of the particular pump or valve or the 
Summer 1973 Addenda, whichever is later.
    (B) Pumps and valves, in facilities whose construction permit is 
issued on or after November 22, 1999, which are classified as ASME Code 
Class 2 and 3 must be designed and be provided with access to enable 
the performance of inservice testing of the pumps and valves for 
assessing operational readiness set forth in editions and addenda of 
the ASME OM Code referenced in paragraph (b)(3) of this section at the 
time the construction permit is issued.
* * * * *
    (4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, pumps and valves which are classified as 
ASME Code Class 1, Class 2 and Class 3 must meet the inservice test 
requirements, except design and access provisions, set forth in the 
ASME OM Code and addenda that become effective subsequent to editions 
and addenda specified in paragraphs (f)(2) and (f)(3) of this section 
and that are incorporated by reference in paragraph (b) of this 
section, to the extent practical within the limitations of design, 
geometry and materials of construction of the components.
* * * * *
    (g) * * *
    (1) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit was issued before January 1, 1971, 
components (including supports) must meet the requirements of 
paragraphs (g)(4) and (g)(5) of this section to the extent practical. 
Components which are part of the reactor coolant pressure boundary and 
their supports must meet the requirements applicable to components 
which are classified as ASME Code Class 1. Other safety-related 
pressure vessels, piping, pumps and valves, and their supports must 
meet the requirements applicable to components which are classified as 
ASME Code Class 2 or Class 3.
* * * * *
    (3) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit was issued on or after July 1, 1974:
    (i) Components (including supports) which are classified as ASME 
Code Class 1 must be designed and be provided with access to enable the 
performance of inservice examination of such components and must meet 
the preservice examination requirements set forth in Section XI of 
editions of the ASME Boiler and Pressure Vessel Code and Addenda 
6 applied to the construction of the particular component.
* * * * *
    (4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) which 
are classified as ASME Code Class 1, Class 2 and Class 3 must meet the 
requirements, except design and access provisions and preservice 
examination requirements, set forth in Section XI of editions of the 
ASME Boiler and Pressure Vessel Code and Addenda that become effective 
subsequent to editions specified in paragraphs (g)(2) and (g)(3) of 
this section and that are incorporated by reference in paragraph (b) of 
this section, to the extent practical within the limitations of design, 
geometry and materials of construction of the components. Components 
which are classified as Class MC pressure retaining components and 
their integral attachments, and components which are classified as 
Class CC pressure retaining components and their integral attachments 
must meet the requirements, except design and access provisions and 
preservice examination requirements, set forth in Section XI of the 
ASME Boiler and Pressure Vessel Code and Addenda that are incorporated 
by reference in paragraph (b) of this section, subject to the 
limitation listed in paragraph (b)(2)(vi) of this section and the 
modifications listed in paragraphs (b)(2)(viii) and (b)(2)(ix) of this 
section, to the extent practical within the limitation of design, 
geometry and materials of construction of the components.
* * * * *
    (iii) Licensees may, but are not required to, perform the surface 
examinations of High Pressure Safety

[[Page 51400]]

Injection Systems specified in Table IWB-2500-1, Examination Category 
B-J, Item Numbers B9.20, B9.21, and B9.22.
* * * * *
    (v) * * *
    (C) Concrete containment pressure retaining components and their 
integral attachments, and the post-tensioning systems of concrete 
containments must meet the inservice inspection, repair, and 
replacement requirements applicable to components which are classified 
as ASME Code Class CC.
* * * * *
    (6) * * *
    (ii) * * *
    (B) Expedited examination of containment.
    (1) Licensees of all operating nuclear power plants shall implement 
the inservice examinations specified for the first period of the first 
inspection interval in Subsection IWE of the 1992 Edition with the 1992 
Addenda in conjunction with the modifications specified in 
Sec. 50.55a(b)(2)(ix) by September 9, 2001. The examination performed 
during the first period of the first inspection interval must serve the 
same purpose for operating plants as the preservice examination 
specified for plants not yet in operation.
    (2) Licensees of all operating nuclear power plants shall implement 
the inservice examinations which correspond to the number of years of 
operation which are specified in Subsection IWL of the 1992 Edition 
with the 1992 Addenda in conjunction with the modifications specified 
in Sec. 50.55a(b)(2)(viii) by September 9, 2001. The first examination 
performed must serve the same purpose for operating plants as the 
preservice examination specified for plants not yet in operation. The 
first examination of concrete must be performed prior to September 10, 
2001, and the date of the examination need not comply with the 
requirements of IWL-2410(a) or IWL-2410(b). The date of the first 
examination of concrete must be used to determine the 5-year schedule 
for subsequent examinations subject to the provisions of IWL-2410(c).
* * * * *
    (C) Implementation of Appendix VIII to Section XI. (1) The 
Supplements to Appendix VIII of 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: 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, 10, 12, and 
13--November 22, 2002.
* * * * *

    Dated at Rockville, MD this 26th day of August, 1999.

    For the Nuclear Regulatory Commission.
William D. Travers,
Executive Director for Operations.
[FR Doc. 99-24256 Filed 9-21-99; 8:45 am]
BILLING CODE 7590-01-P