[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]]
_______________________________________________________________________
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.
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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