[Federal Register Volume 61, Number 154 (Thursday, August 8, 1996)]
[Rules and Regulations]
[Pages 41303-41312]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-20215]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AC93
Codes and Standards for Nuclear Power Plants; Subsection IWE and
Subsection IWL
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to incorporate by reference 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,'' of Section XI, Division 1, of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code) with specified modifications and a limitation.
Subsection IWE of the ASME Code provides rules for inservice
inspection, repair, and replacement of Class MC pressure retaining
components and their integral attachments and of metallic shell and
penetration liners of Class CC pressure retaining components and their
integral attachments in light-water cooled power plants. Subsection IWL
of the ASME Code provides rules for inservice inspection and repair of
the reinforced concrete and the post-tensioning systems of Class CC
components. Licensees will be required to incorporate Subsection IWE
and Subsection IWL into their inservice inspection (ISI) program.
Licensees will also be required to expedite implementation of the
containment examinations and to complete the expedited examination in
accordance with Subsection IWE and Subsection IWL within 5 years of the
effective date of this rule. Provisions have been included that will
prevent unnecessary duplication of examinations between the expedited
examination and the routine 120-month ISI examinations. Subsection IWE
and Subsection IWL have not been previously incorporated by reference
into the NRC regulations. The final rule specifies requirements to
assure that the critical areas of containments are routinely inspected
to detect and take corrective action for defects that could compromise
a containment's structural integrity.
EFFECTIVE DATE: September 9, 1996. The incorporation by reference of
certain publications listed in the regulations is approved by the
Office of the Director of the Office of the Federal Register as of
September 9, 1996.
FOR FURTHER INFORMATION CONTACT: Mr. W. E. Norris, Division of
Engineering Technology, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, telephone (301)
415-6796.
SUPPLEMENTARY INFORMATION: The NRC is amending its regulations to
incorporate by reference the 1992 Edition with the 1992 Addenda of
Subsection IWE and Subsection IWL to assure that the critical areas of
containments are routinely inspected to detect and take corrective
action for defects that could compromise a containment's structural
integrity. The rate of occurrence of degradation in containments is
increasing. Appendix J to 10 CFR part 50 requires a general visual
inspection of the containment but does not provide specific guidance on
how to perform the necessary containment examinations. This has
resulted in a large variation with regard to the performance and the
effectiveness of containment examinations. The rate of occurrence of
corrosion and degradation of containment structures has been increasing
at operating nuclear power plants. There have been 32 reported
occurrences of corrosion in metal containments and the liners of
concrete containments. This is one-fourth of all operating nuclear
power plants. Only four of the 32 occurrences were detected by current
containment inspection programs. Nine of these occurrences were first
identified by the NRC through its inspections or structural audits.
Eleven occurrences were detected by licensees after they were alerted
to a degraded condition at another site or through activity other than
containment inspection. There have been 34 reported occurrences of
degradation of the concrete or of the post-tensioning systems of
concrete containments. This is nearly one-half of these types of
containments. It is clear that current licensee containment inspection
programs have not proved to be adequate to detect the types of
degradation which have been reported. Examples of degradation not found
by licensees, but initially detected at plants through NRC inspections
include: (1) Corrosion of steel containment shells in the drywell sand
cushion region, resulting in wall thickness reduction to below the
minimum design thickness; (2) corrosion of the torus of the steel
containment shell (wall thickness below minimum design thickness); (3)
corrosion of the liner of a concrete containment to approximately half-
depth; (4) grease leakage from the tendons of prestressed concrete
containments; and (5) leaching as well as excessive cracking in
concrete containments.
There are several General Design Criteria (GDC) and ASME Code
sections which establish minimum requirements for the design,
fabrication, construction, testing, and performance of structures,
systems, and components important to safety in water-cooled nuclear
power plants. The GDC serve as fundamental underpinnings for many of
the most safety important commitments in
[[Page 41304]]
licensee design and licensing bases. GDC 16, ``Containment design,''
requires the provision of reactor containment and associated systems to
establish an essentially leak-tight barrier against the uncontrolled
release of radioactivity into the environment and to ensure that the
containment design conditions important to safety are not exceeded for
as long as required for postulated accident conditions.
Criterion 53, ``Provisions for containment testing and
inspection,'' requires that the reactor containment design permit: (1)
Appropriate periodic inspection of all important areas, such as
penetrations; (2) an appropriate surveillance program; and (3) periodic
testing at containment design pressure of the leak-tightness of
penetrations which have resilient seals and expansion bellows. Appendix
J, ``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' of 10 CFR part 50 contains specific rules for leakage
testing of containments. Paragraph III. A. of Appendix J requires that
a general inspection of the accessible interior and exterior surfaces
of the containment structures and components be performed prior to any
Type A test to uncover any evidence of structural deterioration that
may affect either the containment structural integrity or leak-
tightness (Type A test means tests intended to measure the primary
reactor containment overall integrated leakage rate: (1) after the
containment has been completed and is ready for operation, and (2) at
periodic intervals thereafter).
The metal containment structure of operating nuclear power plants
were designed in accordance with either Section III, Subsection NE,
``Class MC Components,'' or Section VIII, of the ASME Code. These
subsections contain provisions for the design and construction of metal
containment structures, including methods for determining the minimum
required wall thicknesses. The minimum wall thickness is that thickness
that would ensure that the metal containment structure would continue
to maintain its structural integrity under the various stressors and
degradation mechanisms which could act on it.
The prestressed concrete containments of most operating nuclear
reactors were designed in accordance with ACI-318 provisions taking
into consideration their unique features in the design of the post-
tensioning system and in determining the prestressing forces. The post-
tensioning system is designed so that the concrete containment
structure will continue to maintain its structural integrity under the
various stressors and degradation mechanisms which act on it. The
liners of concrete containments provide a leak-tight barrier.
These requirements for minimum design wall thicknesses and
prestressing forces as provided in these industry standards used to
design containment structures are reflected in license conditions,
technical specifications, and licensee commitments (e.g., the Final
Safety Analysis Report).
None of the existing requirements, however, provide specific
guidance on how to perform the necessary containment examinations. This
lack of guidance has resulted in a large variation with regard to the
performance and the effectiveness of licensee containment examination
programs. Based on the results of inspections and audits, as well as
plant operational experiences, it is clear that many licensee
containment examination programs have not detected degradation that
could ultimately result in a compromise to the pressure-retaining
capability. Some containment structures have been found to have
undergone a significant level of degradation that was not detected by
these programs.
The Nuclear Management and Resources Council (NUMARC) (which has
since become the Nuclear Energy Institute (NEI)) developed a number of
industry reports to address license renewal issues. Two of those, one
for Pressurized Water Reactor (PWR) containments and the other for
Boiling Water Reactor (BWR) containments, were developed for the
purpose of managing age-related degradation of containments on a
generic basis. The NUMARC plan for containments relies on the
examinations contained in Subsection IWE and Subsection IWL to manage
age-related degradation, and this plan assumes that these examinations
are ``in current and effective use.'' In the BWR Containment Industry
Report, NUMARC concluded that ``On account of these available and
established methods and techniques to adequately manage potential
degradation due to general corrosion of freestanding metal
containments, no additional measures need to be developed and, as such,
general corrosion is not a license renewal concern if the containment
minimum wall thickness is maintained and verified.'' Similarly, in the
PWR Containment Industry Report, NUMARC concluded that potentially
significant degradation of concrete surfaces, the post-tensioning
system, and the liners of concrete containments could be managed
effectively if periodically examined in accordance with the
requirements contained in Subsection IWE and Subsection IWL. The NRC
agrees with NEI that these ASME standards, which the industry has
participated in developing, would be an effective means for managing
age-related containment degradation. Thus, the NRC believes that
adoption of these standards is the best approach.
Background
On January 7, 1994 (59 FR 979), the NRC published in the Federal
Register a proposed amendment to its regulation, 10 CFR part 50,
``Domestic Licensing of Production and Utilization Facilities,'' to
incorporate by reference the 1992 Edition with the 1992 Addenda of
Subsection IWE, and Subsection IWL, of Section XI, Division 1, of the
ASME Code with specified modifications and a limitation.
Five modifications were specified in the proposed rule to address
two concerns of the NRC. The first concern is that four recommendations
for tendon examinations that are included in Regulatory Guide 1.35,
``Inservice Inspection of Ungrouted Tendons in Prestressed Concrete
Containments,'' Rev. 3, are not addressed in Subsection IWL (this
involves four of the modifications, (Sec. 50.55a(b)(2)(ix)(A)-(D)).
Regulatory Guide 1.35, Rev. 3, describes a basis acceptable to the NRC
staff for developing an appropriate inservice inspection and
surveillance program for ungrouted tendons in prestressed concrete
containment structures. The four recommendations contained in
Regulatory Guide 1.35, Rev. 3, which are not addressed by Subsection
IWL, provide positions on issues such as failed wires and tendon
sheathing filler grease conditions. (The ASME Code has considered the
four issues involved and is in the process of adopting them into
addenda of Subsection IWL). The second NRC concern is that if there is
visible evidence of degradation of the concrete (e.g., leaching,
surface cracking) there may also be degradation of inaccessible areas.
The fifth modification (Sec. 50.55a(b)(2)(ix)(E)) requires that
inaccessible areas be evaluated when visible conditions exist that
suggest the possibility of degradation of these areas.
The limitation which was included in the proposed rule specified
the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection
IWL as the earliest version of the ASME Code the NRC finds acceptable.
This is because this is the first edition including addenda combination
acceptable to the NRC staff that incorporates the concept of base metal
examinations and also provides a
[[Page 41305]]
comprehensive set of rules for the examination of post-tensioning
systems. As originally published in 1981, Subsection IWE preservice
examination and inservice examination rules focused on the examination
of welds. This weld-based examination philosophy was established in the
1970s as plants were being constructed. It was based on the premise
that the welds in pressure vessels and piping were the areas of
greatest concern. As containments have aged, degradation of base metal,
rather than welds, has been found to be the issue of concern. The 1991
Addenda to the 1989 Edition, the 1992 Edition and the 1992 Addenda to
Section XI, Subsection IWE, have promoted the incorporation of base
metal examinations.
The proposed rulemaking incorporated a provision for an expedited
examination schedule. This expedited examination schedule is necessary
to prevent the delay in implementation of Subsection IWE and Subsection
IWL (the Summary of Documented Evaluation lists each plant and the
delay in implementation which would be encountered if the subsections
were implemented through routine updates of the ISI programs).
Provisions were incorporated in the proposed rule to ensure that the
expedited examination which would be completed within 5 years from the
effective date of the rule and the routine 120-month examinations did
not duplicate examinations.
On March 4, 1994, the NRC received a request from the Nuclear
Management and Resources Council (which has since become part of the
Nuclear Energy Institute (NEI)) to extend the public comment period
from March 23, 1994 until April 25, 1994, to enable NEI to ``provide
necessary and constructive comments on the proposed rule change.'' This
was granted, and on March 28, 1994 (59 FR 14373), the NRC published in
the Federal Register a notice of extension of the public comment
period.
Summary of Comments
Comments were received from 25 separate sources. These sources
consisted of 15 utilities, one service organization (Entergy
Operations, Inc.) representing five nuclear plants, the Nuclear Energy
Institute (NEI), the Nuclear Utility Backfitting and Reform Group
(NUBARG) represented by the firm of Winston & Strawn, one owner's group
(BWR Owner's Group (BWROG)), one architect and engineering firm (Stone
& Webster Engineering Corporation), one public citizens group (Ohio
Citizens for Responsible Energy (OCRE)), three individuals, and one
consulting firm (VSL Corporation).
Comments received could be divided into three groups. The first
group contains those comments which address the administrative aspects
of the rule (e.g., backfit considerations, effectiveness of current
containment examinations), and the modifications specified by the NRC
in the proposed rule. The second and third groups contain those
comments which address the technical provisions of Subsection IWE, and
Subsection IWL, respectively. The summary and resolution of public
comments and all of the verbatim comments which were received (grouped
by subject area) are contained in the Summary of Documented Evaluation.
The majority of comments generally addressed one of the following
subject areas: (1) The incorporation by reference of Subsection IWE and
Subsection IWL into Sec. 50.55a; (2) the development of guidance
documents instead of regulatory requirements; (3) the rationale for the
proposed backfit; (4) endorsement of the BWROG comments; and (5) the 5-
year expedited implementation. These subject areas encompass the
comments submitted by NEI and NUBARG, and their comments, if any, are
discussed separately in each subject area.
The comments on subject area number one from those that approve of
the incorporation by reference of Subsection IWE and Subsection IWL
into Sec. 50.55a, can be summarized as follows: (1) There is a need for
the periodic examination of containment structures to assure the
containment's pressure-retaining and leak-tight capability; (2) Section
XI requirements define concise, technically sound programs to assure
continuing containment integrity; and (3) input in the development of
these rules was provided by all interested parties involved in
containment inservice inspection--users, regulators, manufacturers,
engineering organizations, and enforcement organizations.
The comments on the other four subject areas are summarized below.
The resolution of public comments contains all of the comments which
were received. Some of the comments resulted in modifications to the
rule, and some of the comments have been transmitted to the ASME for
their consideration. A discussion of the comments which led to
modifications follows the summary of comments on subject area number
five. The resolution of public comments package contains those comments
transmitted to the ASME. Those comments asked for interpretations of
the ASME Code rules.
Regarding subject area number two, eleven commenters believe that
additional specific guidance in the form of a guidance document would
be more appropriate than a regulation. They concur with NEI that
current regulatory requirements for containment integrity and
examinations are already provided by existing regulations (GDC 16 and
53 and Appendix J) and licensee commitments. If more detail on how to
perform containment examinations is needed, the commenters (including
NEI) state that the details could be provided in a regulatory guide,
Information Notice, Generic Letter, or in an industry developed
guidance document. The NRC does not believe that existing regulations
and licensee commitments are adequate. Existing regulations and
licensee commitments have not proved to be adequate to detect the types
of problems which have been experienced in operating reactors. This is
evidenced by the large number of instances of degradation that were
found by the NRC through its inspections or audits of plant structures,
or by licensees because they were alerted to a degraded condition at
another site. Licensee containment inspection programs have generally
not detected the types of degradation being reported (only four of the
32 reported instances of corrosion in Class MC containments were
discovered as a result of the Appendix J general inspection). Further,
the NRC does not believe that providing guidance through a regulatory
guide or industry report would generally improve containment
examination practices. Licensees were made aware of containment
degradation through several industry notices, and yet the staff is
still detecting many of occurrences of degradation. The increasing rate
of occurrence of containment degradation, the number of occurrences,
the extent to which some containments were degraded, the high number of
instances discovered through NRC inspections or by licensees because
they were alerted to a degradation condition at another site, the time-
dependent mechanisms, and the results of the survey performed by the
NRC Regional Offices regarding current containment inspections all
point to the necessity of imposing additional requirements to ensure
that containments comply with design wall thicknesses and prestressing
forces. This is a compliance backfit.
With regard to subject area number three, six general comments were
received from the Nuclear Utility Backfitting and Reform Group (NUBARG)
and from the Nuclear Energy
[[Page 41306]]
Institute (NEI) (which were endorsed by other commenters) regarding the
incorporation by reference of Subsection IWE and Subsection IWL which
are similar in nature. The first comment is that the application of the
compliance exception to this rulemaking is inappropriate, and that the
proposed rule constitutes a backfit for which a cost-benefit analysis
should be performed. The NRC agrees that the rulemaking is a backfit.
However, as discussed under the Backfit Statement, the NRC believes
that the compliance exception to the backfit rule is appropriate.
The second comment was a citation of a paragraph from the Statement
of Considerations to the 1985 final backfit rule which addressed the
compliance exception. That paragraph addressed ``Section 50.109(a)(4)
which creates exceptions for modifications necessary to bring a
facility into compliance or to ensure through immediately effective
regulatory action that a licensee meets a standard of no undue risk to
public health and safety.'' Both NEI and NUBARG assert that the
proposed rule is a new interpretation of how to demonstrate compliance
with existing standards and therefore constitutes a backfit under 10
CFR 50.109(a)(1). The NRC does not believe that the use of the
compliance exception must be confined only to the situation addressed
in the Statement of Consideration to the 1985 final backfit rule--
``omission or mistake of fact.'' In any event, the current
unsatisfactory status of containment inservice inspections can be
characterized fairly as, in retrospect, a mistake about and omission
from the necessary elements of a satisfactory inspection program.
The third comment is that containments must experience corrosion or
degradation that is so unanticipated and excessive so as to constitute
a genuine compliance concern. Another commenter expressed the idea
somewhat differently believing that a broad-based concern with the
operability of containment structures through the industry must be
demonstrated to be a compliance issue. The NRC agrees with those
criteria and concludes, in fact, that there is a broad-based concern
regarding the structural integrity of containment structures. The NRC's
approach focuses on two questions: (1) Is the corrosion such that there
is a basis for reasonably concluding that additional instances of
noncompliance with the relevant GDCs, Appendix J, and/or licensee
commitments at numerous plants; and (2) whether there is a basis for
reasonably believing that the corrosion would have been identified and
properly addressed by the licensees in the absence of additional
regulatory requirements. Based on the: (1) Number of occurrences of
containment degradation; (2) increasing rate of containment
degradation; (3) locations of the degradation; (4) two instances where
containment wall thicknesses were below minimum design wall thickness;
(5) number of corrosion paths which have been reported; and (6) higher
than anticipated corrosion rates in many of the occurrences, the NRC
believes that containments are experiencing corrosion or degradation
that is unanticipated and excessive. Further, based upon factors (1) to
(6) above, the NRC concludes that additional criteria are necessary to
ensure that compliance with existing requirements for minimum accepted
design wall thicknesses and prestressing forces are maintained (and
thereby the ability of the containment to continue to perform its
intended safety function).
The fourth comment by NUBARG and NEI suggested that it is part of
the anticipated process for the industry to rely upon NRC inspections
and audits to identify problems and then alert the industry through NRC
documents such as information notices and generic letters. During the
presentation to the ACRS on February 10, 1995, NEI asserted that ``[i]t
really doesn't matter how the utilities identify these instances of
degradation.'' The NRC believes that inspections conducted by licensees
should be adequate to ensure that containment degradation is identified
without reliance upon NRC inspections.
The fifth NEI and NUBARG comment is that to ensure compliance the
NRC could take individual enforcement action rather than endorse ASME
standards. The NRC believes that the best approach is to adopt the
industry consensus standards (i.e., endorse ASME Section XI Subsection
IWE and Subsection IWL). Containment corrosion and degradation have
been reported since 1986. The patterns of degradation and the
corrective actions were not immediately obvious. Given the number and
the extent of the occurrences, and the variability among plants with
regard to the performance and the effectiveness of containment
inspections, the NRC believes that the best course of action is to
endorse ISI requirements to ensure that containments comply with design
wall thicknesses and prestressing forces.
The sixth comment is that GDC 16 required containments to be
designed and constructed with an allowance for corrosion or degradation
of the containment wall over the projected design life of the plant.
NEI and NUBARG assert that ``[i]t is therefore hardly surprising that,
as noted in the Statement of Considerations, `[o]ver one-third of the
containments have experienced corrosion or other degradation.' ''
Therefore, they believe there is not a broad-based concern with
operability of containment structures. The NRC rejects the argument
that because containments have corrosion allowances and corrosion was
expected to occur that, ipso facto, further inspections are not
necessary and the compliance exception is inappropriate. As previously
pointed out, in many cases, the corrosion rate has been found to be
greater than that for which the containment was designed (in some cases
the rate was twice that predicted). Some of the more extreme cases of
wall thinning occurred in plants with corrosion allowances. The
existence of a corrosion allowance at any given plant is, of course
relevant, but only in the context of determining whether a relevant
requirement or commitment is likely to be violated during the OL term.
A corrosion allowance simply increases the tolerance (time period) for
corrosion. However, once the allowance is eroded, then concern with
compliance becomes relevant. Based upon the staff's finding of the
number and extent of corrosion to date, and the lack of activities to
manage the degradation by many licensees, the NRC concludes that it is
likely that those licensees will be in violation of applicable
requirements for containment structural integrity and leak-tightness
during the OL term, absent the imposition of Subsections IWE and IWL.
Because licensees have been unable to ensure compliance with current
regulatory requirements, the NRC believes that more specific ISI
requirements, which expand upon existing requirements for the
examination of containment structures in accordance with GDC 16, 53,
Appendix A to 10 CFR part 50, and Appendix J to 10 CFR part 50, are
needed and are justified for the purpose of ensuring that containments
continue to maintain or exceed minimum accepted design wall thicknesses
and prestressing forces as provided for in industry standards used to
design containments (e.g., Section III and Section VIII of the ASME
Code, and the American Concrete Institute Standard ACI-318), as
reflected in license conditions, technical specifications, and written
licensee commitments (e.g., the Final Safety Analysis Report). The NRC
believes that the occurrences of corrosion and other degradation would
have been detected by licensees when
[[Page 41307]]
conducting the periodic examinations set forth in Subsection IWE and
Subsection IWL.
With regard to subject area number four, six commenters believe
that the Boiling Water Reactors Owner's Group (BWROG) containment
inspection plan (CIP) will adequately address examinations for the
primary containment when used in conjunction with other existing
examination requirements such as Appendix J. The staff does not believe
that the CIP is a comprehensive containment examination program. In the
CIP, there is a comparison between the CIP and Subsection IWE. The CIP
dismisses seven of the eighteen identified Subsection IWE examinations
as not being justifiable even though some of these areas are likely to
experience accelerated corrosion. The CIP enumerates the conservatisms
and margins against failure in the design of Mark I and II containments
and concludes that in a typical plant probabilistic risk assessment of
failure, the contribution to failure of the containment steel structure
is negligible. The NRC believes that the conservatisms and margins
referred to are not additional tolerances which allow areas of
containments to go unexamined. These conservatisms and margins were
required allowances in the design because of the uncertainties in
loadings, in material properties, in analysis, and in the variation of
steel thicknesses. Examination of large areas of the containment cannot
be dismissed as being non-critical based on conservatisms and margins
when corrosion has clearly eroded the margin of safety in some cases.
In addition, given that only four of the 32 occurrences of corrosion in
metal containments and the liners of concrete containments were
detected during the pre-integrated leakage rate test examination, the
NRC does not believe that the CIP used in conjunction with other
existing examination requirements such as Appendix J will adequately
address examinations for the primary containment as asserted. The
industry initiative that allows a decrease in the frequency of Appendix
J leakage rate testing further erodes confidence in the acceptability
of the BWROG approach.
Comments were received from ten sources on proposed
Sec. 50.55a(g)(6)(ii)(B) which would require a 5-year expedited
examination schedule (subject area number five). Most of these comments
asked for clarifications of the NRC staff intent of this provision.
Some commenters interpreted this provision as a requirement to perform
all of the examinations specified for a 10-year interval in 5 years,
which was not the intent. Sec. 50.55a(g)(6)(ii)(B) has been changed to
clarify that for Subsection IWE, the baseline inspection will be the
inservice examinations which are to be performed during the first
period of the first interval. For Subsection IWL, the baseline
inspection will be the required inservice examinations which correspond
to the year of operation for each unit. The result of the clarification
is that Sec. 50.55a(g)(6)(ii)(B)(1) addresses Subsection IWE and
Sec. 50.55a(g)(6)(ii)(B)(2) addresses Subsection IWL.
Sec. 50.55a(g)(6)(ii)(B)(2) in the proposed rule has become
Sec. 50.55a(g)(6)(ii)(B)(3) and Sec. 50.55a(g)(6)(ii)(B)(3) has become
Sec. 50.55a(g)(6)(ii)(B)(4) in the final rule.
There was one additional comment submitted by NEI. The proposed
rule discussed NEI's (then NUMARC) position on the role of Subsection
IWE and Subsection IWL in license renewal. Subsections IWE and IWL were
referenced many times as one acceptable approach for managing age-
related degradation. The plan for managing age-related degradation
assumes that these examinations are ``in current and effective use.''
NEI commented on the above statements in the proposed rule; ``Although
the BWR and PWR containment IRs [Industry Reports] do reference
Subsections IWE and IWL, their identification in the IRs should not be
misrepresented to imply that Subsections IWE and IWL are being
implemented or that they are required for operating plants during their
initial licensing term.'' The NRC agrees that the IRs were not to be
represented as a requirement for operating licensees to implement
Subsection IWE and Subsection IWL or their equivalent, and that these
subsections were referenced as one acceptable approach of managing age-
related degradation for the license renewal period. However, present
licensee containment examination programs have not proved to be
effective in detecting the types of degradation which have been
reported. The number of occurrences and the extent of degradation
(which includes cases of noncompliance) leads to the conclusion that
additional requirements are needed for managing containment degradation
during the operating term. Because Subsections IWE and IWL were
developed by the ASME with industry input and found to be acceptable by
NEI for managing age-related degradation for the license renewal
period, the NRC believes that adoption of those programs at this time
is the best approach. The NRC also believes that with implementation of
Subsections IWE and IWL, the detrimental effects of containment aging
will be managed during the current operating term, as well as during
the license renewal term.
As a result of the comments received, there is one editorial
change, two clarifications, and four modifications in the final rule.
With respect to the editorial change, a commenter suggested that the
wording of Sec. 50.55a(b)(2)(ix)(D)(2) in the proposed rule be revised
to be consistent with Sec. 50.55a(b)(2)(ix)(D)(1) and
Sec. 50.55a(b)(2)(ix)(D)(3) of the same paragraph.
Sec. 50.55a(b)(2)(ix)(D) addresses the sampling of the grease contained
in post-tensioning systems, and conditions, which if found, are
reportable. The suggested wording has been adopted in the final rule.
One of the clarifications was to proposed Sec. 50.55(g)(6)(ii)(B).
This change was discussed previously in subject area number five.
Sec. 50.55a(g)(6)(ii)(B)(1) and Sec. 50.55a(g)(6)(ii)(B)(2) require
that licensees conduct the first containment examinations in accordance
with Subsection IWE and Subsection IWL (1992 Edition with the 1992
Addenda), modified by Sec. 50.55a(b)(2)(ix) and Sec. 50.55a(b)(2)(x)
within 5 years of the effective date of the final rule. This expedited
examination schedule is necessary to prevent possible delays in the
implementation of Subsection IWE by as much as 20 years and Subsection
IWL by as much as 15 years. Subsection IWE, Table IWE-2500-1, permits
the deferral of many of the required examinations until the end of the
10-year inspection interval. Adding the 10 years that could pass before
some utilities are required to update their ISI plans, a period of 20
years could pass before the first examinations would take place.
Subsection IWL is based on a 5-year inspection interval. Adding the
possible 10 years before update of existing ISI plans, a period of 15
years could pass before the examinations were performed by plants that
have not voluntarily adopted the provisions of Regulatory Guide 1.35,
Rev. 3. Expediting implementation of the containment examinations is
considered necessary because of the problems that have been identified
at various plants, the need to establish expeditiously a baseline for
each facility, and the need to identify any existing degradation.
Paragraphs (g)(6)(ii)(B)(3) and (g)(6)(ii)(B)(4) each provide a
mechanism for licensees to satisfy the requirements of the routine
containment examinations and the expedited examination without
duplication. Paragraph (g)(6)(ii)(B)(3) permits licensees to avoid
duplicating
[[Page 41308]]
examinations required by both the periodic routine and expedited
examination programs. This provision is intended to be useful to those
licensees that would be required to implement the expedited examination
during the first periodic interval that routine containment
examinations are required. Paragraph (g)(6)(ii)(B)(4) allows licensees
to use a recently performed examination of the post-tensioning system
to satisfy the requirements for the expedited examination of the
containment post-tensioning system. This situation would occur for
licensees who perform an examination of the post-tensioning system
using Regulatory Guide 1.35 between the effective date of this rule and
the beginning of the expedited examination.
The four modifications are: (1) Sec. 50.55a(b)(2)(x)(A) expands the
evaluation of inaccessible areas of concrete containments (Class CC) to
metal containments and the liners of concrete containments (Class MC);
(2) Sec. 50.55a(b)(2)(x)(B) permits alternative lighting and resolution
requirements for remote visual examination of the containment; (3)
Sec. 50.55a(b)(2)(x)(C) makes the examination of pressure retaining
welds and pressure retaining dissimilar metal welds optional; and (4)
Sec. 50.55a(b)(2)(x)(D) has been added to provide an alternative
sampling plan. Section 50.55a(b)(2)(x)(E), a clarification, more
clearly defines the frequency of the Subsection IWE general visual
examination.
The first modification, Sec. 50.55a(b)(2)(x)(A), which expands the
evaluation of inaccessible areas of concrete containments (Class CC) to
metal containments and the liners of concrete containments (Class MC),
was the result of a comment received on Sec. 50.55a(b)(2)(ix)(E) of the
proposed rule. The commenter believed that given the number of
occurrences of corrosion in Class MC containments, the proposed
provision (which only addressed concrete containments) should be
expanded in the final rule to include metal containments and the liners
of concrete containments.
The second modification, Sec. 50.55a(b)(2)(x)(B), was added to the
final rule to permit alternative lighting and resolution requirements
for remote visual examination of the containment. Subsection IWE
references the lighting and resolution requirements contained in IWA-
2200. The lighting and resolution requirements contained in IWA-2200
would on a practical basis preclude remote containment examination.
The third modification, Sec. 50.55a(b)(2)(x)(C), makes the
examinations of Subsection IWE, Examination Category E-B (pressure
retaining welds) and Subsection IWE, Examination Category E-F (pressure
retaining dissimilar metal welds) optional. The NRC staff concludes
that requiring these examinations is not appropriate. There is no
evidence of problems associated with welds of this type under the given
operating conditions. In addition, the occupational radiation exposure
that would be incurred while performing these examinations cannot be
justified. It is estimated that the total occupational exposure that
would be incurred yearly in the performance of the containment weld
examinations in accordance with Examination Categories E-B and E-F
would be 440 person-rems.
The fourth modification, Sec. 50.55a(b)(2)(x)(D), provides an
alternative to the ASME Section XI requirements for ``additional
examinations'' (note: additional examinations'' are required during the
same outage when acceptance criteria are exceeded). The alternative
would allow licensees to determine the number of additional components
to be examined based on an evaluation to determine the extent and
nature of the degradation. Five commenters believe that the
requirements for additional examinations used in other subsections of
Section XI is inappropriate for containment components. Additional
examinations are incorporated into Section XI to determine the extent
to which degradation found in one component exists in other similar
components. In some instances, a large number of additional
examinations could be required. The commenters believe that a review of
the operational history of containment components shows that the
degradation is limited to the area in question and is not widespread.
This makes the Section XI requirements for additional examinations
burdensome and inappropriate for application to containments. The NRC
agrees and revised the rule to permit the alternative to the Section XI
requirements for additional examinations.
The NRC believes that these modifications improve the final rule
and will improve the containment inspection program as set forth by
Subsection IWE and Subsection IWL. Some of the public comments cited
failure data which have been accumulated in recent years in support of
various NRC staff activities and industry initiatives. Most of this
data has been accumulated since the ASME committees developed these
subsections. Without the benefit of this recently accumulated
operational data, the ASME committees responsible for developing
Subsection IWE and Subsection IWL modelled those subsections on other
subsections of Section XI and the experience gained from application of
those other subsections. With the additional insights drawn from
analysis of this new data, it is apparent that many aspects of
containments are unique compared to components of other systems. Some
of the containment components which were expected to experience
degradation, based on experience with other systems, have proved not to
be susceptible to the same type of degradation. The ASME working groups
are considering these issues. However, based on initial committee
discussion, it is anticipated that similar changes will be made to
Subsection IWE and Subsection IWL, but the length of the ASME consensus
process precludes the possibility of the changes being adopted into the
ASME Code in the near term. Hence, the NRC has determined to adopt the
1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL
with the modifications which were previously discussed.
Other Provisions Contained in the Final Rule
The following paragraph was contained in the proposed rule and has
not been discussed previously. This paragraph received comments which
resulted in the provision being dropped in the final rule. Section
50.55a(b)(2)(x) was a provision in the proposed rule intended to
provide licensees with a mechanism to merge the Subsection IWE and
Subsection IWL ISI program with their routine 120-month ISI program.
Those licensees who were near the end of their present 10-year ISI
interval when the final rule becomes effective would have been given an
additional 2 years to submit their containment ISI program. Several
commenters responded that due to the time constraints of having to
develop the containment ISI program and then perform the required
examinations within 5 years, the additional 2 years could not be
utilized. Therefore, Sec. 50.55a(b)(2)(x) as it appeared in the
proposed rule has been deleted, and Sec. 50.55a(b)(2)(x) in the final
rule contains the modifications which were added as a result of public
comment on the proposed rule.
The provisions in this paragraph and the following four paragraphs
were contained in the proposed rule and have not changed due to
comments. Section 50.55a(b)(2)(vi) incorporates a limitation specifying
the 1992 Edition with 1992
[[Page 41309]]
Addenda of Subsection IWE and Subsection IWL as the earliest ASME Code
version the NRC finds acceptable. This edition and addenda incorporate
the concept of base metal examinations and also provide a comprehensive
set of rules for the examination of post-tensioning systems. It should
be noted that the wording of this provision has been changed in the
final rule in order to make it consistent with other provisions in
Sec. 50.55a(b).
Section 50.55a(b)(2)(ix) specifies five modifications that must be
implemented when using Subsection IWL. Four of these issues are
identified in Regulatory Guide 1.35, Revision 3, but are not currently
addressed in Subsection IWL. Section 50.55a(b)(2)(ix)(A) requires that
grease caps which are accessible must be visually examined to detect
grease leakage or grease cap deformation. Section 50.55a(b)(2)(ix)(B)
requires the preparation of an Engineering Evaluation Report when
consecutive surveillances indicate a trend of prestress loss to below
the minimum prestress requirements. Section 50.55a(b)(2)(ix)(C)
requires an evaluation to be performed for instances of wire failure
and slip of wires in anchorages. Section 50.55a(b)(2)(ix)(D) addresses
sampled sheathing filler grease and reportable conditions. A comment
was received on this provision which resulted in an editorial change
(this was discussed on page 12). Section 50.55a(b)(2)(ix)(E) requires
that licensees evaluate the acceptability of inaccessible areas of
concrete containments when conditions exist in accessible areas that
suggest the possibility of degradation in inaccessible areas.
Existing Sec. 50.55a(g), ``Inservice inspection requirements,''
specifies the requirements for preservice and inservice examinations
for Class 1 (Class 1 refers to components of the reactor coolant
pressure boundary), Class 2 (Class 2 quality standards are applied to
water- and steam-containing pressure vessels, heat exchangers (other
than turbines and condensers), storage tanks, piping, pumps, and valves
that are part of the reactor coolant pressure boundary (e.g., systems
designed for residual heat removal and emergency core cooling)), and
Class 3 (Class 3 quality standards are applied to radioactive-waste-
containing pressure vessels, heat exchangers (other than turbines and
condensers), storage tanks, piping, pumps, and valves (not part of the
reactor coolant pressure boundary)) components and their supports.
Subsection IWE (Class MC--metal containments) and Subsection IWL (Class
CC--concrete containments) are incorporated by reference into the NRC
regulations for the first time.
Section 50.55a(g)(4) specifies the containment components to which
the ASME Code Class MC and Class CC inservice inspection
classifications incorporated by reference in this rule will apply.
Section 50.55a (g)(4)(v)(A), (v)(B), and (v)(C) specify the
Subsection IWE and Subsection IWL rules for inservice inspection,
repair, and replacement of metal and concrete containments. This is
consistent with the long-standing intent and ongoing application by NRC
and licensees to utilize the rules of Section XI when performing
inservice inspection, repairs, and replacements of applicable
components and their supports.
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.
Finding of No Significant Environmental Impact
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
subpart A of 10 CFR part 51, that this rule is not a major Federal
action that significantly affects the quality of the human environment
and therefore an environmental impact statement is not required.
This final rule is one part of a regulatory framework directed to
ensuring containment integrity. Therefore, in the general sense, this
rule will have a positive impact on the environment. This rule
incorporates by reference into the NRC regulations requirements
contained in the ASME Code for the inservice inspection of the
containments of nuclear power plants. The performance of containment
examinations, as set forth by the provisions of this final rule, for
PWRs, Ice Condensers, and BWR Mark IIs and IIIs is not expected to
result in significant occupational radiation exposure (1.0 person-rems
per year or 0.04 person-rems per unit averaged over 27 examinations
each year). The above categories of plants, for which the occupational
radiation exposure is insignificant, represent the vast majority of
units (89). For BWR Mark I containments, the estimated occupational
radiation exposure which would be incurred per year while performing
BWR Mark I containment examination is 29.4 person-rems per year or 4.2
person-rems per unit averaged over 7 examinations per year. However,
the estimated occupational radiation exposure per unit does not provide
an accurate representation of the actual radiological exposure that
would be incurred by any one individual. 10 CFR 20.101, ``Radiation
dose standards for individuals in restricted areas'' only permits a
whole body dose of 1.25 rem per calendar quarter. As a practical
matter, licensees carefully manage the exposure incurred by any one
individual by practicing and applying ``as low as reasonably
achievable'' (ALARA) principles to protect the health and safety of
personnel. In the performance of the examination of BWR Mark I
containments, this is accomplished by having several individuals
perform the examinations to ``spread out'' the exposure. In this
manner, no one individual will suffer any significant health effects.
It also must be kept in mind that these containment examinations are
scheduled to occur at the interval of once every 3\1/3\ years. This
provides licensees ample time for planning the examinations, and
scheduling personnel in accord with ALARA considerations. Therefore,
the occupational radiation exposure is insignificant given the
relatively low exposure on a unit basis and the licensees' programs for
controlling the impact of exposure for any one individual.
Actions required of applicants and licensees to implement
containment examinations are of the same nature that applicants and
licensees have been performing for many years in other Section XI ISI
programs. Extension of these actions to additional components,
therefore, should not increase the potential for a negative
environmental impact.
The environmental assessment and finding of no significant impact
on which this determination is based are available for inspection at
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the environmental assessment and the
finding of no significant impact are available from Mr. W. E. Norris,
Division of Engineering Technology, Office of Nuclear Regulatory
Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555,
telephone (301) 415-6796.
Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject
[[Page 41310]]
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 collection of information is
estimated to average 4,000 hours per response for development of an
initial inservice inspection plan, and 8,000 hours per response for the
update of the plan and periodic examinations, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
collection of information. The estimate of 8,000 hours for plan update
and performing periodic examinations is a 2,000 hour reduction from the
estimate given in the proposed rulemaking. This reduction results from
changes made in response to public comment. A number of examinations
have been modified or made optional greatly reducing the effort
required to comply with the requirements contained in the final rule.
Send comments on any aspect of this collection of information,
including suggestions for reducing the burden, to the Information and
Records Management Branch (T-6 F33), U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, or by Internet electronic mail
at [email protected]; and to the Desk Officer, Office of Information and
Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and
Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless it displays a currently
valid OMB control number.
Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission hereby certifies that this rule will not have a
significant economic impact on a substantial number of small entities.
This rule affects only the 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. Since these companies are dominant in their service areas, this
rule does not fall within the purview of the Act.
Backfit Statement
The NRC is amending its regulations to incorporate by reference the
1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL
to assure that the critical areas of containments are routinely
inspected to detect defects that could compromise a containment's
structural integrity. Based on a preponderance of reliable information,
the NRC concludes that this rule is a compliance backfit, and therefore
a backfit analysis is not required pursuant to 10 CFR 50.109(a)(4)(i).
A summary of noncompliance is set forth below. The documented
evaluation required by Sec. 50.109(a)(4) to support this conclusion 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 Mr. W.E. Norris, Division of Engineering Technology,
Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, telephone (301) 415-6796.
The rate of occurrence of corrosion and degradation of containment
structures has been increasing at operating nuclear power plants. There
have been 32 reported occurrences of corrosion in metal containments
and the liners of concrete containments. This is approximately one-
fourth of all operating nuclear power plants. Only four of the 32
occurrences were detected by current licensee containment inspection
programs. Nine of these occurrences were first identified by the NRC
through its inspections or structural audits. Eleven occurrences were
detected by licensees after they were alerted to a degraded condition
at another site or through activity other than containment inspection.
There have been 34 reported occurrences of degradation of the concrete
or of the post-tensioning systems of concrete containments. This is
nearly one-half of these types of containments. It is clear that
current licensee containment inspection programs have not proved to be
adequate to detect the types of degradation which have been reported.
Examples of degradation not found by licensees, but initially detected
at plants through NRC inspections include: (1) Corrosion of steel
containment shells in the drywell sand cushion region, resulting in
wall thickness reduction to below the minimum design thickness; (2)
corrosion of the torus of the steel containment shell (wall thickness
below minimum design thickness); (3) extensive corrosion of the liner
of a concrete containment with local degradation at many locations to
approximately half-depth; (4) grease leakage from the tendons of
prestressed concrete containments; and (5) leaching as well as
excessive cracking in concrete containments.
None of the existing requirements for containment inspection
provide specific guidance on how to perform the necessary containment
examinations. This lack of guidance has resulted in a large variation
with regard to the performance and the effectiveness of licensee
containment examination programs. Based on the results of inspections
and audits, and plant operational experiences, it is clear that many
licensee containment examination programs have not detected degradation
that could result in a compromise of pressure-retaining capability.
Most of those occurrences were first identified by the NRC through
its inspections or audits of plant structures, or by licensees while
performing an unrelated activity or, after they were alerted to a
degraded condition at another site. In analyzing the reported
containment degradation, it is apparent that all containments are
subject to certain type(s) of degradation depending on the design.
Information gathered by the staff indicates that many licensees still
have not reacted to this serious safety concern and have not initiated
comprehensive containment inservice inspection. As a result of the rate
of occurrence of containment degradation, and the extent of containment
degradation, the NRC believes that there is a basis for reasonably
concluding that such degradation is widespread and affects virtually
all plants. Because of the serious degradation which has occurred, the
belief that additional occurrences of noncompliance with required
minimum wall thicknesses and prestressing forces will be reported, and
the high likelihood that some of those occurrences could result in loss
of structural integrity and leak-tightness, the NRC has determined that
imposition of these containment inservice inspection requirements under
the compliance exception to 10 CFR 50.109(a)(4)(i) is appropriate.
The NRC believes that the final action would also result in a
substantial safety increase and that the direct and indirect costs of
implementation are justified in view of the significant safety benefit
to be gained. The NRC believes that the inspections contained in
Subsections IWE and IWL will improve significantly the ability to
detect degradation and take timely action to correct degradation of
containment structures. A review of early implementation of the
maintenance rule (10 CFR 50.65) at nine
[[Page 41311]]
nuclear power plants, which is documented in NUREG-1526, indicates that
most licensees assigned a low priority to the monitoring of structures.
Several licensees incorrectly assumed that many of their structures are
inherently reliable. This is true so long as there is no degradation.
However, the degradation of structures can reduce high margins of
safety to a low or negligible margin of safety. As discussed earlier,
such substantial containment degradations have been detected at a large
number of nuclear power plants, and their detection to date can best be
characterized as happenstance. The final rule will provide for improved
periodic examination of containment structures assuring that the
critical areas of containment are periodically inspected to detect and
take corrective action for defects that could compromise the
containment's pressure-retaining and leak-tight capability. The NRC
believes, therefore, that the final action can be justified as a cost-
justified safety enhancement backfit, as well as a compliance backfit.
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. 533, the NRC is adopting the
following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 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 by adding paragraphs (b)(2)(vi),
(b)(2)(ix), (b)(2)(x), (g)(4)(v), and (g)(6)(ii)(B), and revising the
introductory text of paragraphs (b)(2) and (g)(4) to read as follows:
Sec. 50.55a Codes and standards.
* * * * *
(b) * * *
(2) As used in this section, references to Section XI of the ASME
Boiler and Pressure Vessel Code refer to Class 1, Class 2, and Class 3
components of Section XI, Division 1, and include addenda through the
1988 Addenda and editions through the 1989 Edition, and Class MC and
Class CC components of Section XI, Division 1, 1992 Edition with the
1992 Addenda, subject to the following limitations and modifications:
* * * * *
(vi) Effective edition and addenda of Subsection IWE and Subsection
IWL, Section XI. The 1992 Edition with the 1992 Addenda of Subsection
IWE and Subsection IWL shall be used by licensees when performing
containment examinations as modified and supplemented by the
requirements in Sec. 50.55a(b)(2)(ix) and Sec. 50.55a(b)(2)(x).
* * * * *
(ix) Examination of concrete containments. (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 shall 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 anchorages. 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.
(x) 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 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
[[Page 41312]]
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)(x)(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
shall 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
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
shall be performed once each period.
* * * * *
(g) * * *
(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) and the modifications listed
in paragraphs (b)(2)(ix) and (b)(2)(x) of this section, to the extent
practical within the limitations of design, geometry and materials of
construction of the components.
* * * * *
(v) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit was issued after January 1, 1956:
(A) Metal containment pressure retaining components and their
integral attachments must meet the inservice inspection, repair, and
replacement requirements applicable to components which are classified
as ASME Code Class MC;
(B) Metallic shell and penetration liners which are pressure
retaining components and their integral attachments in concrete
containments must meet the inservice inspection, repair, and
replacement requirements applicable to components which are classified
as ASME Code Class MC; and
(C) Concrete containment pressure retaining components and their
integral attachments, and the post-tensioning systems of concrete
containments must meet the inservice inspection and repair 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 shall 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)(ix) by September 9, 2001. The first examination
performed shall serve the same purpose for operating plants as the
preservice examination specified for plants not yet in operation.
(3) The expedited examination for Class MC components may be used
to satisfy the requirements of routinely scheduled examinations of
Subsection IWE subject to IWA-2430(d) when the expedited examination
occurs during the first containment inspection interval.
(4) The requirement for the expedited examination of the
containment post-tensioning system may be satisfied by the post-
tensioning system examinations performed after September 9, 1996 as a
result of licensee post-tensioning system programs accepted by the NRC
prior to September 9, 1996.
(5) Licensees do not have to submit to the NRC staff for approval
of their containment inservice inspection program which was developed
to satisfy the requirements of Subsection IWE and Subsection IWL with
specified modifications and a limitation. The program elements and the
required documentation shall be maintained on site for audit.
* * * * *
Dated at Rockville, Maryland, this 12th day of June 1996.
For the Nuclear Regulatory Commission.
James M. Taylor,
Executive Director for Operations.
[FR Doc. 96-20215 Filed 8-7-96; 8:45 am]
BILLING CODE 7590-01-P