[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