[Federal Register Volume 59, Number 5 (Friday, January 7, 1994)]
[Proposed Rules]
[Pages 979-984]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-341]


[[Page Unknown]]

[Federal Register: January 7, 1994]


                                                     VOL. 59, NO. 5

                                            Friday, January 7, 1994
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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: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) proposes to amend 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 would be required to incorporate Subsection IWE 
and Subsection IWL into their routine inservice inspection (ISI) 
program. Licensees would also be required to expedite implementation of 
the containment examinations and 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 proposed that would 
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. This proposed amendment would specify 
requirements to assure that the critical areas of containments are 
routinely inspected to detect defects that could compromise a 
containment's pressure-retaining integrity.

DATES: Comment period expires March 23, 1994. Comments received after 
this date will be considered if it is practical to do so, but assurance 
of consideration cannot be given except as to comments received on or 
before this date.

ADDRESSES: Written comments or suggestions may be submitted to the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, Attention: Docketing and Service Branch. Deliver 
comments to: 11555 Rockville Pike, Rockville, MD between 7:45 am and 
4:15 pm Federal workdays. Copies of the regulatory analysis, the 
environmental assessment and finding of no significant impact, the 
supporting statement submitted to the Office of Management and Budget, 
and comments received may be examined in the Commission's Public 
Document Room at 2120 L Street, NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT: Mr. W.E. Norris, Division of 
Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, telephone (301) 492-3805, 
or Mr. H.L. Graves, Division of Engineering, Office of Nuclear 
Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 
20555, telephone (301) 492-3813.

SUPPLEMENTARY INFORMATION:

Background

    The NRC is taking the proposed action 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 licensee commitments (e.g., the Final Safety 
Analysis Report). The NRC also believes enhanced ISI examinations are 
needed and are justified to supplement existing requirements specified 
in General Design Criterion (GDC) 16, and GDC 53, appendix A to 10 CFR 
part 50, and appendix J to 10 CFR part 50. Appendix J 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 inspections. In view 
of the increasing rate of occurrences of degradation in containments 
and variability of present containment examinations, the NRC has 
determined that it is necessary to include more detailed requirements 
for the periodic examination of containment structures in the 
regulations to assure that the critical areas of containments are 
periodically inspected to detect defects that could compromise the 
containment's pressure-retaining and leak-tight capability. Recent 
changes and additions to the ASME Code include provisions to address 
the concerns outlined above. The NRC proposes to make these provisions 
mandatory by amending 10 CFR 50.55a to incorporate by reference these 
additional portions of the ASME Code (Subsection IWE and Subsection 
IWL). Subsection IWE and Subsection IWL have not been previously 
incorporated by reference into the NRC's regulations.
    The rate of occurrence of corrosion and degradation of containments 
has been increasing at operating nuclear power plants. Since 1986, 
twenty-one (21) instances of corrosion in steel containments have been 
reported. In two cases, thickness measurements of the walls revealed 
areas where the wall thickness was at or below the minimum design 
thickness. Since the early 1970s, thirty-one (31) incidents of 
containment degradation related to post-tensioning systems of concrete 
containments have been reported. Four recent additional incidents which 
involved grease leakage from tendons have been investigated. In 
addition to grease leakage, these incidents showed signs of leaching of 
the concrete.
    Over one-third of the operating containments have experienced 
corrosion or other degradation. Almost one-half of these occurrences 
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. Examples of degradation not found by 
licensees, but initially detected at plants through NRC inspections 
include: Steel containment shell corrosion in the drywell sand cushion 
region (wall thickness reduced to below minimum design thickness); 
steel containment shell torus corrosion (wall thickness at or near 
minimum design thickness); grease leakage from the tendons of 
prestressed concrete containments, and water seepage, as well as 
concrete cracking in concrete containments.
    There are several GDC criteria 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. 
Criterion 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. Section III and Section VIII of the 
ASME Code, and the American Concrete Institute provide design 
specifications for minimum wall thicknesses and prestressing forces of 
containments, and these are reflected in license conditions, technical 
specifications, and licensee commitments for the operating plants.
    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 V. 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). None of these existing requirements, 
however, provide specific guidance on how to perform the necessary 
containment examinations. This lack of guidance has resulted in a large 
variation in licensee containment examination programs, such that there 
have been cases of noncompliance with GDC 16. 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 
also been found to have undergone a significant level of degradation 
that was not detected by these programs.
    The NRC believes that more specific ISI requirements, which expand 
upon existing requirements for the examination of containment 
structures in accordance with GDC 53 and appendix J , are needed and 
are justified for the purpose of ensuring that containments continue to 
maintain minimum 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). There exists a serious 
concern, based on actual operating experience, regarding continued 
compliance by the operating plants with existing requirements for 
ensuring containment minimum design wall thicknesses and prestressing 
forces if the proposed action is not taken. The NRC also believes that 
the occurrences of corrosion and other degradation discussed above 
would have been detected by licensees implementing the comprehensive 
periodic examinations set forth in Subsection IWE and Subsection IWL of 
the ASME Code proposed for incorporation by reference into 10 CFR 
50.55a.
    The Nuclear Management and Resources Council (NUMARC) has developed 
a number of industry reports to address license renewal issues. Two of 
them, one for PWR containments and the other for 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 five modifications, which are contained in one paragraph of the 
proposed rule, address two concerns of the NRC. The first concern is 
that certain recommendations for tendon examinations that are included 
in Regulatory Guide 1.35, Rev. 3, are not addressed in Subsection IWL 
(this involves four of the modifications, (ix)(A)-(D)). The ASME Code 
has considered these four issues and has adopted them in Subsection 
IWL. These issues will be published in future addenda. The second 
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. This fifth modification ((ix)(E)) 
contains a provision which would require an evaluation of inaccessible 
areas when visible conditions exist that could result in degradation of 
these areas.
    The limitation specifies the 1992 Edition with 1992 Addenda of 
Subsection IWE and Subsection IWL as the earliest version of the ASME 
Code the NRC finds acceptable. This edition and addenda combination 
incorporates the concept of base metal examinations and would provide a 
comprehensive set of rules for the examination of post-tensioning 
systems. As originally published, 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, all have furthered the incorporation of base metal 
examinations.
    The proposed rulemaking incorporates a provision for an expedited 
examination schedule. This expedited examination schedule is necessary 
to prevent a delay in the implementation of Subsection IWE and 
Subsection IWL (Table 4 of Enclosure 2 lists each plant and the delay 
in implementation which would be encountered without an expedited 
implementation schedule). Provisions have been incorporated in the 
proposed rule so that the expedited examination which would be required 
5 years after the effective date of the rule and the routine 120-month 
examinations are not duplicated.
    The NRC has reviewed the 1992 Edition with the 1992 Addenda of 
Subsection IWE and Subsection IWL of Section XI of the ASME Code and 
has found that with the specified modifications these subsections of 
Section XI address current experience and provide a sound basis for 
ensuring the structural integrity of containments. NRC endorsement of 
Subsection IWE and Subsection IWL in its regulations would provide a 
method of improving containment examination practices by incorporating 
rules into the regulatory process that have received industry 
participation in their development and acceptance by the NRC.
    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. 
Neither Subsection IWE (Class MC--metal containments) nor Subsection 
IWL (Class CC--concrete containments) is presently incorporated by 
reference into the NRC regulations.
    Proposed Sec. 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 proposed rule would 
apply.
    Proposed Sec. 50.55a (g)(4) (v)(A), (v)(B), and (v)(C) specify 
Subsection IWE and Subsection IWL rules for repairs and replacements 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 repairs and replacements of 
applicable components and their supports.
    Proposed Sec. 50.55a(b)(2)(vi) would incorporate a limitation 
specifying the 1992 Edition with 1992 Addenda of Subsection IWE and 
Subsection IWL as the earliest ASME Code version the NRC finds 
acceptable. This edition and addenda combination incorporates the 
concept of base metal examinations and provides a comprehensive set of 
rules for the examination of post-tensioning systems.
    Proposed Sec. 50.55a(b)(2)(ix) would specify 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.
    Proposed Sec. 50.55a(g)(4)(v) requires that licensees incorporate 
containment examinations as part of their routine 120-month inspection 
program. It is recognized that when this rule becomes effective, plants 
within 2 years of the end of the 120-month interval may have difficulty 
developing and completing the containment examination program in a 
timely manner. Therefore, proposed Sec. 50.55a(b)(2)(x) specifies that 
licensees with less than 2 years remaining in their present ISI 
interval may complete the Subsection IWE and the Subsection IWL 
portions of their ISI update within 2 years from the end of the present 
ISI interval. This is intended to provide licensees with sufficient 
time to develop the initial ISI plan and to facilitate maintenance of 
one ISI plan instead of two separate plans (i.e, the current Section XI 
ISI plan, and the Subsection IWE and Subsection IWL plan). In order to 
further reduce the burden on licensees and NRC staff, the Subsection 
IWE and Subsection IWL portions of the ISI plan will not have to be 
submitted to the NRC for approval. Licensees may simply retain their 
initial Subsection IWE and Subsection IWL plans at the site for audit.
    Proposed Sec. 50.55a(g)(6)(ii)(B)(1) would require that licensees 
conduct the first containment examinations in accordance with 
Subsection IWE and Subsection IWL (1992 Edition with the 1992 Addenda), 
modified by proposed Sec. 50.55a(b)(2)(ix) 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 
most of the required examinations until the end of the 10-year 
inspection interval. Adding the ten 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.
    Proposed paragraphs (g)(6)(ii)(B)(2) and (g)(6)(ii)(B)(3) would 
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)(2) would permit licensees 
to avoid duplicating 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)(3) would 
allow 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.

Submission of Comments in Electronic Format

    The comment evaluation process will be improved if each comment is 
identified with document title, section heading, and paragraph number 
addressed. In addition to the original paper copy, submitters are 
encouraged to provide a copy of their letter in an electronic format on 
IBM PC compatible 3.5- or 5.25-inch diskettes. Data files should be 
provided as WordPerfect documents. ASCII text is also acceptable or, if 
formatted text is required, data files should be provided in IBM 
Revisable-Form Text/Document Content Architecture (RFT/DCA) format. The 
format and version should be identified on the diskette's external 
label.

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, if adopted, would not be a 
major Federal action significantly affecting the quality of the human 
environment and therefore an environmental impact statement is not 
required.
    This proposed rule is one part of a regulatory framework directed 
to ensuring containment integrity. Therefore, in the general sense, the 
proposed rule would have a positive impact on the environment. The 
proposed rule would incorporate by reference in the NRC regulations 
requirements contained in the ASME Code for the inservice inspection of 
the containments of nuclear power plants. Actions required of 
applicants and licensees to implement the proposed rule are of a 
routine nature that 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, Office of Nuclear Regulatory Research, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 
492-3805, or Mr. H.L. Graves, Division of Engineering, Office of 
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, telephone (301) 492-3813.

Paperwork Reduction Act Statement

    This proposed rule amends information collection requirements that 
are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et 
seq). This rule has been submitted to the Office of Management and 
Budget for review and approval of the paperwork requirements.
    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 10,000 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. Send comments regarding this burden estimate 
or any other aspect of this collection of information, including 
suggestions for reducing this burden, to the Information and Records 
Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the Desk Officer, Office of 
Information and Regulatory Affairs, NEOB-3019, (3150-0011), Office of 
Management and Budget, Washington, DC 20503.

Documented Evaluation

    The Commission has prepared a draft summary of documented 
evaluation on this proposed regulation. The draft evaluation 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, Office of 
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, telephone (301)492-3805, or from Mr. H.L. Graves, 
Division of Engineering, Office of Nuclear Regulatory Research, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, telephone 
(301)492-3813.
    The Commission requests public comment on the draft summary of 
documented evaluation. Comments on the draft evaluation may be 
submitted to the NRC as indicated under the ADDRESSES heading.

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, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This proposed 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 proposed rule does not fall within the purview of the Act.

Backfit Statement

    The NRC is taking the proposed action for the purpose of ensuring 
that containment structures continue to maintain or exceed minimum 
accepted design wall thicknesses and prestressing forces as provided 
for in industry standards used to design containment structures, as 
reflected in license conditions, technical specifications, and licensee 
commitments. Therefore, under 10 CFR 50.109(a)(4)(i) a backfit analysis 
need not be prepared for this rule. A summary of the documented 
evaluation required by Sec. 50.109(a)(4) to support this conclusion is 
set forth below.
    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 V. 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). None of these existing requirements, 
however, provide specific guidance on how to perform the necessary 
containment examinations. This lack of guidance has resulted in a large 
variation in licensee containment examination programs, such that there 
have been cases of noncompliance with GDC 16. 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. The location and extent of corrosion or degradation in a 
containment can be critical to the containment's behavior during an 
accident.
    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 determined so 
that the metal containment structure will continue to maintain its 
structural integrity under the various stressors and degradation 
mechanisms which act on it.
    The American Concrete Institute Standard ACI-318 contains 
provisions for designing and constructing the post-tensioning systems 
of concrete containment structures, including methods for 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.
    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).
    The rate of occurrence of corrosion and degradation of containment 
structures has been increasing at operating nuclear power plants. Over 
one-third of operating containment structures have experienced 
corrosion or other degradation. Almost one-half of the occurrences were 
first identified by the NRC through its inspections or structural 
audits, or by licensees because they were alerted to a degraded 
condition at another site. 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 reduced to below the 
minimum design thickness; (2) corrosion of the torus of the steel 
containment shell (wall thickness at or near minimum design thickness); 
(3) grease leakage from the tendons of prestressed concrete 
containments; and (4) water seepage, as well as concrete cracking in 
concrete containments.
    The NRC believes that more specific ISI requirements, that expand 
upon existing requirements for the examination of containment 
structures in accordance with GDC 53, and appendix J are needed and are 
justified to ensure that containment structures continue to maintain or 
exceed minimum accepted design wall thicknesses and prestressing forces 
as reflected in license conditions, technical specifications, or 
licensee commitments. Based on actual operating experience, a serious 
concern exists regarding continued compliance by the operating plants 
with existing requirements for ensuring containment minimum design wall 
thicknesses and prestressing forces if the proposed action is not 
taken. The NRC also believes that the occurrences of corrosion and 
other degradation discussed above would have been detected by licensees 
when conducting the comprehensive periodic examinations set forth in 
Subsection IWE and Subsection IWL of the ASME Code, as proposed for 
incorporation by reference into 10 CFR 50.55a.
    Recent changes and additions to the ASME Code include provisions to 
address the concerns outlined above; and the staff proposes to make 
these provisions mandatory by amending 10 CFR 50.55a to incorporate by 
reference these additional portions of the ASME Code (Subsection IWE 
and Subsection IWL). The Commission concludes that this proposed 
backfit is necessary to ensure compliance with GDCs 16 and 53, appendix 
J, minimum design wall thicknesses in metal containments, and the 
prestressing forces of concrete containments, which are applicable to 
all licensees through license conditions, technical specifications, and 
licensee commitments.

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 proposing to 
adopt 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. 1244, 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. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, 
Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 
50.54(dd) and 50.103 also issued under sec. 108, 68 Stat. 939, as 
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 
50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also 
issued under 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 paragraph (g)(4) to read as follows:


Sec. 50.55a  Codes and standards.

* * * * *
    (b) * * *
    (2) * * *
    (vi) Effective edition and addenda of Subsection IWE and Subsection 
IWL, Section XI. When using Subsection IWE and Subsection IWL, the 1992 
Edition with the 1992 Addenda is the only acceptable Edition and 
Addenda.
* * * * *
    (ix) Examination of concrete containments. 
    (A) All 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) An Engineering Evaluation Report must be prepared as prescribed 
in IWL-3300(a), (b), (c), and (d) 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.
    (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.
    (D) The licensee shall identify the following conditions, if they 
occur, in the ISI Summary Report:
    (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 may not exceed 10 percent of the tendon net duct 
volume.
    (3) Grease leakage is detected during general visual examination of 
the containment surface.
    (E) 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:
    (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) Subsection IWE and Subsection IWL inservice inspection plans. 
Licensees that have less than 2 years remaining in their present 120-
month inservice inspection interval on (effective date of the final 
rule) may defer completion of the Subsection IWE and Subsection IWL 
portions of the inspection plan for the next 120-month inspection 
interval for up to 2 years from the end of the present interval.
* * * * *
    (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 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), 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 CC; 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 examinations specified for the first inspection interval in 
Subsection IWE and Subsection IWL of the 1992 Edition with the 1992 
Addenda in conjunction with the modifications specified in Sec. 50.55a 
(b)(2)(ix) by (a date will be inserted that is 5 years later than the 
effective date of the final rule).
    (2) The expedited examination may be used to satisfy the 
requirements of routinely scheduled examinations of Subsection IWE 
subject to IWA-2430(c) when the expedited examination occurs during the 
first containment inspection interval.
    (3) The requirement for the expedited examination of the 
containment post-tensioning system may be satisfied by written 
commitments that are in place before (the effective date of the final 
rule) for examinations of the post-tensioning system.
* * * * *
    Dated at Rockville, Maryland, this 3d day of January 1994.

    For the Nuclear Regulatory Commission.
Samuel J. Chilk,
Secretary of the Commission.
[FR Doc. 94-341 Filed 1-6-94; 8:45 am]
BILLING CODE 7590-01-P