ML20058H981
| ML20058H981 | |
| Person / Time | |
|---|---|
| Issue date: | 12/01/1993 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| FRN-59FR979 AC93-1-056, AC93-1-56, AC93-1-57, SECY-93-328, NUDOCS 9312130299 | |
| Download: ML20058H981 (114) | |
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>j RULEMAKING ISSUE (NEGATIVE CONSENT)
December 1, 1993 SECY-93-328
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The Commissioners FROM:
James M. Taylor Executive Director for Operaticns
SUBJECT:
ISSUANCE OF PROPOSED AMENDMENT TO 10 CFR 6 50.55a TO INCORPORATE BY REFERENCE THE ASME BOILER AND PRESSURE VESSEL CODE (ASME CODE), SECTION XI, DIVISION 1, SUBSECTION IWE AND SUBSECTION IWL PURPOSE:
To inform the Commission of intent to publish for public comment the proposed rule endorsing the 1992 Edition of the ASME Code 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," with specified modifications and a limitation.
Subsection IWE provides rules for inspecting the surface of metal containments, the steel liners of concrete containments, pressure-retaining bolts, seals and gaskets, containment vessel welds, and pressure-retaining dissimilar metal welds.
Subsection IWL provides rules for the examination of concrete pressure-retaining shells and shell components, and for the examination of unbonded post-tensioning systems.
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 ISI examinations.
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Contact:
Wallace Norris, RES NOTE:
TO BE MADE PUBLICLY AVAILABLE 492-3805 WHEN THE FINAL SRM IS MA AVAILABLE
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2 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 periodically inspected to detect defects that could compromise a containment's 1
pressure-retaining capability.
CATEGORY:
'1
, This paper covers a minor policy question requiring Commission consideration.
ISSUE:
Should the NRC expand the scope of 10 CFR 50.55a to endorse national standards (ASME) for the inservice inspection of metal and concrete containments, and is 3
i compliance an appropriate backfit justification for the proposed action?
SUMMARY
The staff is proposing this 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 written licensee commitments (e.g.,
1 in the Final Safety Analysis Report). Appendix J to 10 CFR Part 50 requires a
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general visual inspection of the containment but does not provide specific i
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 occurrence of degradation in containments and variability of present containment examinations, the staff has determined that it is necessary to prescribe and endorse more detailed requirements for the periodic examination of containment structures and codify these by 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 staff proposes to make these
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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 staff recommends that the proposed action be treated as a compliance issue. A Summary of the Documented Evaluation as required by 10 CFR 50.109(a)(4) to support this conclusion is provided in Enclosure 2.
The rate of occurrence of corrosion and degradation of containment structures 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. Table 3 of Enclosure 2 lists many of these occurrences of degradation.
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Over one-third of the operating containments have experienced corrosion or other degradation.
Almost one-half of these occurrences were first-identified 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 (1) corrosion of the steel containment shell in the drywell sand cushion region, resulting in reduced wall thickness to below the minimum design thickness; (2) corrosion of the torus of the steel L
containment shell (wall thickness at or near minimum design thickness); (3)
I grease leakage from the tendons of prestressed concrete containments; and (4) water seepage, as well as concrete cracking in concrete containments.
The staff surveyed the NRC Regional Offices to determine the type of inspections being performed on containment structures and to determine the effectiveness of the visual inspection as is currently required in 10 CFR Part 50, Appendix J.
Based on the survey of licensees (performed by NRC regional inspectors) and on experiences of NRC regional inspectors, the staff has determined that there are great differences among plants with regard to the performance and the effectiveness of containment inspections.
The staff believes that more specific ISI requirements, which expand upon l
existing requirements for the examination of containment structures in I
accordance with General Design Criteria ";DC) 53, Appendix A to 10 CFR Part 50, and Appendix J to 10 CFR Part O, are needed and are justified for the purpose of ensuring that containments continue to maintain or exceed l-minimum accepted design wall thicknesses and prestressing forces as provided I
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 staff also believes that the occurrences of corrosion and other degradation i
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 proposed for incorporation by reference into 10 CFR 50.55a.
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UlSCUSSION:
1 The primary function of the containment is to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity into the environment should an accident occur.
The role of an inservice inspection program is to detect any evidence of structural deterioration that may affect i
either the containment structural integrity or leak-tightness. As a result of the increasing rate of occurrence of containment degradation and the 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 integrity.
At present,10 CFR 50.55a specifies requirements for the inservice inspection and inservice testing of ASME Code Class 1, Class 2, and Class 3 components.
While this includes principal components within the nuclear steam supply system, it does not include metal or concrete containments. The proposed rule l
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would expand the scope of '10 CFR 50.55a to include inspection of Class MC
.(metal).and Class CC. (concrete) containments. The staff has found that these new subsections of Section XI of the ASME Code, which is a national consensus standard that is developed with NRC participation, would provide an acceptable method for detecting degradation of metal and concrete containments before margins in structural integrity are seriously compromised.
It is anticipated
.that 6 50.55a would be amended, as appropriate, to include later editions and addenda of Class MC and Class CC similar to the current process for ASME Code Class 1, 2, and 3 components.
Routine 150.55a rulemaking packages, which comprise an update of ASME Code I
addenda, have in the past been signed by the EDO.
However, the staff believes that this rulemaking action should be considered by the Commission-because of the policy nature of the questions involved regarding expanding. the -
scope of 10 CFR 50.55a to incorporate and backfit inservice inspection of metal and concrete containments and the appropriate justification for the backfit.
Further, the staff believes that this action is. an important step, not unlike the maintenance rule, that would strengthen the management of aging degradation for the containment structures and be entirely consistent with the aims of the license renewal objectives of 10 CFR Part 54.
In developing the proposed package, the staff considered two alternatives to the compliance justification for the backfit.
They are: 1) justification as a routine safety enhancement action in accordance with the criteria of 10 CFR 50.109(a)(3), and
- 2) justification under the adequate protection exception,10 CFR 50.109(a)(4)(ii).
The staff concluded that justification does exist for invoking the compliance exception under 10 CFR 50.109(a)(4)(1).
This does not mean, however, that licensees who have not yet adopted the provisions'of Subsection IWE and Subsection IWL for ISI are in non-compliance 'now or until-they do implement these provisions.
A detailed discussion of the staff's consideration of this issue as a justified safety enhancement, including a detailed' cost analysis, is provided in Enclosure 6.
It is clear from that discussion that the proposed action is consistent with the backfit rule criteria for justification as a safety enhancement backfit, (i.e., the proposed action will provide a substantial-safety benefit and the relatively modest costs of-implementation are justified in view of the significant safety benefit to be gained).
The staff believes, however, that the proposed action is not simply a safety enhancement.beyond the level already provided by existing rule requirements.
Based on actual operating experience, a serious concern exists regarding continued compliance by the operating plants with existing requirements (as reflected in license conditions, technical specifications, and written licensee commitments) for ensuring minimum design wall thicknesses and prestressing forces if the i
proposed action is not taken.
The staff believes that the presentation of
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this issue in the rulemaking package should appropriately reflect the compliance concern as the central regulatory justification.
With regard to consideration of the proposed action as an adequate protection issue, an argument can be made for treating this issue as the Pressurized Thermal Shock issue was treated.
In that case, known time-dependent mechanisms could have ultimately threatened the design margins of an important '
barrier in the overall defense-in-depth protection, if appropriate actions to reduce the threat were not taken by the NRC.
An important difference in this case, however, is that the industry has acted in a timely fashion, through the consensus standard process, to establish appropriate guidance and measures C
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that would address the recognized time-dependent mechanisms. The NRC is now acting to codify and augment those measures, and to ensure their timely implementation.
On balance, as a matter.of judgment in view of all relevant circumstances' and considerations, the staff does not believe that the safety concern involved at this point rises to a level that warrants identification in the category of a most compelling safety concern (i.e., adequate protection).
In summary, the staff recommends that Subsection IWE and Subsection IWL be imposed, as a compliance backfit,. on an expedited basis for the purpose of ensuring that containments continue to maintain minimum design wall thicknesses and prestressing forces. Based on actual operating experience, there is a serious concern regarding continued compliance by the operating plants with existing requirements for ensuring containment integrity and leaktightness if the proposed action is not taken.
Recommendation:
That the Commission:
Note that it is my intention to approve the proposed rule for publication j
in the Federal Register (75 day public comment period) within 10 working days from the date of this paper unless instructed otherwise by the Commission.
Coordination:
By memorandum dated November 12, 1992, from Raymond F. Fraley, Executive Director, ACRS, to Eric S. Beckjord, Director, RES, the ACRS advised that they had decided not to review the proposed amendment.
0GC has reviewed the backfit justifications considered for the proposed rulemaking and has concluded it appropriate to justify-the package under the compliance section of 10 CFR 50.109. Consequently, OGC has no legal objection to the proposed rule. The Offices of Nuclear Reactor Regulation, Analysis and Evaluation of Operational Data, and Information Resources Management concur in this proposed rule to amend 10 CFR 50.55a.
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J mes M. Td[ lor h
[
xecutive Director for Operations
Enclosures:
- 1. Federal Register Notice
- 2. Summary of Documented Evaluation
- 3. Environmental Assessment and Finding of No Significant Environmental Impact
- 4. Supporting Statement for Information Collection Requirements 5'. Congressional Committee Correspondence
- 6. Justification of Action as Safety Enhancement
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SECY NOTE:
In the absence of instructions to the contrary, SECY will notify the staff on Friday, December 17, 1993, that the Commission, by negative consent, assents to the action' proposed in this paper.
DISTRIBUTION:
Commissioners OGC OCAA OIG OPA OCA 1
EDO SECY
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ENCLOSURE 1 j
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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.
SUMMARY
The Nuclear Regulatory Commission (NRC) proposes to amend its reg-ulations 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 Co' led Power.
o 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 l
be-required to expedite implementation of the containment. examinations and complete the expedited examination in accordance with Subsection IWE and 1
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.
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DATES:
Comment period expires (75 days after oublication in the Federal Reaister). 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, t
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.
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i 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 j
Commission, Washington, DC 20555, telephone (301) 492-3813.
SUPPLEMENTARY INFORMATION:
Background
The NRC is taking the proposed action for the purpose of ensuring that 4
containments continue to maintain or exceed minimum accepted design wall thicknesses and prestressing forces as provided for in industry standards used i
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 l
license conditions, technical specifications, and. licensee commitments (e.g.,
the Final Safety Analysis Report).
The NRC also believes enhanced ISI l
examinations are needed and are justified to supplement existing requirements l
1 specified in General-Design Criterion (GDC) 16, and GDC 53, Appendix A to j
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 1
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structures in the regulations to assure that the critical areas of
-1 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 f
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 i
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
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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 4
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thickness); steel containment shell.' torus corrosion.(wall thickness at or near 1
. minimum design thickness); grease' leakage from the tendons of prestressed concrete containments, and water seepage, as well a's 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.
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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 o
for postulated accident conditions.
Section III and Section VIII of the ASME l'
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 5
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containment. structures and components be performed prior to any Type A-test to uncover any evidence of structural deterio' ration 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
-l rate: (1) after the containment nas been completed and is ready for operation, t
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 foJ, to
'f have undergone a significant level of degradation that was not de'tected by these programs.
The NRC believes that more specific ISI requirements, which expand upon existing requirements for the examination of containment structures in' f
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 i'
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
.i plants with existing requirements for ensuring containment minimum design wall 1
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thicknesses and prestressing forces ~ if the proposed action is not taken.
The i
NRC also believes that the occurrences of corrosion and other degradation discussed above would have been detected by licensees' implement'ing 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, i
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 i
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 j
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certain'recommen'dations for tendon examinations that 'are included in l
Regulatory Guide 1 35, Rev. 3, are not addressed in Subsection IWL (this
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involves four. of the modifications, (ix)(A)-(D)).
The ASME Code has considered these fout 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
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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.
7 The limitation specifies the 1992 Edition with 1992 Addenda of Subsection l.-
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 l-base metal examinations and would provide a comprehensive set of rules for the I
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 I
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.
l 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 l
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.(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.
l Existing 6 50.55a(g), " Inservice inspection requirements," specifies the requirements for preservice and inservice examinations.for Class 1 (Class'1 l
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-i waste-containing pressure vessels, heat exchangers (other than turbines and condensers), storage tanks, piping, pumps, and valves (not part of the reactor i
coolant pressure boundary)) components and their supports.
Neither Subsection IWE (Class MC -- metal containments) nor Subsection IWL (Class CC -
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- concrete containments) is presently incorporated by reference into the NRC regulations.
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Proposed 5 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 s 50.55a (g)(4)(v)(A), (v)(B), and (v)(C) specify Subsection i
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 6 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 s 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 6 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 10
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. within 2 years of the end of the 120-month _ interval may have difficulty
.i developing and completing the containment examination program in a timely
-l manner. Therefore, proposed 5 50.55a (b)(2)(x) specifies that licensees with less than 2 years remaining in their present ISI interval may complete the j
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.
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Proposed 6 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 s 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 11 4
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.
4 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.
l 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 tha'.
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 i
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 i
this rule and the beginning of the expedited examination.
1 k
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 12 I
H
i i
3.5 or 5.25-inch diskett'es'. Data files-should be provided as Wordperfect documents. ASCll-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
'r the diskette's external label.
)
Finding of No Significant Environmental Impact The Commission has determined under the National Environmental Policy Act i
of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR I
Part 51, that this rule, if adopted, would not be a major Federal action i
significantly affecting the quality of the human environment and therefore an j
1 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
.j rule would have a positive impact on the environment.
The proposed rule would i
incorporate by reference in the NRC regulations requirements contained in the j
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.
j I
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 q
13 up9-,.si m-
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 -
l-average 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per rcsponse for development of an initial inservice inspection plan and 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 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, NE08-3019, (3150-0011), Office of Management and Budget, Washington, DC 20503.
14
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-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 1
i 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 l 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 Administrationaat 13 CFR Part 121.
Since these-companies are dominant in their service areas, this proposed rule does not 15
==.
fal.llwithin the purview of the Act.
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Backfit Statement 3
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)(1) a backfit analysis need not be prepared for this rule.
A-summary of the documented evaluation required by 6 50.109(a)(4) to support
+
this conclusion is set forth below.
GDC 16, " Containment design," requires the provision of reactor con-tainment 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.-
i 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 16 l
4
.I
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 ar 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 nce.essary 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 l
critical to the containment's behavior during an accident.
I i
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, t
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.
17
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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.
t 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 i
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, resultino thickness reduced to below the minimum
.i wo design thickness; 2) corrosior
' " a 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.
18 I
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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.contsinment structures continue to maintain or exceed minimum accepted design wall thicknesses and prestressing forces as reflected in license I
1 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' 1
these additional portions of the ASME Code (Subsection IWE and Subsection
~
<l IWL).
The Commission concludes that this proposed backfit is necessary to
.j ensure compliance with GDCs 16 and 53, Appendix J, minimum design wall thicknesses in metal containments, and the prestressing forces of concrete q
~i containments, which are applicable to all licensees through license-conditions, technical specifications, and licensee ' commitments.
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19
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List of Subjects in 10 CFR Part-50 J
Antitrust, Classified information, Criminal Penalties, Fire protection,.
Incorporation by reference, Intergovernmental relations,-Nuclear power plants
-i and reactors, Radiation protection, Reactor siting criteria,' Reporting and recordkeeping requirements.
i 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 DOMESTIC LICENSING 0F PRODUCTION AND UTILIZATION FACILITIES 1.
The authority citation for Part 50 continues to read as follows:
1 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 I
(42 U.5-C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L.95-601, sec.10, 92 Stat. 2951 (42 N
U.S.C. 5851).
Section 50.10 also issued under secs, 101, 185,-68 Stat. 936.
i 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, 20
. - ~
and 50.56 also issued-under sec. 185, 68 Stat. 955 (42 U.S.C. 2235).
Sectiort 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;fF,, and_ revising the introductory text of paragraph (g)(4) to read as follows:
~9 50,55a Codes and standards.
(h) *
(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.
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(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 t
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.
r (C) When the elongation corresponding to a specific load (adjusted for ef fective wires or strands) during retensioning of t'endons 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:
^
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(1) The sampled sheathing filler grease contains chemically combined water exceeding 10_ percent by weight or the presence of free water; 22 i
(2) The absolute difference between the amount removed and the amount replaced may not exceed 10 percent of the tendon net duct volume.
(1) 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; (1) A description of necessary corrective actions.
(x)
Subsection IWE and Subsection IWL inservice inspection olans, Licensees that have less than 2 years remaining in their present 120-month inservice inspection interval on (insert effective date of the final rule) may defer completion of the Subsection IWE and Subsection IWL portions of the i
inspection plan for the next 120-month inspection interval for up to 2 years
)
from the end of the present interval.
I 23 T
(g)
(4) Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified I
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.
1 f
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e e
s s
e (v)
For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued after January 1, 1956:
1-(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 I
meet the inservice inspection and repair requirements applicable to components which are classified as ASME Code Class CC.
(6)
_i (ii) *
(B) Exoedited examination of containment.
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(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 1
the modifications specified in 150.55a (b)(2)(ix) by (a date will be inserted that is 5 years later than the effective date of the final rule).
1 (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.
i (1) The requirement for the expedited examination of the containment post-tensioning system may be satisfied by written commitments that are in
.i place before (insert the effective date of the final rule)-for examinations of the post-tensioning system.
1
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Dated at this day of 19 For the Nuclear Regulatory Commission.
o i
J Samuel J. Chilk, Secretary of the Commission.
26
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. ENCLOSURE 2
SUMMARY
OF DOCUMENTED EVALUATION-1
' '. h kk
- l 1
9 er 3
SUMMARY
OF DOCUMENTED EVAL.UATION
SUMMARY
The NRC is proposing the subject 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 American Society of Mechanical Engineers (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).
Based on the results of inspections and audits of plant structures, as well as plant operational experiences, the NRC has determined 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 containments behavior during an accident.
Over one-third of the operating containments have experienced corrosion or 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: (1) corrosion of the steel containment shell in the drywell sand cushion region, resulting in reduced wall thickness 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, which expand upon existing requirements for the examination of containment structures in accordance with General Design Criterion (GDC) 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 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, or licensee commitments (e.g.,
the Final Safety Anaylsis Report).
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.
Recent changes and additions to the ASME Code include provisions to address the concerns outlined above; and the NRC proposes to make these provisions mandatory by amending 10 CFR 50.55a to incorporate by reference 2-1
I these additional portions of the ASME Code (Subsection IWE and Subsection IWL).
STATEMENT OF THE OBJECTIVES:
In view of the increasing rate of occurrences of. degradation in containments, the unacceptability of some occurrences, and the 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 routinely inspected to detect defects that could compromise the containment's pressure-retaining integrity.
Specifically, the proposed rule would incorporate by reference the 1992 Edition with the 1992 Addenda of both 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," with specified modifications and a limitation.
Subsection IWE provides rules for inspecting the surface of metal containments,- the steel liners of concrete containments, pressure-retaining bolting, seals and gaskets, containment vessel welds, and pressure-retaining dissimilar metal welds.
Subsection IWL provides rules for the examination of concrete pressure-retaining shells and shell components, and for unbonded post-tensioning systems.
REASONS FOR THE MODIFICATIONS:
An increase in the rate of reported incidents of significant corrosion and degradation of containments has been detected 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 the post-tensioning systems or the concrete of concrete containments have been reported.
Four other recent incidents which involved grease leakage from tendons have been investigated.'
In addition to grease leakage, these incidents showed signs of leaching of the concrete (See Table 3 for a partial listing).
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 (See Table 1). 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 in.to 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 A
2-2
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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 React.rs," 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 ultimately result in a compromise to the pressure-retaining capability.
Some containment structures have also been found to.have undergone an unacceptable level of degradation that was not detected by these programs.
Because of the staff's concern arising from operating experience with regard to the vulnerabilities of highly stressed post-tensioning systems of prestressed concrete containments (PCCs), the staff developed some periodic inspections to monitor the status of these components during the design lives of these containments.
Prior to the issuance of Regulatory Guide 1.35, i
" Inservice Inspection of Ungrouted Tendons in__ Prestressed Concrete Containment Structures," in 1973, these inspection provisions were promulgated on a case-by-case basis.
This regulatory guide was issued to assure _that PCCs'were properly monitored and to assure some consistency with regard 'to industry inspection.
Revisions of the regulatory guide were developed and issued to incorporate increased experience in the examination of post-tensioning systems.
Revision 2 (1976) contained technical updates to-Revision 1, and the costs of implementation of Rev. 2 compared to Rev. I were nearly equal.
Revision 3 (1990) changed the tendon detensioning and sampling requirements such that a considerable cost savings to the industry would be realized (See Tables 5 - 7).
At present, twenty-seven licensees have voluntarily adopted Rev. 3.
Five licensees are committed to Rev. 2, six licensees to Rev._1, and of the five remaining post-tensioned containments, one will start using Rev 3 with their next scheduled surveillance, three would rather use the industry standard (Subsection IWL) than a regulatory guide, and are therefore waiting for the rule to become effective, and the last PCC is a grouted tendon containment to which'these. rules would not apply. As pointed out in NUREG/CR-'
4712, " Regulatory Analysis of Regulatory Guide 1.35 (Revision 3, Draft 2), "
the implementation of Rev. 3'of Regulatory Guide 1.35 (Subsection IWL with modifications) will have very little effect on cost, but a positive impact on 2-3
safety.
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). The ASME Code'has considered these four issues and has adopted them in Subsection IWL.
These issues will be published in the 1993 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 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 staff 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.
The proposed rulemaking 1
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 list 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.
BASIS FOR INV0 KING THE COMPLIANCE EXCEPTION:
The NRC is proposing this 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 1
Final Safety Analysis Report).
Based on actual operating experience, a serious concern exists regarding continued compliance by the operating plants with existing requirements if the proposed action is not taken. The NRC also believes enhanced ISI examinations are needed to supplement existing requirements specified in GDC 53 and Appendix J.
Appendix J requires a j
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.
2-4
4 Based on the results of inspections and audits of plant structures, as well as plant operational experiences, the NRC has determined that many licensee containment examination programs have not detected degradation that could result in a compromise of pressure-retaining capability, i
Over one-third of the operating containments have experienced corrosion or degradation. Almost one-half of these 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 the steel containment shell 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.
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). The NRC believes that the occurrences of corrosion and other depradation discussed above would have been detected by licensees implementing j
the comprehensive periodic examinations set forth in Subsection IWE and i
Subsection IWL of the ASME Code proposed for incorporation into 10 CFR 50.55a.
ADDITIONAL CONSIDERATION:
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 that, NUMARC concluded "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.
i
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I TABLES AND FIGUR$S i
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4 TABLE 1 CONTAINMENT INSPECTION CRITERIA, REOUIREMENTS AND GUIDANCE LCRITERIA
- CRITERIA REQUIREMENTi
. HOWl SUBSECTION IWE'ORJIWL WOULD.'
SATISFY THE. REQUIREMENT Part 50, Structures or components Subsections IWE and IWL would Appendix A, important to safety shall be ensure that specific containment Criterion 1 tested to quality standards components would be inspected to an acceptable standard (i.e.,
the ASME Code)
Part 50, The reactor containment and The containment inspections Appendix A, associated systems provide required by Subsections IWE and Criterion 16 essentially a leak-tight barrier IWL are to ensure the pressure-retaining integrity of the i
containment Part 50, a) Periodic inspection of all Subsections IWE and IWL set forth Appendix A, important containment areas procedures for requirements (a)
Criterion 53 b) an appropriate surveillance and (b).
Appendix J satisfies program; and requirement (c) c) periodic testing, at contain-ment design pressure, of the leak-tightness of penetra-tions Part 50, General inspection of the Subsections IWE and IWL provide Appendix J accessible interior and exterior details for this general surfaces of the containment inspection (e.g.,
what parts of structures and components-the containment structure must be accessible for inspection and personnel qualification require-ments for examiners that are not specified in Appendix J T-1
TABLE 1 (continued)
CONTAINMENT INSPECTION CRITERIA, REOUIREMENTS AND GUIDANCE
' CRITERIA
' CRITERIA REQUIREMENT;
- HOW SUBSECTION IWE ORI IWL WOULD SATISFY THE REQUIREMENT Part 50, Quality assurance requirements Tables IWE-2500-1 and IWL-2500-1 Appendix B for the operation of structures, contain the applicable examination systems, and components which methods.
IWE-4000 and IWL-4000 include inspecting, testing, contain the applicable repair repairing, and modifying requirements, and IWE-7000 and IWL-4000 contain the applicable replacement requirements Standard The structural integrity of the Tendons shall be examined in Technical containment shall be maintained accordance with IWL-2520 Specification at a level consistent with the 3.6.1.6 acceptance criteria in Specification 4.6.1.6 Standard The structural integrity of the Tendons shall be examined in Technical prestressing. tendons of the accordance with IWL-2520 Specification containment shall be demonstrated 4.6.1.6 at the end of 1, 3,
and 5 years following the initial containment structural integrity test and at 5 year intervals thereafter T - 1.1
~~
~ - - -
- ~ - ~ ~ ~
TABLE 1 (continued)
CONTAINMENT INSPECTION CRITERIA. REOUIREMENTS AND GUIDANCE CRITERIA.
CRITERIA l REQUIREMENT HOW SUBSECTIONTIWEiORTIWL4WOULD SATISFY =.THE REQUIREMENT Standard The structural integrity of the Subsections IWB and IWL provide Technical exposed accessible interior and details for this general Specification exterior surfaces of the inspection not specified in 4.6.1.7.1 containment shall be determined Appendix J (e.g.,
what parts of by a visual inspection during the containment structure must be the shutdown for a Type A accessible for inspection, and Containment Integrated Leak Rate personnel qualification l
Test requirements for examiners)
T - 1.2 2-
+
FIGURE 1 LARGE, DRY STEEL SPHERE BIG ROCK POINT NINE MILE POINT 1 OYSTER CREEK DRESDEN 2 & 3 MILLSTONE 1 MONTICELLO PILGRIM 1 MARK 1 STEEL QUAD CITIES 1 & 2 DRYWELL & WETWELL BROWNS FERRY 1,2, a 3 COOPER (22 PUNTS)
DUANE ARNOLD FREE STANDING FERMI 2 STER PRM FITZPATRICK E CREEK 1 PEACH BOTTOM 2 & 3 VERMONT YANKEE MARK 11 STEEL WNP-2 DRYWELL & WETWELL MARK lil REINFORCED PERRY 1 CONCRETE ORYWELt, NI STER WEN (26 PLANTS)
MARK 1 REINFORCED CONCRETE DRYWELL BRUNSWICK 1 & 2
& WETWELL REINFORCED MARK 11 REINFORCED UMERICK I & 2 CONCRETE PRIMARY CONCRETE DRYWELL SUSOUEHANNA 1 & 2.
CONTAINMENT
& WETWELL -
NINE MrLE POINT 2 WITH STEEL UNER MARK ll1 REINFORCED CONCRETE DRYWELL CUNTON1
& WETWELL GRAND GULF 1 (9 PMNTS)
POST-TENSONED MARK 11 REINFORCED CONCRETE PRIMARY CONCRETE DRYWEL1, CONTAINMENT POST-TENSIONED LASALLE 1 & 2 WITH STEEL UNER WETWELL DISTRIBUTION OF BWR CONTAINMENTS BY CONSTRUCTION TYPES F-1
FIGURE 2 LARr'O ORY BARE STEEL P.W ARY SPHERE M4EN C ONTAINMENT STEEL SPHERE (60 PLANTS)
(2 PLANTS)
STEEL SPHERE WITH CONCRETE SAN OME ENCLOSURE BLDG.
nwAuNEC STEEL CYUNDER PRAsRE tuvo : a 2 STEEL CYUNDER WITH CONCRETE onvis.aEsst (7 PLANTS)
ENCLOSURE BLDG.
R LUC 8E 14 2 WATUtSCAD 3 MANE YAreGE REINFORCED CONCRETE m",'s,
REINFORCED CYUNDER WITH HCOAM NECX CONCRETE STEEL UNER No N 2a8 CYUNDER sw_ ARON HMUUS WITH STEEL UNER atwomcoco m ic (t: Ptn(Ts)
CYt#OER WTm SEASRO:* 1 STEEL UtCR M O sec. coNTea.erNT 1 D VERTKAL POST.TENSJOED onNA COPOtETE C1UNDER H.S. ROS#dSON WTTH STEEL USFR DN POST.TENSOED POST-TENSIONED FORT CALHOUN CONCRETE CYUNDER CONCRETE wrTH sTrrt UNFR CYUNDER WITH STEEL UNER 3-D POST TENSIONED A MANsAsis2 CONCRETE CYUNDER 000NC8 8 8 8 8 I
(41 PUWTS)
WITH STEEL UNER h,
CALwRT curra e a 2 PAusADES so Posi TEN 5K)NLO PALD VUOE l.2 & 3 COtORETE CVUNDER WusTOE 2 SAN ON0$ME 2 s a WITH STEEL UNER SELLEFONTE t & 2 SRAN :42 ANo sEc crwTAmwNT symN t a 2 CALLAWAY FARLEY t 4 2 REINFORCED CONCRETE POWT SEACH I 4 2 SEAVER VALLEY : & 2 SOUTHTtxAsta2 CYUNDER WITH NORTH Ase.A 4 4 2 SuuuER SUBATMOSPHERIC STEEL UNER suRRY:a2 TRosAN PRIMARY TuRevpowT3aA CONTAINMENT WoGM i a 2 6ENOmLO GOtusETE (7 PLANTS)
C umd uiLLsTONE 3 s
AND SEC CONTAiNtKNT STEEL CYUNDER SEOuovAM 1 & 2 ICE CONDENSER WITH CONCRETE WAns SAR 14 2 SHIELD BUILDING
[,'g',#
PRIMARY REINFORCED CONCRETE CYUNDER WITH on Cook i a 2 (to PLANTS)
STEEL UNER DISTRIBUTION OF PWR CONTAINMENTS BY CONSTRUCTION TYPES F-2
i s
Table 2 Summary of Potential Containment Decradation Sites CONCRETE LOCATIONS All Concrete Structures ~
MECHANISMS Chemica1 Attack;
. Groundwater-Chemistry;. Sulfates; Acid. Rain; Alkali-Silica Alkali-Carbonate Aggregate Reactions;
-Freeze-Thaw Cycles; Shrinkage; Thermal Cycling >
150'F; Radiation (Internal Heating) ;
Dehydration; Vibration; Differential Settlement INDICATORS Cracking;. Discoloration; Spalling; Pop-out; Loss of Strength; Aggregate Breakdown; Peeling; Discoloration or Delamination of Coatings PROBLEM AREAS Areas with Geometries That Permit Water Accumulation, Condensation, and Micro-biological Attac'k; Areas Exposed to
_i Repeated Wetting-and Drying; Areas' Exposed to Persistent-Leakage; Areas Exposed to Chemical Spills; Areas Below Groundwater Level; Concrete-Steel Inter-j faces; Steel Encased Concrete Struc-tures; " Hot" Penetration Sleeves FAILURE MODES Leakage; Loss of Structural Integrity PREVENTION / REPAIR Visua1 Examination;- Cr.ack Mapping and l
Repair; Coating Repair or Replacement; Replacement of
- Seals,
. Gaskets, and.
Caulked Joints; Tapping of Liner for Hollow Spots; Groundwater Monitoring;-
Core Sampling and Testing; Spill. Pre-vention and Cleanup; Open Drains i
i From R.F.
Sammataro,
" Containment Inservice Inspection and Testing," Lecture Notes, Technical Seminars, Great Neck,- New York, 1990.
h T-2 9
\\
Table 2 (Continued)
Summary of Potential Containment Dearadation Sites POST-TENSIONING SYSTEM LOCATIONS Tendons, Anchorages MECHANISMS Chemical Attack; Hydrogen Embrittlement; Stress Corrosion; Microbiological Cor-rosion; Moisture Intrusion INDICATORS Corrosion; Cracking;. Pitting; Water j
Accumulation in Tendons; Grease Break-down; Anchorage Failures; Loss of Tendon' Stress; Flaking; Blistering, Peeling or Discoloration of Paint or Coating on Anchorages; Mechanical Damage; Wedge Slippage Marks PROBLEM AREAS High Humidity; Areas of. Water Accumula-l tion; Areas Exposed to Chemical Spill FAILURE MODES Loss of Structural Integrity PREVENTION /F W IR Visual Examination; Coating Repair or' Replacement; Tendon Wire and Strand i
Sample _
Examination and Testing; Examination of Tendon Anchorage Areas; Examination of-Corrosion Protection Medium and-Free Water
')
i 1
l
.l l
- 1 i
From.
R'. F.
Sammataro,
" Containment Inservice. Inspection and Testing," Lecture Notes, Technical Seminars, Great Neck, New York, 1990.
R T - 2.1 1
l
_ma-m-
. Table 2 (Continued)
Summary of Potential Containment Decradation Sites S_TEEU LOCATIONS Shell Plating; Penetrations; Hatches; Structural and Nonstructural Attach-ments; Dome and Basemat Liner; Leak Chase Channels; Torus; Supports; Anchor Bars and Studs; Embedments MECHANISMS Corrosion; Fatigue; Chemical Attack; Stress Corrosion Cracking; Micro-biological Corrosion; Galvanic Corro-sion; Radiation Embrittlement INDICATORS Rust; Discoloration,
- Staining, Blistering, and Bubbling of Paint; Spalling of Concrete; Scale Buildup; Buckling or Lift-Off of Liner; Leakage from Drains; Clogged Drains PROBLEM AREAS Areas of Water Accumulation; High Humidity Areas; Areas Exposed to Chemical Spill or Borated Water Spills; Flashed, Caulked or Sealed Joints; Dis-similar Metal Connections; Penetrations; Condensation and Leak Paths; Sand Pockets; Locations with Stray Currents; Heat Trace Areas FAILURE MODES Leakage; Loss of Structural Integrity; Catastrophic Failure Under Severe I
Accident Loads PREVENTION / REPAIR Leakage Testing; Visual Examination; UT Wall Thickness Trending in Critical 1
Areas; Coating Repair or Replacement; Surface NDT of Dissimilar Metal Welds; Replacement of
- Seals, Gaskets, and Caulked Joints; Tap Liner for Hollow Spots; Open Drains i
From R.F.
Sammataro,
" Containment Inservice Inspection and Testing," Lecture Notes, Technical Seminars, Great Neck, New York, 1990.
T - 2.2
Table 2 (Continued) i J
Summary of Potential Containment Deoradation Sites PENETRATIONS LOCATIONS Mechanical and Electrical Penetrations MECHANISMS Mechanical-Corrosion.
Disbonding of Sleeves; Fatigue Cycling at Hot Penetrations; Stress Corrosion of Stainless Steel Penetrations; Seal and Gasket Aging; Bolting Fatigue, Wear and Corrosion Electrical-Environmental.
Qualification Limited' to 40 Years; Disbonding and Cracking of Ceramic Inserts; Physical Damage; Cable Deterioration, Aging, Chemical Breakdown INDICATORS Mechanical-Corrosion Oxidation; Sleeve Motion; Leakage; Cracking Electrical Loss of Electrical Function by i
Shorts, Grounds, Conductance; Leakage PROBLEM AREAS High Humidity Areas; Areas Exposed to Moisture Intrusion or Leakage; Corrosive or Detrimental Environments; Penetra-tions Subject to Cyclic Thermal or_
Mechanical Loadings; Dissimilar Metal Welds FAILURE MODES Leakage; Loss of Mechanical or Electri-i cal Function PREVENTION / REPAIR Visual Examination; Leakage Testing; Surface Examination of Dissimilar Metal
- Welds, Mechanical or!
Electrical Functional Testing; Repair or Replace-ment 2
From R.F.
Sammataro,
" Containment Inservice Inspection and'
?
Testing," Lecture Notes, Technical Seminars, Great Neck, New York, 1990.
T - 2.3 q
l 1
b Table 2 (Contirued)
Summary of Potential Containment Decradation Sites BELLOWS LOCATIONS Penetrations MECHANISMS Corrosion; Fatigue; Chemical Attack; Microbiological Corrosion; Stress Corrosion Cracking INDICATORS Misalignment; Excessive Motions; Debris or Water Accumulation in Convolutions; Physical Damage such as Scratches,. Dents, Arc Strikes; Uneven Convolution Geometry; Discoloration; Leakage PROBLEM AREAS High Humidity Areas; Areas Exposed to Moisture Intrusion or Leakage;-Corrosive or Detrimental Environments; Penetra-tions Subject to Cyclic Thermal - or Mechanical Loadings; Dissimilar Metal i
Welds; Top of External Surfaces; Bottom of Internal Surfaces; Bellows-Collar Joints FAILURE MODES Leakage; Dislocation PREVENTION / REPAIR Visual Examination; Leakage Monitoring and Testing; Surface.Examinationi of Dissimilar Metal Welds _and Surfaces;
,l Maintain Internal and External Cleanli-ness; Observe Motions during Startup; Dimensional Survey; Install Covers Over Bellows; Replacement 3
From R.F.
Sammataro,
" Containment Inservice Inspection and Testing," Lecture Notes, Technical Seminars, Great Neck, New' York,
'1990.
I T - 2.4
m TABLE 3 - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS
-PLANT-
' TYPE-
~ OCCURRENCE:
HOW THE[
~(NUMBER OF T
- D E G R A D A T I O N.-
. IWL' INSPECTION L PLANTS : OF. ;
.WAS FOUNDl
- REQUIREMENTE THE SAME
- THAT.. COVERS -
CONSTRUCTION.
. OCCURRENCE
' TYPE)
Monticello BWR - Mark I Polysulfide seal at A small portion IWE-3513; steel drywell the floor to shell of the drywell Standards for
& wetwell (22 interface had shell was Examination Cate-plants) become brittle excavated as gory E-D,
- Seals, allowing moisture part of a life Gaskets, and to reach the steel extension study Moisture Barriers shell Oyster Creek BWR - Mark I Defective gasket at Corrosion found IWE-3512.2 and steel drywell the refueling pool in the sand IWE-3512.3; Stan-
& wetwell (22 allowed water to cushion during dards for Exami-plants) eventually reach an NRC nation Category the sand cushion inspection E-C, Containment region which Surfaces Requir-corroded the ing Augmented drywell shell Examination Nine Mile BWR - Mark I Corrosion of the Corrosion IWE-3512.2 and Point 1 steel drywell torus discovered IWE-3512.3; Stan-
& wetwell (22 during a dards for Exami-plants) special nation Category announced NRC E-C, containment team inspection Surfaces Requir-ing Augmented Examination T-3 L
e
~
e-n-m-m+
+
<---~.e
+-v-sw w --
.+w-e v.-
, - + -
n,
~
w -r
a TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS PLANT-TYPE.. __
. OCCURRENCE.-
[HOW.THE-..
ISUBSECTIONiIWE[OR (NUMBER _OF_
DEGRADATION:
- IWL
- INSPECTION-
. PLANTS OF;
- WAS-FOUND.
REQUI'tEMENT ?
-THE'SAME CONSTRUCTION
THAT ' COVERS T
- OCCURRENCE:
-TY PE) -
Fitzpatrick BWR - Mark I Degradation of the Technical Spec-IWE-3512.2 sad steel ~drywell torus coating with ification sur-IWE-3512.3; Stan-
& wetwell (22 associated pitting veillance per-dards for Exami-l plants) formed during nation Category an outage E-C, Containment Surfaces Requir-ing Augmented Examination Millstone 1 BWR - Mark I Degradation of the The torus had IWE-3512.2 and steel drywell torus coating had been drained IWE-3512.3; Stan-
& wetwell (22 occurred for modifi-dards for Exami-plants) cations nation Category E-C, Containment Surfaces Requir-ing Augmented Examination Pilgrim BWR - Mark I Degradation of the Inspected as a IWE-3512.2 and steel drywell torus coating had result of BWR IWE-3512.3; Stan-
& wetwell (22 occurred torus corrosion dards for Exami-plants) at other plants nation Category E-C, Containment Surfaces Requir-ing Augmented Examination T - 3.1 e
a.
TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS
. PLANT TYPE.
-OCCURRENCE'
.HOW Ti1E
'SUB ECTIONcIWE'OR (NUMBER.0F
. DEGRADATION
?IWL INSPECTIOll--
PLANTS OF.
'WAS'FOUND-LREQUIREMENT, THE SAME ~
- TilAT_. COVERS 1
. CONSTRUCTION
' OCCURRENCE TYPE)'
s McGuire 2.
PWR - Ice Corrosion'on the Pre-integrated IWE-3510.1; Stan-Condenser; outside of the leak rate test dards for Exami-steel cylinder steel cylinder in inspection nation Category (16 plants) the annular region E-A, Containment Surfaces
[
McGuire 1 PWR - Ice Corrosion on the Inspection as a IWE-3510.1; Stan-Condenser; outside of the result of dards for Exami-steel cylinder steel cylinder in corrosion found nation Category (16 plants) the annular region at McGuire 2 E-A, Containment i-Surfaces Catawba 1 PWR - Ice Corrosion on the Inspection as a IWE-3510.1; Stan-Condenser; outside of the result of dards for Exami-steel cylinder steel cylinder in corrosion found nation Category (16 plants) the annular region at McGuire 2 E-A, Containment Surfaces Catawba 2 PWR - Ice Corrosion on the Inspection as a IWE-3510.1; Stan-Condenser; outside of the result of dards for Exami-steel cylinder steel cylinder in corrosion found nation Category (16 plants) the annular region at McGuire 2 E-A, Containment Surfaces T - 3.2
.m
.. ~m v
~.,*
.n
+
,,s.
TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS PLANT-TYPE
..' OCCURRENCE:
IHOW ' Tile
'SUBSECTIONIIWE OR'
-(NUMBER OE-DEGRADATION:-
DIWL. INSPECTION'
- PLANTS OFf
'WAS!FOUND;
- REQUIREMENT.
~THE SAME CONSTRUCTION'
~
LTHAT.COVERST
. OCCURRENCE:
1 TYPE)
Quad Cities 1 BWR - Mark I Two-ply contain-Excessive leak-Examination Cat-steel drywell ment penetration age from the egory E-P, All
& wetwell (22 bellows leaked due containment was Pressure Retain-plants) to transgranular identified ing Components stress corrosion during a Type A cracking (TGSCC) test Quad Cities 1 BWR - Mark I Two-ply contain-Two other Examination Cat-steel drywell ment penetration bellows in Unit egory E-P, All
& wetwell (22 bellows leaked due 1 have been Pressure Retain-plants) to transgranular replaced due to ing Components stress corrosion TGSCC. Exces-cracking (TGSCC) sive leakage from the bellows was identified Quad Cities 2 BWR - Mark I Two-ply contain-One bellows has Examination Cat-steel drywell ment penetration been' replaced egory E-P,'All
& wetwell (22 bellows leaked due due to.TGSCC.
Pressure Retain-plants) to transgranular Excessive leak-ing Components stress corrosion age from the cracking (TGSCC) bellows was identified I
T - 3.3 i
T TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS PLhNT-TYPE
'OCCURRENCEi HOW~THE 15UBSECTION.$1WE i ORi
"(NUMBER OF, DEGRADATION
' IWLTINSPECTIONj PLANTS.0F WAS'FOUNDL REQUIREMENT <
THE SAME.
THAT<COVERSL
' CONSTRUCTION OCCURRENCE'.
TYPE)'
Dresden 3 BWR - Mark I Two-ply contain-One bellows has Examination Cat-steel drywell ment penetration been replaced egory E-P, All
& wetwell (22 bellows leaked due due to TGSCC.
Pressure Retain-plants) to transgranular Excessive leak-ing Components stress corrosion age was cracking (TGSCC) identified Dresden 2 BWR - Mark I Peeling and Plant personnel IWE-3512.1 and steel drywell discolored interior found that 4 IWE-3512.3; Stan-
& wetwell.(22 shell coatings, de-ventilation dards for Exami-plants) graded electrical hatches in the nation Category cables, and degrad-drywell refuel-E-C, Containment ed valve operator ing bulkhead Surfaces Requir-components due to had been inad-ing Augmented excessive oper-vertently left Examination ating. temperatures closed Crystal River P'WR - Post-Dome delamination Discovered by IWL-2510; Exami-3 tensioned con-due to low quality accident by nation of crete cylinder coarse aggregate, electricians Concrete with steel accompanied by high attempting to liner (35 radial tension secure anchors plants) forces above the top tendons, and compression-tension interaction T.-
3.4 s
,a
+
,v,
_c,a
i TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS PLANT TYPE OCCURRENCE HOW THE SUBSECTION IWE OR-(NUMBER OF DEGRADATION IWL INSPECTION PLANTS OF WAS FOUND
' REQUIREMENT-THE SAME THAT COVERS-CONSTRUCTION OCCURRENCE TYPE)
Farley 1 & 2 PWR - Post-Anchor head corro-Pre-ILRT visual IWL-2524; Exami-tensioned con-sion attributed to inspection nation of Tendon crete cylinder high hardness of Anchorage Areas with steel material, free liner (41 water in grease plants) caps, and high stresses in anchor-heads Hatch 2 BWR - Mark I Brittle fractures Found by plant IWE-3512.1 and steel drywell of 54-inch and 66-personnel dur-IWE-3512.3; Stan-
& wetwell (22 inch vent headers.
ing an inspec-dards for Exami-plants)
The nitrogen inert-tion of the nation Category ing system line in torus interior E-C, Containment torus may have sub-Surfaces Requir-jected material to ing Augmented a temperature below Examination its nil ductility temperature Ginna PWR - Post-Excessive tendon Technical Spec-IWL-2520; Exami-tensioned con-relaxation due to fication sur-nation of Post-crete cylinder tendon restressing veillance per-Tension Systems.
with steel formed during IWL-2524; Exami-liner (41 an outage nation of Tendon plants)
Anchorage Areas T - 3.5
.e
TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS
/ PLANT
- TYPE
. OCCURRENCE"
'HOWITHE-4 SUBSECTIOff 5IWE JOR
' (NUMBER.OF-DEGRADATION IWL; INSPECTION 1
' PLANTS;OF7 WAS.:l FOUND REQUIREMENTi i
THE SAME-
'THAT COVERS-'
'CONSTRUCTIONi
^
, OCCURRENCE 1 TYPE)-
Fort Calhoun PWR - Post-Grease leakage from By plant IWL-2524; Exami-tensioned con-tendon sheathings
. personnel nation of Tendon crete cylinder or at joints Anchorage Areas.
with steel between conduit IWL-2510; Exami-liner (41 lengths nation of Con-plants) crete Three Mile PWR - Post-Grease leakage, and By plant per-IWL-2510; Exami-Island 1 tensioned con-mineral deposits in sonnel and nation of Con-crete cylinder.
lower tendon gal-during NRC site crete with steel lery apparently due visit liner (41 to leaching of plants) concrete '
Trojan PWR - Post-Grease leakage at Found during IWL-2520; Exami-tensioned con-tendon anchors and NRC Audit of nation of Post-crete cylinder through cracks in Civil Engineer-Tension Systems.
with steel containment walls ing Structures IWL-2524;.Exami-liner (41 as well:as leaching nation of Tendon plants)
Anchorage Areas T.- 3.6 I
l
-. 2-
. w
.sw--
,..-y.y, 4..-
-,y-,
w-wy,,,-,.,,,.n,,
w w.
e
TABLE 3 (Continued)
- OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS-(PLANT-
... TYPE L OCCURRENCE-
~HOW Tile SUBSECTION j-IWE.-OR
- (NUMBER OF1-
' DEGRADATION EIWLiINSPECTION
. PLANTS OF
-WAS:FOUND:
REQUIREMENT -;
- Ti!E SAME' TilAT COVERSt
~
CONSTRUCTION'
- OCCURRENCE 4 TYPE):
McGuire 2 PWR - Ice Bare metal found in By plant per-IWE-3510.1; Stan-Condenser; lower containment sonnel before dards for Exami-steel cylinder at various loca-an ILRT nation Category (16 plants) ions; general ves-E-A, Containment sel corrosion and Surfaces some pitting in lower annulus Turkey Point PWR - Post-Low lift-off forces By plant per-IWL-2520; Exami-3 tensioned con-measured on the sonnel during nation of Post-crete cylinder hoop tendons twentieth year Tension Systems.
with steel tendon inspec-liner (41 tion plants) i Cooper BWR - Mark I Spot pitting cor-By plant IWE-3512.2'and steel drywell rosion on the in-personnel IWE-3512.3; Stan-
& wetwell (22 side of the torus dards for Exami-plants) nation Category E-C,, Containment Surfaces Requir-ing Augmented-Examination T - 3.7
TABLE 3 (Continued) - OCCURRENCES OF STRUCTURAL DEGRADATION AT NUCLEAR POWER PLANTS PLANT.
-TYPE.
OCCURRENCE ^
HOW THE
-SUBSECTIONTIWE ORI
.(NUMBER'OF-DEGRADATIONi-IWL -' INSPECTION J iPLANTS OF.
WAS FOUND
! REQUIREMENT: '
I TIIE'SAME TIIAT - COVERS T CONSTRUCTION:
OCCURRENCEL TYPE).
Salem 2 PWR - Rein-Minor corrosion on By plant per-IWE-3510.1; Stan-
' forced con-the liner sonnel before dards for Exami-crete cylinder an ILRT nation Category with steel E-A, Containment liner (11 Surfaces plants)
Beaver Valley PWR - Sub-Original. patches By plant per-IWE-3510.1; Stan-1 atmospheric over construction sonnel before dards for Exami-i primary con-related holes are an ILRT; During nation-Category i
tainment (7 deteriorating and E-A, Containment plants) falling out; rusted Surfaces; areas and peeled IWL-2510; Exami-
)
paint on the liner nation of Con-crete I
l
(
.T - 3.8
.,_.__.._.-.a__.-_--.._---_.-
FIGURE 3 DISTRIBUTION OF CONCRETE COMPONENT PROBLEM AREAS' ORNL-OWG 86-45C8 ETO.
NUMBER OF (DENTIFIED OCCURRENCES CONCRETE (77) 0 5
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(((((([
ANCHORAGE
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[//////
OVERTEMPERATURE N
DOME DE LAMINATION POST. TENSIONING SYSTEM (171 F AILED / CORRODED TENDONS bkhM ANCHORAGE CRACKING V/////b LIFTOFF LOAO b
3 Source: NUREG/CR-4652, " Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear Power Plants," D.J. Naus, p. 39, Figure 18.
F-3
TABLE 4-5 SECTION XI INSERVICE INSPECTION UPDATE SCHEDULE ISI NEXT.
YEARS UNTIL PLANT PERIOD UPDATE IMPLEMENTATIOHI WATTS BAR 1 WATTS BAR 2 l
-BROWNS FERRY 1 1
Mar 87 l
BROWNS FERRY 3 1
Mar 87
~;
ARKANSAS NUCLEAR 1 2
Dec 94 CALLAWAY 1 1
Dec 94 COOPER 2
Jul 94 i
INDIAN POINT 2 2
Jul 94 KEWAUNEE 2
Jun 94
'LASALLE 1 1
Jan 94 LASALLE 2 1
Oct 94 MCGUIRE 2 1
Mar 94
[
OCONEE 1 2-Jul 94 11 j
OCONEE 2 2
Sep 94 j
OCONEE 3 2
Dec 94 SEQUOYAH 1 1
Jul 94 SUMMER 1
Jan 94
[
TURKEY POINT 3 2
Dec 94 TURKEY POINT 4 2
Sep 94 WNP 2 1
Dec 94 ZION 2 2
Sep 94 BYRON 1 1
Sep 95 l
DIABLO CANYON 1 1
May 95 l
DUANE ARNOLD 2
Feb 95
-FITZPATRICK 2
Jul 95 GRAND GULF 1 1
Jul 95 I
MILLSTONE 2 2
Dec 95 PALISADES 2
Dec 95 PILGRIM 1 2
Dec 95 12 PRAIRIE ISLAND 1 2
Dec 95 PRAIRIE ISLAND 2 2
Dec 95 RIVER BEND 1 1
Jun 95 SEQUOYAH 2 1
Jul 95 t
SUSQUEHANNA 2 1
Feb 95 WATERFORD 3 1
Sep 95 I
WOLF CREEK 1
Sep 95 BRUNSWICK 1 2
Mar 96 BRUNSWICK 2 2-Nov 96 13 CATAWBA 1 1
Jun 96
' Based on a 1993 publication ~date for the rulemaking.
T-4 r
m, TABLE 4 (Continued)
SECTION XI INSERVICE INSPECTION ~ UPDATE SCHEDULE ISI NEXT YEARS UNTIL PLANT PERIOD UPDATE IMPLEMENTATION t CATAWBA 2 1
Aug 96 COOK 1 2
Jul 96 COOK 2 2
Jun 96 DIABLO CANYON 2 1
Mar 96 HATCH 1 2
Dec 96 HATCH 2 2
Aug 96 HOPE CREEK 1 1
Dec 96 INDIAN POINT 3 2
Aug 96 LIMERICK 1 1
Feb 96 13 MILLSTONE 3 1
Apr 96 NINE MILE POINT 1 2
Dec 96-NINE MILE POINT 2 1
Mar 96 PALO VERDE 1 1
Jan 96 PALO VERDE 2 1
Sep 96 PALO VERDE 3 1
Jan 96 l
SEABROOK 1 1
Jun 96 TROJAN 2
May 96
--BEAVER VALLEY 1 2
Apr 97 BEAVER VALLEY 2 1
Nov~97 BYRON 2 1
Aug 97 CLINTON 1 1-Apr 97 CRYSTAL RIVER 3 1
Mar 97 FARLEY 1-2 Dec 97 FARLEY 2 2
Jul 97 HADDAM NECK 2
Jan 97 14 PEACH BOTTOM 2 2
Jul 97 PEACH BOTTOM 3 2
Dec 97 PERRY 1 1
Nov 97 SHEARON HARRIS 1 1
May 97 7
.VOGTLE 1 1
May 97 VOGTLE 2 1
Jun 97 BRAIDWOOD 1 1
Jul 98 BRAIDWOOD 2 1
Oct 98 FERMI 2 1
Jan 98 I
NORTH ANNA 1 2
Jun 98' 15 SALEM 3 2
Jun 98 SOUTH TEXAS 1 1
Aug 98 i
ST.LUCIE 1 2
Feb 98
' Based on a 1993 publication date for the rulemaking.
T - 4.1 I
4
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'I TABLE 4 (Continued)
.i SECTION XI INSERVICE INSPECTION UPDATE SCHEDULE i
ISI NEXT YEARS UNTIL PLANT PERIOD UPDATE IMPLEMENTATIONI CALVERT CLIFFS 1 2
May 99.
CALVERT CLIFFS 2 2
Apr 99 16 COMANCHE PEAK 1 1
Dec 99 SOUTH TEXAS 2 1
Jun 99 ARKANSAS NUCLEAR-2 2
Mar 00 GINNA 3
Mar 00 LIMERICK 2 1
Jan 00 MILLSTONE 1 2
Dec 00 17 NORTH ANNA 2 1
Dec 00 POINT BEACH 1 2
Dec 00 3
SAN ONOFRE 1 2
Jan 00 BIG ROCK POINT 2
Dec 01 DAVIS BESSE 1 1
Nov 01 OYSTER CREEK 2
Dec 01 18 SALEM 2 1
Oct 01 a
THREE MILE ISLAND 1 1
Sep 01
. YANKEE-ROWE 3
Jun 01 l
BROWNS FERRY 2 1
Mar 02 DRESDEN 2 2
Jun 02 I
. DRESDEN 3 2
Nov 02 MAINE YANKEE 2
Dec 02 MCGUIRE 1 1
Dec 02 19 l
POINT BEACH 2 2
Sep 02 ROBINSON 2 2
Mar 02 SURRY 1 2
Dec 02 VERMONT YANKEE 2
Nov 02 FORT CALHOUN 1 2-Sep'03 MONTICELLO 2
Jun 03 QUAD CITIES 1 2
Feb 03 QUAD CITIES 2 2
Mar 03 SAN ONOFRE 2 1
Aug 03 SAN ONOFRE 3 1
Apr 03 20 ST.LUCIE 2 1
Aug 03 SURRY 2 _
2 May 03 SUSQUEHANNA 1 1
Jun 03 ZION 1 2
Dec 03 i
' Based on a 1993 publication date for the rulemaking.
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8 TABLE 5 - COMPARISON OF TENDON SAMPLING REOUIREMEHIq2 (LIFT-OFF TESTS)
Containment Tendon Total Number of Tendons Insoected Design Group Number ISI O 1, 3, 5 years ISI O 10, 15... years (Example)
Rev. 2 Rev. 3 Rev. 2 Rev. 3 Ten ns Type I Hoop 489 10 10 3
5 I
(Turkey Dome 165 6
7 3
3 Point 3)
Vert.
180 5
7 3
I 4
Invert Hoop 171 10 7
3 3
Type II Dome 90 6
4 3
3 (Arkansas)
Vert.
102 5
4 3
3 Invert Hoop 152 9
6 3
3 Type III Dome (Trojan)
Vert.
Invert 70 4
4 2
3 2Tendon sampling requirements are presented in terms of total number of tendons to be inspected at each inspection interval.
Representative containment types (I, II, or III) are shown with their typical tendon configurations (Number of Hoop, Dome, Vertical, or Inverted tendons).
The number of tendons to be inspected varies with Revision and with inspection number.
.l 3 From NUREG/CR-4712, D.J. Naus, " Regulatory Analysis of Regulatory Guide 1.35 (Revision 3, Draft 2) - In-Service Inspection of Ungrouted Tendons in Prestressed Concrete Containments."
T-5
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TABLE 68 - COMPARISON OF TENDON TEST REOUIREMENTS2-i Containment Tendon Total Number of Tendons Inspected Design Group Number ISI O 1, 3, 5 years ISI O 10, 15... years (Example) of Tendons Rev. 2 Rev. 3 Rev. 2 Rev. 3 Type I Hoop 489 (Turkey Dome 165 42 27 18 15 Point 3)
Vert.
180 i
Invert Hoop 171 Type II Dome 90 42 18 18 12 (Arkansas)
Vert.
102 Invert Hoop 152 Type III Dome 26 12 10 8
(Trojan)
Vert.
Invert 70 2 In determining the number of tendon tests, if only a tendon lift-off test is
- required, this is considered as one test.
If tendon detensioning if required in addition to a lift-off test, this is considered as two tests.
Revision 2 requires all tendons to be detensioned whereas Revision 3 requires that only one representative tendon of each group be detensioned.
i 3 From NUREG/CR-4712, D.J. Naus, " Regulatory Analysis of Regulatory Guide 1.35 (Revision 3, Draft 2) - In-Service Inspection of Ungrouted Tendons in Prestressed Concrete Containments."
T-6
s f
TABLE 78 ESTIMATED COST SAVINGS IN TENDON SAMPLING AND DETENSIONING REOUIREMENTS RESULTING FROM IMPLEMENTATION OF REVISION 32 containment Tendon Total Design Group Number Estimated Cost Savings ($1000)
(Example) of ISI 9 1, 3, 5 years ISI O 10, 15... years Tendons Type I Hoop 489-(Turkey Dome 165 195 40 Point 3)
Vert.
180 Invert Hoop 171 Type II Dome 90 312 80 (Arkansas)
Vert.
102 Invert Hoop 152 Type III Dome 182 26 (Trojan)
Vert.
Invert 70 2 Using a cost per test of $13,000, the cost savings resulting from-use-of Revision 3 as opposed to Revision 2 can be estimated for each of the example containment types.
l l
l 3
From NUREG/CR-4712, D.J. Naus, " Regulatory Analysis of Regulatory Guide 1.35 (Revision 3, Draft 2) - In-Service Inspection of Ungrouted Tendons in Prestressed Concrete Containments."
T-7
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l Table 8 - Costs of Tendon Surveillance Per Unit 1
i ER$'VISIONZ27 3REVIS5UNY5h 78
!.TENDONSIe W ^
M
$193K (15 tendons)
$165K (9 tendons)
~~
iCdUSDMA5$5$1 greasei:;L.
$18K
$18K igaskets;fshi}mst:Astc/)2i SUPPORTdCR$N57
.1.,
$65K
$65K ll(bbtilonal?someMasss)l:
REEUNT2!
$10K V
$10K
- . SUBTOTAL-
$286
$258
.fMAUPOW5' N6~: weeks,ilo.'; hbur5ayssf6 Uddys/Nk)2 R
S SUPERVISOR.i-
$23,760 12f0C?ENGINEERSS
$47,520 TisiRON/NORSER$5
$142,560
.._......i.......,e P E DIEMJCOSTSf_.(9t 2; $5,000
? people)1 e 5UBTOTA6d
$218,840
. TOTAL;
$504,840 (Rev. 2)
$476,840 (Rev. 3)
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t ADDendix A Details from Some of'the Instances of Dearadation 9
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P OCCURRENCES OF DEGRADATION Specifics on some of the reported occurrences of corrosion and degradation are given below.
Table 3 (which is attached) is a more complete list of reported instances.
Ovster Creek 9
Steel containment shell corrosion was found in the sand cushion region-(elevation 51 feet) of the drywell of Oyster Creek (BWR Mark I) in 1986 during L
an NRC inspection.
The drywell steel shell thickness had been reduced from 1.115 inch to an average thickness of 0.85 inch, with some local spots reduced to about 0.75-inch.
The drywell pressure vessel stresses do not exceed the ASME Code allowable stresses for a minimum local thickness of 0.425 inch provided the mean thickness remains above 0.591 inch. This evaluation of stresses took advantage of actual material properties obtained from Certified l
Mill Test Reports. The root cause of the corrosion is' the presence of very corrosive water solution in the sand cushion. The water leaked.through a' defective rubber gasket from the refueling pool on the top of the drywell into the air gap between the drywell and the concrete shield, then through the gap-forming material, which contains chlorides and sulfides, to the sand cushion.
Corrosion also occurred in the upper portion of the drywell where gap-forming material is present.
b After the discovery of corrosion in the sand cushion region of the steel drywell at Oyster Creek Plant in 1986, the licensee had committed to conduct ultrasonic test (UT) thickness measurements of the drywell sheli at' outages of opportunity (whenever a drywell entry is planned or required). On this basis, measurements were performed in September, 1989 and again in February, 1990.
An evaluation by the licensee of the data collected indicated a corrosion rate of 9.5 to 10 mils per year.
This rate is twice that established on the basis of previous UT measurements at the drywell elevation 51' region.
During a recent licensee examination, some local spots were found to have further degraded having lost up to % inch of the wall thickness.
t Nine Mile Point 1 Corrosion of the torus at Nine Mile Point 1 (BWR Mark I) was discovered i
during an NRC special announced team inspection.
The torus was designed and constructed uncoated. The UT of the torus shell showed several areas where the thickness of steel was at or near the minimum specified wall thickness..
i following the discovery of corrosion at Nine Mile Point 1, a survey of BWRs in NRC Region I was conducted.
Although all plants perform periodic visual inspections per technical specification requirements, the extent of this inspection varies (e.g., some plants only examine the torus above water line, t
and other plants use divers or cameras to inspect the torus underwater).
From these limited inspections, Fitzpatrick, Millstone 1, Pilgrim,' and Oyster Creek have found degradation of the coatings and have had to clean and recoat the surfaces.
Fitzpatrick had some pitting, and Pilgrim experienced flaking of th coating which led to some rusting of the torus.
A-1 l
1
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i Fort Calhoun. TMI-1. & Tro.ian Extensive grease leakage from tendon sheathings or at joints between conduit lengths has been found at Fort Calhoun, Three Mile Island 1, and Trojan plants.
In addition to grease leakage, Three Mile Island I and Trojan showed signs of leaching of the concrete in the tendon gallery.
Farlev 2 During a visual inspection of the Unit 2 containment at Farley, it was i
discovered that the grease cap of the shop-end of a vertical tendon was deformed. When the field-end of the tendon in the tendon gallery was opened for inspection, it was found that the anchor head was broken, allowing the-tendon to detension completely. An extensive investigation conducted to' understand the cause and extent of the occurrence showed that two additional anchor heads were broken, and the tendons were completely detensioned. The anchor heads of 23 other tendons (17 in Unit 2 and 6 in Unit 1) were found to have cracks.
The factors contributing to this event can be summarized as (1) high hardness material of the anchor heads, (2) free water in the grease caps, and (3) high stresses in the anchor heads.
Similar incidents have been reported at two other plants.
Turkey Point 3 1
Turkey Point Unit 3 performed their twentieth year tendon surveillance of the containment post-tensioning system.
Lift-off forces below the predicted lower limit were found in three of the hoop surveillance tendons and in six additional adjacent tendons. The average tendon lift-off force at one buttress was slightly below the minimum requirev grestress force. -Subsequent evaluation showed that higher local internal containment temperatures were responsible.
Even though it was shown that all of the tendon groups provide o
adequate prestress force to maintain containment integrity at this time, the NRC staff is concerned because the relaxation losses are greater than those that were predicted.
The most conservative estimate was that the hoop tendons will provide adequate prestress force to maintain design basis requirements until 1997. Although it is recognized that this was a conservative estimate, the results, should this happen, are that the minimum prestress forces for Unit 3 would be realized some fourteen years earlier than previously predicted.
1 McGuire 2 a
On August 24, 1989, the licensee for McGuire Unit 2 (PWR Ice Condenser)
~
reported base metal corrosion on the outside of the steel containment vessel-(SCV) at plant elevation 725 feet.which was discovered during a' pre-integrated leak rate test inspection.
The shell is enclosed in a reinforced concrete shield building, with a 6 foot annulus.. The failure of the coating has led to the corrosion of the base metal up to 0.123 inch.
The degradation of the shell, which'has a nominal thickness'of 1 inch at elevation 725', is limited to a 37 foot section no higher than 1-1/2 inches above the floor. - Corrosion that is up to 0.03 inch deep was also found in areas below the level of the annulus floor, where concrete was removed to expose the shell surface.
The A-2
's below-floor-corrosion is due to a lack of sealant at the interface between the shell and the floor. The condition that led to the SCV corrosion is attack by.
condensed boric acid coolant leaking from some of the instrumentation lines.
1 Drains are provided in the floor, but they are widely separated, and the floor is not sufficiently graded to prevent pooling of the condensate between.the drain locations. The design is considered adequate to handle large water spills, but leaking instrumentation lines was not considered. -This is a potential problem in the annular area at any BWR or PWR with a steel containment.
McGuire 1. & Catawba 1&2 Subsequently, the McGuire licensee found similar but less extensive corrosion in Unit 1.
The licensee also has two units at Catawba which are PWR Ice Condensers with SCV's similar to those at McGuire. An inspection of the Catawba SCV's found corrosion, but it was less extensive than that at the McGuire plants being limited to about a 15 foot section 1 inch above the annulus floor.
Using ASME Code calculations (NE-3200), the minimum required SCV wall thickness at the base for the Catawba containments was found to be 0.6875 inch as governed by external pressure. Given that the nominal wall thickness is 1 inch at all four units, and using the worst case corrosion rate at McGuire, which is 38.5 mils per year, the time frame for corrective action at McGuire would be 17 months, and would be 24 months for Catawba.
Information Notice 89-79 states that the degradation of the containment shells at the McGuire and Catawba plants is considered significant for several The fact that the corrosion affects four different units indicates reasons.
that other steel containments with similar configurations may be susceptible to the same problem.
Furthermore, the observed rate of corrosion far exceeds the allowance made-for corrosion in the containment design.
This condition leads to the concern that such corrosion could result in undetected wall l
thinning to less than the minimum design thickness, accompanied by a loss of leaktightness or structural integrity.
This problem can be prevented by a containment inservice inspection program that is adequate to ensure early detection and the maintenance of the intended licensing bases through proper corrosion control.
On April 18, 1990, additional areas of degradation of the steel containment at McGuire Unit I were found during an inspection by the licensee.
lhe degradation consists of general coating failures and localized pits having a depth of up to 45 mils. The corrosion is located on the inside surface at the floor level between the upper and the lower containment compartments, in the vicinity of the ice condenser. The corrosion occurs.in a 2-inch floor gap-filled with cork that interfaces with the coated steel containment. The cork ~
contains moisture originating most likely from the ice condenser or from condensation or both.
General surface corrosion, which is presently of no significance, appears throughout the areas accessible for inspection. The l,
licensee had done UT on inaccessible areas, and the worst pitted area still meets minimum thickness requirements.
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NRC Information Notice Number' 89-79, Supplement 1, specifies that -.
although corrosion of the containment shell at the cork interface of the floor expansion joint has been discovered only at-McGuire Unit 1, it is expected that.such corrosion will-likely occur at other plants-with the same design details for the floor expansion joint. There are indications that cork may have been used in foundation level expansion joints -in other plants. The additional corrosion in McGuire Unit I has occurred at locations previously considered not susceptible to corrosion.
The information notice further states that the detection of corroded steel plate material in the drywells and wet wells of BWR plants and corroded steel containments of PWR plants has resulted in the concern that degradation caused by corrosion may be generic to all types of containments.
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i; ENCLOSURE 3
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p ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT t
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_4.
Environmental Assessment ajLd Findino'of No Sionificant Impact Environmental Assessment Identi ficatit,. of Proposed Action The proposed rule would expand existing references in the NRC regulations to the American Society of Mechanical Engineers Boiler and Pressure Vessel. Code (ASME Code).
Specifically, the proposed rule would incorporate by reference the 1992 Edition with addenda through the 1992 Addenda of Subsection IWE,
" Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants," and of Subsection IWL, " Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants," of Section XI, Division 1, of the ASME Code.
Need for the Proposed Action The NRC is proposing this 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 Section_III or_Section VIII of the ASME.
Code, and as provided for'by the American Concrete Institute (e.g., ACI-318),-
(as reflected in license conditions, technical specifications, and' licensee commitments).
The NRC also believes enhanced ISI examinations are needed and are justified to supplement existing requirements specified in GDC 53' and r
Appendix J.
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.
The rate of occurrences of corrosion and degradation of containments.has been increasing at operating nuclear power plants.
In a number of the incidents reported, areas of some containments had degraded to an unacceptable condition.
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 other recent incidents which involved grease leakage from tendons have been~ investigated.
In addition to grease 1
t
.l leakage, these incidents showed signs of leaching of the concrete.
l Over one-third of the operating containments have experienced corrosion or other degradation. Almost one-half of these occurrences were first identified l
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: steel containment.shell corrosion'in the drywell sand cushion region (wall thickness reduced to below minimum design i
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 i
performance of structures, systems, and components important to safety in water-cooled nuclear power plants.
Criterion 16, " Containment design,"
l requires the provision of reactor containment and associated systems to i
estublish an essentially leak-tight barrier against the uncontrolled release i
of radioactivity in to 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 i
are reflected in license conditions, technical specificati' ns, and 1icensee o
commitments for the operating plants.
Criterien 53, " Provisions for containment testing and inspection," requires i
~
that the reactor containment design permit: (1) appropriate periodic i
inspection of all important areas, such as penetrations;. (2) an appropriate surveillance program; and (3) periodic testing at containment design pressure 4
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 icBkage 1
(1) after the containment has been completed and is ready for operation, rate:
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 s
have been cases on 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 an unacceptable level of degradation that was not detected by 2
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I 4
these programs.
The NRC believes that more specific ISI requirements, that expand upon existing requirements for the examination of containment structures in accordance.with General Design Criteria (GDC) 53, Appendix A to 10 CFR Part 50, and Appendix J to 10 CFR-Part 50, are needed and are justified for l
the purpose of ensuring that containments continue to maintain or exceed L
minimum accepted design wall thicknesses and prestressing forces as provided
{
for in Section III or Section VIII of the ASME Code, and as provided for by the American Concrete Institute (e.g., ACI-318), (as reflected in license conditions, technical specifications, and written licensee commitments).
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 lWL of the ASME Code proposed for incorporation by reference into 10 CFR I
50.55a.
There are five modifications, which are contained in one paragraph of the proposed rule, that 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 ~
l result in degradation of these areas.
A limitation specifies the 1992 Edition with 1992 Addenda of Subsection IWE and Subsection IWL as the earliest version of the ASME Code the staff-finds acceptable.
This edition and addenda combination _ incorporates the concept of I
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 pipir:g were the areas of greatest As containments have aged, degradation of base metal, rather than concern.
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 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.
3
I The NRC has reviewed the 1992 Edition'with the 1992i ddenda of Subsection IWE A
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 oy incorporating rules into the regulatory process that.have t
received industry participation in their development and acceptance by the NRC.
i Environmental Impact of the Proposed Action
+
The proposed rule would improve the structurally reliability of Class MC components, and their integral attachments, and of metallic shell -and
~ '
penetration liners, reinforced concrete and the post-tensioning systems of Class CC components, and their integral attachments. Because of the improved structurally reliability of these components, 'the proposed rule could, in the event of an accident, reduce the amounts and types of any effluents that may be released offsite and that there should be no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the Commission concludes that this proposed rule would result in no significant radiological environmental impact.
The proposed rule does not affect non-radiological plant effluents and has 'no other environmental impact.
Therefore, the Commission concludes-that there are no significant non-radiological environmental impacts associated'with the proposed amendment.
Alternative to the Proposed Action The principal alternatives to the technical actions in the proposed _ rule would -
be to not incorporate the new subsections into the NRC regulations.
Since.the Commission has-already concluded that no significant environmental effect would result from the proposed rule, the specified alternatives, which would l
have no safety benefit, would not reduce the environmental impact of plant construction, and inservice examination and inservice testing.
Agencies and Persons Consulted f
The NRC staff prepared the proposed rule in consultation with personnel from the Idaho Engineering Laboratory (Idaho Falls, ID).
Finding of No Sionificant Impact l
The Commission has determined not to prepare an environmental impact statement for the proposed rulemaking.
Based upon the foregoing environmental assessment, the Commission concludes that proposed action will not have a significant effect on the quality of the human environment.
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SuoDortino Statement for Information Collection Recuirements in-10 CFR 50.55a i
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'k SuoDortina' Statement for Information Collection Reouirement's in Proposed Rule. 10 CFR 50.55a (3150-0011)
Dgscriotion of the Information Collection-Under 10 CFR_50.55a, each operating license for a boiling or pressurized l
water-cooled nuclear power facility must meet specific requirements of the ASME Boiler and Pressure Vessel Code. These requirements are incorporated by reference to avoid additional burden to industry and unnecessary duplication of requirements. This rulemaking would incorporate the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL,'of Section XI, Division 1.
Implementation of Subsections IWE and IWL requires the owner to prepare the following:
Plans and schedules for preservice and inservice examination and tests to meet the requirements of Subsection IWE and Subsection IWL; Preservice and inservice inspection summary reports for Class 1 and 2 pressure retaining components and their integral supports.
Records of the examinations, tests, replacements, and repairs.
Specifically, the following recordkeeping requirements are incurred:
IWE-1232 (a)(2). Inaccessible Surface Areas - The procedures for radiography and leak testing, personnel qualifications, and examination results must' be documented _for all. welded joints that are inaccessible for examination.
IWE-1232 (b)(1). Inaccessible Surface Areas :- The procedures for magnetic particle or ultrasonic examination, radiography, and leak testing; personnel qualifications; and examination results must be documented for all portions of Class CC metallic shell and penetration. liners embedded in concrete or otherwise made inaccessible during construction or as a result of repair or replacement.
IWE-2200 (d). Preservice Examination - When a vessel, _ liner, or portion thereof is repaired or replaced during the service lifetime of a plant, the preservice examination requirements for the vessel repiir or replacement must be documented that the repair meets the acceptance _ criteria.
IWE-2200 (e). Preservice Examination -
The procedures, personnel qualifications, and examination. results. must be documented for welds made as part of a repair or a replacement if examined by the magnetic particle or liquid penetrant method.
IWE-2200 (a). Preservice Examination After repaired or replaced welds are examined by the magnetic particle or' liquid penetrant method, if coatings are reapplied, the condition of-the new paint or coating shall be documented in the 'preservice 1
examination records.
IWE-2500 (c)(2). Examination and Pressure Test Recuirements -
Procedures for measurements in accordance with Section V, T-544, personnel qualifications, and examination results must be documented for augmented examinations of surface areas accessible from one side only.
JWE-3112 (a). Acceptance Components containing flaws that cv act exceed acceptable standards are acceptable for service, provided the flaws are recorded in terms of location, size, shape, orientation, and distribution within the component.
IWE-3114.
Repairs and Reexaminations Repairs and reexaminations must be recorded on Form NIS-2, Owner's Report-for Repairs or Replacements, and demonstrate that the repair meets the acceptance standards.
IWE-3122.1. Acceptance by Examination Components with examination results that meet the acceptance standards are acceptable for continued service.
Verified changes of flaws from prior examinations shall be reported in inservice inspection summary reports.
IWE-3122.2. Acceptance by Repair Components whose examination results reveal flaws that do not meet acceptance standards shall be unacceptable for continued service.
Repairs or mechanical Nmoval of unacceptable components must be documented on Form NIS-2, Owner's Report for Repairs or-Replacements.
IWE-3122.3. Acceptance by Replacement - As an alternative to IWE-3122.2, Acceptance by Repair, the component-or the component portion containing the flaw may be replaced.
If welding is required, documentation is required for welding procedures, welder certification and qualifications, and a Certified Material Test Report foi.the welding material.
IWE-3124 Repairs and Reexaminations The results of reexaminations must be documented and demonstrate that the repair meets acceptance standards.
IWE-3130. Inservice Visual Examinations - Components whose visual examination reveals areas that are unacceptable for continued service must be documented that the acceptance requirements-of IWE-3120 are satisfied.
IWE-3510.1 (b). Visual Examinations - Containment Surface 1 --
Prior to conducting a Type A test, conditions that may affect containment structural integrity or leak tightness shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for 2
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Repairs or Replacements.
IWE-3510.2. Visual Examinations. VT-3. on Coated Areas Containment Surfaces Coated areas.that may show signs of flaking, blistering, peeling, discoloration, and other signs of distress shall be accepted by engineering evaluation or
. corrected by repair or replacement and be documented on Form NIS-2, Owner's Report for Repairs or Replacements.
IWE-3510.3. Visual Examinations. VT-3. on Noncoated Areas -
Containment Surfaces Noncoated areas that show signs of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs-or Replacements.
IWE-3511.1. Visual Examinations. VT-3. on Coated Areas '-
Er_ essure Retainino Welds -
Coated areas that show signs of~
flaking, blistering, peeling, discoloration, and other signs of distress shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs or Replacements.
IWE-3511.2. Visual Examinations. VT-3. on Noncoated Areas -
Pressure Retainino Welds - Noncoated areas that show signs of
- cracking, discoloration,
- wear, pitting, excessive corrosion, are strikes, gouges, surface discontinuities, dents, and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs or Replacements.
IWE-3512.1. VT-1 Visual Examinations Coated Areas -
Coated-l areas that show signs of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes,. gouges,. surface discontinuities,
- dents, and other. signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs or Replacements.
IWE-3512.2. VT-1 Visual Examinations Noncoated Areas Noncoated areas that show signs of cracking, discoloration, wear, pitting, excessive' corrosion, arc strikes, gouges, surface discontinuities, dents, and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs or Replacements.
IWE-3512.3. Ultrasonic Examination Containment vessel examinations that reveal material loss exceeding 10% of the 3
nominal containment wall thickness, or material loss that is projected to exceed 10% of the nominal wall thickness prior to the next examination shall be accepted by engineering evaluation or ~ corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs or Replacements.
IWE-3513.1. Visual Examinations. Seals. Gaskets. and Moisture Barriers -
Seals, gaskets, and moisture barriers shall be examined for wear, damage, erosion, tear, surface cracks, or j
other defects that may violate the leak-tight integrity and documented on Form NIS-2, Owner's Report for Repairs or Repl acement s.
IWE-3515.1. Visual Examinations. Pressure Retainino Boltina -
Bolting materials shall be examined in accordance with the material specification for defects which ma" cause the bolted connection to violate either the leak-tight or structural integrity.
Replaced defective items shall be documented on Form NIS-2, Owner's Report for Repali or Replacements.
1 Article IWE-4000, Repair Procedures, are covered by the rules of IWA-4000.
IWA-4130. Repair Program Repair operations shall - be performed in accordance with a program that delineates the essential requirements.
Prior to authorizing a repair, the Owner shall evaluate the suitability of the repair.
Storace and Handlina of Weldino Material Procedures for welding material control shall be included in the repair program. Welding material must be certified by a material test report.
IWA-4340. Defect Removal Procedures for the removal of defects, personnel qualification, and examination results shall be documented.
IWA-4400. Weldinc and Welder Qualifications (Including Weldina Operators)
Welding procedures, welder certifications, l
personnel qualifications, and examination results must be documented.
IWA-4600dxamination - The repaired areas shall be examined to establish a new preservice record. The method that detected the flaw shall be included in the record.
IWL-2523.2. Sample Examination and Testina - Tension tests performed on each removed wire or strand shall be recorded with yield strength, ultimate tensile strength and elongation.
4
1 IWL-2524.1. Visual Examination-- Visual examinations of tendon anchorage areas shall be documented and include the physical-condition of each area.
IWL-2524.2. Free Water Documentation - The quantity of free water contained in the anchorage end cap as well as any which drains from the tendon during the examination process shall-be documented.
IWL-2526. Removal and Replacement of Corrosion Protection Medium - The amount of corrosion protection medium removed for samples shall be measured. The total amount replaced and the difference between the two amounts shall be documented.
IWL-3310. Evaluation Report - The owner shall prepare an Engineering Evaluation Report for items with examination results that do not meet the acceptance standards of IWL-3100~
or IWL-3200.
IWL-7120. Replacement Proaram -A replacement plan must document the removal, reinstallation and replacement of post-tensioning system items for concrete containments.
i In addition, the following requirements, which are modifications to Subsection IWL, must also be submitted by report to the NRC:
50.55a(b)(2)(ix)(B) - An Engineering Evaluation Report, when consecutive surveillances of tendon 'prestressing forces indicate that the tendon force would be less than' the minimLm i
design prestress requirements.
50.55a(b)(2)(ix)(C) - A difference of more than 10% (from that recorded during installation of the tendons).in elongation corresponding to a specific load during detensioning and retensioning of tendons.
50.55a(b)(2)(ix)(D)
Sampled sheathing filler grease' containing chemically combined water exceeding 10% by weight, or replaced grease exceeding 10% of the n, Nct volume.
50.55 afb)(2)(ix)(E) - An evaluation of 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.
A A.
-JUSTIFICATION 1.
Need for the Collection of Information NRC regulations at 10 CFR 1 50.55a incorporate by reference Division 1 rules of Section XI, " Rules for Inservice Inspection of 5
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0 Nuclear Power Plant Components," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
1 Section XI sets forth the requirements to which nuclear power plant components are' tested and inspected.
There are existing recordkeeping requirements in Section XI.
The proposed rule would incorporate by reference the 1992 Edition with Addenda through the 1992 Addenda of Subsection IWE,
" Requirements for Class MC Components of Light-Water Cooled Power Plants," and Subsection IWL,
" Requirements for Class CC Components of. Light-Water Cooled Power Plants," of Section XI (Division 1), of the ASME Code.
Sub'section IWE provides the rules and requirements for inservice inspection of Class MC pressure retaining-components and their integral attachments, ond metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments in i
light-water cooled power plants. Subsection IWL provides the rules and requirements for preservice examination and inservice inspection of the reinforced concrete and the post-tensioning systems of Class CC components.Section XI records are needed to document the plans for and results of inservice inspection and inservice test programs.
The records developed are generally not collected by the NRC, but are retained by the licensee to be made available to the NRC in the event of an NRC audit.
Section XI, Division I,
requirements for inservice inspection records are provided in IWA-6000, " Records and Reports."
The following records and reports identified in IWA-6000 must be maintained for the component or system. These records and reports are:
Index to record file Preservice and inservice inspection plans Preservice and inservice inspection reports Repair records and. reports Replacement records and reports Nondestructive examination procedures Nondestructive examination records IWA-6310 states that the records and reports shall be filed and maintained in a manner which will allow access by the Inspector.
The Owner also shall provide suitable protection from deterioration and damage for all reccrds and reports, in accordance with the Owner's Quality Assurance Program, for the service lifetime of the component or system.
Lifetime retention of'-the above records is necessary to ensure adequate historical information on'the design, examination, and testing of components and systems to provide a basis for evaluating degradation of these components and systems at any time during their service' lifetime.
IWA-6240 requires that ISI Summary Reports be submitted to the regulatory and enforcement authorities having jurisdiction at the plant site.
The requirements of IWA-6240 and-IWA-6310 were incorporated into previous changes to 9 50.55a, and this proposed 6
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rulemaking action, therefore, doe's not require an additional recordkeeping burden.
2.
Acency Use of Information The records are generally historical in nature and provide data on which future activities can be based. The practical utility of the informat ion collection for NRC is that appropriate records are available for auditing by NRC personnel to determine if ASME Code-provisions for construction, inservice inspection, and inservice testing are being properly implemented in accordance with 10 CFR 50.55a, or whether specific enforcement actions are necessary.
3.
Reduction of Burden Throuch Information Technoloov The information being collected represents the documentation for the various plant specific construction, inservice inspection, and inservice testing programs. The NRC has no objection to the use of new information technologies and generally encourages their use.
4.
Effort to Identify Duplication ASME Code requirements are incorporated by reference into the NRC regulations to avoid the need for writing equivalent NRC-requirements.
This amendment will not duplicate the information l
collection requirements contained in any other regulatory requirement.
S.
Effort to Use Similar Information The NRC is using the information reporting requirements specified in j
the ASME Code in lieu of developing its own equivalent requirements.
i 6.
Effort to Reduce Small Business Burden This amendment to 10 CFR 50.55a affects only the licensing and-operation of nuclear power plants.
The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act in the Small Business Size Standards issued by the Small Business Administration at 13 CFR Part 121.
7.
Consecuences of less Frecuent Collection The information is generally not collected, but is retained by the licensee to be made available to the NRC in the event of an NRC 7
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audit.
8.
Circumstances Which Justify Variation from OMB Guidelines The record retention periods for information requested is frequently for the service lifetime of the applicable component. Such lifetime retention of records is necessary to ensure adequate historical information on the examination and testing of components to provide
- l a basis for evaluating degradation of these components and systems at any time during their service lifetime.'
9.
Consultations Outside the NRC The NRC staff prepared the proposed rule in consultation with l
personnel from the Idaho National Engineering Laboratory (Idaho
- Falls, ID),
and ISI Containment Specialists from General Dynamics / Electric Boat Division, Nuclear Engineering (Groton, CT) and Multiple Dynamics Corporation (Southfield, MI).
The proposed rule will be published in the Federal Reaister for comment.
i 10.
Confidentiality of Information
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NRC provides no pledge of confidentiality for this collection of information.
11.
Justification for Sensitive Questions No sensitive questions are involved.
I 12.
Estimated Annualized Cost to the Federal Government NRC inspection personnel who audit plant quality assurance records would include in their audit verification that the above records are being properly prepared and maintained.
The time associated with NRC inspectors verifying these records would be small when the activity is performed as part of a normal quality assurance audit.
13.
Estimate of Burden 7
a.
Number and Tvoe of Resoondents The recordkeeping requirements incurred by 10 CFR 50.55a through incorporation by reference of the ASME Code would apply to the 117 nuclear power plants presently under construction or in operation.
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b.
Estimated Hours Reouired'to Respond to the Collection The incorporation by reference of Subsections IWE and IWL into 10 CFR 50.55a would require each licensee to develop an initial inservice inspection (ISI) plan, implement that ISI plan, and then develop and implement 10-year updates to that ISI plan.
The development of the initial ISI plan is estimated to average 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per year per plant over a 4 year period. Development of the initial inservice inspection plan is a one-time effort.
Total annual burden for the development of the ISI plan is estimated at 117,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> times 117 plants) each year for 4 years.
It is estimated that implementation of the ISI plan would require 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year for each plant performing ISI of the containment.
Assuming that on the average 12 plants per year would be performing ISI of the containment, this would result in an industry burden of 9,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> per year.
The reporting burden of Sections 50.55a(b)(2)(ix)(B), (C), (D), and (E), which are modifications to Subsection IWL, that must also be reported in the ISI summary report are estimated to average 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per plant per year for responding and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per plant per year for reporting. The efore, the total burden estimated for the ISI plan would be 9,840 hours0.00972 days <br />0.233 hours <br />0.00139 weeks <br />3.1962e-4 months <br /> per year or 820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br /> per plant.
Every 10 years each licensee must update the ISI plan.
Update of the plan is estimated to average 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> per plant. Assuming that 12 plants per year would be updating their containment ISI plans, this would result in an industry burden of 2,160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> per year.
I'TEH '
ANNUAd
/ NUMBER OFdtANTUPEM iTOTACINNUAC
- RECORDKEEPING W (HOURS 2
'sYEARS
i' HOURS 1/1 PLANT.
Development' 1000 117 117,000 Periodic ISI 820 12 9,840 Update 180 12 2,160 TOTAL 1000 12,000 2 - Development is not added into the total number of hours as this activity will be completed before the update or periodic ISI activities begin.
c.
Estimated Cost Reouired to Respond to the Collection Based upon the hours specified in Item _ b. above, and a rate of 5132/hr., it is estimated that the cost to the industry for responding to the information collection required by the proposed 9
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T amendment to 9 50.55a is a total of $17,028,000 (117,000 + 2.160 + -
9,720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> X 5132/ hour) for the first four years, and 51,584,000 (12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> X $132/ hour) thereafter.
14.
Reasons for Chanae in Burden i
The change in burden results from incorporation by reference through this proposed amendment into the NRC regulations of the two new ASME s
Code Section XI Subsections, IWE and
- IWL, which contain-recordkeeping requirements.
15.
Publication for Statistical Use This information will not be published for statistical use.
B.
COLLECTION OF INFORMATION EMPLOYING STATISTICAL METHODS Statistical methods are not used in the collection of the ' required information.
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ENCLOSURE 5 CONGRESSIONAL COMMITTEE LETTERS e
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,o The Honorable Richard H. Lehman, Chairman Subcommittee on Energy and Mineral Resources Committee on Natural Resources United States House of Representatives Washington, DC 20515
Dear Mr. Chairman:
Enclosed for the information of the Subcommittee are copies of a notice of a proposed rulemaking to amend s 50.55a of 10 CFR Part 50 which would incorporate by reference national codes and standards for the inservice inspection of nuclear power plant components.
This section of the regulations incorporates by reference Division 1 rules of Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The Nuclear Regulatory Commission (NRC) proposes to amend these 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 ASME Code.
These subsections have not been previously endorsed by the NRC. This proposed rule continues the NRC process of reviewing and, as appropriate, incorporating by reference ASME Code rules for the inservice inspection of components, which until now has been limited to Class 1, Class 2, and Class 3 components.
Endorsement of these subsections at this time is considered necessary because significant corrosion and degradation of containments has occurred increasingly at operating nuclear power plants as evidenced by the number of reported incidents.
The proposed rule would:
e For the first time, incorporate by reference Subsection IWE and Subsection IWL, of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. 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 specified modifications these subsections of Section XI provide an acceptable method for detecting degradation of metal and concrete containments before stuctural integrity is compromised.
Existing regulatory requirements contain general requirements-applicable to containment inspection and surveillance, but these regulations do not provide sufficiently specific guidance on how to perform the necessary containment examinations.
This has resulted in a large variation in licensee containment examination programs.
In spite of present requirements, some containment structures have undergone unacceptable degradation which was not detected by the mandated tests and examinations.
T The Honorable Richard H. Lehman 2
Require licensees to expedite implementation of the Subsection IWE and Subsection IWL containment examinations by completing the first examination within 5 years of the effective date of this rule. This expedited examination schedule is necessary to prevent a delay in the implementation of Subsection IWE and Subsection IWL and~ to establish an early baseline for future examinations.
4 Include modifications to the endorsement of Subsection IWL to address four issues that are addressed in NRC Regulatory Guide I.35, Revision 3, " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures," but are not' currently addressed in Subsection IWL. Because of the importance the NRC attributes to these issues, each issue has been addressed-in the proposed rulemaking in a-modification to the endorsement of Subsection IWL.
Include a modification to the endorsement of Subsection IWE to address the NRC staff concern that Section XI.does not require examination of inaccessible areas.
Investigations of recent incidents of degradation of visible concrete indicate that such degradation could be indicative of degradation in related inaccessible areas.
In view of the routine nature of the amendment, we do not consider that a public announcement is warranted.
Sincerely, l
l-Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosure:
As stated cc: Representative Barbara Vucanovich x
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The Honorable Joseph Lieberman, Chairman Subcommittee on Clean Air and Nuclear Regulation Committee on Environment and Public Works United States Senate Washington, DC 20510
Dear Mr. Chairman:
Enclosed for the information of the Subcommittee are copies of a notice of a proposed rulemaking to amend 9 50.55a of 10 CFR Part 50 which would incorporate by reference national codes and standards for the inservice inspection of nuclear power plant components.
This section of the regulations incorporates by reference Division I rules of Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The Nuclear Regulatory Commission (NRC) proposes to amend these 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 ASME Code.
These subsections have not been previously endorsed by the NRC.
This proposed rule continues the NRC process of reviewing and, as appropriate, incorporating by reference ASME Code rules for the inservice inspection of components, which.until now has been limited to Class 1, Class 2, and Class 3 components.
Endorsement of these subsections at this time is considered necessary because significant corrosion and degradation of containments has occurred increasingly at operating nuclear power plants as evidenced by the number of reported incidents.
The proposed rule would:
For the first time, incorporate by reference Subsection IWE and e
Subsection IWL, of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code.
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 specified modifications these subsections of Section XI provide an acceptable method for detecting degradation of metal and concrete containments before stuctural integrity is compromised.
Existing regulatory requirements contain general requirements applicable to containment inspection and surveillance, but these regulations do not provide sufficiently specific guidance on how to perform the necessary containment examinations. This has resulted in a large variation in licensee containment examination programs.
In spite of present requirements, some containment structures have undergone unacceptable degradation which was not detected by the mandated tests and examinations.
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i The Honorable Joseph Lieberman 2
Require. licensees to expedite implementation of the Subsection IWE and Subsection IWL containment examinations by completing the first examination within 5 years of the effective date of this rule. This expedited examination schedule is necessary to prevent a delay in the implementation of Subsection IWE and Subsection IWL.
and to establish an early baseline for future examinations.
Include modifications to the endorsement of Subsection IWL to address four issues that are addressed in NRC Regulatory Guide I.35, Revision 3, " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures," but are not currently addressed in Subsection IWL.
Because of the importance the NRC attributes to these issues, each issue has been addressed in the proposed rulemaking in a modification to the endorsement of Subsection IWL.
Include a modification to the endorsement of Subsection IWE to address the NRC staff concern that Section XI dcas not require examination of inaccessible-areas.
Investigations of recent incidents of degradation of visible concrete indicate that such degradation could be indicative of degradation in related inaccessible areas.
In view of the routine nature of the amendment, we do not consider that a public announcement is warranted.
Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosure:
As stated cc: Senator Alan K. Simpson i
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p WASHING TON, D. C. 20555 4,...../
The Honorable Philip Sharp, Chairman Subcommittee on Energy and Power Committee on Energy and Commerce United States House of Representatives Washington, DC 20515
Dear Mr. Chairman:
Enclosed for the information of the Subcommittee are copies of a notice of a proposed rulemaking to amend 6 50.55a of 10 CFR Part 50 which would incorporate by reference national codes and standards for the inservice inspection of nuclear power plant components.
This section of the regulations incorporates by reference Division I rules of Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
The Nuclear Regulatory Commissien (NRC) proposes to amend these 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 ASME Code.
These subsections have not been previously endorsed by the NRC.
This proposed rule continues the NRC process of reviewing and, as appropriate, incorporating by reference ASME Code rules for the inservice inspection of components, which until now has been limited to Class 1, Class 2, and Class 3 components.
Endorsement of these subsections at this time is considered necessary because significant corrosion and degradation of containments has occurred increasingly at operating nuclear power plants as evidenced by the number of reported incidents.
The proposed rule would:
For the first time, incorporate by reference Subsection IWE and Subsection IWL, of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. 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 specified modifications these subsections of Section XI provide an acceptable method for detecting degradation of metal and concrete containments before stuctural integrity is compromised.
Existing regulatory requirements contain general requirements applicable to containment inspection and surveillance, but these regulations do not provide sufficiently specific guidance on how to perform the necessary containment examinations. This has resulted in a large variation in licensee containment examination programs.
In spite of present requirements, some containment structures have undergone unacceptable degradation which was not detected by the mandated tests and examinations.
The Honorable Philip Sharp 2
Require licensees to expedite implementation of the Subsection IWE and Subsection IWL containment examinations by completing the first examination within 5 years of the effective date of this rul e.
This expedited examination schedule is necessary to prevent a delay in the implementation of Subsection IWE and Subsection IWL and to establish an early baseline for future examinations.
Include modifications to the endorsement of Subsection IWL to address four issues that are addressed in NRC Regulatory Guide 1.35, Revision 3, " Inservice Inspection of Ungrouted. Tendons in Prestressed Concrete Containment Structures," but are not currently addressed in Subsection IWL.
Because of the importance the NRC attributes to these issues, each issue has been addressed in the proposed rulemaking in a modification to the endorsement of Subsection IWL.
Include a modification to the endorsement of Subsection IWE to a
address the NRC staff concern that Section XI does not require examination of inaccessible areas.
Investigations of recent incidents of degradation of visible concrete indicate-that such degradation could be indicative of degradation in related inaccessible areas.
In view of the routine nature of the amendment, we do not consider that a public announcement is warranted.
Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosure:
As stated cc: Representative Michael Bilirakis t
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ENCLOSURE 6 DISCUSSION OF JUSTIFICATION AS'A SAFETY ENHANCEMENT L
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D_ISCUSSION OF JUSTIFICATION AS A SAFETY ENHANCEMENT SUBSTANTIAL SAFETY IMPROVEMENT RATIONALE 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 e
leaching of the concrete.
Table 3 of Enclosure 2 lists many of these occurrences of degradation.
Over one-third of the operating containments have experienced corrosion or other degradation. Almost one-half of these 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: 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.
The staff surveyed the NRC Regional Offices to determine the type of inspections being performed on containment structures and to determine the effectiveness of the visual inspection as is currently required in 10 CFR Part 50, Appendix J.
Based on the survey of licensees (performed by NRC regional inspectors) and on experiences of NRC regional inspectors, the NRC has determined that there are great differences among plants with regard to the performance and the effectiveness of containment inspections.
The NRC believes that more specific ISI requirements, which expand upon existing requirements for.the examination of containment structures in accordance with General Design Criteria.(GDC) 53, Appendix A to 10 CFR Part 50, and Appendix J to 10 CFR Part 50, will improve significantly the ability to detect degradation and take timely action to correct degradation of containment structures.
One of the containment's functions is to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity into the-environment should an accident occur.
The_ role of an inservice inspection program is to uncover any evidence of structural deterioration that may affect either the containment structural integrity or leak-tightness.
The proposed-action-will provide for improved periodic examination of containment structures and better assure that the critical areas of containments are periodically inspected to detect defects that could compromise the containment's pressure-retaining and leak-tight integrity.
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At present, 10 CFR 50.55a specifies requirements for the inservice inspection and inservice testing of ASME Code Class 1, Class 2, and Class 3 components.
While this includes principal components within the nuclear steam supply system, it does not include metal or concrete containments. The proposed rule would expand the scope of 10 CFR 50.55a to include inspection of Class MC (metal) and Class CC (concrete) containments. The NRC has found that these new subsections of Section XI, which is a national consensus stzndard that is developed with NRC participation, would provide an acceptable method for detecting degradation of metal and concrete containments before margins in structural integrity are seriously compromised.
It is anticipated that 1 l
50.55a would be amended, as appropriate, to include later editions and addenda of Class MC and Class CC similar to the current process for ASME Code Class 1, 2, and 3 components.
l DISCUSSION OF COSTS 1
Incorporating by reference Subsection IWE and Subsection IWL, of Section XI, Division 1, of the ASME Code will establish the NRC staff position on the examination of steel containment structures and metal liners of concrete containments, and concrete containments and reinforcing systems of concrete containments on a generic basis for applicants and licensees, thereby minimizing the need for case-by-case evaluations and reducing the time and effort required for submittal preparations and licensee reviews.
l The value and impact of ASME Code revisions are balanced by the manner in which these revisions are achieved through the American National Standards Institute (ANSI) consensus process. The ANSI consensus process ensures that participation in ASME Code development is open to all persons and organizations that might reasonably be expected to be directly and materially I
affected by the activity, and ensures that such persons and organizations have l
the opportunity for fair and equitable participation without dominance by any l
l single interest. Consensus is established when substantial agreement has been l
achieved by the interests involved. Consensus requires that all views and H
objectives be considered, and that a concerted effort be made toward resolution. ASME Code proposed revisions are published for public comment in l
the ASME Mechanical Enoineerino and ANSI Reporter publications prior-to being i
submitted for final ASME and ANSI approval. Adverse public comments are referred to the appropriate technical committee for resolution.
The consensus process ensures a proper balance between utility, regulatory and other interests concerned with revisions to the ASME Code, and l
ensures that the value of any Code revision is consistent with its impact.
Implementation of the new Code rules requires certain additional i
information collection requirements.
The Supporting Statement for Information Collection Requirements in 10 CFR 50.55a is provided in Enclosure 4 to the SECY Paper.
The proposed rule-will not have a significant economic effect 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 6-2 l
fall within the s_ cope of the' definition of "small entities" set forth in the Regulatory flexibility Act in 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, thi:: proposed rule does not fall in the province of this Act.
Summary of Cottji:
Using the costs for Subsection IWE and Subsection IWL examinations (which are discussed in detail below), and sirrply averaging the cost ~over a ten-year ISI period, shows that the costs per year are minimal.
For Subsection IWE, the averaged costs are approximately $75,000 per year.
For Subsection IWL, the average is approximately 555,000 per year.
The staff believes that the direct and indirect costs of implementation are justified'in view of the reasons for this compliance backfit.
The Estimated Resource Burden on the NRC:
It costs $15,000 - $20,000 to review an entire inservice inspection (ISI) program.
Because the Subsection IWE and IWL ISI programs are new, it is estimated that the costs to the NRC would be $4,000 - $5,000 for each Subsection IWE/lWL review.
Taking into account the augmented examination schedule (submittal of an ISI plan by each licensee within 5 years of the publication date of the rule) and basing the estimation on twenty-reviews per.
year, the costs to the NRC would be $80,000 - $100,000 per year for the first 5 years. Also, based on the expected questions of intent and interpretation which always accompany any new ISI program, another man-year might be required (taking into account regional personnel, writing of Safety Evaluation Reports, etc.).
This would be a total of 2 man-years for the first 5 years. After Subsections IWE and IWL have been-implemented by all of the licensees, it is anticipated that subsequent ISI review and staff time would be approximately 1 man-year per year.
In order to alleviate this burden on the NRC staff, the proposed rule will not require submittsl. of the initial Subsection IWE and Subsection IWL ISI plans, but instead, the Owners will.be required to keep these on site for audit.
Presently, the ASME Code requires that the 10-year ISI plan be submitted to the regulatory-authority for review. The NRC staff is presently reviewing this requirement. The staff believes that this submittal requirement should also be changed, but this change should be made through. the ASME Code committees, and not in this proposed rulemaking.
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f Imoacts Industry Imoacts:
The proposed rule would require each licersee to:
1.
Develop and implement an initial Inservice Inspection Plan (ISI);
2.
Develop and implement 10-year updates to ISI plan.
Following is a description of and cost estimate for the requirements identified above.
Develop and Implement an Initial Inservice Insnection Plan The tasks associated with the initial ISI plan development are identified below for a representative facility.
For most of the tasks, the cost is a direct function of the manpower involved.
Engineering, drafting, and consulting labor has been costed at $66 per hour and clerk labor at $28 per hour. These hourly rates are based on 1984 base wage rates adjusted by a factor of-1.8 for-fringe benefits and plant i
management, with escalation to 1991 dollars' based on the GNP Implicit Price Deflator'. Employees are assumed to work 167 hours0.00193 days <br />0.0464 hours <br />2.761243e-4 weeks <br />6.35435e-5 months <br /> per month (i.e., based on a 2004 hour0.0232 days <br />0.557 hours <br />0.00331 weeks <br />7.62522e-4 months <br /> per year).
The following cost estimates assume no containment inspections are presently taking place, and are therefore conservative.
Estimates are based on information received from utility inservice inspection specialists.
ISI Plan Development Drawing Update - Includes preparation of ISI drawings _ for the containment structure (9 person-months for draftsman) 566 per hr x 167 hrs per month x 9 months.......$99X Computer Database Preparation - Entering components into the 151 computer database program for tracking purposes-(1 Person-Month for Clerk) 528 per hr x 167 hours0.00193 days <br />0.0464 hours <br />2.761243e-4 weeks <br />6.35435e-5 months <br />..........................
55K Video Mapping Containment - For job planning 1
.................................................. 567K NRC analysis of industry labor rates is available in NUREG/CR-4627, Generic Cost Estimates; abstract 6.3, " Industry Labor Rates," June 1986.
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Inspection Plan Preparation - Preparation of the ISI plan to include initial containment inspection, review of as-built data, update the ISI program,-prepare code exemption / relief requests, review construction data,:etc. (6 person-months for engineer).
$66 per hour x.167 hrs per month x.6 months.....$66K Additional Engineer and Consultant Work (4 person-months) 353 per hr x 167 hrs per month x 4 months................................. 544K Clerical Assistance - To assist engineer for review of construction records, typing, archival ~ search, preparation of ISI program updates, etc. (3 person-months for clerk)-
$28 per hr x 167 hrs per month x 3 months.......$15K TOTAL ONE-TIME COST..... 5296K Implementation of Initial Inservice Insoection The proposed rule would result in detailed examinations of containments and post-tensioning systems in accordance with the-proposed ISI plan.
Utility inservice inspection specialists estimate 8000 hours0.0926 days <br />2.222 hours <br />0.0132 weeks <br />0.00304 months <br /> of technicians' (e.g., NDE examiner, Level II technician) time for-the required Subsection IWE inservice inspection (ISI)'of the containment each ten year ISI interval.
For the purposes of this analysis, the staff has assumed that 75% (6000 hours0.0694 days <br />1.667 hours <br />0.00992 weeks <br />0.00228 months <br />) of this inspection effort will occur at the end of each 10 year ISI interval, and the remaining-25% (2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />) will-be distributed every three years corresponding to the approximate 3-year cycle for Appendix J Type A inspections.
These assumptions result in a distribution of examination effort over a 10 year period of 666 hours0.00771 days <br />0.185 hours <br />0.0011 weeks <br />2.53413e-4 months <br /> each in years 3,-6, and 9 (2000-hours or 25% of the 8000 hours0.0926 days <br />2.222 hours <br />0.0132 weeks <br />0.00304 months <br />), and 6000 hours0.0694 days <br />1.667 hours <br />0.00992 weeks <br />0.00228 months <br />'(the remaining 75%) in year 10. Applying an hourly labor rate of 566 results in-a total cost per reactor.on a 1991 present value basis of about-
$105K to $192K for a 10% and 5% real discount rate, respectively.
RegulatodyGuide1.35, Revision 2,"InserviceInspectionof Ungrouted Tendons in Prestressed Concrete Containment Structures,"
was published in January 1976, and'many licensees voluntarily adopted its provisions.
Regulatory Guide 1.35, Revision 3, was issued on July 11, 1990. -Twenty-seven licensees have voluntarily adopted this regulatory guide, and some of these provisions have
~
been incorporated into their Technical Specifications.
For these-licensees, there would be incremental costs of $5K to $12K 6-5
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d associated with adoption of the Subsection IWL tendon surveillance
. program due to specified modifications in the rule.
Five licensees are committed to Regulatory Guide 1.35, Rev. 2, six licensees are committed to Revision 1.
The five remaining post-1 tensioned containments current status is: one will start using Revision 3 with their next scheduled surveillance; three would rather use an industry standard than a regulatory guide, and are therefore waiting for the rule to become effective, and the last is a grouted tendon containment and exempt from the tendon examinations.
If the assumption were made that a licensee currently was not performing any examination of the post-tensioning systen, the costs to perform this examination per Revision 3 of R.G. 1.35 is approximately $475,000.
(Revision 2 contained technical updates to Revision 1, and the costs of implementation of Rev. 2 compared.
to Rev. I were nearly equal.
Revision 3 changed the tendon detensioning and sampling requirements such that a considerable cost savings would be realized (See Tables 5 - 7 of Enclosure 2).
Thus, implementation of Subsection IWL would result in a cost savings to the industry).
Tendon detensioning requirements are changed in Revision 3 relative to Revision 2 because detensioning of only one tendon in a group is required as compared with the previous requirement that all tendons selected for inspection were to be detensioned.
When the effect of this change is combined with the change in tendon sampling requirements, a considerable cost savings is realized by industry.
Tables 5 through 8 compare this reduction in tendon sampling requirements and the associated cost savings.
Using'a typical Type III'. containment (nine inspections), the total cost saving to industry per plant resulting from implementation of the sampling and tendon detensioning requirements in Revision 3 is estimated to be $320,000 (assuming a discount rate of 5% per year).
Assuming a 10% discount rate, the cost savings would be approximately $278,000'.
See Table 5.
Estimated cost savings for industry for a Type I containment, using discount. rates of 5% and 10%, are estimated to be $356,000 and $293,000,.
' s respectively, over the life of the plant.
For a Type II containment, the estimated cost savings, using discount rates of 5% and 10%, are estimated to be
$874,000 and:5468,000, respectively.
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Y periodic Updates to inservice Inscection Plan Updates to the initial ISI plan would be for' subsequent 10-year ISI intervals. Updates would be required in accordance with provisions in 150.55a. Based on a remaining useful life of 30 years, two succeeding 10-year ISI intervals will occur following the initial inservice inspection for which an updates will be necessary..For the purposes of the analysis, implementation of the update is assumed to occur at the midpoint of these 10-year intervals which corresponds to impacts occurring 15 years and 25 years into the future.
Industry inservice inspection specialists have estimated an 1800-hour engineering effort (mainly ISI engineers) per reactor to perform plan updates during a given 10-year ISI interval.
Based on an engineer labor rate of $66 per hour, this results in a cost of $119K cost (1991 dollars) per reactor per 10-year inservice.
interval.
This cost is assumed to occur in the year 2008 and again in the year 2018; corresponding to 15 years and 25 years into the future, respectively. Assuming a 10% real discount rate, the 1991 present value of two ISI plan updates is estimated at
$39K per. reactor.
If a 5% real discount rate is assumed, the 1991 present value cost.is about $93K.
Need to Update Reference to Section XI Edition and Addenda.
P In the past,Section XI has been revised twice a year. These revisions have been published in two addenda each year (i.e., Summer Addenda and Winter -
Addenda).. Starting in 1986, only one. addenda, the 19XX Addenda (e..g., 1988 Addenda) has been published. The revisions are the result of consensus-participants meeting 4-5 times a year for the purpose of' improving the existing rules. The revisions take into account the many lessons learned ~in a-specific area since the development of a particular Code rule. The revisions generally fall into three categories: (i) technical revisions that incorporate new rules in technical areas not previously addressed by the Code; (ii) technical revisions to existing rules; and (iii)~ editorial revisions. When a technical revision is made, it may make the existing set of rules more or less restrictive, or may simply clarify the existing rule without changing its intent.
There are numerous revisions in each addenda.
In general, technical revisions are made to. improve the ASME Code by providing more detailed rules where experience indicates greater guidance is necessary, or relaxing the rules where experience shows equivalent operational safety can be maintained with a reduced burden on the licensee.
Relative to implementation of Section XI, Division 1, 1 50.55a h
specifies that:
(a)
Inservice examinations of components, inservice tests'of pumps and valves, and system pressure tests conducted during the initial 120-month inspection interval shall comply with the requirements-j in the latest edition.and addenda of the ASME Code incorporated by 6-7 l
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t reference on the date 12 months prior to the date of issuance of the operating license, subject to any limitations noted; (i 50.55a(g)(4)(i)).
(b)
Similar to (a.), above, for successive 120-month inspection latest edition and addenda of the ASME Code reference 12 months prior to the start of the 120-month inspection
,4 interval, subject to any limitations noted; (150.55a(g)(4)(ii))
(c)
If a licensee determines that conformance with certain Co 1
notify the Commission and submit information to su determination; (i 50.55a(g)(5)(iii)).
The Commission will.
evaluate licensee determinations that Code requirements are impractical and may grant such relief and may impose alternative requirements giving due consideration to the burden on the licensee; (150.55a(g)(6)(i)).
i The existing requirements in 150.55a specified in items (a) and (b) programs in conformance with updated versions of S The proposed rule updates the reference to the editions and addenda tha e.
addenda to be implemented by licensees consisten re identified in Items (a) and (b), above.
submittal of relief requests by licensees.The existing requirement sp!
It ensures that in those cases where the generic requirements of Section XI are impractical, or are o particular requirement, provided the licensee demo i
that omission of the Section XI requirement believed to be impractical not have an adverse affect on public health and safety.
i will-incorporating new rules to cover areas not previously revising the rules consistent with experience to reduce the number of areas where the Code has been found to be impractical, inadequate, or insufficie clear.
The revisions were developed through the consensus process and, a
y therefore, have been thoroughly reviewed by utilities, designer-constructo and the NRC staff.
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As noted above, l 50.55a presently requires that-licensees update their inservice inspection programs every 10 years to the Section XI rules that w endorsed by the NRC 12 months prior to the start of the next 120-month in-spection interval.
endorsement of the later addenda because it will be these used in subsequent inservice inspection programs.
not be continued from one 120-month inspection interval to anotherObsolete requir t
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4 Implementation of periodic inservice Insoections The following analysis is based on the following assumptions.
The required _ Subsection IWE inservice inspection (ISI) of the containment each ten year ISI interval will require 8000 hours0.0926 days <br />2.222 hours <br />0.0132 weeks <br />0.00304 months <br /> of-i technicians' time, and that 75% of this inspection effort will l
occur at the end of each 10 year ISI interval, and the remaining 25% will be distributed every three years corresponding to the 3-year cycle for Appendix J inspections.
These assumptions result i
in a distribution of examination effort over a 20 year period of 666 hours0.00771 days <br />0.185 hours <br />0.0011 weeks <br />2.53413e-4 months <br /> each in years 12, 15, 18, 21, 24, and 27, and 6000 hours0.0694 days <br />1.667 hours <br />0.00992 weeks <br />0.00228 months <br />
)
each in years 20 and 30. Applying an hourly labor rate of $66 i
results in a total cost per reactor on a 1991 present value basis of about 5210K to 5383K for a 10% and 5% real discount rate, l
respectively.
The costs estimates given for the initial inservice inspection will be the same as the costs for the periodic inservice inspections.
Occupational Exoosure Implementation of this proposed rule is not expected to result'in significant occupational exposure, particularly when compared with other ISI examinations and tests.
For example, the containment liner examination at the Monticello Plant resulted in 20 millirems exposure compared with a total 935 millirems for all testing and surveillance activities conducted for plant life extension studies at this facility. Adopting 20 millirems exposure as representative of the dose per reactor year produces lifetime impacts of 0.6 person-rem and 75 person-rem for an individual reactor and all reactors, respectively.
Dollar quantification based on $1000 per person-rem results in lifetime impacts of 5600 per reactor and $75,000 for the reactor population.
These impacts are viewed as nil when compared to the dollar magnitudes discussed in the previous sections.
Industry Cost - Summary Based on the foregoing analysis, high and low estimates of lifetime costs are summarized below.
Results are presented on a per reactor basis. The low estimates assume a 10% real discount-rate, whereas the high estimate assumes a 5% real discount rate.
All costs are expressed in 1991 dollars and all future costs are present valued.
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J Summary - Lifetime Costs for a Facility Presently j
Using Regulatory Guide 1.35, Rev. 3 (1991 dollars)
J Hioh Estimate Low Estimate J
Cost Per Reactor 51107K 5731K Summary - Lifetime Costs for a Facility Presently Using Regulatory Guide 1.35, Rev. 2 (1991' dollars)
Hiqh Estimate Low Estimate Cost Per Reactor
$787K
$453K l
CONCLUSION:
As discussed in the preceding, the NRC has determined that there is a substantial increase in the overall protection of the public health and safety and that the direct and indirect costs of implementation for that facility are
-justified in view of this increased protection.
In addition, the proposed action is consistent with the Staff Requirement Memorandum (SRM), SECY-93-086
- Backfit Considerations (i.e., the substantial increase criterion are-flexible enough to allow for qualitative arguments, and flexible enough to allow for arguments that consistency with national and international standards, or the incorporation of widespread industry practices, contributes either directly or indirectly to a substantial increase in safety).
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