ML20055J367
| ML20055J367 | |
| Person / Time | |
|---|---|
| Issue date: | 06/13/1989 |
| From: | Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Jordan E Committee To Review Generic Requirements |
| Shared Package | |
| ML20055C236 | List: |
| References | |
| NUDOCS 9008020155 | |
| Download: ML20055J367 (67) | |
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o JUN 131989 l
l 1
HEMORANDUM FOR:
Edward L. Jordan Chairman CommitteetoRevIewGenericRequirements FROM:
Eric S. Beckjord, Director Office of Nuclear Regulatory Research
SUBJECT:
REQUEST FOR CRGR REVIEW OF PROPOSED RULE FOR 10 CFR 50.55A TO
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INCORPORATE BY REFERENCE SUBSECTION IWE OF SECTION XI, DIVISION 1, OF THE ASME BOILER AND PRESSURE VESSEL CODE Enclosed for review and approval by the CRGR is a proposed rule to amend 10 CFR 50.55a to incorporate by reference the 1986 Edition with addenda through the 1987 Addenda of Subsection IWE, " Requirements for Class MC and Me-i tallic Liners of Class CC Components of Light-Water Cooled Power Plants," of i
Section XI, of the American Society of Mechanical Engineers (ASME) Boiler and l
Pressure Vessel Code.
(In a separate but parallel action, a pro)osed rule is under development to amend 650.55a to incorporate by reference tie 1986 Adden.
da and the 1987 Addenda of Section Ill, Division 1, and Section XI Division 1,rulesforClass1, Class 2,andClass3 components;the1986EdItionisal-l l
ready incorporated by reference in the regulation.) Subsection IWE of the i
ASME Code provides the rules and requirements for inservice inspection, re-pair, and replacement of Class MC pressure retaining components and their in-tegral attachments, and of metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments in light-water cooled power plants.
This proposed amendment would provide rules to satisfy, in part, the periodic inspection and surveillance program required by Criter-ion 53 of the General Design Criteria and the general inspection required by the NRC regulation for primary reactor containment leakage testing.
The containment represents the final barrier to protect the public from the
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release of radioactive material in the event of an accident.
PRAs that deal with catastrophic failure of the containment assume the availability of the containment as a structurally adequate component well beyond the design accident pressure.
Yet, age-related degradation of. containments has occurred, and many liners of concrete containments were not designed with corrosion I
allowances.
Specific instances of corrosion are addressed in the regulatory analysis (Enclosure 3).
In view of the potential importance of the containment to the public health and safety, an inservice inspection program which would provide a basis for ensuring the continued operational integrity of these containments is necessary.
FW R
MEETING 167 PDC
2 JUN 131983 Also, we believe that because:
(1) containments are assigned the highest pri-ority in risk studies; (2) recent operating experience indicates a degradation of containments; and (3) there is a total
- absence of any specific inservice inspection requirement for these components, there is adequate justification for requiring implementation of the proposed rule as soon as practicable. The regulatory analysis (Enclosure 2) summarizes the need for containment inspec-tions, and provides an estimate of the associated costs.
Additionally, it should be pointed out that since this will be the first time the NRC has endorsed this Subsection, licensees have not developed an inservice inspection plan to comply with these requirements.
This will result in an estimated one-time cost to a facility of $241K.
Subsequently, to comply with the inspection requirements of Subsection IWE, the cost to a facility over a 10 year period will be roughly $478K.
As addressed in the Backfit Statement of the proposed rule, based on the documented evaluation required by 10 CFR 50.109(a)(4), the provisions of the backfit rule do not apply in this j
particular situation due to the need for maintaining adequate minimum safety, j
Also, because of the potential safety problems found and, because of the lack i
of a containment inservice inspection program, we are requiring a modification-to Subsection IWE.
We are requiring t1at Subsection IWE be implemented expedi-tiously (within 5 years of the final rule being published) in order to ensure continued high containment reliability. This is an acceleration of the ten year schedule prescribed by the ASME rules.
The proposed rule (Enclosure 1) has been approved by OGC, ARM (at that time),
GPA, and AE00.
NRR concurs in the technical need for the rule.
However, NRR has some concerns with the citation of adequate safety as the basis, in terms of the backfit rule.
RES believes that adequate safety is an appropriate ba.
sis for promulgating the rule, and the March 1989 memo from OGC is in conso-nance with the RES position.
The NRR concern is that the requiring of imple-mentation of the rule under the adequate safety provision could result in a plant being shut down unnecessarily.
NRR also believes that there is some basis for the use of the compliance argument as the legal basis for the rule.
This rule package is being sent to the CRGR at this time because both NRR and RES believe that a ventilation of the various options of justification in a i
CRGR meeting would be of benefit to all. We note that a general discussion on this topic is planned for a CRGR meeting (paper to be provided by the CRGR i
staff) and thus it seems timely to take this approach. is the proposed rule. Enclosure 2 contains the Summary of I
proposed Generic Requirements for CRGR review, while Enclosure 3 contains the Regulatory Analysis. contains the transmittal letters from the various NRC offices concerning the adoption of the proposed rule.
is a copy of Subsection !WE.
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JUN13tgpp For further information, contact Wallace Norris, Task Leader, Structural &
i SeismicEngineeringBranch(492-3805).
AA Y
Eric S. Beckjord(LDirector Office of Nuclear Regulatory Research
Enclosures:
As stated cc:
See attached list CRGR(20)
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4 M I I 1989 I
DISTRIBUTION: w/ encl.
ESBeckjord,RES DfRoss, RES-l TSpeis, RES GAArlotto, RES RBosnak, RES JCostello,ES AMurphy R RES TMurley, NRR LShao, NRR JRichardson, NRR CYCheng, NRR KWichman, NRR MHum, NRR FGillesple,RR GJohnson N NRR flitton, NRR WSchwink, NRR RColmar, NRR EJordan, NRR DGrimsley,OD EBrown AE RM/DRR JPartlow, OSP WMcDonald, ARM STreby, OGC EJakel, OGC BMorris, RES HParcover, RMB SDuraiswany, ACRS Sfeld, RES r
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9 (7590-01)
HUCLEAR REGULATORY COMMIS$10N 10 CFR PART 50 Codes and Standards for Nuclear Power Plants; subsection IWE t
i AGENCY:
Nuclear Regulatory Commission, 1
ACTION:
Proposed rule.
SUMMARY
The Nuclear Regulatory Commission (NRC) proposes to ame'nd its reg-ulations to incorporate by reference the 1986 Edition with addenda through the 1987 Addenda of Subsection IWE, " Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants", of Section XI, Division 1, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
This subsection of the ASME Code provides the rules and requirements for inservice inspection, repair, and replacement of Class MC i
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.
Adoption of this amendment would provide rules to satisfy, in part, the periodic inspec-tion and surveillance program required by Criterion 53 of the General Design Criteria and the general inspection required by the NRC regulations for primary l
reactor containment leakage testing.
DATES:
Comment period expires Comments received after this date will be considered if it is practical to do l
so, but assurance of consideration cannot be given except as to comments received on or before this date.
'A date will be inserted allowing 60 days for public comment.
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[759001]
ADDRESSES:
Written comments or suggestions may be submitted to the Secretary t
of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention:
Docketing and Service Branch.
Copies of comments received may be examined in the Commission's Public Document Room at 2120 L Street NW, Washington, DC.
s FOR FURTHER INFORMATION CONTACT:
Mr. W. E. Norris, Division of Engineering, F
Office of Nucient Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington,DC20555, telephone (301)492-3805.
SUPPLEMENTARY INFORMATION In 1977, the NRC requested that the ASME initiate work to establish rt.les for containment inspection.
This request was made as a result of previously identified as-built containment deficiencies.
A working group under the ASME Secion XI Subcommittee on Inservice Inspection was established in response to that request.
In 1979, the working group was elevated to the Subgroup on Con-tainment under the Section XI Subcommittee on Inservice Inspection.
The char.
ter for the Subgroup gave it the responsibility for developing and maintaining the rules in Subsection IWE for Class MC containments.
The Subgroup developed rules which were published as Subsection IWE in 1981.
Since that time, Sub-section IWE has been expanded to provide a comprehensive set of rules for the inservice inspection of Class MC components and the metal liners in concrete containments, in view ci the age-related degradation occurring in contain-i ments, the NRC now proposes to incorporate by reference into 10 CFR Part 50 the 1986 Edition with addenda through the 1987 Addenda of Subsection IWE, "Re-quirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants", of the ASME Code.
For the reasons discussed in
[7590-01) the regulatory analysis, this action is necessary to ensure that the facili-ties continue to provide adequate protection to the health and safety of the public.
The NRC has certain principal design criteria which establish the nec.
essary design, fabrication, construction, testing, and performance require-
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ments for structures, systems, and components important to safety.
These General Design Criteria establish minimum requirements for the principal de-sign criteria for water-cooled nuclear power plants. Implementation of Subsec-tion IWE would satisfy, in part, requirements specified in certain General De-sign Criteria requirements specified in certain Technical Specifications, and in Appendix J of 10 CFR Part 50.
Criterion 1 of the General Design Criteria (Appendix A of 10 CFR Part 50) l requires, among other things, that structures and components important to i
safety be tested to quality standards commensurate with the importance of the safety functions to be performed.
it further states that, where generally rec-ognized codes and standards are used, they shall be supplemented or modified as necessary to ensure a quality product in keeping with the required safety i
- function, implementation of Section XI, Subsection IWE, would ensure that spe-cific containment components would be inspected to an acceptable standard (i.e.,theASMECode).
Criterion 16 of the General Design Criteria requires that reactor con-tainment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the en-q l
vironment and to ensure that the containment design conditions important to l
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[759001) safety are not exceeded for as long as postulated accident conditions require.
The objective of the containment inspettion is to ensure the pressure-retain-ing integrity of the containment throughout the plant lifetime.
Criterion $3 of the General Design Criteria (Appendix A of 10 CFR Part 50) requires that the reactor containment be designed to permit: (1) appropriate periodicinspectionofallimportantareas,suchaspenetrations;(2)enappro-priate surveillance program; and (3) periodic testing, at containment design pressure, of the leak-tightness of penetrations which have resilient seals and expansion bellows.
Subsection IWE sets forth procedures and details for satis-fying the above three requirements.
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 inspec-tion of the accessible interior and exterior surfaces of the containment I
structures and components be performed prior to any Type A test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak-tightness.
Subsection lWE provides details for this general inspection (e.g., what parts of the containment structure must be accessible for inspection and personnel qualification requirements for exami-ners that are not specified in Appendix J.
In Section 4 " Surveillance Requirements" of the Standard Technical Speci-I Type A test means tests intended to measure the primary reactor containment overall integrated leakage rate: (1) after the containment has been completed andisready_foroperation,and(2)atperiodicintervalsthereafter.
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(7590-01) fications, the subsection titled " Containment Surfaces" requires that the structural integrity of the exposed interior and exterior surfaces of the con-tainment, including the liner plate, be determined by a visual inspection dur-ing the shutdown for each Type A containment leakage rate test. The purpose of this inspection is to verify that there are no apparent changes in appearance or other abnormal degradation of these surfaces.
This is an important ccnsid-eration as many liners have very small design allowances for corrosion.
Sub-section IWE gives acceptance criteria for this visual inspection.
Section 6. " containment systems", of the Standard Technical Specifica.
tions requires a periodic inservice surveillance program to ensure the func-tional capability of the containment and associated structures, systems, and components.
Subsection IWE sets forth requirements for this periodic surveil-lance program.
1 This proposed amendment would incorporate by reference into 10 CFR Part 50, Subsection IWE of Section XI, Division 1 of the ASME Code, rules for con-tainnent inservice inspection and would thereby provide systematic examination rules for containment structures.
These rules would serve to satisfy the rel-event requirements of the General Design Criteria, the Technical Specifica-tions, and Appendix J.
The NRC staff has reviewed the 1986 Edition with ad-denda through the 1987 Addenda of Subsection IWE of the ASME Code and found these rules to be acceptable for satisfying the relevant portions of the above requirements with the following modification and limitation.
A modification has been incorporated into the regulation that would re-quire the examinations imposed by Subsection IWE for the first inspection in-5-
[7590-01) terval to be completed within 5 years of the effective date of this rule.
Sub-section IWE as written permits the deferral of most of the required inspec-tions until the end of the inspection interval.
A period of 10 years, there-fore, could pass before the first inspections would take place, in order to establish a baseline for a facility and to identify any existing problems, the modification to the regulation would require all examinations required by Sub.
section IWE to be completed during the first five years of the first 10 year inspection interval. The results of these first inspections would be studied before any decision on the timing of the inspections in the second through fourth inspection intervals would be made.
A limitation is incorporated into the proposed amendment that specifies when using Subsection 1WE, editions and addenda of the ASME Code no earlier than the 1986 Edition with addenda through the 1987 Addenda shall be used.
This would ensure that the latest set of rules would be used.
Subsection IWE,'
which has been expanded since 1981 to provide a comprehensive set of rules for the inservice inspection of Class MC components and the metal liners in con-crete containments.
The modification and the limitation to Subsection IWE requirements are contained in a new paragraph (b)(2)(vi).
Another new paragraph ((g)(7)),
specifies the applicability of ASME Code Class MC and Class CC classifications to containment components to which this rule will apply.
The inservice inspec-l tion requirements for Class MC components and the applicable Class CC cm -
I nents were added to paragraph (g)(4).
I Endorsement of the Subsection IWE rules by the NRC would provide a met' 6
(7590-01)
P of improving containment examination practices by incorporating rules into the regulatory process that are acceptable to the NRC and have received industry participation in their development.
Environmental Impact:
Categorical Exclusion The NRC has determined that this proposed rule is the type of action de-scribedincategoricalexclusion10CFR51.22(c)(3). Therefore, neither an en.
l vironmental impact statement nor an environmental assessment has been prepared for this proposed rule.
Paperwork Reduction Act Statement t
This proposed rule would amend information collection requirements that are subject to the Paperwork Reduction Act of 1980 ('44 U.S.C. 3501 et seq.).
1 i
This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.
Public reporting burden for this collection of information is estimated to average 3,300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> per plant to develop a containment inservice inspection program (this is a one-time burden over a five year period), and 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> per year per plant including the time for reviewing instructions, searching exist-ing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
Send comments regarding this bur-den estimate or any other aspect of this collection of information. including suggestions for reducing this burden, to Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention:
Docketing and Service Branch; 7
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[7590-01) and to the Office of Information and Regulatory Affairs, Office of Management and Budget. Washington, DC 20503.
a.
Regulatory Analysis i
The Commission has prepared a draf t regulatory analysis on this proposed regulation.
The analysis examines the costs and benefits of the alternatives considered by the Commission.
The draf t analysis is available for inspection s
in the NRC Public Document Room 2120 L Street NW, Washington. 0C.
Single copies of the arialysis may be obtained from Mr. W.' E. Norris Division of Engi-neering Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555, telephone (301)492-3805.
The Commission requests public comment on the draf t regulatory analysis.
Comments on the draf t analysis may be submitted to the NRC as indicated under j
l the ADDRESSES heading.
I l
Regulatory flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
l 605(b), the Commission hereby certifies that this rule will not, if promul-l gated, have a significant economic impact on a substantial number of small 1
entities.
This proposed rule 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 l
Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121.
Since these B
[7590-01) companies are dominant in their service areas, this proposed rule does not fall within the purview of the Act.
Backfit Statement The ASME Code defines minimum requirements 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 f
pressure retaining components and their integral attachments in light-water r
cooled power plants. The NRC staff has determined that Subsection IWE provides acceptable minimum requirements for the containment inspections and, therefore, represents responsible application of engineering judgment to ensure adequate i
protection of the public health and safety.
The Commission has concluded, on the basis of the documented evaluation required by 10 CFR 50.109(a)(4), that the provisions of paragraphs (a)(2) and (a)(3) of th$ backfit rule do not ap-ply in this particular situation, because the action is necessary to ensure that facilities continue to provide adequate protection to the health and safety of the public.
The evaluation and justification required by paragraph (a)(4) for imposing the Subsection IWE requirements as adequate protection of the public health and safety is discussed in Appendix B of the regulatory anal-ysis.
List of Subjects in 10 CFR PART 50 Antitrust, Classified information, Fire protection, incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors,
- Penalty, Radiation protection, Reactor siting
- criteria, Reporting and recordkeeping requirements.
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(7590-011 i
Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reor-genization Act of 1974, as amended, and 5 U.S.C. 533, the NRC is proposing to adopt the following amendments to 10 CFR Part 50, PART 50 - DOMESTIC LICENSING OF PRODUCTION At;D UTIL12AT10N FACILITIES T
1.
The authority citation for Part 50 continues to read as follows:
AUTHORITY:
Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat.
936, 937, 938, 948, h 3, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as 1
amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282);
secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 j
U.S.C.5841,5842,5846).
Section 50.7 also issued under Pub. L.95-601 sec. 10, 92 Stat. 2951 (42 U.S.C. 5851).
Section 50.10 also issued under secs. ' 101, 185, 68 Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec.102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332).
Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235).
Sections 50.33a, 50.55a and Appendix Q also issued under sec.102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332).
Sections 50.34 and 50.54 also issued under sec. 204, BB 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 4
Stat. 954, as amended (42 U.S.C. 2234).
Section 50.103 also issued under sec. 108, 68 :, tat. 939, as amended (42 U.S.C. 2138).
Appendix F also issued under sec 187, 68 Stat. 955 (42 U.S.C. 2237).
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[7590-01]
L For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273),
il 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and 50.80(a) are issued under sec.161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); il 50.10(b) and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C.
2201(i)); and 66 50.9, 50.55(e), 50.59(b), 50.70, 50.71, 50.72; 50.73 and 50.78 are issued under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).
2.
Section 50.55a is amended by adding paragraphs (b)(2)(vi) and -(g)(7) and revising paragraph (g)(4) to read as follows:
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650.55a Codes and standards.
(b)(2)(vi)
" Class MC examinations and metallic liner examinations of Class CC (Table IWE-2500-1)".
All light-water cooled nuclear power plant li-l censees shall complete implementation of the examinations in Table IWE-2500-1 I
required during the first inspection interval of Subsection IWE by When using Subsection IWE, editions and addenda no earlier than the 1986.Edi-tion with addenda through the 1987 Addenda shall be used.
5 (9)
I A date will be inserted that is 5 years later than the effective date of the rule.
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[7590-01)
(4)
Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as
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ASME Code Class 1, Class 2, and Class 3 shall meet the requirements, except de-sign and access provisions and preservice examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Ad-denda that become effective subsequent to editions specified in paragraphs (g)(2) and (g)(3) of this section and are incorporated by reference in para-graph (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 inte-gral attachments, and metallic shell and penetration liners which are classi-fied as Class CC pressure retaining compments and their integral attachments shall meet the requirements, except design and access provisions and preserv-ice examination requirements, set forth in Section XI of the ASME Boiler and Pressure Vessel Code and Addenda that are incorporated by reference in para-graph (b), subject to the modification and limitation listed in paragraph (b)(2)(vi) of this section, to the extent practical within the limitations of design, geometry and materials of construction of the components, i
(7)
For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to 1: (1) metal contain-ment pressure retaining components and their integral attachments shall meet 1
the inservice inspection requirements applicable to components which are clas-l l
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1 A date will be inserted that is 30 days after the publication date.
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(F F
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I Sumary of Proposed Generic Requirements for CRGR Review Date:
May 19,
'1989 L
l RES Task Number: - MS 801-1 RES Task Leader: Wallace E. Norris Telephone:
492-3805 1
i Title of Proposed Action:
Codes and Standards: Subsection 1WE, " Requirements for Class MC Components of Light-Water-Cooled Power Plants", of Section XI (Division 1),~of the-American Society of Mechanical Engineers Boiler and~ Pressure Vessel Code (ASME Code).
Type of Action:
Proposed rulemaking.
Cetegory:
Category 2 Statement of the Problem:
- Age-related degradation of containments has occurred.
Many liners of concrete containments were not designed with corrosion allowances.
Erosion of the metal drywell shell at one plant' was found to~ be occurrt g at the rate ot 20 o
mils / year.
The following information is found in' NRv Information Notice No. 88-82, " Torus Shells with Corrosion and-Degraded Coatings in BWR - Contain-ments".
During recent inspections at a particular plant, NRC inservice in-spectors found that the inside surface of the torus shell had corroded.' Thick-l-
ness measurements of the torus shell revealed-several areas in which the thick-ness was at or below the minimum specitied wall thickness.
In tact,. the l
present rate of loss of wall thickness is almost double the loss for which the torus shell was originally designed. A survey of other BWR's in one region
.i l
i revealed three Mark I tori had experienced degradation of the coating and that l
cleaning and recoating will' be required.-
As stated -in the NRC' information Notice, although the torus shell thinning due to corrosion and the torus coat-ing degradation have no immediate effect on plant operation, the NRC staff-considers these deficiencies to-be significant because the measured corrosion rates of torus shells are greater than the corrosion rates assumed.as -part ot the original design..The torus-shell degradation, if 1t continues, may jeop-l ardize containment integrity.
Additional and. potentially. more significant degradation mechanisms can be anticipated as nuclear power. plants age.
.In-view of the importance of the containment to the health and safety of the general public, an inservice inspection program which would provide a basis for assuring the continued operational integrity of these -containments is necessary to adequately protect the public health and-safety.
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In normal operation, three barriers (the fuel rod cladding, the reactor coolant system pressure boundary, and the containment pressure boundary) pro-tect the public from the release of radioactive material generated in the fuel.
In most core meltdown accidents, the first two barriers are progres-4 sively breached and the containment represents the final barrier to mitigating the consequences of a severe accident. The ability of a containment structure to maintain integrity in a severe accident is determined by the magnitude of the loads imposed on the containment and the response of the containment to l
those loads.
Although there is no universally accepted definition of contain-l ment f ailure, typically, for risk purposes, a containment is considered to have failed when the leakage from the containment is great enough that the load will be relieved, or, when the leak rate of radionuclides to the environ-ment is significant. Thus, failure could occur as the result of a gross struc-tural failure of the containment or as the result of a high rate of leakage through a penetration.
Probabilistic risk assessments (PRAs) are largely a snapshot of plant risk in time.
It would be easy to conclude that the identified dominant acci-dent sequences provide a complete and unchanging characterization of plant risk, but this is not the case.
Risk can be decreased in time by modifying systems and procedures.
More importantly, degraded systems, poorly implemen-l few) procedures, inadequate maintenance, or inettective management (to name adrastic ted could magnitude of risk at a plant.
Theref ore, it is very important to provide as-surance that the estimate of risk provided by a PRA is actually realized or even %ttered in practice.
This function would be furnished by an eff ective nal reliability (saf ety) program."
NUREG-1150, " Reactor Risk Ref er-l cr ment", used the underlying assumption in the study of catastrophic e
e
.t tailures was that failure of the containment would not occur un-l J
t l
or' ml accident conditions.
Uncertainty curves show this to be the case up 6 at least 1.5 times the design accident pressure.
HUREG-1150 also looked at containment bypass modes.
To quote the NUREG, "In many PRA analyses, se-quences in which the containment is bypassed by interfacing-system LOCAs are among the most important contributors to overall risk.
Containment bypass se-quences, including those associated with preexisting openings, are among the most plant specific of all contributors since they depend on detailed config-urations of piping, valving, penetrations, control systems, and the like."
ihe importance of the containment in a nuclear power plant to the protection of the public health and safety is evident.
The above study highlights the importance of a comprehensive inservice inspection program in ensuring the continued structural integrity of nuclear power plant systems and components.
Section X1 of the ASME Code provides a comprehensive inservice inspection for such systems and components.
in order to provide a consistent set of rules, with appropriate examination, and repair and replacement details for contain-ment structures, the industry has participated in developing Subsection IWE of Section XI, Division 1, of the American Society of Mechanical Engineers Boiler andPressureVesselCode(ASilECode).
Subsection IWE requires that the critical areas of the containment such as pressure-retaining welds, seals and gaskets, and pressure-retaining bolting maintain their pressure retaining integrity thorughout the plant lifetime.
This Subsection d)es not include inspection of all areas that contribute to overall risk.
However, many of the most important areas (e.g., penetrations, 2
1 containment welds, liner coatings) that need to be inspected are covered by Subsection IWE.
Also, as pointed out in NUREG-1150, an ettective inservice inspection plan is needed to ensure that the " snapshot in time" is maintained.
For the probabilities to have any meaning, the underlying assumption that con-tainment integrity exists prior to an accident must be true.
Subsection IWE inspections will help ensure that the containment will withstand severe acci-dent pressures up to certain limits.
The General Design Criteria (Appendix A of Part 50) of the NRC Regulations require that structures, systems, and components of light-water-reactors be designed, fabricated, erected and tested to quality standards commensurate with i
j the importance of the safety function performed.
Quality Assurance Criteria l
(Appendix B of Part 50) establishes quality assurance requirements for the design, construction and operation of those structures, systems - and compo-nents.. The pertinent requirements of Appendix B apply to all activities af-fecting the safety-related functions of those structures, systems and compo-nents; these activities include inspecting, testing, operating, maintaining, repairing and modifying.
Without a set of specific rules to implenent these quality standards, it would be necessary for each applicant / licensee to de-velop its own program for submittal to the NRC.
Each program would have to reviewed by the staff on a case-by-case basis.
This would increase signifi.
cently the licensing review time and would make inspections by the staff more difficult because of the nonstandard nature of each program.
To provide a consistent set of rules, which the industry has participated in developing, 6 50.55a presently mandates use of Section XI of the ASME Code for inservice inspection of Class 1, Class 2, and Class 3 components and their supports.
Section XI is implemented by licensees of all -light-water cooled reactors.
The NRC tirst endorsed the ASME Code by reference in 10 CFR 50.55a in 1971.
The ASME publishes a new edition of the Code every three years and new addenda yearly.
It has been a continuing policy of the Commission to up-date this section of the regulations to keep the reterences current.
In those cases where an item in the ASME Code is inconsistent with NRC criteria, NRC regulations may take an exception to endorsing that portion of the Code, or supplementary criteria may be incorporated into NRC regulations so that the-item is consistent with staff requirements.
The ASME Code is developed by the consensus process, which ensures that the various interested sectors (e.g., utilities, nuclear steam system suppliers, insurers, regulators) are represented on the standards writing committees and that their viewpoints are considered in the standards writing process.
En-dorsement of the ASME Code by the NRC provides a method of incorporating rules l
into the regulatory process that are acceptable to the NRC and have received industry participation in their development.
The purpose of this proposed amendment is to incorporate. by reference Subsection IWE into the NRC Regula-tions.
Objectives:
The proposed rule would incorporate by reference into i 50.55a of the NRC's regulations Subsection IWE of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code.
Subsection IWE was developed with the ob,iective of en-i turing that the critical areas of the containment (e.g., pressure-retaining I
weIds, seals and gaskets, and pressure-retaining bolting) maintain their pres-3
t sure retaining integrity throughout the plant lifetime.
l Since Appendix J 1eakage tests provide an adequate basis for ensuring con-tinued containment pressure-retaining integrity, Subsection IWE references 4
Appendix J.
Incorporating by reference Subsection IWE, of Section XI, Division 1, of the ASME Code will establish the NRC staff position on the examination of steel-containment structures and liners of concrete containments on a generic basis for applicants and licensees, thereby minimizing the need for case-by-case l
evaluations and reducing the time and effort required for submittal prepa-l rations and licensee reviews.
l In June 1988 Pacific Northwest Laboratory (PNL) submitted to the NRC a report titled. "Prioritization of ' Tirgalex-Recommended Comp (onents for Further Aging l
Research for Implementation by ALEXCC."
(Tirgalex Technical Integration Re-view Group for Aging and Life Extension) is a group that was established in 1986 at the direction of the EDO to develop a plan to integrate the NRC's-ag-ing and life extension activities).
The Tirgalex plan identified the-safety-t related structures and components that should be prioritized for. subsequent evaluation in the the NRC Nuclear Plant Aging Research (NPAR) Program.
Con-tainment structures were given the highest risk importance in the study.
Also, the report showed that currently there is no BWR containment research planned even though problems are presently occurring.
Further, the panel opinion was that the containment is basically uninspected following construc-tion.
The panel's recommendation was for improved surveillance and test meth-ods to detect aging for the risk-significant failure modes in components and structures.
Idaho National Engineering Laboratory (EG&G) published NUREG/CR-4731, "Resid-ual Life Assessment of Major Light-Water Reactor Components-Overview", in June 1987.
In the section titled "In-service. Inspection" EG&G's assessment was that "the establishment of inspection procedures to cover critical areas.where i
adverse environmental conditions such as high temperature, humidity, and/or radiation, and locations subjected to an acidic environment, will be a neces-sary measure to determine the extent of degradation."
Also, the report stated l
that "it should be realized that great safety and economic benefits can be-de-rived if an expanded ISI is implemented to cover -identified degradation sites that may not be frequently inspected."
NUREG-1144, " Nuclear Plant Aging Research (HPAR) Program Plan", states that "it is not the intent of the NPAR Program to do in-depth engineering evalua-tions of: aging and defect characterization, and methods for inspection, sur-veillance, and monitoring of all significant plant elements."
The NPAR pro-gram will support NRR in establishing inspection procedures that are relevant to aging; NRR includes these procedures in the inspection Enforcement-Manual issued to guide the activities of the regions.
For example, some inspection procedures establish guidance for ascertaining that inservice inspection and testing activities are programmed, planned, conducted, recorded, and reported in accordance with Section XI of the ASME Code, in summary, HUREG/CR-4731 stated that by implementing a well planned, meaningful, and comprehensive in-i spection program followed by a firm program of repairs, the service life of containment structures can be extended, probably to at least double the ini-tial licensed term.
4
the proposed rule is intended to ensure the structural reliability of the con-tainnient, and to be consistent with the recommendations summarized above.
Description of the Activity Required by Licensee 4,
l The licensee of each nuclear power plant would incur the: to11owing requiv e-ments as a result of the proposed rule:
1.
Developmentofan-InserviceInspection(ISI) Plan Since this would be a new subsection 'ot Section XI of the. ASME Code, l
licensees have not yet developed an 151 plan to comply with the re-quirements of Subsectio IWE.
This would be a one-time cost to a fa-cility.
i 2.
Periodic updates to ISI plan Because of the problems that have occt.cred at sone facilities and the uncertainty of-the condition of some containments, the NRC staff would require all tacilities to complete.the first-Subsection IWE in-terval in 5 years instead of the normal ten year interval. -It is an-ticipated that from the second interval through the lifetime of a fa-cility the periodic updates to the ISI program will occur at the nor-mal ten year frequency.
3.
Periodic inservice inspections in - conformance with the ISI plan.
The Appendix J inspections would be rquired once every 3-4 years, and the other inspections required by Subsection IWE would be required once every ten years (as mentioned in Item 2 above, NRC would require the first Subsection IWE interval to be completed within 5 years to i
establish a baseline).
Potential Change in the Risk to Public In looking at the risk-dominant accident sequences and containment failure modes, the NUREG-1150 study tound the causes and types of accidents vary considerably among the studied plants.
The results showed that a wide spec-trum of uncertainty exists for the plants studied and that no one uncertainty.
is dominant for the set of plants.
Rather, the sources of important uncer-tainty cover a wide spectrum of both accident frequency and phenomenological c
issues.
These uncertainties carry over in trying to predict off-site conse-quences.
In particular, the ranges of environmental release terms obtained for the containment failure are quite broad.
For the volatile groups of radi-onuclides, the ranges are typically one to two orders of magnitude and, for the more refractory groups of radionuclides, two to-three orders of magni-tude.
The specific source term issues that have the greatest impact on the uncertainties in source terms depend not only on the design of the plant but also on the specific plant damage state.
However, the timing and mode of containment tailure has a major effect on the estimated ranges of consequences.
Even though the uncertainties are signif-icant, the ditterences between early and late containment tailures overwhelm the uncertainty.
Early failures of the containment result in costs that are 5
I approximately 3 orders of magnitude greater than for late containment rupture (early failure is estimated to result in off-site cost in the billions).
With regard to latent cancer fatalities resulting from early failure compared to latent cancer fatalities resulting from late containment failure, the predic-4 ted difference is in the thousands in 30 years.
When compared with the cost to a utility of $241K for the development of the plan, plus approximately $50X over a ten year period for updating the plan, plus approximately $425K over a ten year period for the inservice inspections required, the costs to a utility for implementing Subsection IWE are considered to be reasonable relative to l
the cost impact of a severe accident.
l Potential Impact on Radiological Exposure of Employees implementation of this proposed rule is not expected to significantly in-crease the occupational exposure associated with ISI examinations at nuclear power plants.
For example, the containment liner examination at the Monti-cello Plant resulted in 20 millirems exposure compared with a total 935 milli-rems for all testing and surveillance activities conducted for plant life ex-tension studies at this faciltiy.
Adopting 20 millirems exposure as repre-sentative of the dose per reactor year resulting from Subsection IWE examina-tions produces lifetime impacts of 0.6 person-rem (30 years (assumi presently operating reactors have three more ISI intervals) X 20(10)~gg that.
rem /-
yr) and 75 person-rem (.6 X 124 plants) for an individual reactor and all re-actors, respectively.
Dollar quantification based on $1000 per person-rem results in lifetime im-pacts of $600 per reactor and $75,000 for the reactor population.
These cost i
impacts are viewed as nil when compared to the dollar magnitudes discussed in following sections.
Cost of implementation Following is a brief description and cost estimate for each of the three re-quirements identified previously under " Description of the Activity Required by the Licensee".
1.
Development of Inservice Inspection Plan Based on information received from utility inservice inspection spe-cialists, (the NRC estimates the cost for development of an ISI plan at $241K 1988 dollars) per reactor.
Based on a reactor population of 124, the total industry cost for this requirement is estimated at approximately $29.5M.
This is a one time industry cost which would be incurred in the near-term (i.e., within 5 years of the adoption of this rule).
For the purposes of this analysis, these costs are assumed to be expended in 1988.
The specific tasks associated with ISI plan development are presented below for a representative reactor.
For most of these tasks, the cost estimate is a function of the manpower input requirement.
Engi-neering, draf tsman, and consultant labor has been costed at $53 per hour and clerk labor at $23 per hour.
These estimates are based on 1984 base wage ratet adjusted by a factor of 1.8 for fringe benefits and plant management, and then escalated to 1988 dollars using the p
6
a GNP ImplicitL Price Deflator.I
. 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 based on a 2000 hour-per year level of effort.-
?
Tasks associated with 151' Plan. Development (Costs are ~ to nearest -
$1000)
Drawing Update - Includes Prepara' tion ot'ISI Drawings for the Con-tainment $tructure, Numbering 151 Components (Welds, Penetrations, r
Supports.etc.)andPerformingFieldAs-Builts.AsRequired(9Per-son-Months for Draftsman) i 553 per hr x 167 hrs per month x 9 months............
$80K
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Computer Database Preparation - Entering Components into the -1S1 l
Computer. Dctabase Program for. Tracking Purposes (1 Person-Month for Clerk)
$23 per hr x 167 hours0.00193 days <br />0.0464 hours <br />2.761243e-4 weeks <br />6.35435e-5 months <br />................................
54K L
Video Mapping Containment - For Job Planning' Purposes $54K f
Inspection Plan Preparation - Preparation of the Inspection Plan, Review of As-Built Data, Update:the ISI Program, Prepare Code Ex :
emption/Reliet Requests, Review Construction Data. Etc. (6 Person-Months for Engineer) 4
$53 per hour x 167 hrs per month x 6 months..........
$53K Clerical Assistance - To Assist Engineer for Review of Construc-tion Records, Typing, Archiva1. Search, Preparation of.ISI Program l
Updates,Etc.(3 Person-MonthsforClerk).
523 per hr x 167 hrs per month x 3 months........... $15K-Additional Engineer and Consultant Work (4 Person-Months)'
553 per hr x 167 hrs per month x 4 months...........
535K TOTAL ONE-TIME COST......
5241K i
Summary Cost Impact (1988 dollars)
Cost Per Reactor 5241K Industry-wide Cost 529.8M-2.
Periodic Updates to Inservice Inspection Plan j
Following.the initial development of the ISI plan, updates to the. plan-would.be required for subsequent 10-year ISI intervals.
Such updates-are necessary to incorporate improvements in the methods and knowledge-i INRC analysis of industry labor rates is available in NUREG/CR-4627, Generic Cost Estimates; abstract 6.3, " Industry Labor Rates", June 1986.
j 7
l base for detection and correction of containment defects and are a re-quirement of 10 CFR 50.55a.
Based on a remaining useful life of 30 years, two succeeding 10-year 151 intervals will occur for which updates will be necessary.
For the purposes of the analysis, the updates are as-4 sumed to occur at the midpoint (since some inspections will occur every three to four years) of these 10-year intervals which corresponds to im-i L
pacts occurring 15 years and 25 years into the future.
Industry inservica inspection specialists have estimated an 1800-hour engineering effort per reactor to perform plan updates during a given 10-year ISI interval.
Adopting this estimate and an engineer labor rate of
$53 per hour results in a $95K cost (1988 dollars) per reactor per 10-year inservice interval.
This cost.1s assumed to occur in the year 2003 and again in the year 2018; corresponding to 15 years and 25 years into the future, respectively.
Assuming a 10f real discount rato, the 1988 present value of two ISI plan updates is estimated at $31K per reactor.
If a 5% real discount rate is assumed, the 1988 present value cost is about $74K.
Based on a reactor population of 124, the 1983 present value. industry wide cost assuming 10% and 5% discount rates is $3.8M and $0.2M respect-ively.
Summary Cost Impact (1988 dollars)
Cost Per Reactor 531K to 574K l
Industry-wide Cost 53.8M to 59.2M 3.
Periodic Inservice Inspections Adoption of the proposed rule would result in detailed examinations of containments in accordance with the proposed Subsection IWE ISI plan.
Utility inservice inspection spacialists estimate 8000 hours0.0926 days <br />2.222 hours <br />0.0132 weeks <br />0.00304 months <br /> of techni-cians time for each ten year 151 interval.
For the purpose of this anal-ysis, the staff assumes 75% of this inspection effort will occur at the end of each 10 year ISI interval, and the remaining 25% t!ill be distribut-ed every three years corresponding to the 3-year cycle for Appendix 0 in-spections.
These assumptions result in a distribution of examination ef-fort over a 30 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, 9, 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 10, 20, and 30.
Applying an hourly labor rate of $53 results in a total-cost per reactor on a 1988 present value basis of about $300K to 5550K for a 10% and 5%
real discount rate, respectively.
For the industry at-large, assuming a reactor population of 124, comparable present value estimates are $37.2M and $68.2M. All impacts are expressed in 1988 dollars.
Summary Cost Impact (1988 Dollars) 4 Cost Per Reactor 5300K to 5550K Industry-wide Cost 537.2M to 568.2M 8
Industry Cost...Sumary PAsed on the foregoing analysis, high and low estimates of lifetime costs are sumarized below.
Results-are presented on a per reactor and indus-4 try-wide basis.
The low estimates assume a 10% real discount rate, where-as the high estinate assumes a 5% real discount rate.
All costs are ex-pressed in 1988 dollars and all future costs are present valued.
Sumary - Industry Lifetime Costs (1988 dollars)
High Estimate Low Estimate Cost Per Reactor 5862K 5569K-Industry-wide Cost 5107M 570M Safety Impact of Changes in Facility Operation No change.
Subsection IWE examinations would be performed during planned shutdowns.
Other Impacts of implementations In determining other impacts of implementing Subsection IWE rules, the pres-ent containment examination practices are compared with Subsection IWE exami-nation requirements.
Appendix 0 of the NRC Regulations states the following "V.
Inspection and Reporting of Tests A.
Containment inspection.
A general inspection of the accessible in-terior and exterior surfaces of the containment structures and com-i ponents shall be perforned prior to any Type A test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak-tightness.
If there is evidence of structural deterioration Type A tests shall not be per-l formed until corrective action is taken in accordance with repair procedures, nondestructive exaniinations, and tests as specified in the applicable code specified in b50.55a at the comencement of re-pair work.
Such structural deterioration and corrective actions shall be reported as part of'the test report submitted in accordance with V.B."
Table 1 provides a comparison of the Subsection IWE rules with the present Appendix J rules and utility practices, and sumarizes benefit information, 9
TABLE 1 COMPARISON OF CONTAINMENT EXAMINATION AND TESTNG REQUIREMENTS BENEFIT OF APPENDIX J PRESENTgTILITY' PRACTICE IWE CHANGE
.lWE REQUIREMENT REQUIREMENT Category E-A General General visual VT-1 exam Pressure Re-visual exam, exam, if acces-applied to taining Welds if acces-
- sible, important in' Vessels re-
- sible, welds.
3 quire VT-1 exam-ination on more important welds.
Category E-A-1 General General visual Weld loc-Nonpressure-visual exam, exam,-if acces-tions and Retaining Welds if accessible. sible.-
diagrams require VT-3 documented.
examinations.
Category E-B General General visual VT-1 exam Pressure Re-visual exam, exam, if acces-applied to taining Welds if acces,
- sible, important in Containment sible welds.
Penetrations require VT-1 exam-inations on more important welds.
Category E-C General General visual Weld & com-Pressure Re-visual exam exam, if acces-ponent cov-taining Welds if accessible. sible, erage' doc-in Airlocks &
umented.
Equipment Hatches require VT-3 exam-inations.
Category E-D General General visual
' Assures these Seals & Gaskets visual exam, exam, if acces-important require VT-3 if accessible. sible.
items are examinations, examined.
Category E-E General General visual Accessibil-Integral Attach-visual exam, exam, if acces-ity is re-ment Welds re-if accessi-
- sible, quired, quire VT-3 exam-
- ble, inations.
I See Apperdix A for descriptions of VT-1, VT-2 and VT-3 examinations.
2Present utility practice is a direct result of Appendix J requirements, except for IWE-3000.
10
TABLE 1 (continued)-
COMPARISON OF CONTAlf1 MENT EXAMlHATION AND TESTNG REQUIREMENTS l
APPENDlx J PRESENTyTILITY BENEF11 0F IWE P.E0VIREMENT-REQUIREMENT PRACTICE 1WE CHANGE Category E-F General vis-General visual More sensi-Pressure Retain-ual exam, if exam, it acces-tive exam of ing Dissimilar accessible.
sible.
important-Metal Welds re-welds.
quire surface examination, i
. Category E-G General vis.
General visual Control of Pressure Retain-ual exam, it exam, if acces-bolting in-ing Bolting re-accessible,
- sible, tegrity, quires VT-1 exam-ination and bolt torque or tension test.
Category E-P, All General-vis-General visual Little change Pressure Retaining ual exam, if exam, it acces-from present Components, re-accessible, sible, and leak-practices.
quire VT-3 exam-and leakage age tests.
inations and leak-tests.
age-tests.
IWE 3000-None.
Acceptance stan-More consist-l Acceptance Stan-dards are devel-ency.
dards oped on a case-by case basis.
IWE 4000-ASME Code.
ASME Code.
No change.
Repair Proce-
?
dures IWE 5000 Leakage Leakage test.
No change.
System Pressure test.
Tests.
IWE 7000-ASilE Code.
ASME Code.
No change.
Replacements.
I Present utility practice is a direct result of Appendix J requirements, ex-cept for IWE-3000.
t 11
i
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Subsection IWE, Table IWE-2500-1 Examination Category E-P, All Pressure Retaining Components, references the Appendix J examination requirements and i
7 provides appropriatg detail.
A VT-3 visual test of the pressure retaining boundary and a VT-2 visual test for leakage are specified.
If leak channels j
are used, they are required to be unplugged and tested during the Appendix J.
Type A, containment integrated leakage test or tested independently by a Type B I
local leakage test.
{
A requirement of Subsection IWE is that the VT-2 and VT-3 visual examination personnel are required to be qualified to ANSI N45.2.6.
Appendix J does not specify personnel qualification requirements.
Most of the examinations specified in Subsection IWE, Table IWE-2500-1 are VT-3 visual examinations.
If these examinations were documented during the i
Examination Category E-P VT-3 examination of the pressure retaining boundary, these examinations would meet ~ most of the requirements of the other examina.
tion categories in Table IWE-2500-1.
However, it would be necessary to verify that all welds accessible for examination met Subsection IWE requirements.
Present utility practice does not document coverage of specific welds, and in some cases, locations of welds are not known.
Subsection IWE exempts ineccessible welds from examination provided fabrica-tion requirements specified in Subsection IWE-1221 are met.
For older plants not designed to Section til of the ASME Code, these fabrication requirements may not have been met.
In these cases, it may be necessary for utilities to request relief from the NRC for specific welds not meeting Subsection IWE accessibility requirements.
Implementation of the new ASME Code rules imposes certain additional infor-
-mation collection requirements.
The Supporting Statement for Information Collection Requirements in 10 CFR 50.55a in provided in Appendix C.
Incor-porating Subsection IWE will not impact NRR's intent to have the BWR Mark 1-111 i
type containments ultrasonically examined as Subsection IWE does not require this type of testing.
Additionally, the Tirgalex and NPAR panels refer to the endorsement of Subsection IWE as an action that will help achieve thier goals, NRC Costs NRC staff will be required to review and approve the licensee's ISI plans and examination methods resulting from the proposed rule.
However, since the rule will ensure consistency and uniformity among licensee programs, NRC's review and approval process should be simplified.
Without this proposed rule, it would be necessary for each app'licant/ licensee to develop its own program for submittal to the NRC.
Each program would have to be reviewed by the NRC on a case-by-case basis.
This would increase significantly the licensing review time and would make inspections by the staff more difficult because of the nonstandard nature of each program.
Thus, on balance, adoption of the proposed rule is likely to result in a cost saving to the NRC.
Potential Impact of Differences in Facility Type Subsection IWE will apply to all plants, but it is anticipated that the number 12
~
of relief requests will be relatively few 1n number for the total reactor pop-ulation.
Older-plant designs may not be able' to comply with all of the re-quirements of Subsection IWE.
^
Whether the Proposed Action is Interim or Final This proposed rule is forwarded to the CRGR with the intent of publishing the
{
proposed rule in the Federal Register and soliciting public coments.
- Then, the final rule would be published early 1990.
Subsequently, the NRC would then on a routine basis adopt ASME Code updates to Subsection IWE in the same nenner as it currently adopts updates to other Section III and Section XI Subsections.
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APPENDIX A ASME CODE SECTION XI VISUAL EXAMINATION DESCRIPTIONS l
The descriptions of the VT-1, VT-2 and VT-3 visual examinations are contained in section IWA-2210. " Visual Exeminations".
IWA-2211, titled " Visual Examina-I tion VT-1", contains the following:
i A.
The VT-1 visual examination shall be conducted to determine the con-l dition of the part, component, or surface examined, including such conditions as cracks, wear, corrosion, erosion, or physical damage on the surfaces of the part or components.
l B.
Direct VT-1 visual examination may be conducted when access is suffi-cient to place the eye within 24 inches of the surface to be examined and at an angle not less than 30 degree; to the surface.
Mirrors may be used to improve the angle of vision.
Lighting, natural or artifi-cial, shall be sufficient to resolve a 1/32 inch black line on an 18%
1 neutral gray card.
C.
Remote VT-1 visual examination may.be substituted for direct examina-tion.
Remote examination may use aids, such as telescopes, bore-scopes, fiber optics, cameras, or other suitable instruments, pro-vided such systems have-a resolution capability at least equivalent to that attainable by direct visual examination.
A description of the VT-2 visual examination in contained in 1WA-2212 and con-tains the following:
A.
The VT-2 visual examination shall be conducted to locate evidence of leakage from pressure retaining components, or abnormal leakage from components with or without leakage collection systems as required dur-ing the conduct of system pressure or functional test.
B.
The VT-2 visual examination shall be conducted in accordance with IWA-5240.
1WA-2213 contains the following information on the VT-3 visual examination:
A.
The VT-3 visual examination shall be conducted to determine the gener-al mechanical and structural condition of components and their sup-ports, such as the verification of clearances, setting, physical dis-placements, loose or missing parts, debris, corrosion, wear, erosion, or kthe loss of integrity at bolted or welded connections.
B.
The VT-3 examination. shall include examination for conditions that could affect operability or functional adequacy of snubbers, and con-stant load and spring type supports.
C.
For component support and component interiors, the visual examination may be performed remotely with or without optical aids to verify-the structural integrity of the component.
1 A-1
APPENDlX B 50.109 DOCUMENTED EVALUATION l
paragraph 50.109 (a)(4) provides provisions by which the backfit requirements of paragraph of 50.109 (a)(2) and -(a)(3) are inapplicable.
Following is the documented evaluation for the staff finding consistent with paragraph 50.109 (a)(4)(ii).
l Safety Issue Age-related degradation of containments has-occurred.
Many liners-of con-l crete containments were not designed with corrosion allowances.
Erosion of the metal drywell shell at one plant.was found to be occurring at the rate of 20 mils / year.
The following information is found in NRC Information Notice No. 88-82, " Torus Shells with Corrosion and Degraded Coatings in BWR Contain-ments".
During recent inspections at a particular plant, NRC inservice in-l spectors found that the inside surface of the torus shell had corroded. : Thick-ness measurements of the torus shell revealed several areas in which the thick-j ness was at or below the minimum specified wall thickness, in fact, the pres-ent rate of loss of wall thickness is almost double the loss the torus shell was designed for originally.
A survey of other BWR's in one region revealed three Mark I tori had experienced degradation of the coating and that cleaning and recoating will be required.
As stated'in the NRC Information Notice, al-l though the torus shell thinning due to corrosion and the coating degradation in tori found have no immediate effect on plant operation, the NRC staff con-siders these deficiencies to be significant because the measured corrosion rates of torus shells are greater than the corrosion. rates assumed as part of the original design.
The torus shell degradation, if it continues, may jeopar-dize containment integrity.
Additional and potentially more significant degra-dation mechanisms can be anticipated as nuclear power-plants age.
In view of the importance of the containment to the health and safety of the general public, an inservice inspection program which would provide a basis for assur-ing the continued operational integrity of these containments is _necessary-for the adequate protection of the public health and safety.
Subsection IWE does not include requirements to inspect all of the areas that contribute to overall risk.
But, the staff feels that a number of the most important areas (e.g., penetrations, containment welds, liner coatings) that need to be inspected are covered by Subsection IWE.
As pointed out in NUREG-1150, an effective inservice inspection plan is needed to assure that containment integrity is maintained.
The underlying assumption used in the study _of catastrophic containment f ailures was that failure of the containment would not occur under normal accident conditions.
Uncertainty curves show this to be the case up to at least 1.5 times the design accident pressure.
For the probabilities to have any meaning, the underlying assumption that containment a
integrity exists initially must be true.
Subsection IWE inspections will en-
]
sure that the containment will indeed withstand design basis accident pres-sures up to certain limits, thus, ensuring that the public is adequately pro-tected.
B-1
The importance of the containment in a nuclear power plant to the protec-tion of the public health and safety is evident.
The above studies detail the importance of a comprehensive inservice inspection program in assuring the continued structural integrity of nuclear power plant systems and components.
4 Section XI of the ASME Code provides a comprehensive inservice inspection for such systems and components.
Subsectiorr IWE of Section XI provides rules for the examination of metal containments and the liners of concrete containments.
The purpose of the proposed amendment is to incorporate by reference into the existing regulation, which already establishes requirements for inservice in-spection and testing of Class 1, Class 2, and Class 3 components and their supports, the above rules for inservice inspection of containments.
Potential Impact on Occupational Exposure implementation of this rule is not expected to result in significant oc-cupational 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 faciltiy.
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.
Potential Impact of Differences in facility Type, Design or Age Most of the examinations specified in Subsection IWE, Table IWE-2500-1 are VT-3 visual examinations.
If these examinations were documented during the Examination Category E-P VT-3 examination of the pressure retaining bound-ary, these examinations would meet most of the requirements of the other ex-amination categories in Table IWE-2500-1.
However, it would be necessary to verify that all welds accessible for examination met Subsection IWE require-ments.
Present utility practice does not document coverage of specific welds, and in some cases, locations of welds are not known.
Subsection IWE exempts inaccessible welds from examination provided fabri-cation requirements specified in Subsection IWE-1221 are met.
For older plants not designed to Section I!! of the ASME Code, these fabrication require-ments may not have been met.
In these cases, it may be necessary for utili-ties to request relief from the NRC for specific welds not meeting Subsection IWE accessibility requirements.
Estimated Resowce Burden to the NRC NRC st.ff will be required to review and approve the licensee's 151 plans and examiration methods resulting from the proposed rule.
However, since the rule will ensure consistency and uniformity among licensee programs, NRC's re-view ano approval process should be simplified.
Without this proposed rule, it would be necessary for each applicant / licensee to develop its own program l
for submittal to the NRC.
Each program would have to be reviewed by the NRC i
on a case-by-case basis.
This would increase significantly the licensing re-l view time and would make inspections by the staff more difficult because of l
the nonstandard nature of each program.
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Thus, on balance, ' adoption of the proposed rule is more likely to result in cost saving to the NRC.
Potential Safety impact in Plant or Operational Complexity The ins)ections required by 151 programs typically occur during refueling outages or slutdowns.
Therefore, incorporation by reference of Subsection IWE should not impact on operational complexity.
Implementation of the new ASME Code rules imposes < certain additional in-n formation collection requirements.
The Supporting Statement for Information Collection Requirements in 10 CFR 50.55a in provided-in the Regulatory Analy.
sis.
Conclusion Given the importance of ' the containment in the protection of the public the ASME has seen a need to develop minimum requirements health and safety, inspection of containments.And, with the ongoing degrada-
{
for the inservice tion occuring at some containments, the staff feels that Subsection IWE must be implemented so that the utilities can ensure the adequate protection of the public health and safety.
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r Appendix C St pporting Statement for Information Collection Requirements in 10 CFR 50.55a 1.
Justification 4
a.
Need for the Collection of Information The proposed rule would incorporate by reference the 1986-Edition with Addenda through the 1987 Addenda of Subsection IWE, " Require-ments for Class MC Components of Light-Water Cooled Power Plants", of Section XI (Division 1) of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
Subsection IWE provides the rules and requirements for inservice inspection, re-pair,- and replacement of Class MC pressure retaining components and their integral attachments, and metallic shell and penetration liners of Class CC pressure - retaining components and their integral at-i tachments in light-water cooled power plants.
NRC Regulations in 10 CFR 50.55a incorporate by reference Section XI Division 1, of the American Society of Mechanical Engineers ( ASME}
Boiler and Pressure Vessel Code.
This section of the ASME Code sets forth the requirements to which nuclear power plant components are tested and inspected.
Inherent in these requirements are certain record keeping functions.
Implementation of Subsection IWE requires the owner to do.the follow-ing three items:
- 1) prepare plans-and schedules for pressure and inservice examination and tests to meet the requirements of Subsection IWE;
- 2) prepare records of the examinations, tests, replacements, and re-pairs.
Specifically, the-following recordkeeping requirements are-incurred:
IWE-1232 (a)(2), Inaccessible Welds All inaccessible welded joints of class Mc containment vessels, parts, and appurte-nances must be fully radiographed and tested for leak tightness prior to being covered.
This requires the procedures, person-nel qualifications and examination results to be documented.
IWE-1232 (b)(1), Inaccessible Welds All inaccessible metal-lic shell and penetration liner welded joints must be examined by the magnetic particle method, by ultrasonic exami-nation, or radiographed end then leak tested by either vacuum box method, the solution film test, the halogen oiode method, or helium mass spectrometer method prior to being covered.
This requires the procedures, personnel qualifications and ex-amination results to be docu'nented.
IWE-2200 (e), Acceptance Standards - Welds as a result of re-pair or replacement must be examined by the magnetic particle or liquid penetrant method which requires the procedures, personnel qualifications and examination results to be documented.
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IWE-3112 (a), Acceptance Standards - Components with-acceptable flaws require a Preservice Inspection Report Summary.
IWE-3114, Repairs and Reexaminations Repairs and reexamina-tions require owner's. Report for Repairs or Replacements. Form NIS-2, to demonstrate that repairs meet acceptance standards.
(
Verified changes of IWE-3122.1, Acceptance by Examinations flaws must be recorded in accordance with inservice inspection summary reports.
~
Reexaminations require IWE-3124, Repairs and Reexaminations recorded results demonstrating that the repair meets acceptance-standards.
IWE-3514.1, Visual Examinations - Defective seals and' gaskets shall be replaced or repaired which. requires Owner's Report for Repairs or Replacements, Form NIS-2.
IWE-3517.1, Visual Examinations Defective bolting material shall be replaced which requires Form NIS-2.
IWE-4220 (a), Mechanical Removal Process - Mechanical. removal process requires examination by either magnetic particle or lig-uid penetrant method to assure that the defect has been removed.
This requires that procedures, personnel qualifications and ex-amination results be documented.
IWE-4230 (b), Preparation for Repair Welding Repair welding requires examination by magnetic particle or liquid penetrant method.
This requires that procedures, personnel-qualifications and examination results be documented.
IWE-4321 (e), Butterbead-Temperbead Repair. - Certified Material-Test Report required for welding material.
IWE-4322, Welding Qualification - Welding procedures : and weld-ers require certification with the procedures, personnel quali-fications and examination results being documented.
Prior. to welding, -the magnetic.
IWE-4323, Welding Technique particle or liquid penetrant method is used to examine the area.
.These methods require that the procedure and results be docu-
~
mented.
l IWE-4324 (b), Examination - Completed weld must' examined by mag-netic particle or. liquid penetrant method and results recorded.
IWE-4325, Repair Technique - Defects in weld metal shall be re-paired and examined by either magnetic particle or liquid pene-trant method and results recorded.
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1 IWE-7400, Installation Requiring Welding - Welding procedures j
and welders must be certified which requires documentation of these procedures and qualifications.
IWE-7600, Materials - Certified Material Test Report required for welding material.
and 3) prepare preservice and inservice inspection summary reports for Class 1 and 2 pressure retaining components and their supports.
2.
Agency Use of Information These records are used by the licensees, National Board inspectors, in-surance companies, and the NRC in the review of a variety of activities, many of which affect safety.
The records-are generally historical; in nature and povide data on which future activities can be based.- NRC in--
i spection and enforcement personnel can spot check the records required by l
the ASME Code to determine, for example, if. proper inservice examination test-methods were utilized.
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3.
Reduction-of Burden Through Information Technology The information being collected represents the documentation for - the various plant specific inservice inspection programs.
-The NRC has no objection to the use of new information technologies and generally en-courages their use.
4.
Effort to identify Duplication ASME requirements are incorporated to avoid the need-for writing equiva-lent NRC requirements.
This amendment will not duplicate the information co1Mion requirements contained in 'any Other generic regulatory rr.quirement.
l 5.
Effort to Use Similar Information The NRC is using the information reporting requirements specified in the ASME Code in lieu of developing its own equivalent requirements.
l-6.
Effort to Reduce Small Business Burden This amendment to $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 l
Regulatory Flexibility Act in the Small. Business Size Standards set out in regulations issued by the Small Business Administratio'n at 13-CFR Part 121.
Since these companies are dominant in their service areas, the pro-posed amendment does not fall in he province of this Act.
The proposed rule will have no significant effect on a substantial number of small companies.
7.
Consequences of Less Frequent Collection The information is generally not ' collected, but is retained by the C-3
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licensee to be made available to the NRC in the' event of an NRC audit.
8.
Circumstances Which Justify Variation from OMB Guidelines There is no variance from_0MB guidelines.
9.
Consultations Outside the NRC There were no consultations outside the NRC.
10.
Confidentiality of Information NRC provides no pledge of confidentially. for this collection of information.
11.
Justification for Sensitive Questions No sensitive questions ara involved.
Information collected is simply a documentation of inservice inspection examinations.
12.
Estimated Annualized Cost to the Federal Government NRC inspection personnel who audit plant quality assurance records would i
include in their audit verification that the above records are being prop-erly prepared and maintained.
The time associated with NRC inspectors verifying these records would be extremely small when the activity is performed as part of a normal quality assurance audit.
13.
Estimate of Burden a.
Number and Type of Respondents In general, the information collection requirements incurred by
$50.55a through endosement of the ASME Code apply to the owners of the 17 nuclear power plants under construction and to the owners of the 107 nuclear power plants in operations.
b.
Estimated Hours The information collection requirements inherent in incorporating by reference the 1986 Edition with Addenda through the 1987 Addenda of l
Subsection IWE of Section XI, Division 1 of the ASME Code is the to-tal engineering time given for developing and maintaining a Subsec-tion IWE inservice inspection plan-and ensuing records.
This is estimated at 3,300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> to prepare the inservice inspection plan (this is a one time burden) and 1,800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in each successive 10 year interval for updating the inservice inspection plan.
Based on a reactor population of 124, this would result in a total burden of 409,200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> to prepare the plan, and 22,320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> per year for up-dating the plan.
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c.
Estimated Cost Required to Respond to the Collection-Based upon the hours specified in item 13.b it is estimated that the:
cost to prepare the ISI-plan will be $241,000'per reactor. Based on a reactor population of 124, the total industry. costs for this require-t ment is estimated at approximately $29.5M.
This is a one time indus-l try cost.
Subsequent updates to the inservice inspection plan is estimated to cost $9.5K per reactor per year. The total burden then would be $1178K per year.
d.
Record Retention period The retention period for'information is in accordanc'e with'a' schedule provided in paragrcph )WA-6300 of the ASME Code.
The record:reten-tion periods are for the service lifetime of the component or system.
Lifetime retention of the above records-is necessary to ensure ade quate historical informtion on the design and examination of components and-systems to provide a basis for evaluating degradation of these components and systems at-any time during their - service
- lifetime, 14 Reasons for Change in Bciden
'The ASME Code by a consensus process, which included utility, regulatory and inspection personnel, developed this new subsection.
Subsection IWE provides the rules and requirements for inservice inspection, repair, and replacment attachment of Class MC pressure retaining components and their integral attachments, and of metallic shell and penetrations liners -of Class CC pressure-retaining components and their integral attachments in light-water cooled power plants.
This was done in order to provide a consistent set of rules with appropriate examination details for contain-ment structures and to provide a basis for assuring the continued-opera-tional integrity of these containments.
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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'I I
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'h ENCLOSURE 3 t
REGULATORY ANALYSIS.
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Draft Regulatory Analysis T67 Current Proposed Amendment Revision of 10 CFR 50.55e Codes and Standards 1.
Statement of,the Problem The General Design Criteria (Appendix A of Part 50) of the NRC Regula-tions require that structures, systems, and components of light-water-reactors i
be designed, fabricated, erected and tested to quality standards commensurate with the importance of the safety function performed.
Quality Assurance Cri-j teria (Appendix B of Part 50) establishes quality assurance requirements for a
the design, construction and operation of those structures, systems and compo-nents.-
The pertinent requirements of Appendix B apply to all activities af-fecting the safety-related - functions of those structures, systems and compo-nents; these activities include inspecting, testing, operating, maintaining, repairing and modifying.
Without a set of specific rules to implement these quality standards, it would be necessary for each licensee to develop its own 1
program for submittal to the NRC.
Each program would have to reviewed by the staff on a case-by-case basis.
This would increase significantly the.licens-l ing review time and would make inspections by the staff more difficult because of the nonstandard nature of each program.
To provide a consistent set of rules, which the industry has participated in developing, 9 50.55a mandates use of Section XI of the ASME Code for inser-vice inspection of these components.
Section XI is implemented by applicants and licensees of all light-water cooled reactors.
The NRC first endorsed the ASME Code by reference in 10 CFR 50.55a in 1971.
The ASME publishes a new edi-tion of the Code every three years and.new addenda yearly.
It has been a con-tinuing policy of the Commission to update this section of the regulations to keep the references current.
In those cases where an item in'the ASME Code is inconsistent with NRC criteria, NRC regulations may take an exception.to en-I dorsing that portion of the Code, or supplementary criteria maybe incorporated I
into NRC regulations so that the item is consistent with staff requirements.
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In order to provide a consistent set of rules with appropriate examina-tion, and repair and replacement details for containment structures, the indus-try has participated in developing Subsection IWE of Section XI, Division 1, of the American Society of Mechanical-. Engineers Boiler and Pressure Vessel Code (ASME Code).
The purpose of this proposed amendment is to incorporate by reference Subsection IWE into the NRC Regulations.
The ASME Code is developed by the consensus process, which ensures thai!
the various interested sectors (e.g., utility, nuclear steam system suppliers, insurers, regulators) are represented on the standards writing committees and that their viewpoints are considered in the standards writing process.
En-dorsement of the ASME Ccete by the NRC provides a method of incorporating rules into the regulatory process that are acceptable to the NRC and have received industry participation in their development.
If the NRC did not take action to endorse the ASME Code, the NRC position on the methods for inservice inspection would have to be established on a case-by-case basis and improved methods for inservice inspection might not be imple--
mented.
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Age-related degradation of mtainments has occurred. Many liners of con-crete containments were not designed with corrosion allowances.
Erosion of the metal drywell shell at one plant was found to be occurring at the rate of 20 mils / year.
The following information is found in NRC Information Notice No. 88-82, " Torus Shells with Corrosion and Degraded Coatings in BWR Contain-ments".
During recent inspections at a particular plant, NRC inservice in-spectors found that the inside surface of the torus shell had corroded. Thick-ness measurements of the torus shell revealed several areas in which the thick-ness was at or below the minimum specified wall thickness.
In fact, the pres-ent rate of loss of wall thickness is almost double the loss the torus shell was designed for originally.
A survey of other BWR's.in cr.s region revealed three Mark I tori had experienced degradation of the coating and that cleaning and recoating will be required.
As stated in the NRC Information Notice, al-though the torus shell thinning due to corrosion and the coating degradation in tori found have no immediate effect on plant operation, the NRC staff con-siders these deficiencies to be significant because the measured corrosion rates of torus shells are greater than the corrosion rates assumed as part of the original design. The torus shell degradation, if it continues, may jeopar-dize containment integrity.
Additional and potentially more significant degra-dation mechanisms can be anticipated as nuclear power plants age.
In view of the importance of' the containment to the health and safety of the general public, an inser'vice inspection program which would provide a basis for assur-ing the continued; operational integrity of these containments. is necessary for the adequate protection of the public health and safety.
In normal operation, three barriers (the fuel rod cladding, the reactor coolant system pressure. boundary, and the containment pressure boundary) pro-tect the public from the release of radioactive material generated in -the fuel.
In most core meltdown accidents, the first two barriers are progres-sively breached and the containment represents the final barrier to mitigating the consequences of a severe accident. The ability of a containment structure to maintain integrity in a severe accident is determined by the magnitude of l
the loads imposed on -the containment and the response-of the containment to those loads.
Although there is no universally accepted definition of contain-ment failure, typically, for risk purposes, a containment is considered to have failed when the leakage from the containment is great enough that the load will be relieved, or, when the leak rate of radionuclides to the environ-ment is significant. Thus, failure could occur as the result of a gross struc-tural failure of the containment or as the result of a high rate of leakage through a penetration.
Probabilistic risk assessments (PRAs) are largely a snapshot of plant risk in time.
It would be easy to conclude that the identified dominant acci-dent sequences provide a complete and unchancing characterization of plant risk, but this is not the case.
Risk can be decreased in time by modifying systems and procedures.
More importantly, degraded systems, poorly implemen-ted procedures, inadequate maintenance, or ineffective management (to name a few) could drastically alter both the dominant contributors and the absolute magnitude of risk at a plant.
Therefore, it is very important to provide as--
surance that the estimate of risk provided by a PRA is actually realized or even bettered in practice.
This function would be furnished by an effective operational reliability (safety) program."
NUREG-1150, " Reactor Risk Refer-ence Document", used the underlying assumption in the study of catastrophic containment failures was that failure of the containment would not occur un--
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der normal accident conditions.
Uncertainty curves show this to be the case up to at least 1.5 times the design accident pressure.
NUREG-1150 also looked t
l at containment bypass modes.
To quote the NUREG, "In many PRA analyses, se-quences in which the containment is bypassed by interfacing-system LOCAs are 4
l among the most important contributors to overall risk.
Containment bypass se-l quences, including those associated with preexisting openings, are among the most plant specific of all contributors since they depend on detailed config-t l
urations of piping, valving, penetrations, control systems, and the like."
i Subsection IWE does not include requirements to inspect all of the areas l
that contribute to overall risk.
But, the staff feels that a. number of the most important areas (e.g., penetrations, containment welds, liner coatings) that need to be inspected are covered by Subsection IWE.
Also, as pointed out I
in NUREG 1150, an effective inservice inspection plan is needed to assure that this " snapshot in time" is maintained.
For the probabilities to have any mean-ing, the underlying assumption that containment integrity exists initially must be true.
Subsection IWE inspections will assure that the containment will indeed withstand severe accident pressures up to certain limits, thus, assuring that the public is adequately protected.
The importance of the containment in a nuclear power plant to the protec-tion of the public health and safety is evident.
The above studies detail the importance of a comprehensive inservice inspection program in assuring the continued structural integrity of nuclear power plant systems and components.
Section XI of the ASME Code provides a comprehensive inservice inspection for such systeins and components.
Subsection IWE of Section XI provides rules for i
the examination of metal containments and the liners of concrete containments.
The purpose of the proposed amendment is to incorporate by reference into the I
existing regulation, which already establishes requirements for inservice in-i spection and testing of Class 1, Class 2, and Class 3 components and their supports, the above rules for inservice inspection of containments.
Thus, we feel that the adequate minimum safety argument is justified.
The costs of a severe accident (both on and off site) have been estimated, but are not included here for two reasons.
The most important reason is that this rule is being promulgated under the adequate protection of the public health and safety, and thus, the costs of implementation are irrelevant.
The second reason is that the associated costs of a severe accident are a subject r#
great debate with regard to the magnitude.
However, it is very clear that the costs of a severe accident would be on the average orders of magnitude greater that the costs of implementation c' this rule.
2.
Objectives i
The proposed rule would incorporate by reference into 5 50.55a of the NRC's regulations Subsection IWE of Section XI, Division 1, of the ASME Bc;1er and Pressure Vessel Code.
3.
Alternatives An alternative to incorporating by reference into NRC's regulations the requirements of Subsection IWE would be to take no action.
If the NRC did not take action to endorse Subsection IWE, the NRC position on examination prac-tices for steel containment structures and steel liners of concrete contain-3
ments would have to be established on a case by case basis.
If the NRC does not tale action to include the Subsection !WE rules by reference, improved ex-amination practices for steel containment structures might not be implemented.
This will result in containment examinations being performed solely to the present Appendix J rule.
This is not the best approach since Appendix J is l
primarily concerned with containment leakage testing and does not provide pro-cedures for weld and component examinations.
A second alternative to incorporating by reference the Subsection IWE requirements is to incorporate the entire text into the NRC regulations.
Be.
cause of the volume of this section, this approach is not practicable.
4 Consequences Incorporating by reference Subsection IWE, of Section XI, Division 1, of the ASME Code will establish the NPC staff position on the examination of steel containment structures and liners of concrete containments on a generic besis for applicants and licensees, thereby minimizing the need for case-by-4 case evaluations and reducing the time and effort required for submittal prepa-rations and licensee reviews.
The value and impact of ASME Code revis!cns are balanced by the manner in which these revisions are achieved through the American National Standards in.
l stitute(ANSI)consensusprocess.
The ANSI consensus process ensures that par-ticipation in ASME Code development is open to all persons and organizations that might reasonably be expected to be directly and materially affected by the activity, and ensures that such persons and organizations have the opportunity for f air and equitable participation without dominance by any single interest.
Consensus is established when substantial agreement has been achieved by the interests involved.
Consensus requires that all views and objectives be con-sidered, and that a concerted effort be made toward resolution.
ASME Code pro-posed revisions are published for public comment in the ASME Hechanical Engineering and ANSI Reporter publications prior to being submitted for final A5HE and ANSI approval.
Adverse public comments are referred to the appropri-ate technical committee for resolution, i
The consensus process ensures a pro)er balance between utility, regu-l latory and other interests concerned witi revisions to the ASME Code, and l
ensures tht the value of any Code revision is consistent with its impact.
Implementation of the new Code rules requires certain additional infor-mation collection requirements.
The Supporting Statement for Information i
Collection Requirements in 10 CFR 50.55a is provided in Appendix A.
The proposed rule will not heve a significant economic effect on a sub.
stantial number of small entities. This proposed rule affects only the licen-sing 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 set out in regulations issued by the Small Business Administration at 13 CFR Part 121.
Since these con,panies are dominant in their service areas, this pro-posed rule does not fall in the province of this Act.
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1 a.
Cost of and Benefits of Alternative 1.
Impacts 1.1 Industry impacts:
The licensee of each nuclear power reactor is expected to meet the following requirements as a result of the proposed rule:
1.
DevelopmentofanInserviceInspectionPlan(ISI);
2.
Periodic updates to ISI plan; and 3.
Periodic inservice inspections in conformance with 151 plan.
Following is a brief description and cost estimate for each of the broad requirements identified above.
1.1.1 Development of Inservice Inspection Plan Based on information received by utility inservice inspec-tion specialists, the NRC estimates the cost for develop-ment of an 151 plan at $241K (1988 dollars) per reactor.
Based on a reactor population of 124, the total industry cost for this requirement is estimated at approximately
$29.5M.
This is a one time industry cost which would be incurred in the near-term, closely corresponding to adop-tion of this rule.
For the purposes of this analysis, these costs are assumed '.o be expended in 1988.
The specific tasks associated with 151 plan development l'
are presented below for a representet ce reactor.
For most of these tasks, the cost estimax.
is a function of i
the manpower input requirement.
Engu.eering, draf tsman, and consultant labor has been costed at $53 per hour and clerk labor at $23 per hour.
These estimates are based on 1984 base wage rates adjusted by a fector of 1.8 for fringe benefits and plant management, and then escalgted to 1988 dollars using 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 based on a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year level of effort.
it should be pointed out that the cost estimates use zero as a baseline (assuming no inspections taking place), and are therefore conservative.
TasksassociatedwithISIPlanDevelopment(Costsaretonearest$1000)
I NRC analysis of industry labor rates is available in NUREG/CR-4627, Generic Cost Estimates; abstrer+ 6.3, " Industry Labor Rates", June 1986.
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Drawing Update - Includes Preparation of ISI Drawings for the Con-tainmentStructure,NumberingISIComponents(Welds, Penetrations, i
Supports, etc.) and Performing field As-Builts As Required (9 Per-son-MonthsforDraftsman)
$53 per hr x 167 hrs per month x 9 months............
$80K i
Computer Database Preparation - Entering Components into the ISI Computer Database Program for Tracking Purposes (1 Person-Month forClerk) l
$23 per hr x 167 hours0.00193 days <br />0.0464 hours <br />2.761243e-4 weeks <br />6.35435e-5 months <br />................................
$4K Video Mapping Containment - For Job Planning Purposes. $54K Ins?ection Plan Preparation - Preparation of the Inspection Plan, l
Revtew of As-Built Data, Update the ISI Program, Prepare Code Ex-emption/ Relief Requests, Review Construction Data, Etc. (6 Person-MonthsforEngineer)
$53 per hour x 167 hrs per month x 6 months..........
$53K Clerical Assistance - To Assist Engineer for Review of Construc-Updates. Etc. (3 Person Months for Clerk) paration of ISI Program tion Records, Typing, Archival Search, Pre
$23 per hr x 167 hrs per month x 3 months............
$15K t
Additional Engineer and Consultant Work (4 Person-Months)
$53 per hr x 167 hrs per month x 4 months...........
$35K TOTAL ONE-TIME COST......
1241K Summary Cost Impact (1988 dollars)
Cost Per Fcactor 3241K Industry-wide Cost 329.8M I.1.2 Periodic Update" to Inservice Inspection Plan Following the initial development of the ISI plan, updates to the plan are anticipated for subsequent 10-year ISI intervals.
Such updates are deemed necessary to incorporate improvements in' the methods and knowledge-base for detection bnd correction of conteinment defects.
It is expected that an update will not be necessary during the 10 years following initial implementation of the ISI plan.
However, based on a remaining useful life of 30 vears. two succeeding 10-year ISI intervals will occur for which updates will be necessary.
For the purposes of the analy-sis, the updates are assumed to ocur at the midpoint of these 10-year intervals which corresponds to impacts occurring IS years and 25 years into the future.
Industry inservice inspection specialists have estimated an 1800-hour engineering effort per reactor to perform plan updates during a given 10-year ISI interval. fdopting this estimate and 6
l 1
t an engineer labor rate of $53 per hour results in a $95K cost (1988 dollars) per reactor per 10-year inservice interval. This l
cost is assumed to occur in the year 2003 and again in the year i
2018; corresponding to 15 years and 25 years into the future, respectively.
Assuming a 10% real discount rate, the 1988 pres-ent value of two 151 plan updates is estimated at $31K per re-i actor.
If c 5% real discount rate is assumed, the 1988 present l
value cost is about $74K.
l Based on a reactor population of 124, the 1968 present value in-dustry wide cost assuming 10% and 51 discount rates is $3.8M I
and $9.2M respectively.
Summary Cost Impact (1988 dollars)
Cost Per Reactor 531K to 574K Industry-wide Cost 33.8M to 59.2M 1.1.3 Periodic inservice Inspections it is expected that adoption of this rule will result in de-tailed examinations of containments in accordance with the l
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' time for each ten year ISI interval.
For the pur-poses of this analysis the NRC assumes 75% of this inspection effortwilloccurattheendofeach10yearISIinterval,and the remaining 25% will be distributed every three years corre-sponding to the 3-year cycle for Appendix J inspections.
These assumptions result in a distribution of examination effort over a 30 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, 9, 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 10, 20, and 30.
Applying an hourly labor rate of $53 results in a total cost per reactor on a 1988 present value basis of about $300K to
$550K for a 10% and 5% real discount rate, respectively.
For the industry at-large, assuming a reactor population of 124 comparable present value estimates are $37.2M and $68.2M.
All j
impacts are expressed in 1988 dollars.
1 Summary Cost Impact (1988 Dollars)
Cost Per Reactor 3300K to 5550K l
Industry-wide Cost 537.2M to 568.2R I.1.4 Occupational Exposure Implementation of this rule is not expected to result in sig-nificant occupational exposure, particularly when comparei with 7-
- ~ -
w
i i
other 151 examinations and tests, for example-the containment liner examination at the Monticello Plant resuited in 20 milli-rems exposure compared with a total 935 millirems for all test-ing and surveillance activities conducted for plant life exten-sion studies at this faciltiy.
Adopting 20 millirems exposure i
as representative of the dose ser reactor year produces lifetime impacts of 0.6 person-rem anc 75 person-rem for an individual reactor and all reactors, respectively.
Dollar quantification based on $1000 per person-rem results in l
lifetime impacts of $600 per reactor and $75s000 for the reactor population.
These cost impacts are viewed as nil when compared i
to the dollar magnitudes discussed in the previous sections.
1.1.5 Industry Cost... Summary l
t Based on the foregoing analysis, high and low estimates of life-time costs are summarized below.
Results are presented on a per reactor and industry-wide 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 1988 dollars and all future costs are present valued.
Summary - Industry Lifetime Costs (1988 dollars) i High Estimate Low Estimate Cost Per Reactor 5862K 5569K Industry-wide Cost 3107M 370M
!.2 NRC Costs NRC staff will be required to review and approve the licensee's 151 plans and examination methods resulting from the proposed rule.
However, since the rule will ensure consistency and uni-formity among licensee programs, NRC's review and approval pro-cess should be simplified. Without this proposed rule, it would be necessary for each applicant / licensee to develop its 'own program for submittal to the NRC.
Each program would have to be reviewed by the NRC on a case-by-case basis.
This would in-crease significantly the licensing review time and would make inspections by the staff more diff9 cult because of the nonstand-ard nature of each program.
Thus, on balance, adoption of the proposed rule is more likely to result in cost saving to the NRC.
11.
Benefits The qualitative benefits of incorporating by reference Subsection IWE into the regulations are as follows:
8
~.
1 r
1.
Promotes health and safety -- enhances protection of the l
public health and safety using im) roved methods of con-tainment examination to assure conta1nment integrity.
I 2.
Regulatory Efficiency -- standard plans and examination methods will result in savings in NRC and licensee efforts associated with licensing and approval.
But an even great-er benefit of a cteprehensive and standardized !$1 program is that degradation locations may be identified that in the past were not frequently inspected.
3.
Improvements in Knowledge -- ISI plans based on ASME consen-sus will help ensure that ISI programs are state-of-the-art.
Great safety and economic benefits can be derived by beirg able to arrest degradation in early stages.
I i
4.
Uniformity in !$1 Programs -- ensures consistency among all licensees in their ISL programs.
Experience in these areas can be shared which again, gives saf ety and economic bene.
~
f i t,'.
11.1 Quantitative Benefits In looking at the risk-dominant accident sequences ad containment fail-ure modes, the NUREG 1150 study found the causes and types of accidents vary considerably among the studied plants.
The results showed that a wide spec-l trum of uncertainty exists for the plants studied and that no one uncertainty is dominant for the set of plants.
Rather, the sources of important uncer-l tainty cover a wide spectrum of both accident frequency and phenomenological issues.
These uncertainties carry over in trying to predict off-site conse-quences.
In-particular, the ranges of environmental release terms obtained for the containment failure are quite broad.
For the volatile groups of radi-onuclides, the ranges are typically one to two orders of magnitude and, for the more refractory groups of radionuclides, two to three orders of magni-tude.
The specific source term issues that have the greatest impact on the uncertainties in source terms depend not only on the design of the plant but also on the specific plant damage state.
However, the timing and mode of containment failure has a major effect on the estimated ranges of consequences.
Even though the uncertainties are significant, the differences between early and late containment failures overwhelm the uncertainty.
Early failures of the containment result in costs (estimated to be in the billions) that are approximately 3 orders of magnitude greater than for late containment rupture.
With regard to latent cancer fatal-ities, again, early failure is found to possibly result in many more fatali-ties (thousands) than late containment failure.
When compared with the cost to a utility of $238K for the development of the plan, plus approximately $50K ovor a ten year period for updating the plan, plus approximately.$425K over a tei year period for the inservice inspections required..the costs to a utility fo
- implementing Subsection IWE are reasonable relative to the cost impact of a severe accident, l
j 9
a i
In determining other impacts of implenentating Subsection IWE rules, the present containment examination practices are compared with Subsection IWE ex.
amination requirements.
Appendix J of the NRC Regulations states the follow-ing:
"V.
Inspection and Reporting of Tests A.
Containnent Inspection.
A general inspection of the accessible in-terior and exterior surfaces of the containment structures and com-ponents shall be performed prior to any Type A test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak-tightness.
If there is evidence of structural deterioration Type A tests shall not be per-i formed until corrective action is taken in accordance with repair nondestructive examinations, and tests as specified in procedures, le code specified in ISO.Sta at the comencement of re-the applicab pair work.
Such structural deterioration and corrective actions shall be reported as part of the test report submitted in accordance with V.B.*
Table 1 provides a comparison of the Subsection IWE rules with the pres-ent Appendix J rules and utility practices and sumarizes cost / benefit in-formation.
l 10 l-
k i
t TABLE 1 COMPARISON OF CONTAINMENT EXAMINATION AND TESTNG REQUIREMENTS APPENDIX J PRESENTpTILITY BENEFIT OF IWE REQUIREMENT REQUIREMENT PRACTitE IWE CHANGE Category E-A General General visual VT-1 exam Pressure Re-visual exam, exam, if acces-applied to taining Welds if acces-sible.
important in Vessels re-
- sible, welds.
g quire VT-1 exam-ination on more important welds.
Category E-A-1 General General visual Weld loc.
Nonpressure-visual exam, exam, if acces-tions and Retaining Welds if accessible. sible, diagrams require VT-3 documented, examinations.
Category E-B General General visual VT-1 exam Pressure Re-visual exam, exam, if acces-applied to taining Welds if acces.
- sible, important in Containment sible welds.
Penetrations require VT-1 exam-inations on more important welds.
I Category E-C General General visual Weld & com-Pressure Re-visual exam exam, if acces-ponent cov-taining Welos if accessible, sible, erage doc-in Airlocks &
umented.
Equipment Hatches require VT-3 exam-j inations.
4 Category E-D General General visual Assures these Seals & Gaskets visual exan, exam, if acces-important require VT-3 if accessib!e. sible, items are I
examinations, examined.
Category E-E General General visual Accessibil-i Integral Attach-visual exam, exam, if acces-ity is re-ment Welds re-if accessi-
- sible, quired.
I quire VT-3 exam-
- ble, inations.
I See Appendix A for descriptions of VT-1, VT-2 and VT-3 examinations.
2Present utility practice is a direct result of Appendix J requirements, ex-cept for IWE-3000.
11
TABLE 1(continued)
COMPARISON OF CONTAINMENT EXAMINATION AND TESTNG REQUIREMENTS APPENDIX J PRESENT UTILITY BENEFIT OF IWE REQUIREMENT REQUIREMENT PRACTICE IWE CHANGE Category E-F General vis.
General visual More sensi-ual exam exam, if acces-tive exam of accessible,if Pressure Retain-sible.
important ing Dissimilar Metal Welds re-welds.
quire surface examination.
Category E-G General vis-General visual Control of accessible,f exam, if acces-bolting in-ual exam i Pressure Retain-sible.
tegrity.
ing Bolting re-quires YT-1 exam-ination and bolt torque or tension test.
Category E-P, All General vis-General visual Little change Pressure Retaining ual exam, if exam, if acces-from present Components, re-accessible, sible, and leak-practices, quire VT-3 exam-and leakage age tests.
inations and leak-tests.
age tests.
IWE 3000-None.
Acceptance stan-More consist-Acceptance Stan-dards are devel-ency.
dards oped on a case-by case basis.
IWE 4000-ASME Code.
ASME Code.
No change.
Repair Proce-dures IWE 5000 Leakage Leakage test.
No change.
System Pressure test.
Tests.
IWE 7000-ASME Code.
ASME Code.
No change.
l Replacements, i
12
Subsection IWE, Table IWE-2500-1 Examination Category E-P, All Pressure Retaining Components, references the pppendix J examination requirements and provides appropriatg detail.
A VT-3 visual test of the pressure retaining boundary and a VT-2 visual test for leakage are specified.
If leak channels are used, they are required to be unplugged and tested during the Appendix J.
Type A, containment integrated leakage test or tested independently by a Type B local leakage test.
A requirement of Subsection IWE is that the VT-2 and VT-3 visual examina-tion personnel are required to be qualified to ANSI N45.2.6.
Appendix J does not specify personnel qualification requirements.
~
Most of the examirations specified in Subsection IWE, Table IWE-2500-1 are VT-3 visual examinations.
If these examinations were documented during the Examination Category E-P VT-3 examination of the pressure retaining bound-ary, these examinations would meet most of the requirements of the other ex-amination categories in Table IWE-2500-1.
However, it would be necessary to verify that all welds accessible for examination met Subsection IWE require-ments.
Present utility practice does not document coverage of specific welds, and in some cases, locations of welds are not known.
Subsection IWE exempts inaccessible welds from examination provided fabri-cation requirements specified in Subsection IWE-1221 are met.
For older plants not designed to Section 111 of the ASME Code, these fabrication require-ments may not have been met.
In these cases, it may be necessary for utili-ties to request relief from the NRC for specific welds not meeting Subsection lWE accessibility requirements, b.
Impacts on Other Requirements implementation of the new ASME Code rules imposes certain additional in-formation collection requirements.
The Supporting Statement for Information
(
Collection Requirements in 10 CFR 50.55a in provided in Appendix C.
This amendment to i 50.55a affects only the licensing and operating 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 Administration at 13 CFR Part 121. Since these companies are dominant in their service areas, this amendment does not fall in the province of this Act.
The proposed rule would have no significant effect on a substantial number of small companies.
The following is a discussion of the determination that the proposed rule does not impact other requirements (e.g., plant aging or the maintenance rule).
In June 1988, Pacific Northwest Laboratory (PNL) submitted to the NRC a report titled "Prioritization of Tirgalex-Recommended Components for Further Aging Research for Implementation by ALEXCC."
Tirgalex (Technical Integration Review Group for Aging and Life Extension) is a group that was established in I See Appendix A for descriptions of VT-2 and VT-3 examinations.
l 13
1986 at the direction of the E00 to develop a plan to integrate the NRC's ag-ing and life extension activties.
The Tirgalex plan identified the safety-related structures and components that should be prioritized for subsequent evaluation in the NRC Nuclear Plant Aging Research (NPAR) Program.
Contain-ment structures were given the highest risk importance in the study. Also, the re> ort showed that currently there is no BWR containment research planned even t1ough problems are present v occurring.
Further, the pt.nel opinion was i
that the containment is basical / uninspected following construction.
The panel's recommendation was for improved surveillance and test methods to de-tect aging for the risk-significant failure modes in components and struc-tures, such as the containment.
Idaho National Engineering Laboratory (EG&G) published NUREG/CR-4731, "Re-sidual Life Assessment of Major Light-Water Reactor Compor.ents-Overview", in June 1987.
In the section titled in-service Inspection", EG&G's tssessment was that "the establishment of inspection procedures to cover critical areas where adverse environmental conditions such as high temperature, humidity, and/or radiation, and locations subjected to an acidic environment, will be a necessary measure to determine the extent of degradation."
Also, the report stated that "it should be realized that great safety and economic benefits can be derived if an expanded ISI is implemented to cover identified degradation sites that may not be frequently inspected."
NUREG-1144, " Nuclear Plant Aging Research (NPAR) Program Plan", states that "it is not the intent of the NPAR Program to do in-depth engineering eval-uations of: aging and defect characterization, and methods for inspection, sur-veillance, and monitoring of all significant plant elements."
The NPAR pro-gram will support NRR in establishing inspection procedures that are relevant to aging; NRR includes these procedures in the Inspection Enforcement M6nual issued to guide the activities of the regions.
For example, some inspection procedures establish guidance for ascertaining that inservice inspection and testing activities are programmed, planned, conducted, recorded, and reported in accordance with Section XI of the ASME Code, in sumary, NUREG/CR-4731 stated that by implementing a well planned, meaningful, and comprehensive in-spe: tion program followed by a firm program of repairs, the service life of containment structures can be extended, probably to at least double the ini-l tial licensed term.
l c.
Constraints l
No constraints are anticipated in affecting the implementation of Subsec-l tion IWE.
5.
Decision Rationale From the above analysis, it is concluded that containment integrity is needed, and that the longer that containment integrity is maintained during a severe accident, the lower the costs. Even though the uncertainties are sig-nificant, the differences between early and late containment failures over-whelm the uncertainty.
The estimated cost of this alternative would be ac-ceptable relative to the cost impact of a severe accident, considering the importance of maintaining containment structural integrity.
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f i
l 6.
Implementation
}
No ist.plementation problems are anticipated.
The framework for implemen-tation is already established in both the industry and the NRC.
Subsection IWE examinations can be included with the. updated 151 plan presently prepared to meet 10CFR50.55a(g) requirements.
o 1
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APPENDIX A ASME CODE SECTION XI VISUAL EXAMINATION DESCRIPTIONS The descriptions of the VT-1, VT-2 and VT-3 visual examinations are contained in section IWA-2210. " Visual Examinations".
IWA-2211, titled " Visual Examina-tion VT-1", contains the following:
A.
The VT-1 visual examination shall be conducted to determine the con-dition of the part, component, or surface examined, including such conditions as cracks, wear, corrosion, erosion, or physical damage on the surfaces of the part or components.
B.
Direct VT-1 visual examination may be conducted when access is suffi-cient to place the eye within 24 inches of the surface to be examined and at an angle not less than 30 degrees to the surface. Mirrors may be used to improve the angle of vision.
Lighting, natural or artifi.
cial, shall be sufficient to resolve a 1/32 inch black line on an 18%
neutral gray card.
C.
Remote VT-1 visual examinction may be substituted for direct examina-tion.
Remote examination may use aids, such as telescopes, bore-scopes, fiber optics, cameras, or other suitable instruments, pro-vided such systems have a resolution capability at least equivalent to that attainable by direct visual examination.
A description of the VT-2 visual examination in contained in IWA-2212 and con-tains the following:
A.
The VT-2 visual examination shall be conducted to locate evidence of leakage from pressure retaining components, or abnormal leakage from components with or without leakage collection systems as required dur-ing the conduct of system pressure or functional test.
B.
The VT-2 visual examination shall be conducted in accordance with IWA-5240.
IWA-2213 contains the following information on the VT-3 visual examination:
A.
The VT-3 visual examination shall be conducted to determine the gener-al mechanical and structural condition of components and their sup-ports, such as the verification of clearances, setting, physical dis-placements, loose or missing parts, debris, corrosion, wear, erosion, or kthe loss of integrity at bolted or welded connections.
B.
The VT-3 examination shall include examination for conditions that could affect operability or functional adequacy of snubbers, and con-stant load and spring type supports.
C.
For component support and component interiors, the visual examination may be performed remotely with or without optical aids to verify the structural integrity of the component.
A-1
i APPENDIX B 50.109 DOCUMENTED,EVALUAT10N of paragraph of 50.109 (provides provisions by which the backfit requirements paragraph 50.109 (a)(4) a)(2) and (a)(3-) are inapplicable.
Following is the documented evaluation for the st6ff finding consistent with paragraph 50.109 (a)(4)(ii).
Safety Issue Age-related degradation of containments has occurred.
Many ' liners of con-crete containments were not designed with corrosion allowances.
Erosion of the metal drywell shell at one plant was found to be occurring at the rate of 20 mils / year.
The following information is found in NRC Information Notice No. 88-82, " Torus Shells with Corrosion and Degraded Coatings in BWR Contain-ments".
During recent inspections at a particular plant, NRC inservice in.
spectors found that the inside surface of the torus shell had corroded. Thick-ness measurements of the torus shell revealed several areas in which the thick-ness was at or below the minimum specified wall thickness, in fact, the pres-ent rate of loss of wall thickness is almost double the loss the torus shell was designed for originally.
A survey of other BWR's in one region revealed three Mark I tori had experienced degradation of the coating and that cleaning and recoating will be required.
As stated in the NRC Information Notice, al-though the torus shell thinning due to corrosion and the coating degradation in tori found have no immediate effect on plant operation, the NRC staff con-siders these deficiencies to be significant because the measured corrosion rates of torus shells are greater than the corrosion rates assumed as part of the original design.
The torus shell degradation, if it continues may jeopar-dize containment integrity.
Additional and potentially more significant degra-dation mechanisms can be anticipated as nuclear power plants age.
In view of the importance of the containment to the health and safety of the general public, an inservice inspection program which would provide a basis for assur-ing the continued operational integrity of these containments is necessary for the adequate protection of the public health and safety.
Subsection IWE does not include requirements to inspect all of the areas that contribute to overall risk.
But, the staff feels that a number of the most important areas (e.g., penetrations, containment welds, liner coatings) that need to be inspected are covered by subsection IWE.
As pointed out in NVREG-1150, an effective inservice inspection plan is needed to assure that containment integrity is maintained.
The underlying assumption used in the study of catastrophic containment failures was that failure of the containment would not occur under normal accident conditions.
Uncertainty curves show this to be the case up to at least 1.5 times the design accident pressure.
For the probabilities to have any meaning, the underlying assumption that containment integrity exists initially must be true.
Subsection IWE inspections will en-sure that the containment will indeed withstand design basis accident pres-sures up to certain limits, thus, ensuring that the public is adequately pro-
- tected, i
i B-1
The importance of the containment in a nuclear power plant to the protec-tion of the public health and safety is evident.
The above studies detail the importance of a comprehensive inservice inspection program in assuring the continued structural integrity of nuclear power plant systems and components.
Section XI of the ASME Code provides a comprehensive inservice inspection for such systems and components.
Subsection IWE of Section XI provides rules for the examination of metal containments and the liners of concrete containments.
The purpose of the proposed amendment is to incorporate by reference into the existing regulation, which already establishes requirements for inservice in-spection and testing of Class 1, Class 2, and Class 3 components and their supports, the above rules for inservice inspection of containments.
Potential Impact on Occupational Exposure implementation of this rule is not expected to result in significant oc-cupational exposure, particularly when compared with other 151 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 faciltiy.
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.
Potential Impact of Differences in Facility Type. Design or Age Most of the examinations specified in Subsection IWE, Table IWE-2500-1 are VT-3 visual examinations.
If these examinations were documented during the Examination Category E-P VT-3 examination of the pressure retaining bound-ary, these examinations would meet most of the requirements of the other ex-amination categories in Table IWE-2500-1.
However, it would be necessary to l
verify that all welds accessible for examination met subsection IWE require-ments.
present utility practice does not document coverage of specific welds, and in some cases, locations of welds are not known.
Subsection IWE exempts inaccessible welds from examination provided fabri-cation requirements specified in Subsection lWE-1221 are met.
For older plants not designed to Section 111 of the ASME Code, these fabrication require-ments may not have been met, in these cases, it may be necessary for utili-ties to request relief from the NRC for specific welds not meeting Subsection IWE accessibility requirements.
Estimated Resource Burden to the NRC NRC staff will be required to review and approve the licensee's ISI plans and examination methods resulting from the proposed rule.
However, since the rule will ensure consistency and uniformity among licensee programs, NRC's re-l view and approval process should be simplified.
Without this proposed rule, it would be necessary for each applicant / licensee to develop its own program for submittal to the NRC.
Each program would have to be reviewed by the NRC i
on a case-by-case bads.
This would increase significantly the licensing re-view time and would make inspections by the staff more difficult because of the nonstandard nature of each program.
B-2
Thus, on balance, adoption of the proposed ceie is more likely to result in cost saving to the NRC.
potential Safety Impact in Plant or Operational Complexity The ins)ections required by 151 programs typically occur during refueling outages or slutdowns.
Therefore, incorporation by reference of Subsection IWE should not impact on operational complexity.
Implementation of the new i,$ME Code rules imposes certain additional in-formation collection requirements.
The Supporting Statement for Information Collection Requirements in 10 CFR 50.55a in provided in the Regulatory Analy.
sis.
Conclusion Given the importance of the containment in the protection of the public health and safety, the ASME has seen a need to develop minimum requirements for the inservice inspection of containments.
And, with the ongoing degrada-tion occuring at some containments, the staff feels that Subsection IWE must be implemented so that the utilities can ensure the adequate protection of the public health and safety.
I I
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Appendix C Supporting Statement for Information Collectie Muirements in 10 CFR 50.55a 1.
Justification a.
Need for the Collection of Information The proposed rule would incorporate by reference the 1986 Edition with Addenda through the 1987 Addenda of Subsection IWE, " Require-ments for Class MC Components of Light-Water Cooled Power Plants", of Section XI (Division 1) of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
Subsection IWE provides the rules and requirements for inservice inspection, re-pair, and replacement of Class MC pressure retaining components and their integral attachments, and metallic shell and senetration liners of Class CC pressure retaining components and t1eir integral at-tachments in light-water cooled power plants.
NRC Regulations in 10 CFR 50.55a incorporate by reference Section x1 Division 1, of the American Society of Mechanical Engineers (ASME}
Boiler and Pressure Vessel Code.
This section of the ASME Code sett, forth the requirements to which nuclear power olant components are tested and inspected.
Inherent in these requ'irements are certain record keeping functions.
l Implementation of Subsection IWE requires the owner to do the follow-l ing three items:
- 1) prepare plans and schedules for pressure and inservice examination and tests to meet the requirements of Subsection IWE;
- 2) prepare records of the examinations, tests, replacements, and re-pairs.
Specifically, the following recordkeeping requirements are incurred:
l IWE-1232 (a)(2), inaccessible Welds - All inaccessible welded I
joints of class Mc containment vessels, parts, and appurte-nonces must be fully radiographed and tested for leak tightness prior to being covered.
This requires the procedures, person-nel qualifications and examination results to be documented.
IWE-1232(b)(1),inaccessibleWelds All inaccessible metal-lic shell and penetration liner welded joints must be examined by the magnetic particle method, by ultrasonic exami-l nation, or radiographed and then leak tested by either vacuum box method, the solution film test, the halogen diode method, l
or helium mass spectrometer method prior to being covered.
l This requires the procedures, personnel qualifications and ex.
amination results to be documented.
IWE-2200 (e), Acceptance Standards - Welds as a result of re-pair or replacement must be examined by the magnetic particle or liquid penetrant method which requires the procedures, personnel qualifications and examination results to be documented.
C-1
IWE-3112 (a), Acceptance Standards - Components with acceptable flaws require a Preservice Inspection Report Summary.
Repairs and reexamina-IWE 3114 Repairs and Reexaminations tions require Owner's Report for Repairs or Replacements, Form NIS 2, to demonstrate that repairs meet acceptance standards.
Verified changes of IWE 3122.1, Acceptance by Examinations flaws must be recorded in accordance with inservice inspection summary reports.
t Reexaminations require IWE-3124, Repairs and Reexaminations recorded results demonstrating that the repair meets acceptance standards.
IWE-3514.1, Visual Examinations - Defective seals and gaskets shall be replaced or repaired which requires Owner's Report for Repairs or Replacements, Form NIS-2.
Defective bolting material IWE-3517.1, Visual Examinations shall be replaced which requires Form NIS-2.
IWE-4220 (a), Mechanical Removal Process Mechatiical removal process requires examination by either magnetic particle or liq.
uid penetrant method to assure that the defect has been removed.
This requires that procedures, personnel qualifications and ex-amination results be documented.
Repair welding IWE-4230 (b), Preparation for Repair Welding requires examination by magnetic particle or liquid penetrant i
i method. This requires that procedures, personnel qualifications
(
and examination results be documented.
IWE-4321 (e), Butterbead-Temperbead Repair - Certified Material Test Report required for welding material.
IWE-4M2, Weldir.g Qualification - Welding procedures and weld-er', require certification with the procedures, personnel quali-f' cations and examination results being documented, i
Prior to welding, the magnetic l
1WE-4323, Welding Technique particle or liquid penetrant method is used to examine the area.
These methods require that the procedure and results be docu-mented.
IWE-4324 (b), Examination - Completed weld must examined by mag-netic particle or liquid penetrant method and results recorded.
IWE-4325, Repair Technique - Defects in weld metal shall be re-I paired and examined by either magnetic particle or liquid oene-trant method and results recorded.
C-2
1 I
Welding procedures IWE-7400, Installation Requiring Welding and welders must be certified which requires documentation of these procedures and qualifications.
IWE-7600, Materials - Certified Material Test Report required for welding material.
and 3) prepare preservice and inservice inspection summary reports for Class 1 and 2 pressure retaining components and their supports.
2.
Agency use of Information These records are used by the licensees, National Board inspectors, in-surance companies, and the NRC in the review of a variety of activities, many of which affect safety.
The records are generally historical in nature and povide data on which future activities can be based.
NRC in-spection and enforcement personnel can spot check the' records required by the ASME Code to determine, for example, if proper inservice examination test methods were utilized.
3.
Reduction of Burden Through Information Technology The information being collected represents the documentation for the various plant specific inservice inspection programs.
The NRC has no objection to the yte of new information technologies and generally en-courages their use.
4
[fforttoidentifyDuplication AWE requirements are incorporated to avoid the need for writing equiva-lent NRC requirements.
This amendment will not duplicate the information collection requirements contained in any other generic regulatory requirement.
i 5.
Effort to Use Similar Information The NRC is using the information reporting requirements specified in the ASME Code in lieu of developing its own equivalent requirements.
6.
Effort to Reduce Small Business Burden l
This amendment to $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 set out in regulations issued by the Small Business Administration at 13 CFR Part 121.
Since these companies are dominant in their service areas, the pro-posed amendment does not fall in he province of this Act.
The proposed rule will have no significant effect on a substantial number of small companies.
7.
Consequences of Less reequent Collection The inform 6 tion is generally not collected, but is retained by the C-3
licensee to be made available to the NRC in the event of an NRC audit.
8.
Circumstances Which Justify Variation from OMB Guidelines There is no variance from OMB guidelines.
9.
Consultations Outside the NRC There were no consultations outside the NRC.
10.
Confidentiality of Information NRC provides no pledge of confidentially for this collection of f
information.
- 11. gstificationforSensitiveQuestions s
No sensitive questions are involved.
Information collected is simply a documentation of inservice inspection examinations.
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 prop-erly prepared and maintained.
The time associated with NRC inspectors verifying these records would be extremely small when the activity is performed as part of a normal quality assurance audit.
13.
Estimate of Burden a.
Number and Type of Respondents in general, the information collection requirements incurred by 550.55a through endosement of the ASME Code apply to the owners of the 17 nuclear power plants under construction and to the owners of the 107 nuclear power plants in operations, b.
Estimated Hours The information collection requirtments inherent in incorporating by reference the 1986 Edition with Addenda through the 1987 Addenda of Subsection IWE of Section XI. Division 1 of the ASME Code is the to-tal engineering time given for developing and maintaining a Subsec-tion IWE inservice inspection plan and ensuing records.
This is estimated at 3,300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> to prepare the inservice inspection plan (this is a one time burden) and 1,800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in each successive 10 year interval for updating the inservice inspection plan.
Based on a reactor population of 124, this would result in a total burden of 409,200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> to prepare the plan,-and 22,320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> per year for up-dating the plan.
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Estimated Cost Required to Respond to the Collectio,n Based upon the hours specified in Item 13.b it is estimated that the cost to prepare the ISI plan will be $241,000 per reactor. Based on a reactor population of 124, the total industry costs for this require-ment is estimated at approximately $29.5M. This is a one time indus-try cost.
Subsequent updates to the inservice inspection plan is estimated to cost $9.5K per reactor per year. The total burden then would be $1178K per year, d.
Record Retention Period The retention period for information is in accordance with a schedule provided in paragraph IWA-6300 of the ASME Code.
The record reten-tion periods are for the service lifetime of the component or system.
Lifetime retention of the above records is necessary to ensure ade-quate historical informtion on the design and examination of components and systems to provide a basis for evaluating degradation of these components and syttems at any time during their service lifetime.
14 Reasons for Change in Burdy The ASME Code by a consensus process, which included utility, regulatory and inspection personnel, developed this new subsection.
Subsection IWE provides the rules and requirements for inservice inspection, repair, and replacment attachment of Class MC pressure retaining components and their integral attachments, and of metallic shell and penetrations liners of i
Class CC pressure retaining components and their integral attachments in light-water cooled power plants.
This was done in order to provide a consistent set of rules with appropriate examination details for contain-ment structures and to provide a basis for assuring the continued opera-tional integrity of these containments.
- 15., Publication (or Statistical Use This infor. nation will not be published for statistien1 use.
B.
COLLECTION OF INFORMATION EMPLOYING STATISTICAL METHODS I
Statistical methods are not used in the collection of the required information.
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APPEND 1X B 50.109 DOCUMENTED EVALUATION Paragraph 50.109 (a)(4)(provides provisions by which the backfit requirements i
of paragraph of 50.109 a)(2)and(a)(3)areinapplicable.
Following is the documented evaluation for the staff finding consistent with paragraph 50.109 (a)(4)(ii),
1 Safety Issue 1
Age-related degradation of containments has occurred. Many liners of con-crete containments were not designed with corrosion allowances.
Erosion of the petal drywell shell at one plant was found to be occurring at the rate of 20 mils / year. The following information is found in NRC Information Notice 4
No. 88-82 " Torus Shells with Corrosion and Degraded Coatings in BWR Contain-ments".
During recent inspections at a particular plant NRC inservice in-spectors found that the inside surface of the torus shell had corroded.
T hick-ness measurements of the torus shell revealed several areas in which the thick-ness was at or below the minimum specified wall thickness.
In fact, the pres-ent rate of loss of wall thickness is almost double the loss the torus shell was designed for originally. A survey of other BWR's in one region revealed three Mark i tori had experienced degradation of the coating and that cleaning and recoating will be required.
As stated in the NRC Information Notice, al-though the torus shell thinning due to corrosion and the coating degradation in tori found have no immediate effect on plant operation, the NRC staff con-siders these deficiencies to be significant because the measured corrosion rates of torus shells are greater than the corrosion rates assumed as part of the original design. The torus shell degradation, if it continues, may jeopar-dize containment integrity. Additional and potentially more significant degra-dation mechanisms can be anticipated as nuclear power plants age.
In view of the importance of the containment to the health and safety of the general public, an inservice inspection program which would provide a basis for assur-ing the continued operational integrity of these containments is necessary for the adequate protection of the public health and safety.
4 Subsection 1WE does not include requirements to inspect all of the areas that contribute to overall risk.
But, the staff feels that a number of the mostimportantareas(e.g., penetrations,containmentwelds,linercoatings) that need to be inspected are covered by Subsection IWE.
As pointed out in NUREG-1150, an effective inservice inspection plan is needed to,ssure that containment integrity is maintained. The underlying assumption used in the study of catastrophic containment failures was that failure of the containment would not occur ur. der normal accident conditions.
Uncertainty curves show this to be the case up to at least 1.5 times the design accident pressure.
For the probabilities to have any meaning, the underlying assumption that containment integrity exists initially must be true.
Subsection IWE inspections will en-sure that the containment will indead withstand design basis accident pres-sures up to certain limits, thus, ensuring that the public is adequately pro-tected.
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l The importance of the containment in a nuclear power plant to the protec-tion of the public health and safety is evident.
The above studies detail the importance of a comprehensive inservice inspection program in assuring the continued structural integrity of nuclear power plant systems and components.
Section XI of the ASME Code provides a comprehensive inservice inspection for such systems and components.
Subsection lWE of Section XI provides rules for the txamination of metal containments and the liners of concrete containments.
The purpose of the proposed amendntnt is to incorporate by reference into the existing regulation, witch already establishes requirements for inservice in-spection and testing of Class 1. Class 2, and Class 3 components and their supports, the above rules for inservice inspection of containments.
Potential Impact on Occupational Exposure Implementation of this rule is not expected to result in significant oc-cupational exposure, particularly when compared with other ISI examinations and tests.
For example the containment liner examination at the Monticello Plantresultedin20milliremsexposurecomparedwithatotal935 millirems for all testing and surveillance activities conducted for plant life extension studies at this faciltiy. 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.
Potential Impact of Differences in Facility Type. Design or Age Most of the examinations specified in Subsection IWE, Table IWE-2500-1 are VT-3 visual examinations, if these examinations were documented during the Examinatior Category E-p VT-3 examination of the pressure retaiaing bound-ary, these enminations would meet most of the recuirements of the other ex-amination rotegories in Table IWE 2500-1. However, it would be necessary to l
verify that all welds sccessible for examination met Subsection IWE require-mento.
Present utility practice does not document coverage of specific welds, and in some cases, locations of welds are not known.
Subsection IWE exempts inaccessible welds from examination provided fabri-cation requirements specified in Subsection IWE-1221 are met.
For older plents not designed to Section 111 of the ASME Code, these fabrication require-ments may not have been met.
In these cases, it may be necessary for utili-ties to request relief from the NRC for specific welds not meeting Subsection IWE accessibility requirements.
Estinated Resource Burden to the NRC NRC staff will be required to review and approve the licensee's ISI plans and examination methods resulting from the proposed rule.
However, since the rule will ensure consistency and uniformity among licensee programs, EC's re-view and approval process should be simplified.
Without this proposed rule, it would be necessary for each applicant / licensee to develop its own program for submittal to the NRC.
Each program would have to be reviewed by the NRC on a case-by-case basis. This would increase significantly the licensing re-view time and would make inspections by the staff more difficult because of 1
the nonstandard nature of each program.
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Thus, on balance, adoption of the proposed rule is more likely to result in cost saving to the NRC, potential Safety Impact in Plant or Operational Complexity The insoections required by ISI programs typically occur during refueling Therefore incorporation by reference of Subsection IWI outages or siutdowns.
should not impact on operational, complexity, Implementation of the new ASME Code rules imposes certain additional in-formation collection requirements. The Supporting Statement for Information Collection Requirements in 10 CFR 50.55a in provided in the Regulatory Analy-sis.
Conclusion Given the importance of the containment in the protection of the public health and safety, the ASME has seen a need to develop minimum requirements for the inservice inspection of containments. And, with the ongoing degrada-tion occuring at son;a containments, the staff feels that Subsection IWE must be implemented so that the utilities can ensure the adequate protection of the public health and safety.
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ENCLOSURE 4 i
0FFICE CONCURRENCES 7
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