ML20058P854

Request for OMB Review & Supporting Statement Re 10CFR50, Domestic Licensing of Production & Utilization Facilities. Estimated Respondent Burden Is 117,000 H

ML20058P854
Person / Time
Issue date:	12/09/1993
From:	Cranford G NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
To:
References
OMB-3150-0011, OMB-3150-11, NUDOCS 9312270304
Download: ML20058P854 (42)

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I PART III.-Compista This Pcrt Only if ths Request is for Appreycl cf a Collection

~ i cf infermatien Undir thi Ptperwirk Reducti:n Act cnd 5 CFR 1320.

13. Abst act-Descr'De reeds, uses and affected pubhc in 50 words or less "Nuc1 ear Facilities, Nuclear Power P1 ant Constructior The proposed amendment would incorporate by reference the 1992 edition with the 1992 Addenda of Subsections IWE and IWL of Section XI, Division 1 of the ASME Code with their -attendant information collection requirements.

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• "" h* V tow 7.n (Oneck enly one)

,ntsm s non altections nct contained in rules 2 C Emergency submission (certi6 cat;on at*actrec)

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J_.'esMe egaancepo;aerge proposec) 6 T nal or intenm fmal without prior NPRM 7 Enter date of expected or actual recerai 4 3 'w; me :* m c.csec r emamg (NDDM)

A U Regular suomissen Reg' ster pubhcaten at this stage of rulemaking so W P'. ns orev.c s;, c.onsned B C Emergency ssbm ssion (certAcaron ar*ased)

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4 C Reirstatement of a previously approved coitect#ori t0F which approvai i I % s c c 3 cerently aporcved conection as m ed

_ E#emsn c tre emrat Dr cate of a current'y approved cohection 5 C Existmg conecten in use without an oMB controt number e

r.'Qqrje *' t*e s.Dstance er m tne 'nethod of collection i

.6. %en a mcm 7 ~ %mte%(mclude standard oct oral form number (s))

22. Purpose of enformation collectson (check as many as apply)

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i 1 O Apphcation for benefits N/A 2 O Program evaluation j

T' % # epc$1 er 5 sc!csre burcen 3 0 cenerai purpose statistics -

I67 2 uoer at <esporce-ts 4

Regu'atory or cornphance 2 Ner.er of est.crses ce* responcent 2R 8 5 C Program planning or management 3 Tcta !mnua' respo"ses(!,re I fees Ane 2) 4 313 6 C Research 842.7 7 O Audit

. rws ae >esarse 5 Tcta ac M eJr~esime o 3_634_365

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23. Frequency of recordkeepmg or reporting (crteca att tnat apply) 1 Neoer c' recercacepers 167 1 E Recordkeepmg 2 Ars a hors per eccoceper 12.849.4 Reporting 3 :ta' recoceepmg neurs (1,re 1 tres hec 2)

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3 Me er:eiMe 1 ess "ne 2; bonana tion ct dot *erence 8

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23. CU"e*! f ?c5t rece*t) CM6 control number or comment number

24. Respondents' obhgat on to cornply (checA the strongest oCl.gation rt:stapphesl g g__Qpgj 10 voiuntary 21 %c.g e:.m:w care 2 O Reauired to obta'n or retain a benefit June _3L 1994 3 U "'""*' '

2 L ke r e esorcents prear;!y educational agencies or institutions or is the pnmary purpose of the collect on related to Federal education programs? C Yes @ No

26. D:es re gency use samsung to setect respondents or does the agency recommend or prescribe the use of samphng or statistical analys:s e, <escar. cents >

0 res G No 2 7. Depa*0ry a. thor:ty *0r tre ;nformation collecton 10 crn 50

or FR

or,other (specity)

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P aperwork Certification P 5,.0t!i' -g this reAest fo-OMB apptcwal, the agency head. the senior official of an authorged representative, certifies that the requirer"ents of 5 CFR 1320. the Prway At statstica! stancarcs or $rectses, and any other apphcab6e information pohcy directives have been comphed with.

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Supportino Statement for Information Collection Recuirements in l

ProDosed Rule. 10 CFR 50.55a (3150-0011)

Descriotion of the Information Collection i

Under 10 CFR 50.55a. each operating license for a boiling or pressurized water-cooled nuclear power facility must meet specific requirements of the i

ASME Boiler and Pressure Vessel Code. These requirements are incorporated by reference to avoid additional burden to industry and unnecessary duplication of requirements. This rulemaking would incorporate the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL. of Section XI. Division 1.

Implementation of Subsections IWE and IWL requires the owner to prepare the following:

Plans and schedules for preservice and inservice examination and tests to meei. the requirements of Subsection IWE and Subsection IWL:

Preservice and inservice inspection summary reports for Class 1 and 2 pressure retaining components and their integral supports.

Records of the examinations, tests, replacements, and repairs.

Specifically, the following recordkeeping requirements are incurred:

t IWE-1232 (a)(2). Inaccessible Surface Areas - The procedures for radiography and leak testing, personnel qualifications, and examination results must be documented for all welded joints that are inaccessible for examination.

IWE-1232 (b)(1). Inaccessible Surface Areas - The procedures for magnetic particle or ultrasonic examination, radiography, and leak testing: personnel qualifications: and examination results must be documented for all portions of Class CC metallic shell and penetration liners embedded in concrete or otherwise l

made inaccessible during construction or as a result of repair or replacement.

i l

IWE-2200 (d). Preservice Examination - When a vessel, liner, or portion thereof is repaired or replaced during the service lifetime of a plant. the preservice examination requirements for the vessel repair or replacement must be documented that the repair meets the acceptance criteria.

IWE-2200 (e). Preservice Examination The ' procedures, personnel qualifications, and examination results must be documented for welds made as part of a repair or a replacement if examined by the magnetic particle or liquid penetrant method.

IWE 2200 (a). Preservice Examination After repaired or i

replaced welds are examined by the magnetic particle or liquid penetrant method, if coatings are reapplied, the condition of the new paint or coating shall be documented in the preservice examination records.

IWE-2500 (c)(2). Examination and Pressure Test Recuirements -

1 Procedures for measurements in accordance with Section V. T 544

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j personnel qualifications, and examination results must be i

documented for augmented examinations of surface areas accessible from one side only.

1 IWE-3112 (a). Acceptance Commnents containing flaws that do j

not exceed acceptable standarcs are acceptable for service.

provided the flaws are recorded in terms of location, size, j

shape. orientation, and dist: <bution within the component.

IWE 3114.

Repairs and Reexaminations Repairs and l

1 reexaminations must be recorded on Form NIS-2. Owner's Report s

t for Repairs or Replacements, and demonstrate that the repair j

meets the acceptance standards.

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j IWE-3122.1. Acceptance by Examiratica Comoonents with examination results that meet the acceptance standards are acceptable for continued service.

Verified changes of flaws i

from prior examinations shall be reported in inservice j

inspection summary reports.

IWE-3122.2.

Acceptance by Repair Components whose examination results reveal flaws that do not meet acceptance standards shall be unacceptable for continued service. Repairs or mechanical removal of unacceptable components must be documented on Form NIS-2. Owner's Report for Repairs or i

Replacements.

IWE-3122.3. Acceptance by Replacement - As an alternative to IWE-3122.2. Acceptance by Repair, the component or the component j

portion contabling the flaw may be replaced.

If welding is required. documentation is required for welding procedures.

welder certification and qualifications, and a Certified Material Test Report for the welding material.

4 IWE-3124. ReDairs ant Reexaminations The results of 1

reexaminations must b6 documented and demonstrate that the 1

repair meets acceptance standards.

a i

IWE-3130. Inservice -Visual Examinations Components whose visual examination reveals areas that are unacceptable for continued service must be documented that the acceptance j

requirements of IWE-3120 are satisfied.

IWE 3510.1 (b). Visual Examinations - Containment Surfaces Prior to conducting a Type A test, conditions that may affect containment structural integrity or leak tightness shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE-3510.2. Visual Examinations. VT-3. on Coated Areas Containment Surfaces Coated areas that may show signs of flaking, blistering, peeling, discoloration, and other signs of 3

distress shall be accepted by engineering evaluation or 2

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corrected by repair or replacement and be documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE-3510.3. Visual Examinations. VT-3. on Noncoated Areas -

Containment Surfaces Noncoated areas that show signs of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS-2, Owner's Report for Repairs or Replacements.

IWE-3511.1. Visual Examinations. VT-3.

on Coated Areas -

Pressure Retainina Welds -

Coated areas that show signs of flak.ing, blistering, peeling, discoloration, and other signs of distress shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE-3511.2. Visual Examinations. VT-3. on Noncoated Areas -

Pressure Retainina Welds - Noncoated areas that show signs of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE 3512.1. VT-1 Visual Examinations Coated Areas -

Coated areas that show signs of cracking, discoloration, wear, pitting, excessive corrosion, arc

strikes, gouges, surface discontinuities. -dents, and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE-3512.2. VT-1 Visual Examinations Noncoated Areas Noncoated areas that show signs of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents.-

and other signs of surface irregularities shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE-3512.3. Ultrasonic Examination Containment vessel examinations that reveal material loss exceeding 10% of the nominal containment wall thickness, or material loss that is projected to exceed 10% of the nominal. wall thickness prior to the next examination shall be accepted by engineering evaluation or corrected by repair or replacement and documented on Form NIS 2, Owner's Report for Repairs or Replacements.

IWE 3513.1. Visual Examinations. Seals. Gaskets. and Moisture Barriers -

Seals, gaskets. and moisture barriers shall be examined for wear, damage, erosion, tear, surface cracks, or other defects that may violate the leak tight intt7rity_ and 3

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documented on Form NIS 2, Owner's Report for Repairs or j

Replacements.

IWE-3515.1. Visual Examinations. Pressure Retainino Boltino -

Bolting materials shall be examined in accordance with the i

j material specification for defects which may cause the bolted connection to violate either the leak-tight or structural integrity. Replaced defective items shall be documented on Form d

l NIS-2. Owner's Report for Repairs or Replacements.

1 Article IWE 4000 Repair Procedures, are covered by the rules of IWA-4000.

I IWA 4130. Repair Procram Repair operations shall be performed in accordance with a program that delineates the essential requirements.

Prior to authorizing a repair, the Owner shall evaluate the suitability of the repair.

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IWA 4210.

Storace and Handlino of Weldino Material Procedures for welding material control shall be included in the '

repair program.

Welding material must be certified by a material test report.

IWA-4340. Defect Removal Procedures for the removal of defects, personnel qualification, and examination results shall be documented.

IWA-4400. Weldino and Welder Qualifications (Includino Weldino Welding procedures, welder certifications, Operators) j personnel qualifications, and examination results must be

' documented.

j IWA-4600. Examination The repaired areas shall be examined to establish a new preservice record. The method that detected i

the flaw shall be included in the record.

IWL-2523.2. Samole Examination and Testino - Tension tests 1

i performed on each removed wire or strand shall be recorded with j

yield strength, ultimate tensile strength and elongation.

IWL-2524.1. Visual Examinatiqn - Visual examinations of tendon anchorage areas shall be documented and include the physical condition of each area.

IWL-2524.2. Free Water Documentation The quantity of free water contained in the anchorage end cap as well as any which drains from the tendon during the examination process shall. be documented.

IWL 2526. Removal and Reolacement of Corrosion Protection Medium

- The amount of corrosion protection medium removed for samples shall be measured. The total amount replaced and the difference between the two amounts shall be documented.

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t IWL-3310. Evaluation Report - The owner shall prepare an Engineering Evaluation Report for items with examination results that do not meet the acceptance standards of IWL-3100 or IWL-3200.

IWL-7120. Replacement Proaram - A replacement plan must document the removal, reinstallation and replacement of post-tensioning system items for concrete containments.

In addition. the following requirements, which are modifications to Subsection IWL, must also be submitted by report to the NRC:

50.55a(b)(2)(ix)(B) - An Engineering Evaluation Report, when consecutive surveillances of tendon prestressing forces indicate that the tendon force would be less than the minimum design prestress requirements.

50.55a(b)(2)(ix)(C) - A difference of more than 10% (from that recorded during installation of the tendons) in elongation corresponding to a specific load during detensioning and retensioning of tendons.

l 50.55a(b)(2)(ix)(D) - Sampled sheathing filler grease containing chemically combined water exceeding 10% by weight, or replaced grease exceeding 10% of the net duct volume.

50.55a(b)(2)(ix)(E)

An evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas i

that could indicate the presence of or result in degradation to such inaccessible areas.

A.

JUSTIFICATION 1.

Need for the Collection of Information NRC regulations at 10 CFR S 50.55a incorporate by reference Division 1 rules of Section XI, " Rules for Inservice Ir.5pection of Nuclear Power Plant Components," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).Section XI sets forth the requirements to which nuclear power plant components are tested and inspected. There are existing recordkeeping recuirements in Section XI. The proposed rule would incorporate by reference the 1992 Edition with Addenda through the 1992 Addenda of Subsection IWE, " Requirements for Class MC Components of Light Water Cooled Power Plants," and Subsection IWL, " Requirements for Class CC Components of Light-Water Cooled Power Plants," of Section XI (Division 1), of the ASME Code.

Subsection IWE provides the rules and requirements for inservice inspection of Class MC aressure retaining components and their integral attachments, and meta'lic shell and penetration liners of Class CC pressure retaining components and their integral attachments in light-water cooled power plants.

Subsection IWL provides the rules and requirements for preservice examination and inservice inspection of the reinforced concrete and the post-tensioning systems of Class CC 5

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~ f components.Section XI recoros are needed to document the plans for and results of inservice inspection and inservice test programs. The records developed are generally not collected by the NRC. but are retained by the licensee to be made available to the NRC in the event j

of an NRC audit.

Section XI. Division I, requirements for inservice inspection records and reports are provided in IWA-6000

" Records and Remrts."

The following records and reports identified in IWA 6000 must a maintained for the component or system. These records and reports are:

Index to record file Preservice and inservice inspection plans Preservice and inservice inspection reports Repair records and reports Replacement records and reports Nondestructive examination procedures Nondestructive examination records IWA 6310 states that the records and reports shall be filed and maintained in a manner which will allow access by the Inspector. The Owner also shall provide suitable protection from deterioration and damage for all records and reports, in accordance with the Owner's Quality Assurance Program, for the service lifetime of the component or system.

Lifetime retention of the above records is necessary to ensure adequate historical information on the design, examination, and testing of components and systems to provide a basis for evaluating degradation of these components and systems at any time during their service lifetime.

IWA-6240 requires that ISI Summary Reports be submitted to the regulatory and enforcement authorities having jurisdiction at the plant site. The requirements of IWA-6240 and IWA-6310 were incorporated into previous changes to S 50.55a. and this proposed rulemaking action, therefore, does not impose additional recordkeeping or reporting burden.

2.

Acency Use of Information The records are generally historical in nature and provide data on which future activities can be based.

The practical utility of the information collection for NRC is that appropriate records are available for auditing by NRC personnel to determine if ASME. Code provisions for construction, inservice inspection, and inservice testing are being properly implemented in accordance with 10 CFR 50.55a, or whether specific enforcement actions are necessary.

3.

Reduction of Burden Throuah Information TechnoloQv The information being collected represents the documentation for the s

various plant specific construction, inservice inspection, and 6

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l inservice testing programs. The NRC has no objection to the use of new information technologies and generally encourages their use.

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4.

Effort to Identify Duplication 4

1 ASME Code requirements are incorporated by reference into the NRC regulations to avoid the need for writing equivalent NRC requirements.

This amendment will not duplicate the information collection requirements contained in any other regulatory requirement.

t 5.

Effort to Use Similar Information I

e The NRC is using the information collection requirements specified in the ASME Code in lieu of developing its own equivalent requirements.

j 6.

Effort to Reduce Small Business Burden This amendment to 10' CFR 50.55a affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act in the Small Business Size Standards issued by the Small Business Administration at 13 CFR Part 121.

7.

Consecuences of less Frecuent Collection The information is generally not collected, but is retained by the licensee to be made available to the NRC in the event of an NRC audit.

8.

Circumstances Which Justify Variation from OMB Guidelines The record retention periods for information requested is frequently for the service lifetime of the applicable component.

Such lifetime retention of records is necessary to ensure adequate historical information on the cumination and testing of components to provide a basis for evaluating degradation of these components and systems at any time during their service lifetime.

9.

Consultations Outside the NRC The NRC staff prepared the proposed rule in consultation with personnel from the Idaho National Engineering Laboratory (Idaho Falls. ID) and I

ISI Containment Specialists from General Dynamics / Electric Boat Division. Nuclear Engineering (Groton. CT) and Multiple Dynamics Corporation (Southfield. HI). The proposed rule will be published in the Federal Reaister for comment.

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10.

Confidentiality of Information NRC provides no pledge of confidentiality for this collection of information.

11.

Justification for Sensitive Questions No sensitive questions are involved.

12.

Estimated Annualized Cost to the Federal Government NRC inspection personnel who audit plant quality assurance records would include in their audit verification that the above records are being properly prepared and maintained. The time associated with NRC inspectors verifying these records would be small when the activity is performed as part of a normal quality assurance audit.

13.

Estimate of Burden a.

Number and Tvoe of Respondents The recordkeeping requirements incurred by 10 CFR 50.55a through incorporation by reference of the ASME Code would apply to the 117 nuclear power plants presently under construction or in operation.

b.

Estimated Hours Recuired to Respond to the Collection The incorporation by reference of Subsections IWE and IWL into 10 CFR 50.55a would require each licensee to develop an initial inservice inspection (ISI) plan, implement that ISI plan and then develop and implement 10 year updates to that ISI plan.

The development of the initial ISI plan is estimated to average 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per year per plant over a 4 year period. Development of the initial inservice inspection plan is a one-time effort. Total annual burden for the development of the ISI plan is estimated at 117.000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> times 117 plants) each year for 4 years.

It is estimated that implementation of the ISI plan would require 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year for each plant performing ISI of the containment.

Assuming that on the average 12 plants per year would be performing ISI of the containment, this would result in an industry burden of 9.600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> per year. The reporting burden of Sections 50.55a(b)(2)(ix)(B).

(C). (D). and (E). which are modifications to Subsection IWL. that must also be reported in the ISI summary report are estimated to average 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per plant per year for recordkeeping and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per plant per year for reporting. Therefore, the total burden estimated for the ISI plan would be 9.840 hours0.00972 days <br />0.233 hours <br />0.00139 weeks <br />3.1962e-4 months <br /> per year or 820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br /> per plant.

Every 10 years each licensee must update the ISI plan. Update of the plan is estimated to average 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> per plant.

Assuming that 12 8

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t plants per year would be updating their containment ISI plans, this would result in an industry burden of 2.160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> per year.

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ITEM-....

i ANNUAL NUMBER OF PLANTS PER TOTAL ANNUAL RECORDKEEPING.

YEAR HOURSf HOURS / PLANT-i

)

Development 1000 117 117.000 l

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Periodic ISI 820 12 9.840 l

Update 180 12 2.160 4

TOTAL 1000 12.000 1

t 2

j Development has not been included into the total number of hours as this activity i

j will be completed after the first four years.

The total reflects the continuing annual burden after development is complete.

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c.

Estimated Cost Recuired to Respond to the Collection Based upon the hours saecified in Item b. above, and a rate of

$132/hr., it is estimatec that the cost to the industry for responding to the information collection required by the proposed amendment to S 50.55a is a total of $17.012.000 (117.000 + 2.160 +

9.720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> X l

$132/ hour) for the first four years, and $1,584.000 (12.000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> X 4

]

$132/ hour) thereafter.

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14.

Reasons for Chance in Burden The change in burden results from incorporation by reference through this proposed amendment into the NRC regulations of the two new ASME Code Section XI Subsections. IWE and IWL. which contain recordkeeping requirements.

15.

Publication for Statistical Use This information will not be published for statistical use.

B.

COLLECTION OF INFORMATION EMPLOYING STATISTICAL METHODS l

Statistical methods are not used in the collection of the required information.

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[7590-01]

NUCLEAR REGULATORY COMMISSION 10 CFR PART 50 RIN 3150-AC93 i

Codes and Standards for Nuclear Power Plants:

Subsection IWE and Subsection IWL i

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Proposed rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) proposes to amend its reg-l ulations to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsection IWE. " Requirements for Class MC and Metallic Liners of Class CC l

Components of Light-Water Cooled Power Plants."

and Subsection IWL.

" Requirements for Class CC Concrete Components of Light-Water Cooled Power f

Plants." of Section XI. Division 1. of the American Society of Mechanical j

Engineers Boiler and Pressure Vessel Code (ASME Code) with specified modifications and a limitation.

Subsection IWE of the ASME Code provides i

rules ' for inservice inspection, repair, and replacement of. Class MC pressure retaining components and their integral attachments and of metallic shell and penetration liners of Class CC pressure retaining components and thei r integral attachments in light-water cooled power plants.

Subsection IWL ~ of '

the ASME Code provides rules for inservice inspection and repair of the reinforced concrete and the post-tensioning systems of Class CC components.

Licensees would be required to incorporate Subsection IWE and Subsection IWL

j

~

~

l j

into their routine inservice inspection (ISI) program.

Licensees would also i

be required to expedite implementation of the containment examinations and complete the expedited examination in accordance with Subsection IWE and l

Subsection IWL within 5 years of the effective date of this rule.

Provisions have been proposed that would prevent unnecessary duplication of examinations l

between the expedited examination and the routine 120-month ISI examinations.

I Subsection IWE and Subsection IWL have not been previously incorporated by reference into the NRC regulations.

This proposed amendment would specify requirements to assure that the critical areas of containments are routinely inspected to detect defects that could compromise a containment's pressure-retaining integrity.

l DATES:

Comment period expires (75 davs after oublication in the Federal j

Reaister).

Comments received after this date will be considered if it is i

practical to do so. but assurance of consideration cannot be given except as t

to comments received on or before this date.

}

1 i

l ADDRESSES:

Written comments or suggestions may be submitted to the Secretary of the Commission. U.S. Nuclear Regulatory Commission. Washington. DC 20555.

Attention:

Docketing and Service Branch.

Deliver comments to:

11555 1

t l

Rockville Pike. Rockville. MD between 7:45 am and 4:15 pm Federal workdays.

l l

l Copies of the regulatory analysis, the environmental assessment and finding of l

no significant impact, the supporting statement submitted to the Office of j

t l

Management and Budget, and comments received may be examined in the 1

Commission's Public Document Room at 2120 L Street NW.

(Lower Level).

l Washington. DC.

2 l

_-m

-.n y,

,...y-y..v,-,--

y...,. -,-.,.. -

i i

j 1

1 j

FOR FURTHER INFORMATION CONTACT:

Mr. W. E. Norris. Division of Engineering.

i Office of Nuclear Regulatory Research. U.S. Nuclear Regulatory Commission, i

l Washington DC 20555. telephone (301) 492-3805, or Mr. H. L. Graves. Division

)

of Engineering. Office of Nuclear Regulatory Research. U-.S. Nuclear Regulatory Commission. Washington. DC 20555. telephone (301) 492-3813.

1 i

i l

SUPPLEMENTARY INFORMATION:

j

Background

3 j

The NRC is taking the proposed action _ for the purpose of ensuring that l

containments continue to maintain or exceed minimum accepted design wall j

i

]

thicknesses and prestressing forces as provided for in industry standards used to design containments (e.g..Section III and Section VIII of the ASME Code.

and the American Concrete Institute Standard ACI-318), as reflected in license conditions. technical specifications, and licensee commitments (e.g..

1 the Final Safety Analysis Report).

The NRC also believes enhanced ISI j

examinations are needed and are justified to supplement existing requirements i

j specified in General Design Criterion (GDC) 16. and GDC 53. Appendix A to j

10 CFR Part 50. and Appendix J to 10 CFR Part 50.

Appendix J requires a general visual inspection of the containment but does not provide specific j

guidance on how to perform the nece: ary containment examinations.

This has j

resulted in a large variation with regard to the performance and the 4

{

effectiveness of containment inspections.

In view of the increasing rate of occurrences of degradation in containments and variability of present 4

t

)

l j

containment examinations, the NRC has determined that it is necessary to 1

j include more detailed requirements for the periodic examination of containment structures in the regulations to assure that the critical areas of containments are periodically inspected to detect defects that could compromise the containment *s pressure-retaining and leak-tight capability.

Recent changes and additions to the ASME Code include provisions to address the concerns outlined above.

The NRC proposes to make these provisions mandatory by amending 10 CFR 50.55a to incorporate by reference these additional portions of the ASME Code (Subsection IWE and Subsection IWL).

Subsection IWE and Subsection IWL have not been previously incorporated by

]

reference into the NRC's regulations.

The rate of occurrence of corrosion and degradation of containments has been increasing at operating nuclear power plants.

Since 1986 twenty-one (21) instances of corrosion in steel containments have been reported.

In two i

cases, thickness measurements of the walls revealed areas where the wall thickness was at or below the minimum design thickness.

Since the early 1

l 1970s. thirty-one (31) incidents of containment degradation related to post-I tensioning systems of concrete containments have been reported.

Four recent additional incidents which involved grease leakage from tendons have been j

investigated.

In addition to grease leakage. these incidents showed signs of leaching of the concrete.

Over one-third of the operating containments have experienced corrosion or other degradation.

Almost one-half of these occurrences were found by the NRC through its inspections or audits of plant structures, or by licensees l

because they were alerted to a degraded condition at another site.

Examples of degradation not found by licensees, but initially detected at plants 4

1 f

through NRC inspections include: steel containment shell corrosion in the j

drywell sand cushion region (wall thickness reduced to below minimum design thickness); steel containment shell torus corrosion (wall thickness at or near minimum design thickness): grease leakage from the tendons of prestressed j

concrete containments. and water seepage, as well as concrete c racking in i

concrete containments.

i i

There are several GDC criteria and ASME Code sections which establish j

minimum requirements for the design. fabrication. construction. testing, and '

performance of structures. systems. and components important to safety in water-cooled nuclear power plants.

Criterion 16.

" Containment design."

[

requires the provision of reactor containment and associated systems to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity into the environment and to ensure that the containment r

design conditions important to safety are not exceeded for as long as required for postulated accident conditions.

Section III and Section VIII of the ASME Code. and the American Concrete Institute provide design specifications for minimum wall thicknesses and prestressing forces of containments, and these l

l are reflected in license conditions. technical specifications, and licensee l

commitments for the operating plants.

I i

l Criterion 53.

" Provisions for containment testing and inspection."

4 requires that the reactor containment design permit: (1) appropriate periodic inspection of all important areas. such as penetrations: (2) an appropriate surveillance program; and (3) periodic testing at containment design pressure of the leak-tightness of penetrations which have resilient seals and expansion bellows.

Appendix J. " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." of 10 CFR Part -50 contains specific rules for leakage i

1

~

testing of containments.

Paragraph V.

A.

of Appendix J requires that a

{

general inspection of the accessible interior and exterior surfaces of the containment structures and components be performed prior to any Type A test to uncover any evidence of structural deterioration that may affect either the containment structural integrity or leak-tightness.

(Type A test means tests

)

intended to measure the primary reactor containment overall integrated leakage j

rate: (1) after the containment has been completed and is ready for operation.

1 i

and (2) at periodic intervals thereafter).

None of these existing l

requirements. however. provide specific guidance on how to perform the i

necessary containment examinations.

This lack' of guidance has resulted in a l

large variation in licensee containment examination programs, such that there have been cases of noncompliance with GDC 16.

Based on the results of inspections and audits. as well as plant operational experiences. it is clear

)

that many licensee containment examination programs have not detected degradation that could ultimately result in a compromise to the pressure-t retaining capability.

Some containment structures have also been found to have undergone a significant level of degradation that was not detected by J

i these programs.

The NRC believes that more specific ISI requirements. which expand upon existing requirements for the examination of containment structures in s

accordance with GDC 53 and Appendix J. are needed and are justified for the purpose of ensuring that containments continue to maintain minimum design wall th;cknesses and prestressing forces as provided for in industry standards used to design containments (e.g.

Section III and Section VIII of the ASME Code.

and the American Concrete Institute Standard ACI-318), as reflected in license conditions. technical specifications. and written licensee commitments (e.g.

the Final Safety Analysis Report).

There exists a serious concern. based on 6

- _. _ _ _ = _ _, _...

i j

i actual operating experience, regarding continued compliance by the operating plants with existing requirements for ensuring containment minimum design wall thicknesses and prestressing forces if the proposed action is not taken.

The l

NRC also believes that the occurrences of corrosion and other degradation discussed above would have been detected by licensees implementing the comprehensive periodic examinations set forth in Subsection IWE and Subsection IWL of the ASME Code proposed for incorporation by reference into 10 CFR 50.55a.

The Nuclear Management and Resources Council- (NUMARC) has developed a i

number of industry reports to address license renewal issues.

Two of them.

one for PWR containments and the other for BWR containments, were developed for the purpose of managing age-related degradation of containments on a generic basis.

The NUMARC plan for containments relies on the examinations contained in. Subsection IWE and Subsection IWL to manage age-related degradation. and this plan assumes that these examinations are "in current and effective use."

In the BWR Containment Industry Report. NUMARC concluded that "On account of these available and established methods and techniques to adequately manage potential degradation due to general corrosion of l

freestanding metal containments, no additional measures need to be developed l

and. as such. general corrosion is not a license renewal concern if the containment minimum wall thickness is maintained and verified." Similarly, in the PWR Containment Industry Report.

NUMARC concluded that potentially significant degradation of concrete surfaces the post-tensioning system. and the liners of concrete containments could be managed effectively if periodically examined in accordance with the requirements contained in l

Subsection IWE and Subsection IWL.

l I

7 I

l l

I I

{

The five modifications, which are contained in one paragraph of the proposed rule, address two concerns of the NRC.

The first concern is-that certain recommendations for tendon examinations that are included in I

Regulatory Guide 1.35. Rev. 3.

are not addressed in Subsection IWL (this involves four of the modifications.

(ix)(A)-(D)).

The ASME Code has considered these four issues and has adopted them in Subsection IWL.

These issues will be published in future addenda.

The second concern is that if there is visible evidence of degradation of the concrete (e.g..

leaching, l

surface cracking) there may also be degradation of inaccessible areas.

This I

fifth modification ((ix)(E)) contains a provision which would require an evaluation of inaccessible areas when visible conditions exist that could result in degradation of these areas.

t i

The limitation specifies the 1992 Edition with 1992 Addenda of Subsection IWE and Subsection IWL as the earliest version of the ASME Code the NRC finds j

acceptable.

This edition and addenda combination incorporates the concept of l

base metal examinations and would provide a comprehensive set of rules for the examination of post-tensioning systems.

As originally published. Subsection l

I l

IWE preservice examination and inservice examination rules focused on the examination of welds.

This weld-based examination philosophy was established in the 1970s as plants were being constructed.

It was based on the premise l

that the welds in pressure vessels and piping were the areas of greatest concern.

As containments have aged. degradation of base metal. rather than welds. has been found to be the issue of concern.

The 1991 Addenda to the 1989 Edition the 1992 Edition and the 1992 Addenda to Section XI. Subsection IWE. all have furthered the incorporation of base metal examinations.

l The proposed rulemaking incorporates a provision for an expedited 8

t

l examination. schedule.

This expedited examination schedule is necessary to l

prevent a delay in tne implementation of Subsection IWE and Subsection IWL-(Table 4 of Enclosure 2 lists each plant and the delay in implementation which j

would be encountered without an expedited implementation schedule).

)

Provisions have been incorporated in the proposed rule so that the expedited examination which would be required 5 years after the effective date of the l

rule and the routine 120-month examinations are not duplicated.

The NRC has reviewed the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL of Section XI of the ASME Code and has found that with the ser-ified modifications these subsections of Section XI address current exper._.

1d provide a sound basis for ensuring the structural integrity of containments.

NRC endorsement of Subsection IWE and Subsection IWL in its regulations would provide a method of improving containment examination practices by incorporating rules into the regulatory process that j

have received industry participation in their development and acceptance by i

the NRC.

i

[

Existing S 50.55a(g). " Inservice inspection muirements." specifies the l

requirements for preservice and inservice examinations for Class 1 (Class 1 l

refers to components of the reactor coolant pressure boundary). Class 2 (Class i

j 2 quality standards are applied to water-and steam-containing pressure l

vessels. heat exchangers (other than turbines and condensers), storage tanks.

piping. pumps. and valves that are part of the reactor coolant pressure boundary (e.g.. systems designed for residual heat remaval and emergency core cooling)). and Class 3 (Class 3 quality standards are applied to radioactive-waste-containing pressure vessels heat exchangers (other than turbines and condensers), storage tanks. piping pumps, and valves (not part of the reactor l

9 l

coolant pressure boundary)) components and their supports.

Neither Subsection IWE (Class MC -- metal containments) nor Subsection IWL (Class CC -

- concrete containments) is presently incorporated by reference into the NRC regulations.

{

Proposed S 50.55a(g)(4) specifies the containment components to which the ASME Code Class MC and Class CC inservice inspection classifications incorporated by reference in this proposed rule would apply.

Proposed S 50.55a (g)(4)(v)(A)

(v)(B). and (v)(C) specify Subsection IWE and Subsection IWL rules for repairs and replacements of metal and concrete i

containments.

This is consistent with the long-standing intent and ongoing application by NRC and licensees to utilize the rules of Section XI when performing repairs and replacements of applicable components and their I

supports.

Proposed S 50.55a(b)(2)(vi) would incorporate a limitation specifying-the 1992 Edition with 1992 Addenda of Subsection IWE and Subsection IUL as ins earliest ASME Code version the NRC finds acceptable. This edition and addenda combination incorporates the concept of base metal examinations and provides a comprehensive set of rules for the examination of post-tensioning systems, Proposed S 50.55a(b)(2)(ix) would specify five modifications that must be implemented when using Subsection IWL.

Four of these issues are identified in Regulatory Guide 1.35. Revision 3.

but are not currently addressed in l

Subsection IWL.

i Proposec S

50.55a(g)(4)(v) requires that licensees incorporate 10 L

. ~. ~

containment examinations as part of their routine 120-month inspection program.

It is recognized that when this rule becomes effective, plants I

within 2 years of the end of the 120-month interval may have _ difficulty f

developing and completing the containment examination program in a timely j

l manner.

Therefore. proposed S 50.55a (b)(2)(x) specifies that licensees with less than 2 years remaining in their present ISI interval may complete the Subsection IWE and the Subsection IWL portions of their ISI update within 1

2 years from the end of the present ISI interval. This is intended tn provide licensees with sufficient time to develop the initial ISI. plan and to i

I facilitate maintenance of one ISI plan instead of two separate plans (i.e. the current Section XI ISI plan. and the Subsection IWE and Subsection IWL plan).

In order to further reduce the burden on licensees and NRC staff. the Subsection IWE and Subsection IWL portions of the ISI plan will not have to be submitted to the NRC for approval.

Licensees may simply retain their initial Subsection IWE and Subsection IWL plans at the site for audit.

f i

l Proposed S 50.55a(g)(6)(ii)(B)(1) would require that licensees conduct the first containment examinations in accordance with Subsection IWE and i

Subsection IWL (1992 Edition with the 1992 Addenda), modified by proposed f

S 50.55a(b)(2)(ix) within 5 years of the effective date of the final rule.

' This expedited examination schedule is necessary to prevent possible delays in the implementation of Subsection IWE by as much as 20 years and Subsection IWL by as much as 15 years.

Subsection IWE. Table IWE-2500-1, permits the 5

deferral of most of the required examinations until the end of the 10-year inspection interval.

Adding the ten years that could pass before some utilities are required to update their ISI plans, a period of 20 years could pass before the first examinations would take place.

Subsection IWL is based on a 5-year inspection interval.

Adding the possible 10 years before update 11 i

m...._..-,. ~,, - - -, - -. -.. -. - - '

of existing ISI plans. a period of 15 years could pass before the examinations were performed by plants that have not voluntarily adopted the provisions of Regulatory Guide 1.35. Rev. 3.

Expediting implementation of the containment examinations is considered necessary because of the problems that have been identified at various plants. the need to establish expeditiously a baseline for each facility. and the need to identify.any existing degradation.

Proposed paragraphs (g)(6)(ii)(B)(2) and (g)(6)(ii)(B)(3) would each provide a mechanism for licensees to satisfy the requirements of the routine containment examinations and the expedited examination without duplication.

Paragraph (g)(6)(ii)(B)(2) would permit licensees to avoid duplicating examinations required by both the periodic routine and expedited examination programs.

This provision is intended to be useful to those licensees that would be required to implement the expedited examination during the first periodic interval that routine containment examinations are required.

Paragraph (g)(6)(ii)(B)(3) would allow licensees to use a recently performed examination of the post-tensioning system to satisfy the requirements for the expedited examination of the containment post-tensioning system.

This situation would occur for licensees who perform an examination of the post-tensioning system using Regulatory Guide 1.35 between the effective date of this rule and the beginning of the expedited examination.

Submission of Comments ;a Electronic Format The comment evaluation process will be improved if each comment is identified with document title, section heading, and paragraph number 12

addressed.

In addition to the original paper copy. submitters are encouraged to provide a copy of their letter in an electronic format on IBM PC compatible 3.5-or 5.25-inch diskettes.

Data files should be provided as Wordperfect documents.

ASCII text is also acceptable or, if formatted text is required.

data files should be provided in IBM Revisable-Form Text / Document Content Architecture (RFT/DCA). format.

The format and version should be identified on the diskette's external label.

Finding of No Significant Environmental Impact The Commission has determined under the National Environmental Policy Act of 1969 as amended. and the Commission's regulations in Subpart A of 10 CFR Part 51 that this rule. if adopted, would not be a major Federal action significantly affecting the quality of the hum?.n environment and therefore an environmental impact statement is not required.

This proposed rule is one part of a regulatory framework directed to ensuring containment integrity. Therefore, in the general sense, the proposed rule would have a positive impact on the environment. The proposed rule would incorporate by reference in the NRC regulations requirements contained in the ASME Code for the inservice inspection of the containments of nuclear power plants.

Actions required of applicants and licensees to implement the proposed rule are of a routine nature that should not increase the potential for a negative environmental impact.

The environmental assessment and finding of no significant impact on 13

which this determination is based are available for inspection at the NRC Public Document Room. 2120 L Street NW. (Lower Level). Washington. DC. Single copies of the environmental assessment and the finding of no significant impact are available from Mr. W. E. Norris. Division of Engineering. Office of Nuclear Regulatory Research. U.S. Nuclear Regulatory Commission. Washington.

DC 20555.

telephone (301)492-3805.

or Mr.

H.

L.

Graves.

Division of Engineering. Office of Nuclear Regulatory Research. U.S. ' Nuclear Regulatory Commission. Washington. DC 20555. telephone (301)492-3813.

Paperwork Reduction Act Statement This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq).

This rule has been submitted to the Office of Management and Budget for review and approval of the peperwork requirements.

The public reporting burden for this collection of information is estimated to average 4.000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per response for development of an initial inservice inspection plan and 10.000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per response for the update of the plan and periodic examinations, including the time for reviewing instructions.

searching existing data sources, gathering and maintaining the data needed.

and completing and reviewing the collection of information.

Send comments regarding this burden estimate or any other aspect of this collection of l

information.

including suggestions for reducing this

burden, to the 1

Information and Records Management Branch (MNBB-7714). U.S. Nuclear Regulatory Commission. Washington. DC 20555-0001. and to the Desk Officer. Office of 14

l 4

Information and Regulatory Affairs.

NE0B-3019.

(3150-0011).

Office of Management and Budget. Washington. DC 20503.

4 Documented Evaluation The Commission has prepared a draft summary of documented evaluation on this proposed regulation.

The draft evaluation is available for inspection in the NRC Public Document Room. 2120 L Street NW. (Lower Level). Washington DC.

Single copies of the analysis may be obtained from Mr. W. E. Norris. Division of Engineering. Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission. Washington DC 20555, telephone (301)492-3805, or from Mr. H. L.

Graves. Division of Engineering. Office of Nuclear Regulatory Research. U.S.

Nuclear Regulatory Commission. Washington DC 20555. telephone (301)492-3813.

~

The Comission requests public comment on the draft summary of j

1 documented evaluation.

Comments on the draft evaluation may be submitted to j

the NRC as indicated under the ADDRESSES heading.

i Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980. 5. U.S.C.

d 605(b).

the Commission hereby certifies that this rule will

not, if j

promulgated, have a significant economic impact on a substantial number of small entities.

This proposed rule affects only the operation of nuclear power plants.

The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory j

l 15 a

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 2

l companies are dominant in their service areas, this proposed rule does not i

fall within the purview of the Act.

i Backfit Statement i

The NRC is taking the proposed action for the purpose of ensuring that containment structures continue to maintain or exceed minimum accepted design

]

wall thicknesses and prestressing forces as provided for in industry d andards used to design containment structures, as reflected in license conditions, technical specifications, and licensee commitments Therefore, under 10 CFR 50.109(a)(4)(i) a backfit analysis need not be prepared for this rule.

A summary of the documented evaluation required by S 50.109(a)(4) to support this conclusion is set forth below.

l i

GDC 16. " Containment design." requires the provision of reactor con-tainment and associated systems to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity into the environment and to ensure that the containment design conditions important to safety are not exceeded for as long as required for postulated accident conditions.

Criterion 53.

" Provisions for containment testing and inspection."

requires that the reactor containment design permit: (1) appropriate periodic inspection of all important areas. such as penetrations: (2) an appropriate surveillance program; and (3) periodic testing at containment design pressure 16

of the leak-tightness of penetrations which have resilient seals and expansion bellows.

Appendix J. " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors " of 10 CFR Part 50 contains specific rules for leakage testing of containments.

Paragraph V.

A.

of Appendix J requires that a general inspection of the accessible interior and exterior surfaces of the containment structures and components be performed prior to any Type A test to uncover any evidence of structural deterioration that may affect either the containment structural integrity or leak-tightness (Type A test means tests intended to measure the primary reactor containment overall integrated leakage rate: (1) after the containment has been completed and is ready for operation.

and (2) at periodic intervals thereafter).

None of these existing requirements. however. prsvide specific guidance on how to perform the necessary containment examinations.

This lack of guidance has resulted in a large variation in licensee containment examination programs, such that there have been cases of noncompliance with GDC 16.

Based on the results of inspections and audits. and plant operational experiences. it is clear that many licensee containment examination programs have not detected degradation that could result in a compromise of pressure-retaining capability.

The location and extent of corrosion or degradation in a containment can be critical to the containment's behavior during an accident.

The metal containment structure of operating nuclear power plants were designed in accordance with either Section III. Subsection NE. " Class MC Components." or Section VIII, of the ASME Code.

These subsections contain provisions for the design and construction of metal containment structures, including methods for determining the minimum required wall thicknesses.

The minimum wall thickness is determined so that the metal containment structure 17

l i

will continue to maintain its structural integrity under the various stressors a

and degradation mechanisms which act on it.

The American Concrete Institute Standard ACI-318 contains provisions for designing and constructing the post-tensioning systems of concrete containment y

structures, including methods for determining the prestressing forces.

The post-tensioning system is designed so that the concrete containment structure will continue to maintain its structural integrity under the various stressors and degradation mechanisms which act on it.

l These requirements for minimum design wall thicknesses and prestressing forces as provided in these industry standards used to design containment structures are reflected in license conditions technical specifications, and licensee commitments (e.g.. the Final Safety Analysis Report).

3 j

The rate of occurrence of corrosion and degradation of containment 1

structures has been increasing at operating nuclear power plants.

Over.one-third of operating containment structures have experienced corrosion or other 1

3 degradation.

Almost one-half of the occurrences were first identified by the NRC through its inspections or structural audits, or by licensees because they were alerted to a degraded condition at another site.

Examples of degradation I

not found by licensees, but initially detected at plants through NRC inspections include 1) corrosion of steel containment shells in the drywell sand cushion region, resulting in wall thickness reduced to below the minimum I

design thickness: 2) corrosion of the torus of the steel containment shell a

(wall thickness at or near minimum design thickness): 3) grease leakage from l

the tendons of prestressed concrete containments: and 4) water seepage. as 4

18 j

i

.. - ~

well as concrete cracking in concrete containments.

l The NRC believes that more specific ISI requirements. that expand upon existing requirements for the examination of containment structures in accordance with GDC 53, and Appendix J are needed and are justified to ensure that containment structures continue to maintain or exceed minimum accepted design wall thicknesses and prestressing forces as reflected in license conditions. technical specifications, or licensee commitments.

Based on actual operating experience, a serious concern exists regarding continued compliance by the operating plants with existing requirements for ens,uring containment minimum design wall thicknesses and prestressing forces if the proposed action is not taken.

The NRC also believes that the occurrences of i

corrosion and other degradation discussed above would have been. detected by licensees when conducting the comprehensive periodic examinations set forth in Subsection IWE and Subsection IWL of the ASME Code.

as proposed for incorporation by reference into 10 CFR 50.55a.

1 Recent changes and additions to the ASME Code include provisions to address the concerns outlined above: and the staff proposes to make these provisions mandatory by amending 10 CFR 50.55a to incorporate by reference these additional portions of the ASME Code (Subsection IWE and Subsection IWL).

The Commission concludes that this proposed backfit is necessary to ensure compliance with GDCs 16 and 53. Appendix J.

minimum design wall thicknesses in metal containments, and the prestressing forces of concrete containments, which are applicable to all licensees through license conditions. technical specifications, and licensee commitments.

1 19 4

e -- -

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.,,,rn,

.,.w,-,.,e a-,m emw,.,wr,m-,-

I I

J i

List of Subjects in 10 CFR Part 50 t

4 Antitrust. Classified information. Criminal Penalties. Fire protection.

Incorporation by reference. Intergovernmental relations. Nuclear power plants and reactors. Radiation protection. Reactor siting criteria. Reporting - and l

recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the l

Atomic Energy Act of 1954. as amended, the Energy Reorganization Act of 1974.

as amended, and 5 U.S.C. 533. the NRC is. proposing to adopt the following s

amendments to 10 CFR Part 50.

i PART 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES l

1 i

i 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.

i 1

936. 937. 938. 948. 953. 954. 955. 956. as amended, sec. 234. 83 Stat. 1244, as amended (42 U.S.C. 2132, 2133. 2134. 2135. 2201, 2232. 2233. 2236. 2239.

2282); secs. 201. as amended. 202. 206. 88 Stat. 1242, as amended. 1244.

246 (42 U.S.C. 5841. 5842. 5846).

Section 50.7 also issued under Pub. L.95-601'. sec.10. 92 Stat. 2951 (42 U.S.C. 5851).

Section 50.10 also issued under secs. 101. 185. 68 Stat. 936, 955 as amended (42 U.S.C. 2131. 2235): sec. 102. Pub. L.91-190. 83 Stat. 853 20

_ _., _, _.. _,.. _...,, _ _. -.. _,.. _....... _ _....,.. _,,,. ~,. _ _, _ _..., _

(42 U.S.C. 4332).

Sections 50.13, 50.54(dd) and 50.103 also issued under sec.

I

}

108. 68 Stat. 939. as amended (42 U.S.C. 2138)

Sections 50.23, 50.35, 50.55.

i and 50.56 also issued under sec.185. 68 Stat. 955 (42 U.S.C. 2235).

Sections 50.33a. 50.55a and Appendix 0 also issued under sec. 102. Pub. L.91-190. 83 Stat. 853 (42 U.S.C. 4332).

Sections 50.34 and 50.54 also issued under sec.

[

204. 88 Stat.1245 (42 U.S.C. 5844).

Sections 50.58. 50.91 and 50.92 also issued under Pub. L.97-415. 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 l

also issued under sec. 122. 68 Stat. 939 (42 U.S.C. 2152).

Sections 50.80-50.81 also issued under sec. 184. 68 Stat. 954. as amended (42 U.S.C. 2234).

i Appendix F also issued under sec. 187. 68 Stat. 955 (42 U.S.C. 2237).

i 2.

Section 50.55a is amended by adding paragraphs (b)(2)(vi).

l (b)(2)(ix). (b)(2)(x). (g)(4)(v). and (g)(6)(ii)(B). and revising the introductory text of paragraph (g)(4) to read as follows:

6 50.55a Codes and standards.

e (b) *

(2) *

(vi) Effective edition and addenda of Subsection IWE and Subsection IWL.

Section XI.

When using Subsecticr IWE and Subsection IWL. the 1992 Edition with the 1992 Addenda is the only acceptable Edition and Addenda.

1 21

4 i

(ix) Examination of concrete containments.

i (A) All grease caps that are accessible must be visually examined to detect grease leakage or grease cap deformations.

Grease caps must be removed for this examination when there is evidence of grease cap deformation that indicates deterioration of anchorage hardware.

(B) An Engineering Evaluation Report must be prepared as prescribed in IWL-3300(a). (b), (g), and (d) when evaluation of consecutive surveillances of f

prestressing forces for the same tendon or tendons in a group indicates a trend of prestress loss such that the tendon force (s) would be less than the minimum design prestress requirements before the next inspection interval.

i (C) When the elongation corresponding to a specific load (adjusted for effective wires or strands) during retensioning of tendons differs by more than 10 percent from that recorded during the last measurement, an evaluation must be performed to determine whether the difference is related to wire failures or slip of wires in anchorages, A difference of more than 10 percent must be identified in the ISI Summary Report.

(D) The licensee shall identify the following conditions. if they occur, in the ISI Summary Report:

1 (1) The sampled sheathing filler grease contains chemically combined 22

water exceeding 10 percent by weight or the presence of free water:

i i

I (2) The absolute difference between the amount removed and the amount replaced may not exceed 10 percent of the tendon net duct volume.

(3) Grease leakage is detected during general visual examination of the containment surface, l

1 I

J i

l (E) The licensee shall evaluate the acceptability of inaccessible areas I

l when conditions exist in accessible areas that could indicate the presence of j

or result in degradation to such inaccessible areas.

For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report:

l (1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation:

l (2) An evaluation of each area. and the result of the evaluation, and; l

(3) A description of necessary corrective actions.

(x) Subsection IWE and Subsection IWL inservice insoection olans.

Licensees that have less than 2 years remaining in their present 120-month inservice inspection interval on (insert effective date of the final rule) may defer completion of the Subsection IWE and Subsection IWL portions of the 23 l

i I

i l

1 inspection plan for the next 120-month inspection interval for up to 2 years from the end of the present interval.

(g)

(4) Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1. Class 2. and Class 3 must meet the requirements. except design and access provisions and preservice examination requirements, set forth in Section XI of editions of the ASME wiler and Pressure Vessel Code and Addenda that become effective subsequent to editions specified in i

)

paragraphs (g)(2) and (g)(3) of this section and are incorporated by reference in paragraph (b) of this section, to the extent practical within the limitations of design.

geometry and materials of construction of the components.

Components which are classified as Class MC pressure retaining components and their integral attachments, and components which are classified as Class CC pressure retaining components and their integral attachments must meet the requirements. except design and access provisions and preservice i

examination requirements set forth in Section XI of the ASME Boiler and Pressure Vessel Code and Addenda that are incorporated by reference in

)

i paragraph (b). subject to the limitation listed in paragraph (b)(2)(vi) and the modifications listed in paragraphs (b)(2)(ix) and (b)(2)(x) of this section, to the extent practical within the limitations of design. geometry and materials of construction of the components.

24

4 L

i (v)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued after January 1.1956:

(A) Metal containment pressure retaini_ng components and their integral 1

attachments must meet the inservice inspection, repair, and replacement requirements applicable to components which are classified as ASME Code Class MC:

(B) Metallic shell and penetration liners which are pressure retaining components and their integral attachments in concrete containments must meet the inservice inspection. repair and replacement requirements applicable to components which are classified as ASME Code Class CC; and I

(C) Concrete containment pressure retaining components and their integral attachments. and the post-tensioning systems of concrete containments must meet the inservice inspection and repair requirements applicable to components which are classified as ASME Code Class CC.

8 (6) 25 I

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i (ii) * *

• l (B) Exoedited examination of containment.

l (1) Licensees of all operating nuclear power plants shall implement the i

examinations specified for the first inspection interval in Subsection IWE and Subsection IWL of the 1992 Edition with the 1992 Addenda in conjunction with the modifications specified in S 50.55a (b)(2)(ix) by (a date will be inserted that is 5 vears later than the effective date of the final rule).

l i

(2) The expedited examination may be used to satisfy the requirements of routinely scheduled examinations of Subsection IWE subject to IWA-2430(c) when the expedited examination occurs during the first containment inspection interval.

l l

(3) The requirement for the expedited examination of the containment post-tensioning system may be satisfied by written commitments that are in I

place before (insert the effective date of the final rule) for examinations of j

the post-tensioning system.

l l

Dated at this day of 19_.

For the Nuclear Regulatory Commission.

26

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Samuel J. Chilk.

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Secretary of the Commission.

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[7590-01]

NUCLEAR REGULATORY COMMISSION Documents Containing Reporting or Recordkeeping Requirements; Office of Management and Budget (0MB) Review AGENCY:

Nuclear Regulatory Commission (NRC).

ACTION:

Notice of the OMB review of information collection.

SUMMARY

The Nuclear Regulatory Commission has recently submitted to OMB for review the following proposal for collection of information under the provisions of the Paperwork Reduction Act (44 U.S.C.

Chapter 35).

1.

Type of submission, new, revised, or ex+ension:

Revision l

2.

The title of the information collection:

Codes and l

Standards for Nuclear Power Plants; Subsection IWE and l

Subsection IWL" 3.

The form number if applicable:

Not applicable.

4.

How often is the collection required: The American Society of Mechanical Engineers Boiler and Pressure Vessel Code i

1

O (ASME Code) requires that Subsection IWE and Subsection IWL reports be submitted to the regulatory and enforcement authorities having jurisdiction at the plant site. The inservice inspection reports of Subsection IWL must be submitted every five years.

The inservice inspection reports of Subsection IWE must be submitted every three and a half years. The modified requirements at 50.55(b)(2)(ix) which are reported in the ISI Summary Report would be submitted with the Subsection IWL report. The ASME Code requires that these records and reports be retained for the service lifetime of the component or system.

5.

Who will be required or asked to report:

Nuclear power plant licensees.

6.

An estimate of the number of respondents:

117 7.

An estimate of the number of hours annually needed to complete the requirement or request:

117,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> over the first four years (1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per plant) for development of the inservice inspection plan, plus 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (1,180 per plant) for the periodic update of the plan.

8.

An indication of whether Section 3504(h), Pub. L.96-511 applies:

Applicable t

2

i f

9.

Abstract:

The Nuclear Regulatory Commission (NRC) proposes j

]

to amend its regulations to incorporate by reference the.

1992 Edition with the 1992 Addenda of Subsection IWE, f

t

" Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants," and Subsection IWL, " Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants," of Section i

XI, Division 1, of the American Society of Mechanical

~

Engineers Boiler and Pressure Vessel Code (ASME Code) with specified modifications and a limitation. Subsection IWE of i

the ASME Code provides rules for inservice inspection,-

repair, and replacement of Class MC pressure retaining components and their integral attachments and of metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments in light-water cooled power plants.

Subsection IWL of the ASME Code provides rul"s for inservice inspection and repair of the reinforced cc rete and the post-tensioning systems of Class CC components.

Provisions have been proposed that would prevent unnecessary duplication of examinations between the expedited examination and the routine 120-month ISI examinations. Subsection IWE and Subsection IWL have not been previously incorporated by reference into the NRC regulations. This proposed amendment would specify i

requirements to assure that the critical areas of containments are routinely inspected to detect defects that 3

l i

J could compromise a containment's pressure-retaining integrity.

Copies of the submittal may be inspected or obtained for a fee from the NRC l

Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

l I

l Comments and questions can be directed by mail to OMB reviewer:

I Tim Hunt Office of Information and Regulatory Affairs (3150-0011) l NE08-3019 Office of Management and Budget l

Washington, DC 20503 l

Comments may also be communicated by telephone at (202) 395-3084.

l The NRC Clearance Officer is Brenda Jo. Shelton, (301) 492-8132.

Dated at Bethesda, Maryland, this day of w b v, 1993.

l For the Nuclear Regulatory Commission.

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1 8

/ Gerald F. Crhnford, Designped Senio l

Official for Information Resources Management 1

4