ML20206F072

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Technical Evaluation Rept on Third 10-Year Interval ISI Program Plan,Entergy Operations,Inc,Arkansas Nuclear One, Unit 1
ML20206F072
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 03/31/1999
From: Mary Anderson, Beth Brown, Charles Brown
IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:
NRC (Affiliation Not Assigned)
Shared Package
ML20206F067 List:
References
INEEL-EXT-99, INEEL-EXT-99-00, INEEL-EXT-99-00049, INEEL-EXT-99-49, NUDOCS 9905050328
Download: ML20206F072 (26)
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INEEL/ EXT 99 00009

. Technical Evaluation Report on the Third 10-Year interval inservice inspection Program Plan:

i Entergy Operations, Inc.,

l Arkansas Nuclear One, Unit 1, i Docket Number 50-313 l

M. T. Anderson, B. W. Brown, C. T. Brown, S. G. Galbraith, A. M. Porter l

A Published March 1999 I

idaho National Engineering and Environmental Laboratory l Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 l

l Prepared for the  !

Division of Engineering i Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 FIN No. J2229 (Task Order A25) l

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r 1 ABSTRACT

' His report presents the results of the evaluntion of the Third Ten-Yearinterm! Inservice inspection l

Plan / Program For Arkansas Nuclear One, Unit 1, submitted June 25,1997, and December 9,1998, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. The Third Ten.

YearInternt inservice inspection Plan / Program For Arkansas Nuclear One, Unit 1 is evaluated in Section 2 of this report. He inservice inspection (ISI) plan / program is evaluated for (a) compliance with

, the appropriate edition / addenda of Section XI, (b) acceptability ofexamination sample, (c) correctness of l the application of system or component exanunation exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. De requests for relief are evaluated in Section 3 of this report.

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SUMMARY

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ne licensee, Entergy Operations, Inc., prepared the Third Ten-YearInterm! Inservice inspection Plan / Program For Arkansas Nuclear One, Unit 1 to meet the requirements of the 1992 Edition, with a '

portion of the 1993 Addenda, of the American Society ofMechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI. De licensee was authorized to use the 1992 Edition /1993 Addenda by an NRC Safety Evaluation Report dated December 12,1996. De third 10-year interval began June 1, 1997, and will end on May 31,2007.

ne information in the Third Ten-Yearinterwr! Inservice inspection Plan! Program For Arkansas Nuclear One, Unit i submitted June d PC' and December 9,1998, was reviewed. De review included l requests for relief from the ASME Code Section XI requirements that the licensee has determined to bc l

l impractical. As a result of this review, a request for additional information was prepared describing the '

l information and/or clarification required from the licensee in order to complete the review. He licensee l provided the requested information in a submittal dated December 9,1998.

l Based on the review of the plan / program, the licensee's response to the Nuclear Rel=*~y i

Commission's RAI, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, multiple discrepancies from regulatory requirements or commitments were id*ntified in the Third Tennear Intern! Inservice Inspection i Plan / Program For Arkansas Nuclear One, Unit 1. nese are described in Section 2.2.2 of this report.

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c CONTENTS

< i AB STRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii S UM M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I

.' 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . . . . . . . . . . . . . 3 2.1 Documents Evaluated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 Compliance with Code Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.2 Acceptability of the Examination Sample . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.3 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7. . . . . . . . . . . . . . . .

3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1.1 Reactor Pressure Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1.1.1 Request for ReliefNo.97-004, Use of Code Case N-521, Alternative Rules For Deferral ofInspections of Nonle-to-Vessel Welds, Inside Radius Sections, and Nonle-to. Safe End Welds of a Pressurized Water Reactor PWR Vessel . . . . . . 8 l 3.1.2 Pressu rizer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

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' 3.1.4 Piping Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 1 3.1.5 Pump Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.6 Valve Pressure b*y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.7 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.7.1 Request for Relief No. 97 005, Use of Code Case N 533, Alternative Requirements for VT 2 Visual Examination of Clus 1 Insulated Pressure-Retaining Bolted Connections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.2 Class 2 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.2.1 Pressu re Vessels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3 .2.2 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.2.2.1 Request for Relief No.97-003, Examination Category C-F-1, Items C5.11, C5.12, and C5.21, Class 2 Piping Welds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3 .2.3 Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3 . 2.4 Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.2.5 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.3 Class 3 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.3.1 Pressure Vessels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.3.2 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

. 3.3. 3 P umps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3 .3 .4 Val ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.3.5 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4 Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.1 Class 1 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.2 Class 2 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.3 Class 3 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.4 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3 .5 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 Iv hs & mem eil me ; _ m ense.im M y a

p 3.5.1 Ultrasonic Examination Techniques . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.5.2 Exempted Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 g

3.5*3h.....*.........................................................17

4. CONCLUS ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . 13
5. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 9

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. TECHNICAL EVALUATION REPORT ON THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

ENTERGY OPERATIONS, INC.,

ARKANSAS NUCLEAR ONE, UNIT 1, DOCKET NUMBER 50-313 i

. 1. INTRODUCTION Broughout the service life of a water-cooled nuclear power facility, its components (including -

supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class I,2, and 3 are required by 10 CFR 50.55a(g)(4) (Reference 1) to meet the requirements, except the design and access provisions and the preservice examination requirements, of the ASME Code, Section Xl, Rulesfor Inservice Inspection ofNuclear Power Plant Comp nents, (Reference 2) to the extent practical within the limitations of design, geometry, and materials of construction of the components. His section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. He components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitatica and modifications listed therein, and subject to Nuclear Regulatory Commission (NRC) approval. He licensee, Entergy Operations, Inc., has prepared the Third Ten-Year intervalinservice Inspection Plan / Program For Arkansas Nuclear One, Unit I (Reference 3) to meet the requirements of the 1992 Edition of the ASME Code,Section XI. De third 10-year interval began June 1,1997, and will end on May 31,2007. Use of the 1992 Edition of ASME Section XI with the portions of the 1993 Addenda was authorized by the Nuclear Regulatory Commission in a Safety Evaluation Report (SER) dated December 12,1996 (Reference 4).

Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the Code requirements may be used when authorized by the NRC. He licensee must demonstrate either that the proposed alternatives provide an acceptable level of quality and safety, or that Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. Pursuant to 10 CFR 50.55a(g)(5)(iii), if the licensee determines that conformance with certain Code examination requirements is impractical for its facility, the ,

licensee shall submit information to the NRC to support that determination. Pursuant to .

l 10 CFR 50.55a(g)(6)(i), the NRC will evaluate the licensee's determination that Code requirements are i impractical. The NRC may grant relief and may impose attemative requirements that it determines to be l ,

authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the j

requirements were imposed on the facility.  !

ne information in the Third Ten-Year Intervalinservice Inspection Plan / Program For Arkansas Nuclear One. Unit 1 submitted June 25,1997, and December 9,1998, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. His review was performed using the standard review plans of NUREG-0800, Section 5.2.4, j l

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" React:r Coolant Boundary Inservice Inspections and Testing," and Section 6.6, ' Inservice Inspection of Class 2 and 3 Components" (Reference 5).

In a letter dated December 2,1997 (Reference 6), the NRC requested additional information that was necessary to complete the review of the inservice inspection (ISI) plan / program. De requested information was provided by the licensee via a [[letter::1CAN129801, Forwards Response to NRC 971202 RAI Re ANO-1 Third 10-yr ISI Program.Rev 1 to Third 10-Yr Interval ISI Program for ANO-1, Encl|letter dated December 9,1998]] (Reference 7). In this response, Entergy Operations, Inc. provided the requested information, including the Third Ten-Year intervalinservice inspection Program For Arkansas Nuclear One, Unit 1.

ne Third Ten-YearIntervalinservice inspection Plan / Program For Arkansas Nuclear One, Unit 1 is evaluated in Section 2 of this report. He ISI program plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous reviews. De requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI,1992 Edition. Inservice test programs for snubbers and for pumps and valves are being evaluated in other reports.

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2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN

%c evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertment to ISI activities. His section describes the submittals reviewed and the results of the avview.

2.1 Documents Evaluated

. Review has been completed on the following information from the licensee:

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Third Ten-Year Interallnservice inspection PlaWProgram For Arkansas Nuclear One, Unit 1 dated June 25,1997 and December 9,1998 (Reference 3)

Entergy Operations Response To Request For AdditionalInformation, dated D~% 9,1998 (Reference 7).

2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions  !

Inservice inspection program plans are to be based on Section XI of the ASME Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). The third interval at Arkansas Nuclear One, Unit 1, began June 1,1997; therefore, the Code applicable to the third interval ISI program is the 1989 Edition.

However, as stated in Section I of this report, the licensee was authorized to use the 1992 Edition, with a portion of the 1993 Addenda, and has prepared the Third Ten-YearInterm/Inserviceinspection Plan / Program For Arkansas Nuclear One, Unit I to meet the requirements of 1992 Edition of the Code and a portion of the 1993 Addenda In accordance with 10 CFR 50.55a(c)(3),10 CFR 50.55a(d)(2), and 10 CFR 50.55a(e)(2), ASME Code cases may be used u attematives to Code requirements. Code cases that the NRC has approved for use are listed in Regulatory Guide 1.147, Inservice inspection Code Case Acceptability, (R.,.h 8) with any additional conditions the NRC may have imposed When used, these Code cases must be implemented in their entirety. The licensee may adopt an approved Code case by providing wntten notification to the NRC. Published Code cases awaiting approval and subsequent listing in Regulatory Guide 1.147 may be adopted only if the licensee requests, and the NRC authorizes, their use on a case by-case basis.

De licensee's third 10-year ISI program includes the Code cases listed below. Dese code cases either have been approved for use in Regulatory Guide 1.147 or are included as requests for relief.

Code Case N-416-1 Alternatin Pressure Test Requirementfor WeldedRepairs orInstallation of Replacementitems by Welding. Class 1, 2, and 3. (This Code Case addresses both pressure testing and repair / replacement.) (Relief Request IS12-006 approved in SER dated September 4,1997.)

Code Case N-435 l Alternattw Enmination Requirementsfor Vessels with Wall Thickness 2in. orLess Code Case N-460 Alternatin Dnmination Cowragefor Class 1 and Class 2 Welds Code Case N-46l Alternattw Rulesfor Piping Calibration Block Thickness 3

Code Case N 4S1 Alternative Examination Requirementsfor Cast Austenitic Pump Casing Code Case N-508 1 Rotation ofServicedSnubbers andPressure ReliefValusfor the Purpose ofTesting Section E. DMston 1 (Relief Request 1S12 004 approved in SER dated July 30, 1997)

Code Case N 509 Alternatin Rulesfor the Selection and Eramination ofClass 1, 2, and 3 Integrally

  • Walded Attachments (Relief Request 1S12-003 approved in SER dated November 19, 19%7)

, Code Case N 521 Alternatin Rulesfor Deferral ofInspection ofNonle-to-Vessel Welds, inside Radius Sections, andNonle-to-Safe End Welds ofa Pnssurized Water Reactor (PWR) Vessel (Evaluated in Section 3.1.1.1 of this report)

Code Case N-524 Alternatin Examination Requirementsfor Longitudinal Welds in Class 1 and 2 Piping (Relief Request 1512-005 approved in SER dated November 19,1997)

. Code Case N-532 Alternattw Requirements to Repair and Replacement Documentation Requirements andInservice Summary Report Preparation and Submission as Required byIWA-4000 andlWA-6000 (Relief Request ISI2-007 approved in SER dated Sc7 M 4, 1997)

Code Case N-533 Alternatin Requirementsfor YT-2 VisualExamination ofClass 1 Insulated Pressure Retaining Bolted Connections (Evaluated in Section 3.1.7.1 of this report)

Code Case N 546 Alternatin Requirementsfor Qualifcation of VT-2 Examin.stion PersonnelSection XI, Division 1 (Relief Request 1S12-002 approved in SER dated November 19,1997) i, 2.2.2 Acceptability of the Examination Sample Inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1,2, and 3 components and their supports using sampling schedules described in Section XI of the ASME Code and 10 CFR 50,55a(b). Sample size and weld selection procedures have been implemented in ace,,4-.c4 with the Code and 10 CFR 50.55a(b). Review of the licensee's examination samples resulted in the l following apparent discrepancies. It should be noted that while these discrepancies appear in the program plan, the licensee may have further insight describing the condition.

Item Numbers B3.130 and B3.140 require 100% volumetric examination coverage of all primary side steam generator nozzle-to-vessel welds and nozzle inside radius sections. Table 4.1 of the licensee's examination program shows a total of six primary side SG nozzle-to-vessel welds and six primary side SG inside radius sections. From the " Summary of all nird Interval Inspections" tables, it appears that j

, only three primary side SG nozzle-to vessel welds and three primary side SG inside radius sections have been selected for examinatica.

Item Number B5.40 requires 100% volumetric and surface examination coverage of all pressuriser nozzle-to-safe end butt welds greater than NPS 4. Table 4.1 of the licensee's examination program ,

shows a total of two pressurizer nozzle-to-ssfe end butt welds greater than NPS 4. Frun the  !

" Summary of all Hird Interval Inspections" tables it appears that only one pressurizer nozzle-to-safe end butt weld greater than NPS 4 has been selected for examination. '

4 i

5 Item Number B6.190 requires visual examination, VT-1, cf pump flange surfaces when connections are disassembled. Table 4.1 of the licensee's examination program shows a total of 4 pump flange surfaces. From the " Summary of all Hird Interval Inspections" tables it appears that no flarige surfaces have been selected for examination.

Item Number B7.70 requires visual examination, VT-1, ofvalve bolts, studs, and nuts. Bolts, studs, and nuts may be examined; in piace, under tension, when the connection is disassembled or when the bolting is removed. Table 4.I of the licensee's examination program show a total of 23 components.

- From the " Summary of all Third Interval Inspections" tables, it appears that no bolts, studs, or nuts have been scheduled for examinaten Item Number B9.31 requires a surface and volumetric examination on branch pipe connection welds NPS 4 and larger. B9.32 requires a surface examination on branch pipe connecten welds less than NPS 4. De 25% of total number of branch connection welds must be examined Table 4.1 of the licensee's examination program shows a total of two B9.31 components and fourtoon B9.32 components. From the "Surnmary of all Dird Interval Inspections" tables, it appears that no branch connection welds have been selected for examination.

Item Numbers B10.10 and B10.20 in Code Case N 509 require a surface examination on pressure vessel integrally welded attachments and piping integrally welded attachments. Table 4.1 of the licensee's examination program shows a total of twenty one B10.10 components and one B10.20 component. From the " Summary of all Third Interval Inspections" tables, it appears that no B10.10 or B10.20 components have been selected for exammation Item Number C' > reiuires a surface and volumetric examination on circumferential pipe welds of carbon or low alloy steel piping 23/8" nominal wall thickness for piping >NPS 4. Item Number C5.81 requires a surface examination on circumferential pipe welds of carbon or low alloy steel piping >NPS

2. De welds selected for examination of C-F-2 welds shall include 7.5%, but not less than 28 welds. It is unclear why the licensee bas selected 18 of 193 welds for Item Number C5.51, and 1 of 18 welds for Item Number C5.81.

Item Fl.20, requires a visual, VT-3, examination on 15% of Class 2 piping supports. It is unclear why the licensee has selected 47 of 376 supports (12.5%) for examination.

De licensee should re evaluate the discrepancies identified above and verify that all examination categories are in carpliance with Section XI Code requirements regarding examination samples.

2.2.3 Exemption Criteria De criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC 1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). It appears that the exemption criteria have been applied by the licensee in accordance with the Code, and appear to be correct.

., 2.2.4 Augmented Examination Commitmants in addition to the requiremmt:; specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations: ~>

a Ultrasonic examinations on the reactor pressure vessel in accordance with U.S. NRC Regulatory Guide 1.150, Rev.1, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations."

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. . - . . _ . _ . . . . . + . - -

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High Energy Line Break (HELB) and Moderate Energy Line Break (MELB) examinations in accordance with Upper Level Document ULD-0-TOP-07, "HELB/MELB Topical ULD,"

Calculation 86D 105-29, Appendix B, and ANO-1 Technical Specification 4.15.

Visual inspection of High Pressure Injection (HPI) suppon MU-167 and associated piping in asih with ANO Condition Repon CR-1-96-0502.

The examination of stagnant, borated water systems (i.e., NRC IE Bulletin 79-17) as Anainead in Relief Request Number 97 003.

Enhanced ultrasonic examinations of 17 HPI welds and visual inspection of two segments ofHPI piping in accordance with Entergy Operations' January 31,1991, response to NRC IE Bulletin 88-08 (OCAN019102) l Surface and volumetric examinations of reactor coolant pump flywheels in accordance with ANO-1 Technical Specifications 4.2.6.

Ultrasonic examinations on one pressurizer upper level tap in asih with ANO Calculation l 86E-0074-103, memorandum ANO-92-00507, and the NRC submittal dated August 5,1993 (ICAN089302).

a Ultrasonic examinations of reactor coolant pump shaAs and surface examinations of reactor l coolant pump shaA covers each time a pump is disassembled in accordance with Plant Impact l Evaluation (PIE) 87-0082B and Byron Jackson Technical Service Bulletin 8710 80-009.

Surface examinations and visual inspections of emergency feedwater riser welds in accordance with Babcock & Wilcox letter No. APL-85-349," Water Hammer Concern in Auxahary Feedwater Headers "

l Visual VT-2 inspections ofpressurizer level taps in accordance with ANO Condition Report CR-190-0853 07 and memorandum No. ANO-92-02496.

Surface examinations and VT-3 visual inspections of special lifting devices in asih with Entergy Operations' June 8,1984, response to NUREG 0612, " Control of Heavy Imds at Nuclear Power Plants."(OCAN068402)

Ultrasonic examinations of HPI nozzle knuckles and radiography of HPI nozzle thermal sleeves

' in accordance with ANO Condition Repons CR-1-89-0035 and CR-1-89-0508 and Entergy Operations' submittal to the NRC dated April 22,1985 (1CAN048501).

Ultrasonic examinations and thickness measurements of four main steam moisture and reheater piping welds in accordance with ANO Condition Report CR-1-89-0621.

Visual inspections of VSC-24 dry spent fuel storage casks per 1.3.2 and 1.3.3 of the storage cask's " Certificate of Compliance", which was issued in accordance with 10CFR72.

Visual inspections of reactor coolant pump seal injection piping and supports in asih with ANO Condition Report CR-1-90-0514-il.

I a

Ultrasonic examinations ofpressurizer relief valve piping in accordance with ANO Condition Reports CR-1-92-0244 and CR-1-91-0131 Item 8.

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Ultrasonic examinations cf the pressurizer surge line elbow as discussed in NRC letter N2.

ICNA029402 (TAC No. M72108) dated February 24,1994.

2.3 Conclusion Based on the review of the documents listed in Section 2.1, it is concluded that the Third Ten-Year intervalinservice inspection Program For Arkansas Nuclear One, Unit 1, contains multipie discuepancies as noted in Section 2.2.2, Acceptabit/ty ofthe Eramination Sample, of this report. The INEEL staff believes that the licensee should recaluate the discrepancies listed in Sectioa 2.2.2. Note that this report does not include a review of the implementation of the augmented exammations, it merely records that the licensee has committed to perform them.

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3. EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the third 10-year inspection interval are evaluated in the following sections. j 3.1 Class 1 Components 3.1.1 Reactor Pressure Vessel 3.1.1.1 Request for Rollef No. 97 004, Use of Code Case N-521, Alternative Rules For Deferral of Inspections of Norale to Vessel Welds, inside Radius Sections, and Norrie to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel Code #equirement-Section XI, Table IWB-2500-1, Examination Category B-D, items B3.90 and B3.100 require that, for reactor pressure vessel (RPV) nozzle welds and inner radius sections, at least 25%

but no* more than 50% (credited) of the nozzles shall be examined by the end of the drst inspection period and the remainder by the end of the inspection interval. Examinstion Category B-F, Item B5.10, Note (1) states that the reactor vessel nozzle-to-safe end weld exam *mations may be performed with the vessel nozzle examinations.

L/censee's proposed A/temat/ve-4n accordance with 10 CFR 50.55a(a)(3), the licensee has proposed to use Code Case N-52i Alternative Rulesfor DeferralofInspections ofNostle-to-Vessel Welds, inside Radius Sections. andNozzle-to-Safe End Welds ofa Pressurized Water Reactor Vessel. On the following welds:

RL22 B11QQ B5.100 01 011 01-0!!R 01-025 01-012 01-012R 01-26 .01-013 01013R 01-014 01-014R 01-015 01-015R 01-016 01016R 01-017 61-017R 01-018 01-018R

'!he licensee stated:

"Entergy Operations shall complete the required nozzle-to-vessel weld examinations, the noule inside radius section examinations, and the nozzle-to-safe end weld examinations concurrent with the reactor vessel ten-year examinations during the third period of the third ten-year inservice inspection interval,

, in accordance with Code Case N-521."

Licensee's Basis for Proposed Attemative-

"Reliefis requested to defer 100 percent of the reactor vessel nozzle-to-vessel weld examinations, the nozzle inside radius section examinations, and the nozzle-to-safe end weld examinations to the third period of the third ten-year inspection interval at ANO-1.

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  • Entergy Operations believes that perf:rming 25 percent to 50 percent cf the reactor vessel nonic examinations in the first period cf the third inspection interval is impractical for the followmg reasons:
1) "The vendor cost (not including site training, plant support, or potential critical path time) to perform these examinations with automated tooling in the first inspection period is currently estimated at $250,000. The cost to perform these same examinations at the end of the third inspection interval concurrent with the reactor vessel ten-year examination is estimated at only

$25,000. The major expense associated with the first inspection period examinations is the added

  • equipment and personnel mobilization costs and equipment assembly and disassembly costs.
2) "Appmximately three to four man-rem exposure is currently expended for automated equipment assembly and disassembly in the reactor cavity area. In addition to exposure, there are approximately two to three cubic feet of solid radwaste generated during iwiformance of automated examinations in the reactor vessel. Under current Code rules, this iwi,ce.;w: exposure and radwaste generation would be incurred twice; once for the noule first inspection period examinations and again for the reactor vessel examinations at the end of the inspection interval.

Performing the nonle examinations concurrent with the reactor vessel ten year exambtians will save approximately three to four man-rem exposure and two to three cubic feet ofsolid radwaste."

"For reasons listed below, Entergy Operations believes that deferral of 100 percent of the reactor vessel nonle examinations to the end of the third inspection interval will provide an acceptable level ofsafety and quality.

1) "All Reactor Vessel nonle-to-vessel welds, nonle inside radius sections, and nonle-to safe end welds were examined in 1995 during the third period of the second ten-year inspection interval.

No indications or relevant conditions were discovered that required successive inspections in

- accordance with Paragraph IWB 2420(b). Furthermore, no inservice repairs or replacements by welding have ever been performed on any of the nonle-to-vessel welds, nonle inside radius sections, or nonle-to-safe end welds at ANO-1.

2) . The pressurizer and primary steam generator nonle-to-vessel welds, inside radius sections, and aceto-safe end welds are similar in configuration, material properties, weld process pars octers, and operate in the same reactor coolant system environment as the reactor vessel nonles. Due to this similarity, distribution of the pressurizer and steam generator nozzle examinations in accordance with Examination Category B-D and Examination Category B-F will further substantiate the integrity of the reactor vessel nozzles until they are examined at or near the end of the third inservice inspection interval.
3) " Performing all the automated reactor vessel exammations during a single refueling outage improves consistency of the examinations by utilizing the same equipment, personnel, and procedures. Moreover, this improves the reliability and reproducibility of the examinations."

Evaluat/on-The Code requires the examination of at least 25%, but not more than 50% (credited) of RPV nonles and associated inside radius (IR) sections and nonle safe ends during the first inspection period The licensee has requested to use Code Case N 521, which defers the examination of these areas until the end of the third 10-year interval ---

Code Case N 521 states that the examination of RPV nonles, IR sections, and nonle-to-safe end welds may be deferred provided (a) no inservice repairs or replacements by welding have ever been performed on any of the subject areas, (b) none of the subject areas contain identified flaws or relevant conditions that currently require successive inspections in accordance with IWB-2420(b), and (c) the unit is 9

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not in the first interval. The licensee has confirmed that these conditicas have been met. An additio requirement imposed by the NRC is that all subject areas be scheduled for examination such that the new l

sequence of examinations will not exceed 10 years between examinations. The licensee examined all the subject areas during the third period of the second 10-year interval. By examining the nozzles, associated IR sections, and nozzle-to-safe end welds at the end of the previous 10-year interval, the licensee has established a new sequence of examinations and will not exceed 10 years between examinations. By 3

meeting the conditions in the Code Case and by repeating the examinations at the end of the previous 1 interval, the licensee's proposed alternative will provide an acceptable level ofquality and safety since the maximum time of 10 [ Code] years bs w. inspections will not be exceeded.

1 j

Conciuston -Considering that the licensee met all the conditions stated in the Code Case and examined all of the affected areas at the end of the previous interval, a new sequence ofexaminations has been established. Furthermore, since the time between examinations will not exceed 10 (Code] years, the licensee's proposed attemative will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed attemative be authorized pursuant to 10 CFR 50.55a(aX3Xi).

The use of Code Case N 521 should be authorized for the third 10 year interval at Arkansas Nuclear One, Unit 1, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. AAer that time, the licensee may continue to use the Code Case with the limitations, if any, listed in Regulatory Guide 1.147.

3.1.2 Pressurizer No reliefrequests 3.1.3 Heat Exchangers and Steam Generators No reliefrequests 3.1.4 Piping Pressure Baundary No reliefrequests 3.1.5 Pump Pressure Boundary No reliefrequests I

3.1.6 Valve Pressure Boundary  :

No reliefrequests 3.1.7 General

, 3.1.7.1 Request for Relief No.97-005, Use of Code Case N 533, Alternative Requirements for VT 2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections

. . . . . . . . ,3 Code Reguirement-4WA 5242(a) requires that insulation be removed from pressure-retaining bolted connections for VT-2 visual examination in systems borated for the purpose of controlling reactivity.

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Licens?o's prop;s d A/t:rn:riva--In accordance with 10 CFR 50.55a(a)(3), the licensee proposed to use Code Case N-533, Alternattw Requirementsfor VT-2 Visual Examination ofClass 1 insulated

. Pressure-Retaining Bolted Connections. The licensee stated:

"A system pressure test with a minimum four hour hold time and W-2 visual examination shall be performed each refueling outage without removal ofinsulation on systems borated for the purpose of controlling reactivity.

"Each refueling outage, the insulation shall be removed from the bolted connectens in systems borated for the purpose of controlling reactivity, and a W-2 visual examination shall be performed on each of the connections. During this VT-2 examination, the connections are not required to be pressurized.

Any evidence ofleakage shall be evaluated in acevid.nce with IWA 5250.

"These altemative examination requirements are the same as those specified in ASME Section XI Code l

Case N 533 as approved by the Board of Nuclear codes and Standards, with the additional four hour j hold time provision stated above. I I

The licensee stated in their response to the NRC request for additional information dated December 9, 1998.

" Insulation removal will be performed for Class I borated systems each refueling outage while l insulation removal for Class 2 borated systems on a seriodic (40 morth) basis similar to draft Code Case N 533-1, which is at the ASME Section main committee awaiting approval. The intent of the request for the attemative to the Code-required removal ofinsulation on bolted connections within Class 2 systems borated for the purpose of controlling reactivity, is so that the requirements of Code Case N-533 could be applied to the Clus 2 system on a periodic (40 month) basis. During a refueling or during normal plant operation, if a Class 2 system is scheduled for a Code required periodic system leakage test and the system is not at pressure and temperature at the time of the pressure test, then the rules as written for Class 1 in Code Case N 533 could be applied to Class 2 systems. This means that the first VT-2 inspections would be performed at atmospheric or static pressure with insulation removed to look for any evidence ofleakage. Any evidence ofleakage would be evaluated in accordance with IWA-5250. The second VT-2 would be performing the Code required pressure test without removing the insulation, while the system is at nominal operating pressure and t ..g4.re."

Licensee's Basis for Proposed Attemative-I

" Systems which are borated for the purpose of controlling reactivity at ANO-1 include reactor coolant, i

decay heat removal, high pressure injection and make-up. These systems encompass a large portion of the overall ISI program and physically cover a large expanse of the reactor building. Many areas in which this piping and the associated bolted connections are located are difficult to access (e.g., scaffold and/or ladder installation is required) and many of these areas are located such that significant radiation exposures would be encountered.

"In order to identify leakage to be repaired during the outage, the preferred time frame to perform this inspection is in the beginning of the outage subsequent to depressurization of the reactor coolant i system. To perform these inspections at pressure would involve holding the reactor coolant system at

.. operating pressure and temperature for an extended period of time to allow for scaffold constmetion, insulation removal and VT-2 inspection.

"This is normally a relatively short time frame when the unit is transitioning to cold shutdown Holding the unit at normal operating temperature and pressure for an extended period of time would i result in a significant delay in going to cold shutdown.  ;

  • "In addition, the removal and reinstallation ofinsulation with plant equipment in operation at system pressure and temperature increases the risk of personnel injury and presents a safety concern to plant 11 I l

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personnel. The personnel risk and radiation exposure is significant for the removal and reinstallation of insulation at these bolted connections during pressure testmg activities.

"A VT-2 visual examination with the system depressurized would still provide adequate detection of l leakage because boric acid residue can be easily detected with insulation removed at the bolted <

connocuan.

~

" Based on the previously stated reasons, Entergy Operations requests relieffrom the inspection at

  • operating pressure requirements detailed in IWA 5242 (a). In lieu of these requirements, Entergy Operations proposes the alternative examination requirements that follow." '

EvaAmt/on-The Code requires the removal of all insulation from pressure-n:tainirig bolted conre in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations duririg I

system pressure tests. As an alternative, the licensee has proposed to perform a system pressure test and associated VT-2 visual examination without removal ofinsulation from bolted connections on Class 1 and 2 systems. The system pressure tests will be augmented with a minimum 4-hour hold time prior to the VT-2 visual examination. De frequency ofexaminations will be in accordance with the requirements in Table j IWB 2500-1 for Class I systems (each refueling outage) and Tables IWC-2500-1 for Class 2 systems

]

(each period, not exceeding 40 months). In addition, with the systems depressurized, insulation will be removed from the bolted connections for direct visual examination each ofueling outage for Clas I systems, and each period (40 months) for Class 2 systems.

The licensee's proposed altemative is essentially equivalent to Code Case N-533, Altematlur Requirements I f>r YT-2 Visual Examination ofClass 1 Insulated Pressure-Retaining Bolted Connections, except the proposed alternative was extended to address Code Class 2 bolted connections. Code Case N-533 is currently under review by the NRC staff and has not yet been approved for use by incorporation into Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability.

For Class I and 2 systems, the licensee's proposed alternative provides a thorough approach to ensuring the leak-tight integrity of systems borated for the purpose of controlling reactivity. First, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time should allow potential leakage to penetrate the insulation, thus providing a means of Mg significant leakage with the insulation in place. Further, by subsequently removing the insulation each refueling outage for Class I bolted connections, and each period (not exceedmg 40 months) for Class 2 bolted connections, the licensee will be able to detect minor leakage indicated by the presence of boric acid residue. Therefore, it is concluded that this two-phased approach will provide an acceptable level of quality and safety for bolted connections in borated systems.

l Conclus/on-The licensee's proposed alternative, to use Code Case N-533 with the additional visual examination and scheduling requirements for Clus 1 and 2 bolted connections, will provide an acceptable level of quality and safety. Derefore, it is recommended that the licensee's proposed attemative be j authorized pursuant to 10 CFR 50.55a(a)(3)(i). He use of Code Case N-533 should be authorized for the third 10-year interval at Arkansas Nuclear One, Unit 1, or until the Code Case is approved for general use

. by reference in Regulatory Guide 1.147. After that time, the licensee may continue to use the Code Case

, with the limitations, if any, listed in Regulatory Guide 1.147.

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3.2 Class 2 Components 3.2.1 Pressure Vessels No relief requests 12

3.2.2 Piping No relief requests 3.2.2.1 Request for Relief No.97-003, Examination Category C-F-1, items C5.11 C5.12 -

and C5.21, Class 2 Piping Wolds Code #equ/rement--Examination Category C-F-1, Items C5.!!, C5.12, and C5.21, require that surface and/or volumetric examinations be performed on the welds selected. The welds selected for examination shall include 7.5%, but not less than 28 welds, of all austenitic stainless steel or high alloy welds not exempted by IWC-1220. (Some welds not exempted by IWC-1220 are not required to be nondestructively examined per Examination Category C-F-1. These welds, however, shall be included in the total weld j count to which the 7.5% sampling rate is applied.) The examinations shall be distributed as follows:

(a) the examinations shall bc distributed among the Class 2 systems prorated, to the degree practicable, on the number of nonexempt austenitic stainless steel or high alloy welds in each system (i.e., if a system contains 30% of the nonexempt welds, then 30% of the nondestructive examinations required by Examination Category C-F-1 should be performed on that system);

l (b) within a system, the examinations shall be distributed among terminal ends and structural l discontinuities prorated, to the degree practicable, on the number of nonexempt terminal ends and

structural discontinuities in that system; and l

(c) within each system, examinations shall be distributed between line sizes prorated to the degree practicable.

Licensee's Proposed Afternative--in accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to inspect a minimum of 7.5% of all non-exempt C F-1 piping welds regardless ofpipe wall thickness. He licensee stated "A uniform 7.5% sampling rate will be applied to all Examination Category C-F-1 piping welds regardless of nominal wall thickness. The examination requirements shall be as follows:

I

! 1) " Piping 23/8" thick will be subject to volumetric and surface examinations as stated in i

ASME Section XI.

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2) " Piping < 3/8" thick which is not subject to IE Bulletin 79-17 will be subject to a surface examination.
3) " Piping < 3/8" thick which is subject to IE Bulletin 79 17 will be subject to a volumetric examination.

" Systems considered subject to IE Bulletin 79 17 will be those that meet the definition of a stagnant, oxygenated, borated water system," as stated in the bulletin.

"The piping welds selected for examination will still be subject to the distribution requirements stated

, in ASME Section XI, Table IWC 2500-1, Examination Category C-F 1, Note (2)."

Licensee's Basis for Proposed Attemative-

" Code rules require that a 7.5% sampling rate be applied to all Examination Category C-F-1 piping i welds not exempted by IWC-1220. The total weld count to which the sampling rate is applied includes both those welds required to be examined (i.e., 23/8" nominal wsil thickness) and those welds for which examination is not required (i.e., <3/8" nominal wall thickness). De total number of welds required to be examined are then distributed, in a prorated manner, among those systems requiring examination. Rose piping welds less than 3/8" nominal wall thickness, while not requiring 13 l

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examination, will have an impact on the number cf examinations required in systems with piping 23/8" nominal wall thickness.

"At ANO-1 a number of the Class 2 systems have piping with nominal w311 thickness <3/8". This includes portions of the Make up, Decay Heat Removal, Iow Pressure Injection and Building Spray systems. In the past, the Nuclear Regulatory Commission has expressed concems at a number of nuclear facilities, including ANO 2, about examinations not being pedormed on sections of piping because they were less than 3/8" nominal wall thickness.

"In addition to this issue, most of the same Class 2 stainless steel piping is subject to the commitments l made in response to Nuclear Regulatory Commission IE Bulletin 79-17, which stipulates examination of pipes susceptible to cracking from stagnant borated systems in Pressurized Water Reactors. 'Ihe l purpose of this Request is to propose alternative examination requirements for Class 2 piping which addresses both the " thin wall piping" and the IE Bulletin 79-17.

" Based on the reasons listed below, the proposed altemative examinations will ensure that an acceptabic level of quality and safety will be met.

1) " Applying a sampling rate of 7.5% to all Exammation Category C-F-1 piping welds regardless of wall thickness will ensure that all Class 2 rystems will undergo examinations.
2) "During the second ten year interval at ANO-1 piping that was originally subject to IE Bulletin 79-17, but not the ASME Code, was upgraded to. Class T2 (i.e., treated as Class 2 for inspection and testing purposes). Therefore, this piping is included in the sampling plan and subject to examination.
3) "Intergranular Stress Corrosion Cracking (IGSCC) in Class 2 piping is not the significant concern that it was in 1979. Since IE Bulletin 79-17 was first issued, numerous examinations have been performed. In addition, the cause of the cracking, which was the introduction of sodium thiosulfate, has long since been removed. It is logical to conclude that welds that have not shown signs of cracking thus far would remain defect-free because there is no longer a degradation mechanism to initiate or propagate flaws.
4) "The examinations stipulated are conservative, but appropriate. Volumetric examinations are specified on piping which could be susceptible to IGSCC because they are more appropriated than surface examinations for detecting cracking initiating from the inside surface of the pipe. However, even IE Bulletin 79-17 doesn't require volumetric examinations on piping less than 0.250" thick."

Evaluation-Category C-F-1, Items C5.ll, C5.12 and C5.21, require that surface and/or volumetric l , examinations be performed on the welds selected. The welds selected for examination shall include 7.5%, -

! but not less than 28 welds, of all austenitic stainless steel of high alloy welds not exempted by IWC-1220.

(Some welds not exempted by IWC-1220 are not required to be nondestructively exammed per Examination Category C-F-1. 'these welds, however, shall be included in the total weld count to which the 7.5% sampling rate is applied.)

The licensee has proposed to inspect a minimum of 7.5% of all non-exempt C-F-1 piping regardless of pipe wall thickness. The piping welds that comply with the C-F-1 requirement (equal to or greater than 3/8-inch NWT) will receive a full Code examination (volumetric ami surface). The piping welds that are less than i

14

3/8-inch NWT and are subject to IE Bulletin 79-17 will receive a v:lumetric examination. He piping welds that are less than 3/8 inch NWT and are not subject to IE Bulletin 79-17 will receive a surface examination. He proposed change reduces the number of examinations on piping welds greater than 3/8-inch [ wall thickness) but does not affect the overall examination sample size for Class 2 piping welds since piping welds less than 3/8-inch will now be included in the examination sample.

He INEEL staff believes that it is technically prudent to examine a representative sample of Class 2 thin-walled piping welds. This can be accomplished by either augmenting the examination sample or by substituting thin-walled welds for heavy-walled welds. Ec licensee essentially has elected to substitute thin-walled piping in lieu of heavy-walled piping. De licensee has proposed to perform volumetric examinations on piping less than 3/8-inch wall thickness that is subject to IE Bulletin 79-17, and surface examinations on piping ! css than 3/8-inch wall thickness that is not subject to IE Bulletin 79-17. Whde it is understood that volumetric examinations oflarge-bore, thin-walled Class 2 piping is not required by the Code, the INEEL staff believes that portions of the large-bore, thin-walled Class 2 piping represent a significant group of welds with unique operational characteristics (i.e., lower pressures, lower flow rates, stagnant borated fluid), and should receive examinations commensurate to the heavy-wall piping examinations. However, the licensee's proposed attemative, to perform only volumetric examinations on

' piping with less than 3/8 inch wall thickness for piping which is subject to IE Bulletin 79-17, or only surface examinations on piping with less than 3/8 inch wall thickness for piping which is not subject to IE Bulletin 79 17, is inconsistent with thin-walled and heavy-walled examination Code requirements of C-F-1 piping welds. For example C5.ll, and C5.21, require both surface and volumetric examinations. De examinations proposed will provide partial assurance of detection ofinservice flaws, however the INEEL staff believes that adequate assurance of detection of both ID and OD initiated flaws will not be provided.

Additionally, it appears that the licensee has made commitments to IE Bulletin 79 17, and that the licensee's submittal, in part, is seeking relief from those commitments. He INEEL staff does not believe that relief from IE Bulletin 79-17 commitments falls within the scope of this evaluation.

Conclus/on-He licensee's proposed attemative includes the performance of thin walled C-F-1 piping examinations in lieu of heavy-walled piping examinations. De licensee's proposed alternative, includes in part, to perform only volumetric examinations on piping < 3/8" thick which are subject to IE Bulletin 79-17 and only surface examinations on piping < 3/8" thick which are not subject to IE Bulletin 79-17.

Performance of volumentric examinations only, or surface examinations only, is inconsistent with typical C-F-1 requirements. C F 1 requirements for circumferential pipe welds include both a surface and volumetric examination requirement. He licensee has not provided adequatejustification to eliminate surface or volumetric examinations for the subject piping welds. Additionally, it is unclear whether the licensee is seeking relief (in part) from commitments made conerning IE Bulletin 79-17. Herefore based on the above conditions it is recommended that the licensees proposed alternative not be authorized.

3.2.3 Pumps No relief requests

. 3.2.4 Velves No reliefrequests 3.2.5 Generel No relief requests 15

__ _ _ = ;- -

r.-

~*

l i .-

l 1-

!- 3.3 Class 3 Components 3.3.1 Pressure Vessels No reliefrequests l

, 3.3.2 Piping No reliefrequests 1 3.3.3 Pumps ,

No reliefrequests 3.3.4 Velves No relief requests i 3.3.5 General No relief requests 3.4 Pressure Tests - 3.4.1 Cliss 1 System Pressure Tests . v -

5ff ,

b '

( -

3.4.2 Class 2 System Pressure Tests No relief requests 3.4.3 Class 3 System Pressure Tests No reliefrequests 3.4.4 General No reliefrequests

. 3.5 General i 3.5.1 Ultrasonic Examination Techniques No reliefrequests 3.5.2 Exempted Componenta 1

, No relief requests 16

b 3.5,3 Cther No relief requests f

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l 17

. '*w*esam#.., ,,

  • s=

'~ - -

4. CONCLUSION Pursuant to 10 CFR 50.55a(a)(3), it is concluded that for Relief Requests97-004 and 97-005, the licensee's proposed alternatives will (a) provide an acceptable level of quality and safety, or (b) Code compliance will result in hardship or unusual difficulty without a compensating increase in safety. It is secc.. ..ci,ded that the proposed alternatives be authorized.

For Relief Request 97-003 , the licensee did not provide a proposed alternative that would provide an

, acceptable level of quality and safety. Therefore, it is recc.Ts.cr.ded that relief be denied.

=

'this technical evaluation has not identified any practical method by which Entergy Operations, Inc.

can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the existing Arkansas Nuclear One, Unit 1. Compliance with all of the Section XI examination requirements would necessitate redesign of a significant number ofplant systems, procurement of replacement components, installation of the new components, and performance of baseline examinations for these components. Even after the redesign efforts, complete compliance with the Section XI examination requirements probably could not be achieved. Therefore, it is concluded that the public interest is not served by imposing provisions of Section XI of the ASME Code that have been determined to be impractical.

Entergy Operations, Inc. should continue to monitor the development ofnew or improved ext.mination techniques. As improvements are achieved, Entergy Operations, Inc. should incorporate these techniques in the ISI program plan examination requirements.

Based on the review oithe Third Ten-Yearinterm! Inservice inspection Program For Arkansas Nuclear One, Unit 1, the Entergy Operations, Inc.'s response to the NRC's request for additional information, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, deviations from regulatory requirements or commitments were identified in Relief Request 97-003 and the discrepancies noted in Section 2.2.2, Acceptabiltry ofthe Examination Sample, of this report.

l e

e 18

~

- - - - - ~ ~ - -- - -- - - - - -

. .i .

5. REFERENCES
1. Code of Federal Regulations, Title 10, Part 50.
2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1, 1992 Edition.
3. Third Ten-YearInterwlinservice Inspection Plan / Program For Arkansas Nuclear One Unit 1, submitted June 25,1997, and December 9,1998.
4. Letter, dated December 12,1996, W. Beckner (NRC) to J. G. Dewcase (Entergy Operation Inc.)

, containing NRC SER.

5. NUREG-0800, Standard Review Planfor the Review ofSafety Analysis Reportsfor Nuclear Power Plants, Section 5.2.4, " Reactor Coolant Boundary Inservice inspection and Testing," and Section 6.6,

" Inservice Inspection of Class 2 and 3 Components," July 1981.

6. Imer dated December 2,1997 G. Kalman (NRC) to C. R. Hutchinson (Entergy Operation Inc.)

containing request for additional information.

7. Letter dated December 9,1998 to Document Control Desk (NRC), containing response to the NRC RAI dated December 2,1997.
8. NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, Revision 11, October 1994.

l 19

. e' .  !

6 NRC Form 333 U.S. Nulsar Regulatory Comnussaan

1. REPORTNUMBER

- NPCu llo2 (A=igned by NRC, Add Vol., Supp., Rev, and i

  • 32o1,32o2 Addendum Numbers,ifany)

BIBLIOGRAPHIC DATA SHEET INEEUEXT-99 00049 2.1Tlu AND SUBTTIM

3. DATE REPORT PUBtasHED Technical Evaluation Report on the third 10-Year Interval Inservice Inspection Maab Y'ar  ;

Program Plan: Entergy Operttions, Inc. Arkansas Nuclear One, Unit 1, Docket Number 50 313 March 1999 i

4. 71N OR ORANT NUMBER JCNJ2229(TWA A25)
3. ALTTHOR(S) 6. TYPE OF REPORT M.T. Anderson, B. W. Brown Technical C. T. Brown
7. PERIOD COVERED (lactusive Deses)

S.G. Galbraith A. M. Porter i

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Communion, and mailing addreer, ifconancsur. provide name and mailing address)

Idaho National Engineering and Emironmental Laboratory /LMITCO P.O. Box 1626 l Idaho Falls,ID 83415 i

9. SPONSORING ORGANIZATION . NAME AND ADDRESS (If NRC, type "Same as above*';ifcontraaor, provide NRC Division, OHice or Regi Nuclear Regulatory Commission, and mailing address) i l

Civil and Geosciences Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington D.C. 20555

10. SUPPLEMENTARY NOTES
11. ABSTRACT (200 words orlas)

This report presents the results of the evaluation of the 7hird Ten YearIntervalinserviceinspectioa Plan / Program ForArkansas Nuclear One, Unit I submitted June 25,1997 and December 9,1998, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be l

impractical. The Third Ten-YearIntervalinservice inspection Plan / Program For Arkansas Nuclear One, Unit 1 is evalusted in i Section 2 of this report. The inservice inspection (ISI) plan / program is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component l examination exclusion criteria, and (d) compliance with ISI-related commitments identified during presions Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.

12. KEY WORDS/DESCRIPTORS (Ust words or phrases that will assist recarchers in locating the rePart) 13. AVAtt. ABILITY STATEMENT

, J Unlimited  !

14. SECURTTY CLASSIFICATION l (this page) Unclassified i

(This report) Unclassified

15. NUMBER OF PAGES I
16. PRICE l

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