L-96-034, TER on Third 10-Yr Interval ISI Program Plan:Nppd Cns
| ML20212E953 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/31/1997 |
| From: | Mary Anderson, Feige E, Hall K IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20212E074 | List: |
| References | |
| CON-FIN-J-2229 INEL-96-0348, INEL-96-0348-01, INEL-96-348, INEL-96-348-1, NUDOCS 9711040119 | |
| Download: ML20212E953 (70) | |
Text
i INEL-96/0348 March 1997
[ /daho Technical Evaluation Report on the
? National Third 10-Year interval inservice 7,8',"',*[,l"8 Inspection Program Plan:
, Nebraska Public Power District, Cooper Nuclear Station, 1
Docket Number 50-298 M. T. Anderson E. J. Feige K. W. Hall I
i L O C K N E E D M A R T I Nf ENCLOSURE 2 E u' 286!! Ma883,e P PDR
INEL 96/0348 Technical Evaluation Repc,rt on the Third 10-Year Interval inservice inspection Program Plan:
- Nebraska Public Power District, Cooper Nuclear Station, Docket Number 50-298 i
M. T. Anderson, E. J. Folge, K. W. Hall l
Published March 1997 Idaho National Engineering Laboratory Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for the Division of Engineering j Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JCN No. J2229 (Task Order TWA A12) i
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9 ABSTRACT
.This report presents the results of the evaluation of the Cooper Nuclear Station, Third 10-Year Intervalinservice inspection (ISI) Program, Revision 1, submitted by letter dated.
April 11,1996, including the requests for relief from the American Society of Mechanical Engineers (ASMd) Boiler and Pressure Vessel Code Section XI requirements that the licensee has dett mined to be impractical. The Cooper Nuclear Station, Third 10-Year
/nterval/nservice /nspection Program, Revision 1, is evaluated in Section 2 of this report.
The ISI Program Plan is evaluated for (a) compliance with the appropriate editionladdenda 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 previous Nuclear Regulatory Comm!ssion (NRC) reviews.
The requests for relief are evaluated in Section 3 of this report.
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i This work was funded under:
U.S. Nuclear Regulatory Commission JCN No. J2229, Task Order TWA-A12 Technical Assistance in Support of the NRC Inservice inspection Program ii
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SUMMARY
The licensee, Nebraska Public Power District, has prepared the Cooper Nuclear Station, Third 10 Yest Intervalinservice inspection Program, Revision 1, to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and
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Pressure Vescel Code, except that the extent of examination of Class 1 piping welds has
- been determined by the 1974 Edition with Addenda through Summer 1975 as permitted by 10 CFR 50.55a(b). The third 10-year intarval began March 1,1996, and ends February.
' 28,2006.
i The Cooper Nuclear Station, Third 10-YearIntervalinservice inspection Program,
- Revision 0, submitted by letter dated October 18,1995, and Revision 1 to the program,
- submitted in a letter dated April 11,1996, were_ reviewed, as were the requests for relief from the ASME Code Section XI requirements that the licensee has determined i be impractical. As a result of this review, reouests for additional information (RAl's) were=
- prepared describing the information and/or clarification required from the licensee in order
- to complete the review.- The licensee provided the requested information in submittals dated April 11,1996,' August 5,1996, and December 31,1996. In addition, the licensee revised Request for Relief PR-03 by letter dated February 7,1997.
Based on the review of the Cooper Nuclear Station. Third 10-YearIntervalinservice inspection Program,~ Revision 1, the licensee's response to the Nuclear Regulatory - -!
Commission's RAl's, 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, no deviations from regulatory requirements or commitments were identitled in the Cooper Nuclear Station, Third 10-Year Intervalinservice inspection Program, Revision 1, with the exception of Requests for Relief RI 06 (Revision 1), RI-09, RI 11. RI 17, and RI-24.
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CONTENTS ABSTRACT ...................................................... il
SUMMARY
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- 1. I NT R O D U CTI O N . . = . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . . . . . 4 2.1 Docume nt s Evaluated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2 Compliance with Code Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . . . . . . . . . . . 4 2.2.2 Acceptability of the Examination Sample . . . . . . . . . . . . . . . . . . . . . . 5 2.2.3 Exemption Criteria ...................................... 6 l- 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . . . . . . . . . . 6 -
l 2. 3 C o n c lu si o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
- 3. EVALUATION OF RELIEF RE QUESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . 9 3.1 Cla ss 1 C om ponents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1.1 Reactor Pressure Vessel ................................. 9 3.1.1.1 Request for Relief RI-03, Examination Category B-D, Item B3.100, Reactor Pressure Vessel Top Head Nozzle Inner Radius Examinations . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1.1.2 Request for Relief RI 06, Revision 1, Examination Category B-A, items B1.11, B1.12, Pressure Retaining Welds in the Reactor Pressure Vessel, items B1.21 and B1.22. Reactor Pressure Vessel Circumferential and Meridional Head Welds ........................... 10 3.1.1.3 Request for Relief RI-16, Examination Category B G 1, item B6.10, Surface Examination of the Reactor Vessel Closure Head Nuts ' . . . . . . . . . . . . . . . . . ........... 11 3.1.1.4 Request for Relief Al-07. Examination Category B-H, Reactor Pressure Vessel Support Skirt to Bottom Head W eld . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
-3.1.1.5 Request for Relief Rl-15, Examination Category B-O, item B14.10. CRD Housing Welds . . . . . . . . . . . . . . . . . . . 14 3.1.1.6 Request for Relief RI 19, Ravision 1, Examination Category B-D, IWB-2420(b), Successive Examination R e quir e m e nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.1.1.7 Request for Relief, Rl-21, Revision 1, Examination Category B-D, items B3.90 and B3.100, Reactor Vessel Nozzle-to-Vessel Welds . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3 .1. 2 Pr e s s u ri z e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 j 3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . . . . . . . . . 17 !
3.1.4 Piping Pressure Boundary ................................ 18 i iv u
3.1.4.1 Request for Relief RI 08, Pa.ragraph IWB-2430, Expansion Criteria for Welds Goverr ed by Generic Letter 88 01 and NUREG 0313, Rev. 2 . . . . . . . . . . . . . . . 18 3.1.4.2 Relief Request RI 20, Revision 1, Examination Category B F, item B5.130, Dissimilar Metal Butt Weld in Piping ..... 19 3.1.4.3 Request for Relief RI 22, Revision 1, Examination Category B-J, item B9.31, Examinatica of Class 1 Pipe Branch Connection Welds ....................... , 20 .
3.1.5 Pump Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.1.6 Valve Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3 .1. 7 G e n e r a l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 .
3.2 Class 2 Components . . . . . . . . . . . . . . . . . . . 22 3.2.1 Pressure Vessels ..................................... 22 3.2.1.1 Request for Relief RI 05, Examination Category C A, item C1.30, Residual Heat Exchanger, Heat Exchanger Tubesheet to-Shell Welds . . . . . . . . . . . . . . . . . . . . . . . . . 22 3 . 2. 2 Pi pi r ig . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.2.2.1 Request for Relief RI 23, Revision 1, Examination Category C-F 2, item C5.51, Class 2 Piping Welds . . . . . . . 23
. 3.2.3 Pumps.............................................. 23 3.2.3.1 Request for Relief RI 18. Examination Category C C, item C3.30, Integrally Welded Attachments to the Residual Heat Removal Pump Casings . . . . . . . . . . . . . . . . 23 3 . 2. 4 V al v e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3 . 2. 5 G e n e r a l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.3 Cla ss 3 Com ponents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
- 3. 4 Pr e s sur e Te st s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.4.1 Class 1 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.4.1.1 Request for Relief PR-02, Revision 2, Examination Category B-P, item B15.10, Reactor Vessel Pressure-Ret aining S ound a ry . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . 25 3.4.2 Class 2 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 3.4.2.1 Request for Relief PR-04, Examination Category B P, items B15.50 and B15.51. Pressure Testing of the Reactor Pressure Vessel (RPV) Head Flange Seal Leak Dete ction S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 3.4.2.2 Request for Relief PR 05, Revision 1, Examination .
Category C-H, items C7.30 through C7.80, Class 2 Containment Penetration Piping and Valves . . . . , . . . . . . . 28 3.4.2.3 Request for Relief PR-06, IWA-5244(b), Examination -
b Category D A, item D1.10, VT-2 Visual Examination of s
Redundant Systems for Buried Components . . . . . . . . . . . . 31 3.4.2.4 Request for Relief PR-09, IWA 5211(d), Examination Category C-H, item C7.40, Hydrostatic Test of HPCI and RCIC Discharge Piping . . . . . . . . . . . . . . . . . . . . . . . . 33 3.4.3 Class 3 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 l
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_ _ _ _______________________________.________E_-_i__~_^EZ_E___^______'_~'._____._____.____.__.___
3.4.3.1 Request for Relief PR-08, Exemination Category " A, item D1.10, Functional and Hydrostatic Pressure Testing for the Main Steam Relief Valve Discharge Lines . . . 34 3.4.4 General............................................. 35 3.4.4.1 Request for Relief PR 01, IWA 4700(a) and (b),
Alternative Pressure Test Requirements For Code Class 1, Class 2, and Class 3 Systems Following Repair,
- Replacements, and Modifications . . . . . . . . . . . . . . . . . . . 35 3.4.4.2 Request for Relief PR-07, Examination Categories B P, C-H, and D A,10 Year Hydrostatic Pressure Test Requirements for Clast 1,2, and 3 Systems ...........38 3.5 General.................................................. 41 3.5.1 Ultrasonic Examination Techniques . . . . . . . . . . . . . . . . . . . . . . . . . 41 3.5.1.1 Request for Relief RI 02, Revision 1, Appendix ill, Calibration Block Matsrial Specification Requirements . . . . . 41 3.5.1.2 Request for Relief RI-09, Paragraph IWA 2311(b),
Appendix Vll Ultrasonic Examination Personnel Qualification Requirementa . . . . . . . . . . . . . . . . . . . . . . . . . 43 3.5.2 Exempted Components . . . . . . . ........................ 44 3.5.3 Other .............................................. 44 i
3.5.3.1 Request for Relief RI 10, IWA, IWB, iWC, and IWF-
- 4000 (IWX 4000), Repair Procedures, IWA, IWB, IWC, l
and IWF 7000 (IWX-7000), Replacements . . . . . . . . . . . . . 44 3.5.3.2 Request for Relief RI 11, IWB 2420 and IWC 2420, Successive Examinations of Class 1 and 2 Vessels . . . . . . . 47 t
3.5.3.3 Request for Relief RI-12. Examination Category B-J, b ltem B9.12 and Examination Categories C F 1 and C F-2, items C5.12, C5.22, C5.42, C5.52. C5.62 and C5.82, Examination of Class 1 and 2 Longitudinal Pi pin g Weld s . . . . . . . . . . . . . . , , . . . . . . . . . . . . . . . . . . 49 3.5.3.4 Request for Relief RI 13, Examination of Code Class S n u b be r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 3.5.3.5 Request for Relief RI 14, Use of Code Case N 509 for Selection and Examination of Class 1. 2. and 3 Integrally Welded Attachments . . . . . . . . . . . . . . . . . . . . . 50 3.5.3.6 Request for Relief Rl 17, Class 1 and Class 2 Integral!y WeJded Shear Lugs on Piping . . . . . . . . . . . . . . . . . . . . . . 52 3.5.3.7 Request for Relief RI 24, IWB-2412(a) and IWB 2420(a),
Successive Examination Requirements and The Sequence of Examination for Class 1 Bolting . . . . . . . . . . . 53 3.5.3.8 Request for Relief PR-03, IWA 5250(a)(2), Corrective Action Resulting from Leakage at Bolted Connections . . . . . 54
- 4. C O N C L U SI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 5.
REFERENCES................................................. 58 vi
TECHNICAL EVALUATION REPORT ON THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DOCKET NUMBER 50-298 1, INTRODUCTION Throughout the service life of a water cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1 Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code Section XI, Rules for Inservice inspection of Nuclear Power Plant Components (Reference 2), to the extent practical within the limitations of dasign, geometry, and, materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals shall comply with the requirements in the latest edition end 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. The components fincluding 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 limitations and modifications lined therein, and subject to Nuclear Regulatory Commission (NRC) approval. The i
licensee, Nebraska Public Power District, prepared the Cooper Nuclear Station, Third 10-Year Intervalinservice Inspection Program, Revision 0 (Reference 3), to meet the requirements of the 1989 Edition, except that the extent of examination of Class 1 piping welds has been determined by the 1974 Edition through Summer 1975 Addenda as permitted by 10 CFR 50.55a(b). The third 10 year interval began March 1,1996, and ends February 28,2006.
As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain Code examination requirements are impractical and requests relief from them, the licensee shall submit information and justification to the NRC to support that determination.
Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's determination that Code requirements are impractical to implement. The NRC may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the 1
public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Alternatively, pursuant to 10 CFR 50.55ata)(3), the NRC will evaluate the licensee's determination that either (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. Proposed alternatives may be used when authorized by the NRC.
The information in the Cocper Nuclear Station, Third 10-Year Intervalinservice Inspection Program, Revis6n 0, submitted by letter dated October 18,1995, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. The review of the ISI Program Plan was performed using the Standard Review Plans of NUREG 0800 (Reference 4),
Section 5.2,4, "Reactc r Coolant Boundary inservice Inspections and Testing," and Section 6.6, " Inservice inspection of Class 2 and 3 Components."
in a letter dated February 8,1996 (Reference 5), the NRC requested additional information that was required to complete the review of the ISI Program Plan. The requested information was provided by the licensee in the Response to Request for Information Related to the Inservice Inspection Program Plan dated April 11,1996 f (Reference 6), in this response, Nebraska Public Power District provided additional documentation and clarification regarding questions on the program and submitted Revision 1 (Reference 7) to the program. In addition, the licensee submitted 9e new relief requests and revised one.
In a letter dated June 3,1996 (Reference 8), the NRC requested additional information regarding the April 11,1996, submittal that was required to complete the aview of the ISI Program Plan. The requested information was provided by the licensee in the Response to Request for Information Related to the Inservice inspection Program Plan dated August 5, 1996 (Reference 9). In this response, Nebraska Public Power District provided additional clarification regarding questions on Revision 1 and withdrew Request for Relief RI 23.
. In a letter dated October 24,1996 (Reference 10), the NRC requested additional clarification on the licensee's submittals. The requested information was provided by the licensee in the Response to Request for Information Related to the Inservice Inspection Program Plan dated December 31,1996 (Reference 11). In this response, Nebraska Public ,
Power District provided additional clarification regarding the program, requests for relief, revised nine requests for relief, and submitted Request for Relief RI 24.
By letter dated February 7,1997 (Reference 12) the licensee submitted Revision 1 to Request for Relief PR 03.
The Coopar N.sclear Station Third 10-Year IntervalInservice Inspection Program is evaluated in Sect.on 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate editionladdenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion 2
. . - _ _ . . _ . . _ . - _ _ _ _ _ _ _ _ . _ _ . . . . ~ . . _ _ . _ .___.- .. . -_._. - -
4 criteria, and (d) compliance with ISI related commitments identliiod during the NRC's previous reviews.
The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASMJ Code,Section XI,1989 Edition.- Specific ~
inservice test (IST) programs for pumps and valves are being evaluated in other reports.
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- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This 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 pertinent to ISI activities. This section describes the submittals reviewed and the results of the review.
2.1 Documents Evaluated -
Review has been completed on the following information from the licensee:
- Cooper Nuclear Station. Third 10 Year Intervalinservice Inspection Program, Revision 0, dated October 18,1995 (Reference 3).
- Cooper Nuclear Station, Third 10-Year IntervalInservice Inspection Program, Revision 1, dated April 11,1996 (Reference 7).
- Response to tht request for additionalinformation dated April 11,1996 (Reference 6).
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- Response to the request for additionalinformation dated August 5,1996 (Reference 9).
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- Response to the request for additionalinformation dated December 31,1996 (Referenct 11).
- Submittal of additionalinformation dated February 7,1997 (Reference 12).
2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions The inservice inspection Program Plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of March 1, 1996, the Code applicable to the third interva! ISI program is the 1989 Edition. However, as permiticd by 10 CFR 50.55a(b), the extent of examination of Class 1 piping welds has been determined,by the 1974 Edition through Summer 1975 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 as alternatives to Code requirements. Code cases that the NRC has approved for use are listed in Regulatory Guide 1.147, /nservice Inspection Code Case Acceptabl/ity, (Reference 13) with any additional conditions the NRC may have imposed. When used, these Code cases must be implemented in their entirety. 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.
The licensee's third 10 year ISI program includes the Code cases listed below. These Code cases either have been approved for use in Regulatory Guide 1.147 or are included as requests for relief.
Code Case N 3071 Revised Ultrasonic Examination Volume For Class 1 Botting, Table IWB 25001, Examination Category B-G-1, When the Examinations Are Conducted From the Center Drilled Hole Code Case N 408 2 Alternative Rules for Examination of Class 2 Piping Code Case N 416 Alternate Rules for Hydrostatic Testing of Repair or Replacement of Class 2 Piping Code Case N-457 Qualification Specimen Notch Location For Ultrasonic Examination of Bolts and Studs Code Case N-458 Magnetic Particle Examination of Coated Materials Code Case N 460 Alternative Examination Coverage For Class 1 and 2 Welds Code Case N 461 Altemative Rules for Piping Calibration Block Thickness
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Code Case N 4631 Evaluation Procedures and Acceptance Criteria for Flaws in Class 1 Ferritic Piping that Exceeds the Acceptance Standards of IWB 3514.2 Code Case N-491 Alternative Rules for Examination of Class 1, 2, 3 and MC Component Supports of Light Water Cooled Power PlantsSection XI, Division 1 Code Case N-496 Helical-coil Threaded Inserts Code Case N-498 Alternative Rules for Ten Year Hydrostatic Pressure Testing for Class 1 and 2 Systems These Code cases are listed in Regulatory Guide 1.147, Revisien 11, and are acceptable
, for use as w*itten, except that Code Cases N-408 2, and N-461 are acceptable for use only if the conditions stated in Regulatory Guide 1.147 are satisfied.
12.2 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 samp!ing schedules described in Section XI of the ASME Code and 10 CFR 50.55a(b). The sample size and weld 5
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selection have been implemented in accordance with the Coda and 10 CFR 50.55a(b) and appear to be correct.
2.2.3 Exemption Criteria The 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). The exemption criteria have been applied by the licensee in accordance with the Code, as .
discussed in the ISI Program Plan, and appear to be correct.
2.2.4 Augmented Examination Commitments in addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:
- Ultrasonic examination of the feedwater nozzle safe ends, bores, and inside blend radii, and visual inspection of the feedwater spargers per Table 2 and Section 4.3.2.4 of NUREG 0619. In lieu of the dye penetrant examination of feedwater nozzles per NUREG 0619, automated UT of the nozzles is performed (Reference 14).
- Visual inspection of the Core Spray spargers and the Core Spray piping inside the RPV each refueling outage. (Reference IE Bulletin No. 8013).
- Ultrasonic examinations of the jet pump hold down beams. These examinations will be performed once during the third ten-year interval and may be deferred to the end of tho interval. (Reference NUREG CR3052).
- Examination of welds subject to Generic Letter 88 01, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC)in BWR Austenitic Steinless Steel Piping (Reference 15). All accessible welds will be examined in accordance with CNS GL 88-01 commitments. Added requirements for weld crown conditioning for UT (future welds) will be applied per General Electric SIL No.117R3.
- Visual inspection of steam dryer channel welds during each inspection period (Reference General Electric SIL No. 474).
- Visual inspection of jet pump nozzles and mixer inlets each inspection period in conjunction with jet pump inspection. (
Reference:
General Electric Sll No. 465 S1).
- Based on the results of previous examinations, an ultrasonic examination of the shroud support access hole covers will be performed once every five years, and a visual examination will be performed once each refueling outage. (
Reference:
General Electric SIL Nol 462, Supplement 3).
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- Visualinspection of the core spray T-junction box welds inside the reactor vessel.
(
Reference:
General Electric SIL No. 289, R1,' S1).
- Visual inspection of the reactor recirculation (RR) pumps' shafts, pump covers, impeller! shaft attachment region (including bolts), and hydrostatic bearings (including baffle plate). (
Reference:
General Electric SIL No. 459 and RICSIL No.
038).
Visual examination of all accessible areas of the intermediate range monitor (IRM) and source range rnonitor (SRM) dry tubes during the sixth refueling outage after replacement and every third refueling outage thereafter. (
Reference:
General Electric SIL No. 409 R1).
- Ultrasonic examination of all remaining old design, creviced, inconel 600 shroud head bolts (HSHBs) each refueling outage. (
Reference:
General Electric SIL No.
433 S1).
Augmented inservice inspection of the REC system as required by CNS CR94-0485.
Visual examination of jet pump sensing lines and sensing line support brackets each inspection period in conjunction with scheduled ISI examinations. (Reference General Electric SIL No. 420)
Visual examination of the steam separator, once per inspection interval as described by the CNS Invessel Visual Examination Procedure.
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- Ultrasonic inspection of the core shroud per BWRVIP and GL 94 03. (Supersedes GE SIL NO. 572 and RICSIL No. 68).
Inspection of instrument r > tie safe ends in conjunction with scheduled ISI examinations. (Reference - s.: SIL No. 571).
- Visual inspection of the top guide. (
Reference:
Visualinspection of the top guide and core plate. (Reference GE SIL No. 588 R1)
Visual inspection of jet pump riser brace each inspection period. (
Reference:
Visualinspection of jet pump adjusting screw each inspection period. (
Reference:
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2,3 Conclusions Based on the review of the documents listed above, no deviations from regulatory '
requirements or commitments were identified in the Cooper Nuclear Station, Third 10 Year
- Interval /nservice /nspection Program, Revision 1, with the exception of those discussed in the following section.
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- 3. EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has l determined to be impractical for the third 10 year inspection interval are evaluated in the l following sections.
3.1 Class 1 Components j 3.1.1 Reactor Pressure Vessel 3.1.1.1 Request for Relief Rl-03, Examination Category B D, item B3.100, Reactor Pressure Vessel Top Head Nozzle Inner Radius Examinations Code Requirement-Section XI, Table IWB 2500-1, Examination Category B-D, item B3.100 requires a 100% volumetric examination of all reactor vessel nozzle inner radius sections each inspection interval as defined by Figure IWB 2500 7.
Licensee's Code Reile/ Request-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Code required volumetric exar.linc; ion of the inner radius sections for the head spray nozzle (N6A), head instrument nozzle (N6B), and head vent nozzle (N7).
Licensee's Basis-
"The three nozzles specified in this request for relief are located in the reactor vessel closure head. The nozzles are of such size ar.d geometric configuration that a volumetric examinution, performed from the O.D. surface of the nozzle, will neither yield meaningful results nor provide for complete coverage of the required weld volume "
Licensee's Proposed Attemative Examination-
"As an alternate examination, CNS will perfcrm a surface examination of the subject nozzle inner radius sections in lieu of the required volumetric examination.
Since the intent of the volumetric examination is to detect cracking at the inner surface of the nozzle radius section, a surface examinatien performed when the closure head is removed during a refueling outage, would provide more sensitive results and provide for complete coverage of the required area subject to examination. Relief from the volumetric examination requirements and performing a surface examination of the nozzle inner radius sections was granted during the second ten-year inspection intervalin Relief Request No. 3 of the CNS ISI Program."
Evaluation-The Code requires thct the subject nozzle i ner radius sections be 100%
volumetrically examined. However, because of the s: and geometry of the reactor pressure vessel head nozzles, it is difficult to obtain a meaningful ultrasonic examination of the inner radius sections without the use of computer modeling and multiple, angle beam transducers. The licensee h5s proposed to perform a surface examination of the inner 9
radius for the area labeled M to N as defined by Figure IWB 2500 7,in lieu of a volumetric examination. Boiling water reactor closure heads are typically unclad, and access to the subject nozzle inner radius section from beneath the reactor pressure vessel head is possible when located on the head stand during refueling. For the inner radii of these nozzles, inservice flaws will likely initiate on the vess31inside surface. The licensee's alternative, performing a surface examination on the inner radius section, will verify that flaw initiation at the nozzle inside radius is not occurring. Therefore, the INEL staff believes that the su?f ace examination will provide an acceptable level of quality and safety.
Conclusion-Because the licensse's proposed inside surface examination, in lieu of a volumetric examination, shculd detect flaws initiating at the inner radius surfaces, the INEL -
staff believes that an acceptable level of quality and safety is provided. Therefore, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(alt 3)(i). ',
3.1.1.2 Request for Relief RI 06, Hevision 1, Examination Category B A, items B1.11, B1.12 Pressure Retalning Welds in the Reactor Pressure Vessel, items B1.21 and B1.22, Reactor Pressure Vessel Circumferential and Meridional Head Welds
/ Code Requirement-Section XI, Table IWB-25001, Examination Category B-A, items B1.11 and B1.12 require that essentially 100% of all circumterential and longitudinal reactor pressute vessel shell welds be volumetrically examineo as defined in Figures IWB-2500-1 and IWB-2500-2. Items B1.21 and B1.22 require that cssentially 100% of the accessible length of all reactor pressure vessel circumferential and meridional head welds be volumetrically cwamined as defined in Figure IWB 2500 3.
Licensee's Code Re/lef Request-The licensee requested relief f'ont the Code-required 100% volumetric coverage of the tollowing welds: HMB-BB ~l, HMB BB-2, HMB BB 3, HMB-BB 4, HMB BB 5, HMB BB 6, VCB-BA-2, VLA-BA 1, VLA-BA-2, VLA-BA 3, VLB-BA-
- 1. VLB-BA 2, VLB BA 3, HMC BB-1, VCB-BC 51, VCB BC 5-2, and VCB-BC-5 3.
Licensee's Basis for Requesting Relief-
'The Cooper Nuclear Station construction permit was issued before the effective date of implementation for ASME Section XI and thus the plant was not designed to meet the requirements of inservice inspection; therefore,100% compliance is not feasible or practicable.
" Access to the reactor vessel beltline region from the exterior is not possible. The reactor vesse is insulated with permanent reflective insulation and surrounded by a concrete biological shield. The annular space between the inside diameter of the insulation and the outside diameter of the reactor vesselis a nominal 2 inches.
There is no working space to remove the insulation panels from the vessel, which precludes both direct and remote examination of the outside surface.
"The interior surf ace is clad and the vesselinternals, shroud and jet pu'1ps make an internal volumetnc examination of these woldc difficu!t. Parts of longitudinal seams VLA BA-1,2, and 3 however appear to be accessible from openings arour'd the 1
10
_ recirculation riser nozzles N2A, N2E, an N2H respectively. Again these seams are
- not 100% accessible. -In order 'e scan the weld a minimum of 17 inches of surface area from the weld would be re9 ired. This surface area is only available for_ a few inches close to a nozzle."
" Access to the bottom head circumferential weld HMC BB 61-is limited due to the proximity of the vessel skirt. The configuration limits scanning with the 600 probe. ,
, The total composite coverage is approximately 86%.
" Access to the shell to flange circumferential weld VC8-BC 51, 2, & -3, is hmited
.. due to the flange configuration and oroximity of the vessel thermocouples. T_he configuration limits scanning with the O*,45*, and 60' probes. The thermocouples limit scanning with both the 45' and 600 probes. The total
- composite coverage is approximately 74%"
Licensee's Proposed Altemative Examination-
"As an alternative examination, CNS will develop and implement a
- Reactor Vassel (RPV) Examination Plan" in accordance with 10 CFR 50.55a(g)(6)(A). Upon l_ ccmpletion of the examinations, this relief request will be revised as necessary to
- dccument the weld lengths accessible for future examination.
Evaluation-The Code requires that all RPV shell welds and the accessible length of RPV head welds receive essentia!!y 100% volumetric examination. . Because the licensee has not performed the examinations to determine actual coverages for compliance with the
. Augmerced Reactor Pressure Vessel Rule as required by the Code of Federal Regulations, 10 Cr A 50.55a(g)(6)(ii)(A), issued September 8,1992,it is recommended that relief be denis 1.
Conclusion-For the subject reactor pressure vessel welds,-thelicensee should submit
. requests for relief based.on information developed following actual examinations where coverages are less than 90%. Therefore, it is recommended that relief be denied.
3.1.1.3 Request for Relief RI-16. Examination Category B G 1, item B6.10, Surface Examination of the Reactor Vessel Closure Head Nuts
- Code Requirementt-Section XI, Table IWB-25001, Examination Category B-G-1, item
.,. _ B6.10 requires a 100% sunace examination of all reactor vessel closure head nuts.
~ Licensee's Code Re//e/Recuest-Fursuant to 10 CFR 50.55ata)(3)(i), the licensee proposed an alternative to the Code-required surface examination of the reactor vessel closure head nuts as specified in. Table'lWB-25001 of the 1989 Edition of ASME Section XI.
Licensee's Basis-
" Table IWB 2500-1 cf the 1989 Edition of ASME Section XI requires a surface examination to be performed on the reactor vessel closure head nuts. However, Table IWB-2500-1 does not provide the corresponding " Examination 11
F
- Requirements / Figure Number" and " Acceptance Standard". - These provisions were stillin the course of preparation.
L " Provisions for the " Examination Requirements / Figure Number" and " Acceptance Standard" for the reactor vessel closure head nuts were later incorporated in the 11989 Addenda of ASME Section XI. This Addenda also changed the examination method to a VT-1 visual examination.
"It would bs impracticable to follow the incomplete examination requirements for the reactor vessel closure head nuts delineated in the 1989 Edition of ASME Section XI, when the 1969 Addenda has incorporated the complete examination .
requirements.
" Based on the above, CNS requests relief from the requirements specified in Table IWB 2500-1 of the 1989 Edition of ASME Section XI for reactor vessel closure
' head nuts."
. Licensee's Proposed Attemative Examination- !
"As an attemate examination, CNS will perform a VT-1 visual examination of the surface of all reactor closure head nuts, utilizing the acceptanco criteria of IWB-3517, as delineated in the 1989 Edition of ASME Section XI "
Evaluat!on-The licensee has requested relief frnm performing the Code-required surface
- examination on the reactor pressure vessel closure head nuts. As an alternative, the licensee proposes to perform a VT 1 visual examination. Based on a review of examination requirements for Examination Category B-G-1,'it is noted that with the
= exception of the reactor pressure vessel closure head nuts and the cbsure studs (when
- removed), a VT-1 visual examination and volumetric examination (as applicable) is required.
. Typical relevant conditions that would require corrective action prior to putting closure
. head nuts back into service would include corrosion, deformed or sheared threads,
- deformation, and degradation mechanisms (i.e., boric acid attack). The applicable Code examination requirernent for closure head nuts is a surface examination. Surface
- examination procedures are typically qualified for the detectionif linear flaws (cracks), and have acceptance criteria only for rejectable linear flaw lengths. When performing surface examinations in accordance with the 1989 Edition of the Code, item B6.10, the surface- .
examination acceptance criteria is not provided. Without clearly defined acceptance criteria, relevant conditions that require corrective measures may not be adequately-addressed.
Article IWB 3000, Acceptance Standards, IWB 3517.1, Visual Examination, VT 1, describts relevant conditions that require corrective action prior to continued service of
- botting and essociated nuts. Included for corrective action in IWD-3517.1 is the requirement to compare crack like flaws to the flaw standards of IWB 3515 for acceptance. Surface examine' ion acceptance criteria are typically limited to linear flaws (i.e., cracking, aligned pitting, and corrosion). Because the VT-1 visual examination 12
. - - ~ - - - - -.- -..- - ._ . .-. - - - . .. - - _ . -
4 4
L l" Performance 'of this partial coverage volumetric examination, coupled with a surface
[ examination on the A B side of the weld, will provide adequate assurance of the structuralintegrity of the weld.-
7 q " Based on the above,'CNS requests relief from the ASME Section XI, surface '
! examination requirements of Figure _lWB 250013.'
3- "
Relief from the examination requirements of Esction XI was granted for this weld '
< Licensee's Proposed Attemative- .
!: "As an alternate 6xamination, CNS will perform a surface examination on the j ' accessible side of the subject weld (area A B), and an ultrasonic examination of i- area A B C-D, as shown on Figure RI-07.2, each impection interval."
~
I ' Eve /uetion-Based on a review of the subject examination _ area, it is apparent that access ,
to perform a surface examination _of area C D inside the spport skirt is limited. The
- iicansee's proposed alternative is to examine area C-D to the extent possible with an
[ Lultrasonic technique. Based on this review, the INEL staff believes that the licensee's L - proposed alternative to examine area A B C D ultrasonically, in combination with the Code- -
i required surface examination of the weld area A B outside of the support skirt, will provide ,
an acceptable level of qual.ity and safety.--
h Conclusion-The licensee's proposed alternative, to perforra a volumetric examination of area A B C D, in combination with a surface examination of area C-D, will provide an .
L acceptable level of quality and safety. Therefore, it is recommended that the proposed -
- alternstive volumetric examination be authorized, pursuant to 50.55aia)(3)(i).
4-
- i. -. .
3.1.1.5 Request for Relief RI 15, Examinction Category B-0, item B14.10, CRD Housing -
i . Welds
,~ .
[- -
~
LCode Asquirement-Table lWB 25001, Examination Category B-0, item B14.10 requires a
~
{~ surface examination to be performed on 10% of the peripherr.1 CRD housing welds.
t I I Licensee's Code RelistRequest-Pu .uant to 10 CFR 50.55a(g)(5)(iii), the licensee
' requests relief from performing 10J% surface (;Xaminations on 10% of the periphery CRD
! housing _ welds. ,
- Licensee's Basis for Requesting Relief- ,
"There are' thirty.six CRD housings on the periphery. Each housing has an upper
- -and lower weld. A surface examination of 10% of these welds would require the
_w' olds in four housings be examined. The upper CRD housing welds are located
~
inside the reactor vessel skirt. The twelvo inch diameter hole in the reactor vessel support' skirt is too small to permit access for a surface examination. The low'er
- ~
CRD housing welds are partially accessible, however, the adjacent CRD housings
_ prevent surface examination of approximately 50% of the weld."
U 14 t
~
, ,n, .,. ,, .- -- - -l - , , . -
-.. - , . . . , - - , - . ,~ ,.
f acceptance criteria include the requirement for evaluation of crack like indications and other relevant conditions requiring corrective action, such as deformed or sheared threads, localized corrosion, deformation of part, and other degradation mechanisms,~ it can be concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure head nut. As a result, the INEL staff believes that VT-1 visual examination provides an acceptable level of quality and safety. In addition, it is noted that the 1989 Addenda of Section XI changes the requirement for the subject reactor pressure vessel closure head nuts from surfact to VT 1 visual examination and provides appropriate acceptance criteria.
. Conclusion-The licensee has proposed, as an alternative to the Code required surface examination of reactor pressure vessel closure head nuts, to perform a VT-1 visual
- examination. Based on the comprehensive assessment that the VT-1 visual examination provides, and considering that the 1989 Addenda and later editions of the Code require only a VT 1 viwat examination on reactor pressure vessel closure head nuts, an acceptable level of quality and safety is provided. Therefore,it is recommended t ' the proposed alternative VT 1 visual examination be authorized pursuant to 10 CFR ut,.55a(a)(3)(i).
3.1.1.4 Request for Reliaf RI ti7, Examination Category B H, Reactor Presaure Vessel Support Skirt to Bottom Head Weld Code Requirement-Table IWB-25001, Examination Category B H, requires 100%
volumetric or surf ace examination, as defined by Figures IWB-250013, -14, or -15, as applicable.
Licensee's Code Re/le/ Request-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee ptoposed an alternative to performing the surface examination of the areas A B and C D of the reactor pressure vessel support skirt weld.
Licensee's Basis-
"The reactor vessel bottom head was constructed with a weld build up around the circumference of the head that was designed as the attachment point for the reactor vessel support skirt (Figure RI 07.1'). This weld build-up was machined and heat treated along with the reactor vessel.
"The support skirt is attached to the weld build-up by means of a full penetration butt weld. As can be seen in Figure Ri-07.2, the design of this weld is such that the surface examination requirements of Figure IWB 250013 (area C-D) can not be met due to lack of accessibility. Additionally, the configuration c,; .he weld precludes the performance of a full coverage volumetric examination of volume A-B-c C-D from one side (A-B) of the weld. However, a partial coverage volumetric examination of this volume can be performed from one side (A-B) of the weld.
9 Based upon a review of the fabrication drawings,it is estimated that an ultrasonic examination can be performed on volume A-B-C-D, as shown on Figure RI-07.2,
- 1. Figures provided by the licensee are not included with this evaluation.
13
Licensee's rhoposed Attemstive-
"In lieu of performing the Code required examinations, CNS proposes to examine 50% of eight peripheral CRD lower housing welds during the inspection interval and visually examine (VT 2) the remaining CRD housing welds (upper and lower) in -
l _ conjunction with the Class 1 system leakage test after each refueling outage."
l
)- EveAustion-The Code requires a 100% surface examination of a 10% sample of the subject peripheral CRD housing welds. However, based on review of the documentation provided by the licensee, it has been determined that clearances between the support skirt and the CRD's restrict access for examination personnel, inside the support skirt, making
, the Code-require _d surface examination impractical. To perform the Code required surface examination, the CRD s and reactor vessel support skirt would require design modification to allow access for examination. Imposition of this requirement would cause t, considerable burden on the licensee. Surface examination of 50% of the eight peripheral CRD lower housing welds (equivalent to 100% of 4 CRDs required to be examined by Code) and the VT-2 visual examinations of CRD housing welds in conjunction with Class 1 system leakage testing after each refueling outage will detect a pattern of degradation, if present. As a result, reasonable assurance of operational readiness will be provided.
Conclusion-Examination of the CRD housing welds to the extent required by the Code for CNS is impractical due to inaccessibility. The licensee will perform surface examinations
- of 50% of the eight peripheral CRD lower housing welds-(equivalent to 100% of 4 CRDs required to be examined by Code) and VT 2 visual examinations of CRD housing welds in conjunction with Class 1 system leakage testing after each refueling outage, providing reasonable assurance of operational readiness. Therefore,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.1.6 Request for Relief RI 19, Revision 1 Examination Category B D, IWB 2420(b),
- Successive Examination Requirements Request for Relief RI 19, Revision 1,is not within the scope of this review and is being evaluated in a separate report by the NRC.
3.1.1.7 Request for Relief, RI 21, Revision 1. Examination Category B D, items B3.90 and B3.100, Reactor Vessel Nozzle to Vessel Welds Code Requirement _-Section XI, Table IWB 25001, Examination Category B-D, item B3.90 requires _100% volumetric examination ci reactor vessel nozzle-to-vessel welds as defined by Figure IWB-2500 7. Item B3.100 requires.100% volumetric examination of nozzle
. inner radius sections as defined by Figure IWB-2500 7.
Licensee's Code Re//e/ Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required volumetric examinations of the following nozzle-to-shell welds, and the nozzle inner radius section ior the N9 CRD nozzle:
15
1 Nonle Norrle % :
Number Description Access Restrictions Examined i
N1A,B Recirculation Inlet Nozzle configuration, and insulation 32%
support frame N2A H, J&K Recirculation Nozzle configuration, ar.d insulation 40% 7 Outlet support frame ,
inA D
~
Main Steam Norrie configuration 35%
N4A&C Feedwater Norrie configuration, insulation 27% -
support frame, and thermoccuple pads N4B&D F)edwater Norzle configuration, insulation 31 %
support fratus, and thermocouple pads ;
N5A,B Core Spray Nozzle configuration, lotulation 31 %
support frame, and thermocouple pads NBA,B Top Head Spray Norzle configuration 50 %
Top Head Vent Norrie c0nfiguration 30%
. 4,8 Jet Pump Nozzle configuration, and insulation 76%
Instrumentation support frame N9 CRD Return Nozzle configuration, and insulation 70 %
Nozzle and Inner support frame Radius Licensee's Basis for Requesting Relief-
The Cooper Nuclear Station construction permit was issued before the effective date of implementation for ASME Section XI and thus the plant was not designed to meet the requirements of inservice inspection; therefore,100% compliance is not feasible or practical. The configuration of the norWs, the design of the vessel insulation support rings and th? nozzle access hatchos, and interferences from thermocouple pads, instrument lines, etc. prevents 100% examination of the required weld volumes. Alternate angles were used to the extent practical to increase tha volume examined. The extent of the nozzle restrictions and the total '
volume examined is summarized on table RI 21 1.'"
In the August 5,1996, response to a %fest for additionalinformation, the licensee -
provided the following:
" Request for Relief Rl 21, Rev. O, Code item B3.100, Nozzle Inner Radius, applies
- to the Inner Radius of the CRD Retern Nozzle, N9 only, it does not apply to any of 2, ~ Not included with this evaluntion.
, .16
the other nozzles listed. 70% of the CRD Return Line Nozzle Inner Radius was examined as indicated in the relief request.
- Typical nozzle ultrasonic examinations incluCed O',40', and 60' scans.
Supplemental 70' scans were used in zone 1 of the inner radius. Supplemental manual exams were used where possible to increass coverage when automated examinations were performed. The percent examined lei Table RI 21 1 represents the total composite coverage. The extent of coverage *nt individual scans is documented in the examination record. Examinat. lor, etsults wer iindependently reviewed by a UT Levelill examiner and by the Authorizeo Nuclear Inservice inspector.
Licensee's Proposed Alternative Examination-
"In lieu of performing the Code required exarninationi,, CNS propos3s to examine the accessible portions of reactor vessel nozzlo to vesse! welds."
Evaluation-The licensee has requested relief from the Code's volumetric examination requirements for the subject reactor vessel nozzle +to vessel welds and the CRD nozzle inside radius section. Based on a review of the information provided, it has been determined that the component configuration, insulation support rings, nozzle access hatches, and thermocouple pads and instrument linis limit scanning, precluding complete volumetric examination of the nozzle to vessel weids and the N9 CRD nozzle. ?-3 a result, compliance with the Code requirement is impractical. Providing access to the examination areas would require design modifications and, therefore, imposition of this requirement would cause a considerable burden on the licensee.
The licensee is examining the accessible portions of the subject nozzle to shell welds and the CRD nozzle inner radius section to the extent practical, which results in coverages from 27% to 70E Based on the percent of combined coverages obtainable for the subject nozzles, in conjunction with the VT 2 visual examination, it is reasonable to conclude that degradation, if occurring, will be detected. As a result, reasonable assurance of structuralintegrity is provided.
Conclusion-Examination of the subject nozzle areas is impractical due to component configuration and obstructions. The limited examinations,in combination with remote visual examinations of the accessible internal surf aces and hydrostatic testing, should provide reasonable assurance of operational readiness. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55atg)(6)(i).
. 3.1.2 Pressurizer Not Applicable to BWR's 3.1.3 Heat Exchangers and Steam Generators No relief requests, t
17
l 3.1.4 Piping Pressure Boundary 3.1.4.1 Request for Rollef Rl 08 Paragraph IW8 2430, Expansion Criteria for Wolds Governed by Generic Letter 88 01 and NUREG 0313, Rev. 2 Code Megu/rement-Section XI, Paragraph IWB 2430 states that:
(a) Additional examinations shall be performed during the current cutage when .
indications exceed the acceptance standards of Table IWB 34101. The additional examinations shallinclude the remaining welds, areas, or parts included in the inspection item listing and scheduled for this and the subsequent period. -
(b) If the additional examinations required above reveolindications exceeding the acceptance standards of Table IWB 34101, the cxaminations shall be further extended to include additional examinations at thit; outage. The additional examinations shallinclude all the welds, areas, or parts of similar design, size, and function.
(c) For the inspection period following the period in which the examinations of (a) or (b) above were completed, the examinations shall be performed as normally scheduled in accordance with IWB 2400.
L/censee's Code Melie/ Request-Pursuant to 10 CFR 50.55ata)(3)(l), the licensee proposed an alternative to IWB 2430 additional examination requirements for all full-penetration circumferential and branch pipe connection welds in austenitic stainless steel piping that are NPS 4 or larger and contain reactor coolant at a temperature greater than 200*F during power operation, licensee's Basis-
- Each of the subject welds fall under the augmented inspection prngram required by Generic Letter 88 01, and NUREG 0313, Rev. 2. This program governs examination methods, examination frequency, and sample expansion. The sample expansion requirames of this program are designed such that additional examinations are limned to welds that have the same susceptibility to Intergranular Stress Corrosion Cracking (IGSCC) as the weld in which the flaw was found. This rnethodology ensures that welds at a high risk for cracking are examined during the same refueling outage, while not requiring expenditure of the Man rtm and outage .
- time associated with examining additionallow risk welds.
"In many instances, the examinations performed to meet the requirements of Generic Letter 88 01 are also used to meet the requirements of ASME Section XI.
In these cases it is r'ot practical to apply the expansion criteria of both Generic Letter 88 01/NUREG 0313 and ASME Section XI when unacceptable IGSCC flaw indications are identified.
18
- Baserd on the rhove, CNS requests relief from the ASME Section XI requirements for ad kiona! exorninations when unacceptable flaw indications are identified in the suWet.' nelds."
Licensee's Proposed Altemellve Eneminellon-
- CNS will perform sample expansions as required by Generic Letter 88 01 and NUREG 0313, Rev. 2 when unacceptable IGSCC flaw indications are identified in
, the sub}ect welds.*
l Evaluation-The Code states that examinations that revealindications exceeding acceptance randards shall be extended to the remaining wolds, areas, or parts included in the inspection item listing and scheduled for examination during this and the subsequent petiod. NUREG 0313 states that for welds susceptible to intergranular Stress Corrosion Cracking (lGSCC) an additional sample of the welds in the sop < priate category (Categories l
A. B, 04 C) should be inspected. The additional sample should oe apprWmately equalin
! number to the original sample and should be similar in distribution (pipe size, system, and t
location) to the original sample. The NRC staff has found it acceptable to take Section XI l
credit for the augmented volumetric examinations performed in accordance with Generic Letter 88 01 and NUREG 0313, Rev. 2 provided that the Code required surface l examination is also performed.
The licensee has requested authorization to use the sample expansion criteria for additional examination areas provided in NUREG 0313. This sample expansion methodology is a systematic approach to determination of potential f ailure trends since the sample is selected from components with similar characteristics. In addition, the structure of the NUREG 0313 scheduling criteria essentially doubles the number of welds susceptible to iG5dd that receive volumetric examination during the 10 year interval, which tends to offset the smaller number of additional examinations that may be required if IGSCC is detected.
Conclus/ou-The licensee's proposed alternative in combination with the Code required surface examinations will provide an acceptable level of quality and safety because the additional examination areas selected will more closely relate to the welds where IGSCC is detected. Therefore,it is recommended that the nroposed afternative be authorized pursuant to.10 CFR 50.55a(a)(3)(i).
, 3.1.4.2 Relief Request RI.20. Revision 1, Examination Category B F, item B5.130, Dissimilar Metal Butt Welds in Piping Code Requ/rement-Table IWB 25001 Examination Category B r, item B5.130 requires 100*A surface examination of nozzle to safe end welds as defined by Figure IWB 2500 8.
Licensee's Code Ae/le/#equest-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code required 100'A surface examination of Weld RVD BF 14.
19 -
e licensee's Basis for Requesting ReNef- -
- A rigid support i&cated adjscent to this 2 inch NPS stainless steel to carbon steel butt weld in the reactor drain line. The design of the support does not permit disassembly without cutting the support members. The support limits access to one side of the weld crown such that only 1/4 inch of the adjacent bar.e metal can '
be examined. The total surface examined is approximately 83%."
Licensee's Proposed Alternative Examinellon-
"In lieu of performing the Code required examinations, CNS proposes to examine the accessible portions of Weld RVD BF 14."
Evoluet/on- The Code requires that the subject we!d receive a 100% surface examination.
However, due to the proximity of a rigid support, access to one side of the weld is restricted. As a result, the complete Code required examination is impractical. To istisfy Code requirements, the rigid support must be removed by cutting or the line must be modified, either of which would result in a burden on the licensee.
The licensee is examining the accessible portion of the subject weld, which results in 83% coverage. Based on the significant percent of coverage obtainable,it is reasonable to conclude that, degradation, if occurring, will be detected. As a result, reasonable assurance of structuralintegrity is provided.
Conclusion-Based on f.he above evaluation,it is concluded that performing the Code-required examination is impractical. The examination, when performed to the extent practical results in 83% coverage. Based on the significant percent of coverage obtainable, it is reasonable to conclude that, degradation, if occurring will be detected. As a result, reasonable assurance of structural integrity is provided. Therefore, it is recommended that reliet be granted pursuant to 10 CFR 50.55a(a)(6)(i).
3.1.4.3 Request for Relief RI 22, Examination Category B J, item B9.31, Examination of Class 1 Pipe Branch Connection Wolds Code Requ/rement-Table IWB 25001, Examination Category B J, item B9.31 requires 100% surface and volumetric examination of oranch pipe connection welds > 4 inches NPS as definett in Figure IWB 250010.
L/censee's Code Melle/#equest-Pursuant to 10 CFR 50.55a(gM5Hiii), the licensee ,
requested relief from 100% volumetric examination of Class 1 pipe branch connection Welds FWA BJ 81, RAS BJ 10, and HBS BJ 6A.
Licensee's Basis for Requesting Relief-
"The configuration of the 8 inch branch weldolet to 18 inch pipe weld FWA BJ 81, prevents complete ultrasonic examination of the code required volume. The configuration of the weldolet prevents examination from the branch side of the weld and limits the examination to one direction from the other side. No st.pplemental examinations to increase the coverage were possible. The total volume examined is approximately 82%.
20
"The configuration of the 6 inch branch to 20 inch elbow weld RAS BJ 10, prevents complete ultrasonic examination of the code required volume. The branch connection is wcided to the extrados of the elbow. The angle between the branch connection and the elbow changes as the probe is scanned c{rcumferentially around the branch. Where the angle is most acute, the sound can not reach the root of the weld and the configuration of the branch prohibits the use of larger angles.
Supplemental examinations vare performed to increase the coverage. The total
, volume examined is approximately 82%.
"The configuration of the 4 inch branch weldo;et to 28 inch pipe weld RBS BJ 6A,
. prevents complete ultrasonic examination of the code required volume. The configuration of the weldolet prevents exarr9 nation frem the branch side of the weld and limits the examination to one direction from the other side. No supplemental examinations to increase the coverage were passible. The total
- volume examined is approximately 67%."
fJcensee's Proposed Alternative Examination-
- "In lieu of performing the Code required examinations, CNS proposes to examine i the accessible portions of branch connection welds FWA BJ 81, RAS BJ 10, and RBS BJ 6A."
l l Eyslust/on-The Code requitos that the subject branch connection welds receive 100%
volumettic and surface examinations. The licensee stated that 67% to 82% volumetric coverege of the subject examination areas can be obtained. Based on the review of the information provided, the Code required volumetric examination is impractical due to the I
configuration of the branch connections. To perform the complete volumetric examination, design modifications or replacement of the branch connections with those of a configuration that provides for complete coverage would be required, imposition of this reauirement would cause a considerable burden on the licensee.
The licensee proposes to perform the volumetric examinations to the extent practical, resulting in at least 67% volumetric coverage. Based on the coverage that can be ubtained, in combination with the surfar'e examination, significant degradation, if present, should be detected. As a result, reasonable assurance of structuralintegrity is provided.
Conclus/on-The Code reyJired volumetric examination for the subject branch connection
, welds is impractical for Cooper Station. Considering the percent of the examinations that will te completed for these br6nch connection wc!da,in conjunction with the surface examinations, it can be .;oncluded that significant degradation, if present, should be detected. As a result, reasonable assu'ance of structuralintegrity will be provided.
Therefore,it is recommended that reDet be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.5 Pump Pressure Boundary No relief requests.
21
3.1.6 Valve Pressure Boundary No relief requests.
3.1.7 General No relief requests.
3.2 Class 2 Components 3.2.1 Pressure Vessels ,
3.2.1.1 Request for Relief RI 05, Examination Category C A, item C1.30. Residual Heat Exchanger, Tubesheet to Shell Welds l Code Aeou/rement-Section XI, Examination Category C A, item C1.30, requires a 100%
volumetric examination of tubesheet to shell welds as defined by Figure IWC 2500 2.
l l L/censee's Code Melle/#eguest-Pursuant to 10 CFR 50.t7)a(g)(5)(lii), the licensee requested relief from the Code required volumetric examina.lon of the following Residual Heat Exchanger tubesheet to shell welds:
RHR Heat Exchanger 1 A, Weld No RHR CA 3A RHR Heat Exchanger 1B, Weld No RHR CA 3B licensee's Basis for Requesting Relief-
"The RHR heat exchanger tubesheet to shell welds as shown in Figure RI 05.18 are designed with a geometry that provides a corner trap for ultrasonic signals and has limited accessibility. The geometric reflectors inherent in this design prevent a meaningful ultrasonic es, amination from being performed on these welds.
"An investigation into the feasibility of performing ultrasonic examinations on the subject welds was conducted during the second ten year interval of the Inservice inspection Program for CNS. Various ultrasonic examination techniques tried during the investigation cont..uded that a meaningful ultrasonic examination can not be performed on this joint configuration. The investigation determined that the tubesheet to shell weld configuration was not accessible for performing either a -
volumetric or surface examination. As a result CNS applied for specific relief from the examination requirements of Table IWC 25001 which was granted by the NRC for the second inspection interval. *
" Based on the above. CNS requests relief from the ASME Section XI examination requirements for performing a volumetric examination of the RHR heat exchanger tubesheet to shell wwids."
- 3. Not included with this evaluation.
22
Licensee's P<oposed Attemative-
"As an alternate examination, CNS will perform a visual examination, VT 1 of the applicable weld each inspection interval. Additionally, a visual examination, VT 2, at the required frequency specified by Table IWC 25001, Category C H, will bd performed on the shell side of the best exchanger "
Eve /uet/on-The Code requires a 100% volumetric examination of the subject RHR
, tubesheet to shell welds. However, due to the geometry of the weld and associated component design, it is impractical to perform a volumetric examination to the extent required by the Code, in addition, a surface examination of the subject weld is impractical due to limited secess. The licensee has proposed a VT 1 visual examination of the subject weld. The INEL staff believes that the proposed alternative will provide reasonable assurance of operational readiness.
Conclus/on-The INEL staff has determined that the Code required examinations of the subject weld are impractical without disassembly of the component. As an alternative to the Code required examination, the licensee will perform a VT-1 visual examination on the subject welds. The INEL staff believes that this alternative,in comb! nation with the Code-required VT 2 visual examination, will provide reasonable assurance of operational readiness. Therefore,it is recommended that relief from the Code required volumetric examination be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.2.2 Piping 3.2.2.1 Request for Relief RI 23, Examination Category C F 2, item C5.51, Class 2 Piping Welds By letter dated August 5,1996, the licensee withdrew Request for Relief RI 23.
3.2.3 Pumps 3.2.3.1 Request for Relief RI 18 Examination Category C C, item C3.30, Integrally Welded Attachments to the Residual Heat Removal Pump Casings Code Requirement-Table IWC 25001, Examination Category C C, item C3.30 requires a surface examination of 100% of the length of the pump integral attachment welds as
, defined by Figure IWC 2500 5.
L/censee's Code Relief Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested ralief from performing the Code required surface examination on the integral attachment welds of Residual Heat Removal Pumps 1 A,1B,1C, and 1D.
Licensee's Basis for Requesting Relief - ,
" Specific relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety."
23
"Each RHR pump has an integrally welded attachment connecting the pump to the pump baseplate located on the underside of the pump as shown on Figure RR 181'. This weld is completely inaccessible and examination is not possible."
Licensee's Proposed Attemative-
'As an alternate examination, CNS will perform a VT 2 visual examination of the applicable pump and baseplate in conjunction with Class 2 system pressure test .
required by Category C H."
l Eve /uetion-The Code requires a 100% surface examination of the subject integral attachment welds. However, based on review of the figure provided by the licensee,it is eviuant that the design of the base plate support and location of the welded attachments make the surface examination impractical. To perform the Code required surface
. examinations, design modification to allow access for examination of the supports and welded attachments would be required, imposition of this requirement would cause a considerable buiden on the licensee. Examination of other welded attachments and the VT 2 Code requirad visual examination performed on the subject RHR pump attachment welds will detect significant degradation, if present. As a result, reasonable assurance of operational readinass will be provided.
Conclusion-Surf ace examination of the subject integral attachments welds is impractical.
Examination of otbei welded attachments and the VT 2 Code required visual examination performed on the subject RHR pump attachment welds will detect significant degradation, if present. As a result, reasonable assurance of operational readiness will be provided.~
Therefore,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.2.4 Valves No relief requests.
3.2.5 General No relief requests.
3,3 Class 3 Components No relief requests.
- 4. Not included with this evaluation.
24
. i 3.4 Pressure Tests t
i !
- 3.4.1 Class 1 System Pressure Tests 3.4.1.1 Request for Relief PR 02, Revision 2, Examination Category S P, item 815.10, ?
- Reactor Vessel Pmssure Retalning Boundary l 9 .- Code #e$;.=st-Section Xi, Table IWB 25001, Examination Category B P, item '
h 815.10 requires a VT 2 visual examination during system pressure tests and system !
, hydrostatic tests of the Class 1 reactor vessel pressure retaining boundary components -
each refueling outage, l
"a Licensee's Code #P#e/#eguest-Pursuant to 10 CFR 50.55a(a)(3)(ll), the licensse - !
proposed an alternative to the Code required system pressure test of the RPV pressure retaining boundary.:
+
, licensee's Basis- I
- "It is impracticable to perform a system laakage test during actual reacter startup.-
[ The high dose rates associated with reactor operation, the temperature levels in the !
.drywell, and that large amount of piping and components to be examined makes this examination a hardship. In lieu of a system leakage test during reactor startup-a hydrostatic test is performed at the pressure associated with 100% rated reactor'
, power.
4 "Section XI states that the boundary for the system leakage test shall be the j, reactor coolant pressure boundary with all valves in the position required during reactor startup. It is impracticable to extend the boundary this far during pressure -
testing._
"a) Three'of the four Feoowater Chack valves will be closed for the system pressure test followin'g a refueling outage. The inboard check on one Feedwater line is kept l e
. open by Reactor Water Cleanup (RWCU) flow. The RWCU system is kept in service during the pressure tests. The outboard check valves are the class 1 r boundary valves.
"b)' The four outboard Main Steam isolation Valves (MSIV) will be closed for the ;
. - system pressure test _and the ten year hydrostatic pressure test. The inboard l
- MSlV's are opened to pressurize the system to the outboard valves. The outboard ;
i -
MSIV's are the class 1 boundary valves.
4 "c) Both High Pressure Coolant injection (HPCI) and both Reactor Core Isolation Cooling (RCIC) steam supply valves will be closed for the system pressure test following a refueling outage. These velves close automatically on low steam supply pressure. Dur ng tho ten year hydrostatic pressure test, the system will be 4 - pressurized te the outboard valves. The' outboard valves are the cless 1 boundary i valves.
. , 25 i N
mer - - * . . - , m,-,-.E- - c..+--., m--.,
"In addition to the extensive line ups required to perform a system leakage test of this mognitude, special measures would be required to temporarily support steam lines due to the excess weight of the water.
"The position of the valves for the system leakage test as described above is consistent with the intent of footnote 1 to Table IWB 25001, Category B P.
Abnormallineups and installation of jumpers is not required for the system leakage test. The valves described above are normally open during a reactor startup. In ,
order to pressurize the reactor coolant pressure boundary for testing, these valves must be closed. Except as described above, the Class 1 boundary is pressurized as required by the Code. The VT 2 inspection includes the entire reactor coolant pressure boundsry.
"Since the portions of the piping between the valves described above are operated at or above reactor pressure during normal operation, any through wallleakage would be detected by the drywellleakage collection system, or by operations personnel on normal rounds.
"In summary, it is unsafe to perform e VT 2 visual examination during actual reactor startup durine reactor operation and it is impracticable to perform a system leakage test of the reactor coolant pressure boundary with all valves in the position required for normal reactor startup.
" Based on the above, CNS requests relief from the ASME Section XI requirements for performing a system leakage test using the boundaries stated in Note 1 of Table IWB 25001."
ikensee's Proposed Altemative-
"The Feedwater Check valves, the outboard MSIV's, the Main Steam Line Drain valves, and HPCI and RCIC Steam Supply Valves will be closed during the system leakage test, but will be included in the VT 2 examination. A VT-2 visual examination will be performed during the system leakage test at a pressure not less than that associated with 100% rated power.
"In addition, a system hydrostatic test will be performed once per interval at a pressure not leus than that associated with 100% rated reactor power. The pressurization boundary and the VT 2 visual examination for this test will extend to .
the Class 1 boundary."
Eva/ust/on-Paragraph IWB 5221(a) requires that system leakage tests be performed at a test pressure not less than the nominal operating pressure associated with 100% rated reactor power. To c.'atain nominal operating pressure for CNS, the reactor must achieve 100% power.
To perform a system leakage test at 100% power for nonisolable portions of a system is a major effort requiring many manhours from skilled maintenance and inspection personnel while causing excessive radiation exposure and personnel safety concerns. As 26
I 1
l l
an alternative to the system pressure test, the licensee proposes to perform a VT 2 visual examination during the system leakage test at a pressure not less than that associated with 100% rated power and with systems in their normal line up to the extent practical, extending to the outermost valves in these systems. The INEL staff believes that the proposed alternative will provide an acceptable level of safety and quality.
The INEL staff believes that requiring the licensee to perform a system pressure test at
, 100% reactor power will result in a hardship without a compensating increase in quality and safety. Based on the proposed alternative, reasonable assurance of operational readiness will be provided.
Conclus/on-For CNS requiring a VT 2 visual examinatio e at 100% reactor pewer would resuit in a hardship without a compensating increase in quality and safety. A VT 2 visual examination performed during the system leakage test at a pressure not less than that associated with 100% rated power wili povide reasonable assurance of the continued operational readiness of mechanical connections, extending to the Class 1 boundary.
Therefore, it is recommended that the proposed alternative, VT 2 visual examination during the system leakage test at a pressure not less than that associated with 100% rated power, be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
3.4.2 Class 2 System Pressure Tests 3.4.2.1 Request for Relief PR 04, Examination Category B P, items B15.50 and B15.51, Pressure Testing of the Reactor Pressure Vessel (RPV) Hand F:hnge Seal Leak Detection System Code Requ/rement-Section XI, Table IWB 25001, Examination Category B P, items B15.50 and B15.51 require a VT 2 visual examination during system leakage and system hydroststic tests of Class 2 components.
Licens, rs Code Re/lef Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code required system pressure test of the RPV Head Flange Seal Leak Detection System, Line Number 1 MS 1521.
Licensee's Basis for Requesting Relief-
"The Reactor Vessel Head Flange Leak Detection Line is separated from the reactor
, pressura boundary by one passive membrane, a silver plated 0 ring located on the vessel flange. A second 0 ring is located on the opposite side of the tap in the vessel flange (See Figure PR-04.1)'. This line is required during plant operation in
~dar to indicate failure of the inner flange seal 0 ring. Failure of the 0 ring would result in the annunciation of a High Level Alarm in the control room. Upon receipt of this alarm, control room operators would quantify the leakage rate from the O ring and then isolate the leak detection line from the drywell sump. Failure of the inner 0 ring is the only condition under which this line is pressurized.
- 5. Not included with this report.
27
"The conf!guration of this system precludes hydrostatic testing while the vessel head is removed because the odd configuration of the vessel tap coupled with the high test pressure requirement (1000 psig minimum), prevents the tap in the flange from being temporarily plugged. Adequate testing cannot be performed when the head is installed because the seal prevents complete filling of the line, which has no available vent. Operational testing of thic line is precluded because the line will only be pressurized in the event of a failure of the inner 0 ring. It is impracticable to purposely fail the inner 0 ring in order to perform a pressure test. . ,
" Based on the above, CNS requests relief from the ASME Section XI requirements for static and operational pressure testing of the Reactor Vessel Head Flange Seal -
Leak Detection System.'
Licensee's Proposed Alternative-
"A VT 2 visual examination will be performed on the line when the reactor cavity is flooded during refueling outage. The hydrostatic head developed due to the water above the vessel flange during refueling outages will allow for the detection of any gross indications in the line. This examination will be performed with the frequency specified by Table IWB 25001 for an IWB 5221 test (once each inspection period).*
Evaluation-The Code requires that system pressure tests be conducted for those systems required to function during normal plant operation. The RPV Head Flange Leak Detection Line is pressurized only when the inner 0 ring fails. The design of this line makes the Code required system pressure test impractical. To perform the system pressure test in accordance with the requirements, the RPV Head Flange Leak Detection System and the RPV flange would have to be redesigned, f abricated, and installed.
The licensee has committed to perform a VT 2 visual examination on the RPV Head Flange Leak Detection Line when the reactor cavity is flooded. The proposed alternative will provide adequate assurance that gross inservice flaws in the subject line will be detected if any have developed.
Conclus/on-The systerr) pressure test required by Section XI for the subject Class 1 line is impractical because of the possibility of damage to the RPV Head flange 0 ring seats. The VT 2 visual examination of the RPV Head Flange Leak Detection Line when the reactor cavity is !!ooded will provide reasonable assurance that if gross inservice flaws have -
developed in the subject line, they will be detected. Therefore,it is recommended that '
rclief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.4.2.2 Request for Relief PR 05, Examination Category C H Items C7.30 through C7.80, Class 2 Containment Penetration Piping and Valves Code Requ/rement-Table IWC 25001, Examination Category C H Items C7.30 through C7.80 require a system pressure test, inservice or functional (IWC 5221), and a system hydrostatic test (IWC 0222) in conjunction with a VT 2 visual examination of pressure-retaining components. Ol ping, and valves.
28
- i L/censee's Code Relie/ Request-Pursuant to 10 CFR 50,55a(a)(3)(i), the licensee '
proposed an alternative to the Cods required tests for the following piping that penetrates the containment and is attached to non Code Class piping.
System Penetration Number instl Air 22,30E,30F,33E,33F PCC, NI, & SBNI 2,25,26,45C,SIE,51F,203A,203B,205,220 RPVINST 40A, 408, 40C, 40D, 47(All)
Reactor Recire. 29E, 209 A, 2098, 209C, 2090, 213 A 213B, 215, 229A, 2298, 229C, 229 D, 229E, 229F, 229G, 229 H, 229), 229K, 229M Drains 18,19,43,44 Domin Water 20 H202 36,203A,203B REC 23,24 Service Air 21 TIP -
35A,358,35C,35D,35E licensee's Basis-
"The portion of piping that penetrates containment and the associated inboard and outboard containment isolation valves are required to be constructed in accordance with Class 1 or Class 2 design requirements, in instances where the piping penetration is for a nonsafety related system, the sole safety function of the penetration piping 9nd associated valves is to provide conta% ment isolation and maintain containment integrity la the event of a failure of the attached nonrafety-related piping, in all cases during normal plant operation, the isolation valves associated with these penetrations are maintained in the locked closed position, are administratively closed (controlled procedurally), or they close automat!cally upon receipt of a containment isolation signal or on Ims of flow. The integrity of these punetrations is verified by 10 CFR 50, Appendix J, leakage testing,
" Additionally, Code Case N 522, ' Pressure Testing of Containment Piping Section XI, Division 1,' has determined that pressure testing of these containment penetration: par 10 CFR 50 Appendix J,la an acceptable alternative to the requirements of Table IWC 25001 Category C H,
" Performing system pressure tests each inspection period and a hydrostatic test each inspection interval per Section XI, Table IWC 25001, Category C H, is redundant to Appendix J testing. Code pressure testing provides no commensurate increase in quality or safety with cost benefit, Pressure testing of piping in nonsafety related systems penetrating containment pursuant to requirements of 20
10CFR50, Appendix J,in lieu of Section XI pressure testing requirements provides an acceptable level of quality and safety."
Lice.1see's Proposed Alternative-
"As an siternative to Section XI pressure testing requirements for piping penetrating containment attached to a nonsafety related system, CNS will adopt the provisions of ASME Section XI Code Case N 522.
- Pressure testing of the below listed penetrations'shall be performed in accordance with the requirements and frequency specified in 10CFR50, Appendix J, in lieu of the additional requirements specified in Table IWC 25001,
- Category C.H. The test will be conduc'ied et the peak calculated containment pressure. If the leak rate exceeds the Appendix J acceptance criteria, additional measurer will be applied to verify that the test failure is not due to a through wall leak.
Eve /ust/on-The licenseo proposes to implement the alternatives contained in Code Case N 522, Pressure Testing of Containment Penetration Piping, in lieu of the Code roquired pressure tests for portions of the subject lines that are Class 2 at the containment penetration. The licensee stated that the test will be conducted at the peak calculated containment pressure, if the leak rate exceeds the Appendix J acceptance criteria, additional measures will be applied to verify that the test failure is not due to a through-Walllesk. These segments of lines are safety related only because they function as part of the containment pressure boundary and are relied on for containment integrity. Therefore, it is logical to test the penetration piping portion of the associated systems to the containment test criteria found in 10 CFR50.55a Appendix J.
Appendix J pressure tests verify the leak tight integrity of the primary reactor containment and of systems and components that penettete centainment by localleak rate and integrated leak rate tests, in addition, Appendix J test frequencies provide assurance that the containmen't pressure boundary is being maintained at an acceptable level while monitoring for deterioration of seals, valves, and piping.
The Class 2 containment isolation valves (C!Vs) and connecting pipe segment must withstand the peak calculated containment internal oressure related to the maximum design containment pressure. The INEL finds that the pressure retaining integrity of the CIVs and connecting piping and their essociated safety functions may be verified with an .
Appendix J, Type C test if conducted at the peak calculated containment pressure. The seal between the connecting pipe segment and containment may be verified using an Appendix J. Type B test. Therefore, wher' tM connecting pipe segment is subjected to either : .se B or C test,its safety function is verified by the Appendix J test.Section XI, IWC 521' t requires that where air or gas is used as a testing medium, the test procedures whallinclude methods for detection and location of through wall leakage in components of the system tested. Because an Appendix J, Type C test most likely uses
- 6. See Table under Licensee's Code Relief Request. l 30
_ . _ _ . . _ __ _ _ _ - _ _ _ _ _ . _ _ - _ . _ _ _ ~ _ _ _
1 air as a testing medium, the licensee's test procedure should meet the above requirement for the CIVs and pipe segments between the CIVa.
Conclus/on~The INEL staff believes that an acceptable level of quality and safety will be pruvided by Appendix J tests when the licensee performs the leak test at the peak calculated containment design pressure and implements e test procedure that provides for detection and location of through wall leakages in the pipe segments that are being tested.
therefore, pursuant to 10 CFR 50.55ala)(3)(i), it is recommended that the licensee's proposed altamative be authorized. The use of the Code case in combination with the licensee's proposal should be authorized for the current interval or until such time as the
. Code Case is published in a future revision of Regulatory Guide 1,147. At that time,if the licent4e intends to continue to implement this Code Case, the licensee is to follow all provisions in Code Case N 522 with limitations issued in Regulatory Guide 1,147, if any, 3.4.2.3 Request for Relief PR 06,lWA 5244(b), Ex6mination Category D A, item D1.10, VT 2 Visual Examination of Redundant Systema for Buried Components Code #equ/rement-lWA 5244 (b) requires that for redundant systems where the buried components are nonisolable, the VT 2 visual examination shall consist of a test that determines the change in flow between the ends of the buried pipe.
L/censee's Code Me//e/#equest-Pursuant to 10 CFR 50.55ata)(3)(i), the licensee proposed an alternative to the flow measurement requirements of the Code for the buried service water critical supply headers leading from the service water building to the control building, l/censee's Bes/s-
- For the buried portion of the service water critical supply headers, isolation valves
- are installed in this rec'undant system. Buried components in redundant systems that are isolable are not addressed in lWA 5244. However, leakage testing of the buried piping is impracticable because the isolation valves located in the service water building and the control building that isolate the buried piping are large butterfly valvas which are not suitable for performing a pressure isolation function.
Each critical header supplies two RHRSW booster pumps, one REC heat exchanger and one diesel generator. A butterfly isolation valve is installed in the main header in the service water building and in ea:h of these branch supply lines in the control building, However, since these valves are not designed to be leak tight, these five butterfly valvos would provide multiple leakage paths. Leakage testing of this buried piping and determining the rate of pressure loss would require extensive valve seat maintenance and would not provide conclusive test results.
- Current Code rules allow determining a change in flow between the ends of the buried components llWA 5244(a) and 5244(bil. Flow instruments are installed in the service water lines in the control building, However, no flow instrumentation is installed in the system upstream of the buried piping, Accurate flow measurements using temporary flow instrumentation (e.g., ultrasonic flow meters) are not possible due to insufficient runs of straight pipe between the pump discharge and the buried 31
,. . .- - . l. - ---- -. -. -- - -- - -
piping. Therefore, direct measurement of the change in flow between the ends of the buried piping is not practical.
'The instellation of permanent flow instrumente would require significant system modifications which would be burdensome. The cost of these modifications, when weighed against the benefits, are not justifiable. The following proposed alternative would provide reasonable assurance that any significant leakage from the buried piping will be detected." .
Licensee's Proposed Altemative-
"In lieu of performing a visual examination VT 2 in accordance with IWA 5244, CNS shall utillre existing plant instrumentation for the determination of buried pipe integrity. Discharge pressure is indicated by pressure gauges provided at each individual pump (SW Pl 360 A, B, C, & D). Servico water pumps A & C discharge to a common header, as do pumps B & D. Each header is provided with pressura indication prior to exiting the intake stiucture (SW PI 383 A&B). When these headers resurf ace in the control building, pressure indication (SW Pl 384 A & B) and flow indication (SW F1385 A & B and SW Fl 364 A & B) are provided.
"The integrity of the buried piping is verified during qu6fterly pump testing. Using the downstream flow instruments, flow rate is act at the fixed test reference value and documented in the test record. The pump discharge pressure is then measured and used to datermine the head produced by the pump. Head and flow rete are interdependent variables which, together, define pump hydraulle performance. As the pump degrades, the developed head will decrease at the reference flow rate.
However, due to the location of the flow rate instruments (downstream of the buried piping), a decrease in pump head during testing may also indicate side-stream leakage into the isolated non critical header or through wall leskoge in the buried portion of the service water system piping. This is because the head developed by the pump decreases as flow rate increases. Significant through wall leakage would be evident because the total flow rate would increase even though the downstream i.ndicated flow rate is set at the reference value. Therefore, a satisfactory quarterly service water pump test elso verifies the integrity of the buried system supply piping.
"Should the pump test results fallin the required action range of the Code, then additional tests and evaluations will be performed to determine whether the ,
unsatisf actory test results are due to side stream leakage past butterfly isolation valves, degraded purap performance, or through wall leakage."
Evaluatlon-The Code requires a VT 2 visual examination consisting of a test that determines the change in flow between tho ends of the buried pipe. The licensee has proposed, as an alternative, to test this buried pipe in conjunction with quarterly testing of pumps. This test verifies flow by use of the existing monitoring systems for pressure and flow rate. Significant through wall leakage would be detected in a similar manner to that expected with the Code required flow measuremont test. Based on the licensee's 32
4 proposed alternative, the INEL staff believes that an acceptable level of quality and safety will be provided.
Conclusion-The licensee has proposed to test the subject piping in conjunction with quarterly testing of pumps. This test will verify that the subject p; ping will perform its design function, thereby providing an acceptable level of quality and safety. Therefore,it is recommended that the proposed alternative be authorized pursuant to
. 10 CFR 50.55ata)(3)(i).
3.4.2.4 Request for Relief PR-09,lWA 5211(d), Examination Category C H, item C7.40, Hydrostatic Test of HPCI and RCIC Discharge Pip!ng Code Requ/rement- lWA 5211(d) requires that preuure retaining components within each system boundary be subjected to a system hydrostatic pressuro test.
L/censee's Code Melle/#eguest-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Class 2 hydrostatic pressure test requirements for the HPCI and RCIC discharge piping between the injection check valve and the upstream isolation valve.
l Licensee's Basis-
"The HPCI and RCIC systems discharge through check valves into separate loops of the Feodwater system, and then through the associated Feedwater check valves to the fleactor Vessel. The Class 2 to Class 1 boundaries of these systems are the HPCI and RCIC check valves. Since tnese valves will open when the system pressure is higher than the reactor pressure, these valves can not serve as the boundaries for the Class 2 system hydrostatic pressure tests. The maintenance j isolation valves on the Feedwater system can not be used for the test boundary I since they are located between the inboard Feedwater check valves and the reactor. The Class 2 hydrostatic test pressure would exceed the Class 1 hydrostatic test pressure for these components.
"There is an isolation valve on each system upstream of the HPCI and RCIC injection check valves. These valves can be used for the Class 2 pressure test boundaries. The piping down stream of these valves up to the check valves can be tested with Class 1 piping."
Licensee's Proposed Altemative-
"In lieu of performing a Class 2 system hydrostatic pressure test on the HPCI and RCIC discharge piping between the injection check valve and the upstream isolation valve, CNG shall hydrostatically pressure test these sections of piping each interval with the Class 1 pressure test per Table IWB 2500, Category B P and Code Case N 498.*
Evaluation-The licensee has requested relief from the Class 2 hydrostatic test requirements for the portions of HPCI and RCIC system piping between the injection check valves and the upstream isolation valves. The licensee proposes to test these portions of 33
~
~
the subject systems in conjunction with the Class 1 piping in accordance with Code Caso N 498. Because the licensee's proposed alternative, to test the subject portion ci piping in N conjunction with the Class 1 piping, will subject the subject Class 2 piping to pressures equal to the Class 1 piping, the INEL staff believes that an acceptable level of quality and safety will be assured.
Conclus/on-The licensee proposes to pressure test portions of the Class 2 HPCI and RCIC piping in conjunction with Class 1 pressure tests. This test will provide an acceptable level ,
of quality and safety. Therefore,it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
3.4.3 Class 3 System Pressure Tests 3.4.3.1 Request for Relief PR 08, Examination Category D A, item D1.10 Hydrostatic Pressure Testing for the Main Steam Relief Valve Discharge I.Ines Code Requirement-Section XI, Table IWD 25001, Examination Category D A, item D1.10 requires c inservice pressure test and/or a system hydrostatic test to be performed.
IWD 5222(f) states that for safety or relief valve piping that discharges into the containment pressure suppression pool, a pneumatic test (at a pressure of 90% of the pipe submergence head of water) that demonstrates leakage integrity shell be performed in lieu of a system hydrostatic test.
L/censee's Code Relief Request-Pursuant to 10 CFR 50.b6a(a)(3)(i), the licensee proposed an alternative to performing a VT 2 visual examination of the Main Steam Relief Valve discharge piping under normal plant operating conditions and from the hydrostatic test requirements to perform a pneumatic test at 90% pipe submergence head once every inspection interval.
Licensee's Basis-
"These relief valves are currently actuated once each operating cycle commensurate with Reactor Vessel pressure > 100 psig. Suppression Pool temperature and levels monitored during this test substantiate the integrity of the discharge piping by its ability to direct flow from the relief valve to the suppression pool.
"The Code required 10 year pressure test of the discharge piping with a pneumatic test at a pressure of 90% of the pipe submergence head of water equates to an appi;ed pressure of approximately 1.17 psig equivalent to the 3 feet of submerged piping.
- This Code requirement has been removed from the 1994 Addenda of ASME Section XI 1992 Edition.
" Current test parameters significantly exceed Code requirements in piping pressurization and frequency. Performance of the current Code required testing 34
4 would not increase the margin of asswance for safety beyond current test parameters, and would only serve as a redundant inferior test requirement."
L/censee's Proposed Afternat/ve-
'In lieu of performing a hydrostatic pressure test at a pressure of 90% of the pipe submergence head of water, as required by IWD 5222(g), CNS shall use existing plant surveillance tests of the operability of each Main Steam Relief Valve to
, demonstrate the integrity of the discharge piping."
Eve /uat/on-The Code requires that safety or relief valve piping that discharges into the containment pressure suppression pool receive a pneumatic test (at 90% of the pressure of the pipe submergence head of water) that demonstrates leakage integrity in lieu of a system hydrostatic test. The licensee has noted that 90% of the pressure of the pipe submergence head of water is approximately 1.17 psig for the subject system. This pressure is considered insignificant. The licensee submits that the existing plant surveillance test of the operability of each main steam relief valve subjects the piping to much greater pressure than that required by Code. Based on this information, the INEL staff believes that the license 6's proposed pressure test of the subject piping, in conjunction with the relief valve testing, provides an acceptable level of quality and safety.
Conclus/on-The main steam relief valve lif t test subjects the discharge lines to a pressure higher than that associated with the submergence head. Verification of discharge lin6 integrity at the next opportunity in conjunction with the lift test provides reasonable assurance of operational readiness. Therefore, it is recommeno,d that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
3.4.4 General 3.4.4.1 Request for Relief PR 01,IWA 4700(a) and (b), Alternative Pressure Test t
Requirements For Code Class 1, Class 2, and Class 3 Systems Following Repair, Replacements, and Modifications Code Requ/rement-lWA 4700(a), Pressure Test, requires that, after repair by welding on the pressure retaining boundary, a system hydrostatic test shall be performed in accordance with IWA 5000. IWA 5214, Repairs and Replacements, requires that a repaired or replaced componant be pressure tested prior to resumption of service if required by IWA 4400 and IWA 4600. The test pressure and temperature for a system hydrostatic test subsequent to component repair or replacement shall be the system test pressure and temperature specified in IWB 5222, IWC 5222, or IWD 5223, as appropriate for the system that contains the repaired or replaced component.
L/censee's Code Melle, Taquest-Pursuant to 10 CFR 50.55ata)(3)(ii), the her nsee proposed an alterr.ative 'o performing a hydrostatic test following repairs, replacements, and modifications on Ccta Class 1, Class 2, and Class 3 systems.
35
Licensee's Dosis-
" Elevated pressure hydrostatic tests are difficult to perform and often represent a true hardship. Some of the difficulties associated with the elevated pressure testing include the following
- Hydrostatic testing of ten requires complicated or abnormal valve line ups in order to properly vent, fill, and isolate the component requiring testing.
- Re5ief valves with setpoints lower than the hydrostatic test pressure must be gagged or removed and blind flanged. This process requires the draining and refilling of the system. -
- Valvos that are not normally used for < solation (e.g., normally open pump discharge valves) are often required to provide pressure isolation for an elevated pressure hydrostetic test. These valves frequently require time consuming seat rnaintenance in order to allow for pressurization.
- The radiation exposure required to perform a hydrostatic pressure test is high (in comparison to operational pressure testing) due to large amount of time required to prepare the volume for testing (i.e. installing relief valve gags, performing appropriate valve line ups, etc.)
"The difficulties encountered in performing a hydrostatic pressure test are >
prohibitive when weighed against the benefits, industry experience, which is corroborated by CNS's experience, shows that most through wallleakage is detected during system operation as opposed to during elevated pressure tests, such as the ten year hydrostatic tests.
'Little benefit is gained from the added challenge to the piping system provided by an elevated pressure hydrostatic test (when compared to an eperational test),
especially when one considers that the piping stress experienced during a hydrostatic test does not include the quite significant stresses affiliated with the thermal growth and dynamic loading associated with design basis events. As an industry, it has been histodcally documented that leakage will occur and be detected at nominal operatng pressures of a system. Elevating pressure 10 25%
has no meaningful impact.
"Use of hydrostatic test defenals, which are presently allowed per Code Case N 416 for Class 2 components,is not a satisfactory solution because the required test must be eventually performed, and it is the performance of the test itself that is burdensome.
- These arguments are also supported by NRC endorsement of Code Case N 498,
" Alternative Rules for 10 Year Hydrostatic Pressure Testing for Class 1 and 2 Systems,Section XI, Division 1". This relief request is a logical extension of that Code Case.
36
)
" Based on the above, CNS requests relief from the ASME Section XI requirements for performing elevated pressure hydrostatic tests on Class 1,2, and 3 repaired / replaced components."
Licensee's Proposed Attemative-
"CNS proposes to perform pressure testing on Clacs 1,2, and 3 repaired / replaced components in accordance with the requirements of ASME Section XI Code Case
, N 4161. This Code Case offers an acceptable attemative to Section XI requirements. In addition to the NDE requirements of the Code Case, MS will perform a surface examination of the root pass layer of a repair or replacement
. weld on Class 3 components in accordance with the NDE requirements of ASME Section 111.
"With the pressures currently required by Section XI, elevated pressure hydrostatic tests do not offer a commensurate increase in safety with cost benefit and places undo burden upon a licensee to perform these tests."
l '
Evaluarlon-Section XI of the Code requires a system hydrostatic test to be performed in l accordance with IWA 5000 after repairs by welding on the pressure retaining bour 2ty.
1he licensee proposes to implement the alternative to hydrostatic pressure tests contained in Code Case N 4161 for Code Class 1,2, and 3 repairs / replacements. In addition, for L Class 3 repair / replacement welds or welded areas, the licensee will supplement the pressure test with an additional surf ace examination on the root pass layer.
Hardships are generally encountered in the performance of hydrostatic testing in accordance with the Code. For example, special equipment, such as temporary attachment of test pumps and gages, and unique valve lineups are often needed due to the test pressure, which is higher than the nominal operating pressure. Hydrostatic testing only subn is the piping components to a smallincrease in pressure over the design pressure and, therefore, does not present a significant challenge to pressure boundary integrity. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leak detection during the examination of components under pressure. rather tnan as a measure of the structuralintegrity of the components.
Code Case N 4161 specifies that NDE of the welds be performed in accordance with the applicable subsection of the 1992 Edition of Section Ill. This Code Case also allows a VT 2 visual exarnination to be performed at nominal operating pressure and temperature in conjunction with a system leakage test. in accordance with paragraph IWA 5000 of the 1992 Edition of Section XI,in lieu of the hydrostatic test. The 1989 Edition of Sections 111 and XI are the latest Code editions referenced in 10 CFR 50.55a. Comparison of the system pressure test requirements of the 1992 Eoition of Section XI to those of the 1989 Edition of Section XI shows that:
- 1) The test frequencies and the pressure conditions associated with these tests have not changed:
- 2) The hold times have either remained unchanged or increased:
37
I i
l 3
- _ 3) The terminology associated with the system pressure test requirements for all three Code clastss has been clarified and streamlined; and
- 4) The NDE requirements for welded repairs remair' the same.
Following welding, the Code requires volumetric examination (depending on wall l thickness) of repairs or replacements in Code Classes 1 and 2 piping components, but only requires a surface examination of the final weld past la code Ciets 3 components. There are no ongoing NDE requirements for Code Class 3 components except for VT 2 visual examination for leaks in conjunction with the 10 year hydrortatic tests and the periodic ;
pressure tests.
i Considering the NDE performed on Code Class 1 ano 2 systems and considering that i hydrostatic pressure tests rarely result in pressure boundary leaks that would not occur during system leakage tests, the INEL staff believes that increased assurance of the integrity of Class 1 and 2 welds is not commensurate with the burden of performing hydrostatic testing. Further, it is also believed that the added assurance provided by a hydrostatic test of Class 3 welds is not commensurate with the burden of performing hydrostatic testing when a surface examination is performed on the root pass layer of butt and socket welds and a system pressure test is performed.
Conclus/on-Compliance with the Code hydrostatic testing requirements for welded repairs' >
or replacements of Code Class 1,2, and 3 components would result in hardship without a !
compensating increase in the level of quality and safety. Therefore, it is recommended that the proposed alternative be authorized, pursuant to 10 CFR 50.55a(a)(3)(ii), including the licensee's proposal to pcrform a surface examination on the root pass layer of Class 3 butt and socket welds. Use of Cods Case N 4161, with the licensee's proposed -
augmented surf ace examination, should be authorized for the current interval or until such
- time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time,if the licensee intendr to continue to implement this Code Case, the licensee should follow all the provisions in Code Case N 4161, with lim 9ations issued in Regulatory Guide 1.147, if any.
1.4.4.2 Request for Relief PR 07, Examination Categories B P, C H, and D A,10 Year Hydrostatic Pressure Test Requirements for Cless 1,2, and 3 Systems
- Code Requirement-lWA 5211(d) requires a system hydrostatic pressure test at an elevated pressure and accompanying VT 2 visual examination at least once aach inspection interval.
L/censee's Code Relief Aequest-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an alternative to performing the Code required hydrostatic tests at elevated
- pressures for Cless 1,2, and 3 systems.
38
_ _ __-__ _ a __ - - - . _ . _ ~ _
Licensee's Basis-
"ASME Code Case N 498 currently provides an alternative for Class 1 and 2 system hydrostatic testing allowing use of a reduced pressure equal to system nominal operstmg pres 3ure. Recently published Code Case N 4981, while repeating these alternative pressure requirements for Class 1 and 2, also adopted and included rules for Class 3 systems. Also, Code Case N 4981 clarified the intent of using installed plant instrumentation without the need for test gauging or imposing the requirements of IWA 5260 when p9tforming these nominal operating pressure tests.
- lt is CNS's position the' conducting system pressure tests on Class 1 and 2 systems consistent with the requirements of N 4981,in conjunction with performing the applicable volumetrit:, surf ace, and visual examinations in accordance with the owners ISI Program, provides a level of quality and safety equivalent to, or greater than, that provided by the Code hydrostatic test pressure and instrumentation requirements.
'CNS employs a very active erosion / corrosion monitoring and control program which periodically measures wall thickness in selected Class 3 piping and I components. This program primarily focuses on those portions of piping which are l most susceptible to erosion, microbiologically influenced corrosion (MIC) and other
} identified corros!on mechanisms which are inherent to the service water and like systems. The screening criteria for selection of piping and components to be chosen for " Thickness Examination" includes: (1) sections susceptible to wall thinning by erosion (2) low flow sections, and (3) intermittent or no flow sections.
- lt is CNS's intention to select those portions of piping and components for examination most susceptib!a to erosion and corrosion thereby giving a conservative representation of overall pressurc boundary integrity.
- lt is CNS's position that performing system pressure tests on Class 3 systems consistent with the requirements of N 4981, together with augmented test p*ograms (e.g. erosion / corrosion monitoring for piping determined to be most susceptible to erosion and corrosion), provides a level of quality and safety equivalent to, or greater than, that provided by the Code hydrostatic test pressure and instrumentation requirements."
Licensee's Proposed Allemative-
"As an alternative to existing Section XI requirements, CNS will adopt the provisiare of Code Case N 4981.
- In lieu of performing a hydrostatic pressure test at a pressure above nominal operating pressure or system pressure for which overpressure protection ir required, as required by Table IWA 52101, Examination Categories B P, C H, D A, D B, and D C, a system pressure test at nominal operating pressure and temperature shall be performed.
39
"In lieu of instrumentation requirements specified in IWA 5260, existing plant instrumentation will be used per IWA 5212(b), Where gaging may be required and does not exist, the rules of IWA 5260 shall be used. For Class 3 Systems, CNS shall also continue to maintain and implement an erosion / corrosion monitoring program for piping determined to be most susceptible to erosion and corrosion, as previously described."
Evs/ustion-The Code requires the performance of a system hydrostatic test once per ,
intervalin accordance with the requirements of IWA 5000 for Class 1,2 and 3 pressure-retaining systems. In lieu of the Code required hydrostatic testing requirements, the licensee proposes to implement the alternatives to Code requirements contained in Code Case N 4981, Alternative Rules for 10 Year System Hydrostatic Testing for Class 1,2, and 3 Systems, dated May 11,1994.
The system hydrostatic test, as stipulated in Section XI,is not a test of the structural integrity of the system but rather an enhanced leakage test (Reference 15). Hydrostatic testing only subjects the piping components to a smallincrease in pressure over the design pressure; therefore, piping dead weight, thermal expansion, and seismic loads present far greater challenges to the structuralintegrity of a system. . Consequently, the Section XI hydrostatic pressure test is primarily regarded as a means to enhance leak detection during the examination of components under pressure, rather than as a method to determine the structural integrity of the components. In addition, the industry experience indicates that leaks are not being discovered as a result of hydrostatic test pressures causing a preexisting flaw to propagate through the wall-in most cases leaks are being found when the system is at noimal operating pressure.
Code Case N 498, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1 and 2 Systems, was previously approved for general use on Class 1 and 2 systems in Regulatory Guide 1.147, Rev. 9. For Class 3 systems, Code Casm N 4981 specifies requirements identical to those for Class 2 components (for Class 1 and 2 systems, the alternative requirements in N 4981 are unchanged from N 498), in lieu of 10 year hydrostatic pressure testing at or near the end of the 10 year interval, Code Case N 4981 requires a VT 2 visual examination at nominal operating pressure and temperature in conjunction with a system leakage test performed in accordance with paragraph IWA-5000 of the 1992 Edition of Section XI.
Class 3 systems do not normally receive tha amcr.t and/or type of nondestructive ,
examinations that Class 1 and 2 systems reep,ve. While Class 1 and 2 system failures are relatively uncommon, Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a review of Class 3 system failures requiring repair during the last 5 years,' the most common causes of f ailure are erosion-corrosion (EC),
microbiologically induced corrosion (MIC), and general corrosion, in general, licensees have implemented programs for the prevention, detection, and evaluation of EC and MIC;
- 7. Documented in Licensee Event Reports and the Nuclear Plant Reliability Data System databases. l 40 1
+
therefore, Class 3 systems receive inspection commensurate with their functions and expected f ailure mechanisms.
System hydrostatic testing entails considerabis time, radiatinn dose, and money. The safety assurance provided by the enhanced leakage detection gained from a slight increase in system pressure during a hydrostatic test may be offset or negated by the necessity to gag or remove safety and/or relief valves (placing the system, and thus the plant,in an off-
, nortnal state), erect tamporary supports in steam lines, and expend resourc es to set u9 testing with special equipment and gages. Therefore, system hydrostatic testing represents a considerable burden. Giving consideration to the minimal amount of increased assurance provided by the increased pressure associated with a hydrostatic test over that of a system leakage test, and the hardship associated with performing the Code-required hydrostatic test, the INEL staff finds that compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety.
Conclus/on-CompUance with the Section Xl hydrostatic testing requirements results in hardship and/or unusual difficutty without a compensating increase in the level of quality and safety. Performing a system leakage test in accordance with Code Case N 4981 will provide reasonable assurance of operational readiness. Therefore,it is recornmended that the licensee's proposed alternative, to implement the pressure testing rules of Code Case N 4981 for Code Class 1,2, and 3 components, be authorized for CNS pursuam to 10 CFR 50.55a(a)(3)(ii). Nebraska Public Power District's alternative should be authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this code case, the licensee is to follow all provisions in Code Case N 4981, with limitations issued in Regulatory Guide 1.147,if ary.
3.5 General ,
3.5.1 Ultrasonic Examination Techniques 3.5.1.1 Request for Relie' RI 02, Revision 1. Appendix lil, Calibration Block Material Specification Requirements Code Requirement-Section XI, IWA 2232 states that ultrasonic examinations shall be conducted in accordance with Appendix 1. Appendix 1,12200 states that ultrasonic examination of vessel welds less than 2 inches thick and eli piping welds shall be conducted in accordance with Appendix lit, as supplemented by Appendix 1. Appendix ill, Paragraph Ill 3411 outlines the material specification requirements for calibration blocks.
It requires calibration block: to be fabricated from material of the same specification as the piping being joined by the weld, it also states that if material of the same specification is not available, material of similar chemical analysis, tensile properties, and metallurgical structure tr.ay be used.
41 l
Licensee's Code Relie/Meguest-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed sa alternative to the Appendix 111, Paragraph 1113411, requirements for calibration block material specifications.
Licensee's Bes/s-
"Several of the calibration blocks currently being used at CNS lack the docurnentation necessary to demonstrate compliance with the material soecification requirements of Appendix ill. This is because the documentation requirements .
existing at the time of their fabrication did not require traceability to the material's chemich? or physical certifications. Consequently, the only documentation available for these existing calibration blocks is verification of the appropriate P number grouping.
"It would be impractical to iabricate's new set of calibration blocks in order to satisfy the documentation requirements of the current Code. Existing records, which indicate the appropriate P number grouping, provide adequate assurance that the blocks will establish the proper ultrasorde calibration and sensitivity.
" Based on the above, CNS requests relief from the ASME Section XI, Appendix lil rcquirements for cahbration block material specifications, in order to allow the contirued use of the existing calibration blocks."
4 Licensee's Proposed Altemative-
"All future calibration blocks will meet the material specification requirements of ASME Section XI, Appendix ill and will be provided with the documentation necessary to demonstrate compliance with these requirements. Additionally, when using existing calibration blocks that lack the appropriate documentation, a comparison will be made between the attenuation and material velocity of the calibration block and the material being examined."
Evaluation-The' material specification documentation required by the 1989 Edition was not required by the original f abrication code; the original calibration blocks were fabricated based on P-number groupings. The procurement of new cali'o ration blocks of the same matcrial specifications would result in an unusual difficulty without a compensating increase in the level of quality and safety. The licensee has committed to compare the attenuation of the estibration block and material velocity of the material being examined.
This additional comparison will provide adequate assurance that the existing blocks will provide the proper ultrasonic calibration and sensitivity.
Conclus/on-The licensee proposes to continue to use existing calibration blocks. The licensee has committed to compare the attenuation of the calibration block and material velocity of the material being examined. Acquiring materials for new calibration blocks to satisfy cuirent Code requirements is an unusual hardship for CNS. The imposition of this requirement would create a burden on the licensee without a compensating increase in quality and safety. Therefore, pursuant to 10 CFR 50.55ata)(3)(ii), it is recommended that the licensee's proposed alternative be authorized.
42
3.5.1.2 Request for Relief RI 09, Paragraph IWA 2311(b), Appendix Vil Ultrasonic Examination Personnel Qualification Requirements Code Regu/rement-Section XI, Paragraph IWA 2311(b) requires that the training, qualification, and certification of ultrasonic examination personnel shall also comply with the requirements specified in Appendix Vll. Appcndix Vil states requirements for the employer's wntten practice, qualificction of ultrasonic examiners, qualification records, and
, the minimum content of initial training court >es for the ultrasonic examination method.
Licensee's Code Reflef Request-The licensee requested relief from the Appendix Vil requirements for the qualification of nondestructive examination personnel for ultrasonic examination. 4 Licensee's Est/s for Requesting Relief-
" Appendix Vil was first introduced in the 1988 Addenda to Section XI. This Appendix represents a dramatic change from previous Code editions and current industry practices in the requirements for qualification of ultrasonic examination personnel. For instance, new training programs must be developed and taught by trained it'structors, employer's written practices must be completely rowritten,
! examination quet ion banks must be developed, and specimen banks of at least 15 l specimens (with 5 containing actual or simulated flaws) must be developsd and
! purchased.
" Implementation of this Appendix will require a massive industry effort. Although the industry is currently working towards compliance with Appendix Vil, full implamentation has still not been achieved. In fact, since Appendix Vil allows for the use of specimens prepared for ultrasonic performance demonstrations per Appendix Vill, many NDE vendors are developing these two programs simultaneously in order to avoid purchasing dual specimens.
" Based on the above, CNS requests relief from the ASME Section XI, Appendix Vil s requirements for the qualification of nondestructive examination personnel for ultrasonic examination."
Licensee's Proposed Attemative-
"CNS will utilize ultrasonic examination personnel qualified in accordance with the, requirements of IWA 2300, with the exception of IWA 2311(b). Additionally, personnel utilized to perform ultrasonic examinations on IGSCC susceptible welds
- will be qualified in accordance with the latest EPRI guidelines."
Eva.'uaffon- Aptrendix Vil was incorporated in the 1988 Addenda to the 1986 Edition of ASME Section XI to enhance ultrasonic examination flaw detection. This appendix specifies administrative and examination qualification requirements. Although Appendices Vil and Vill both have requirements related to flaw detection in ultrasonic examinations, their concurrent implementation is not necessary. Certain requirements of Appendix Vill may strengthen the efforts of Appendix Vil, but they are not necessary for its implementation.
43
The INEL ttaff believes that the licensee has had sufficient time to develop an Appendix Vil program. Other utilities have already committed to foltowing Appandix Vil.
Although Appenix Vill will further improve confidence in flaw detection, its implementation is not required in conjunction with Appendix Vll. An Appendix V!I program willincrease quality and safety and is not considered impractical.
Conclusion-The INEL statf believes that compliance with the requirements of Appendix Vil will enhance the overall quality of Code Section XI required volumetric examinations. .
Therefore, it is recommended that relief be denied.
3.5.2 Exempted Components -
No relief requests.
3.5.3 Other 3.5.3.1 Request for Fielief RI 10, IWA, lWB, IWC, and IWF-4000 (IWX-4000), Repair Procedures, and IWA, IWB, IWC, and IWF 7000 (IWX 7000), Replacements Code Requirement-Section XI, IWX 4000 provides the rules and requirements for repair of pressure retaining Class 1,2, and 3 components and their supports, and for the attachment of replacements to the system by welding or brazing. lWX 7000 provides the p l rules and requirements for specification and construction of items to be used for J l replacement.
Licensee's Code Re//e/ Request-Pursuant to 10 CFR 50.55ata)(3)(i), the licensee proposed an ahemative to the 1989 Code rules and requirements for repair of pressure-retaining Class 1,2, and 3 components and supports defined by IWX 4000 and IWX-7000, licensee's Basis -
"The 1989 Addenda to Section XI made cevoral changes to Articles IWX-4000 and IWX-7000. Very few of these changes were technicalin nature, instead, the changes restructured some of the requirements, clarified others that were difficult to interpret, and eliminated redundant requirements. Of the actual technical changes made, these changes either add enhancements to the program, add requirements not applicable to CNS, or delete requirements for the use of Section til for installation of non welded piping joints and allow the use ,i the original code of construction." (
"The following is a detailed aummary of each of the changes made to IWX-4000 and IWX 7000 in the 1980 Adder.ca of Section XI.
- lWA-4130: This section was restructured to differentiate between a repair program and a repair plan. The repak program is the document or set of documents that defines the managerial and administrative control for the completion of repairs. The repair plan is the document that identifies the 44
l u
+
essential requirements for completion of the repair. This section also includes additional items that must be ider)tified in the repair plan. These ite'ms include::
1 The Code Edition of Section XI governing the repair.
- 2. The original construct lon code for the item being repaired.
- 3. The construction code applicable to the repair,
_m 4.- A description of the work to be performed. ,
S. __ Material requirements, t . :- "lWA 4322: This section was clarified to specify that material must be mechanically rernoved from thermally processed areas.
"lWA 4700: Seatheids were added to the items exempted from hydrostatic testing.
Also, the statement identifying repairs not exempted from hydrostatic testing was deleted. There was no need for this statement since this section already identifies the only repairs'that could be exempted.
"lWB-4300: This section on heat exchanger tube sleeving was added._ However, _{
since CNS has no Class 1 heat exchangers, this change has no effect on '
the program.
/
"lWA 7320: The title of this section was changed from " Welding" to "Installatico".
The !.ection was changed to address individual requirements for installation by welding or brazing and installation by mechanical methods. Also, it now delineates specific requirements for pressure testing mechanical connections. Prior to the 198g Addenda, the requirements for pressure testir.g mechanical connections were only inferred by IWA 5214.
"lWB 7100 The scope was changed from " Installation not Requiring Welding" to
" Mechanical Joints and Connections".
"lWB-7320: This section (Bolted Connections) was deleted. This change allows the use of criginal construction code for determining bolt size and torquing loads, in lieu of the methods specified in Section ill, Appendix E.
"lWB 7400: This section (Inst'allation Requiring Welding) was deleted. There was no need for this section since the same requi:ements are already identified in IWA 7320.
"lWB 7600: This section (Materials) was deleted. .There was no need for this section since the same requirements are already identified in IWA-7200.
"lWC 7200: This number was changed to IWC 7100. Also, the section was changed to state that the rules of IWA 7000 apply. The only technical difference this change makes is that it allows the use of-the original construction 45 2
u L code for determining bolt size and torquing loads, in lieu of the methods .
specified in Section til, Appendix E. This is because the change eliminates the requirement to follow the rules of IWB 7320.
"lWC 7300: The section (Non welded Piping Joints) was deleted. This change allows non-welded pi~ing joints to meet the requirements of the original construction code, in lieu of thoso specified in NC 3671.
"lWC 7600: This section (Materials) was deleted. There was no need for this section since the same requirements are already identified in IWA 7200.
' LWD 7200: This number was changed to IWD 7100. Also, the section was changed to state that the rules of IWA 7000 apply. The only technical difference this change makes is that it allows the use of the original construction code for determining bolt size and torquing loads, in lieu of the methods specified in Section lil, Apundix E, This is because the change eliminates the requiremer.t to follow the rules of IWB 7320.
" LWD 7300: This saction (Non-wended Piping Joints) was dcleted. This change allows non welded piping joints to meet the req"!rements for the original construction code, in lieu of those specified in NC 3671. ,
" LWD 7600: This section (Matc,*ls) was deleted. There was no need for this section since the same ret, rements are already identified in IWA 7200.
"lWF-7000: The title of this section was changed from "E:: ope" to " General Requirements". Also, the section was changed to state that the rules oi IWA-7000 apply, "lWF 7300: This section (Installation not Requiring Welding), which was simply a title, was deleted.
"lWF 7310: This section (Mechanical Joints) was deleted. There w'as no need for this section since the same requirements are already identified in IWA-7200/
"lWF 7400: This section (Installation Requiring Welding) was deleted. There was no need for this section since the same requirements are already identified in IWA 7320.
"lWF-7600: This section (Materials) was deleted. There was no need for this section since the same requirements are already identified in IWA-7200."
"The use of the aforementioned Edition and Addenda of Section XI will provide the basis for an enhanced Inservice inspection Program."
46
1 Licensee's hoposed Attemative-
'CNS will use the 1989 Edition of ASME Section XI, as amended by the 1989 J Addenda, to govern Repair Procedures (IWX-4000) and Replacements (IWX 7000)."
EveAustion-The licensee has requested to implement the 1989 Addenda of Section XI to
- govern repair procedures. Addenda to the Code typically provide enhancements or clarification to existing Code requirements. Based on a review of the changes in the 1989
, Addenda to Section XI, the INEL staff concurs with the licenses that the changes incorpo_ rated by the 1989 Adoenda enhance and/or clarify the intent of the rules for the implementation of repair and replacement programs and plans. The changes appear to be primarily administrative in nature and do not alter the technical content of the Code rules.
Therefore, the INEL staff believes that changes incorporated by the 1989 Addenda continue to provide an acceptable level of quality and safety.
Conchssion-The INEL staff believes that implementation of the repair procedures incorporated by the 1989 Addenda will provide an acceptable level of quality and safety.
Therefore,it is recommended that the proposed alternative, use of the 1989 Addenda, be authorized pursuant to 10 CFR 50.55ata)(3)(i).
'3...5 3 2 Mequest for Ree li f RI 11, IW8 2420 and/WC-2420, Successive Examinations of Class 1 ami2 Vessels Code Requirement-lWB 2420, Successive inspections, states:
"(b) If flaw indications or relevant conditions are evaluated in accordance with IWB-3132.4 or IWB 3142.4, respectively, and the component qualifies as acceptable for continued service, the areas containing such flaw indications or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the inspection programs of IWB 2410.
"(c) If the reexaminations required by (b) above reveal that the flaw indications remain essentially unchanged for three successive inspection periods, the component examination schedule may revert to the original schedule of successive inspections.- '
"(d) For steam generator tubing, the successive examinations shall be governed by the plant Technical Specification."
lWC 2420, Successive Inspections, states:
'"(b),if component examination results require evaluation of flaw indications in ac:cdance with IWC 3000, and the component qualifies as conditionally '
acceptable for continued service, the area containing such flaw indications shall be re examined during the' next inspection period listed in the schedules of the inspection programs of IWC-2411 and IWC-2412.
"(c) If the reexaminations required by (b) above reveal that the flaw indications remain essentially unchanged for three successive inspection periods, the 47
component examination schedule may revert to the original schedule of successive inspections,"
Licensee's Code Melief #equest-The licensee requested relief from the ASME Section XI requirements for examining flaws during successive periods.
I Licensee's Besis for Mequesting Relief-
" Relief is requested on the basis that the alternatives would provide an acceptable level of quality and safety,
" Industry experience has shown that most vessel flaws located during inservice inspection volumetric examinations are not planar or crack like. They are embedded volumetric anomalies resulting from material manufacture or component fabrication,
, e.g., lamination, mid plate segregates, slag, side walllack of fusion, etc. SimHarly, most of these flaws are located mid wall or in a neutral zone with regard to stresses. Analysis shows these types of flaws to be non-propagating or benign for growth considerations. The industry's ultrasonic examination capability for flaw identification is available, and has been readily demonstrated The expense and
- additional radiation exposure to perform out of interval or unscheduled examinations of benign fabrication flaws are extensive and do not offer any commensurate increase in safety with cost benefit."
Licensee's Proposed Attemative-
"As an alternate to IWB 2420 and IWC 2420, CNS will not perform successive examinations on vessel flaws which, through analysis, have been determined to originate from material manufacture or fabrication provided:
a) The flaw is characterized as subsurface in accordance with IWA 3310(b);
b) The NDE technique and evaluation which identified and characterized the flaw as
- - originating from material manufacture or fabrication are documented in the flaw evaluation report; and c) The flaw has been determined to be acceptable for continued service in accordance with IWB 3132.4, or IWC-3122.4, and demonstrated to have growth within acceptable limits until the next scheduled inspection, or the end of service life of the component." -
Evaluation-For each examination that would result in successive examinations, specific information to establish the position that the flaw is a manufacturing defect and will not grow must be developed. The licensee should request relief from successive examinations only when flaws that are detected can be verified as manufacturing defects based on the fabrication radiograph and/or the inservice flaw size correlates well with the fabrication or baseline examination. This information should be submitted to regulatory authorities for evaluation on a case "by" case basis.
48
Conclus/on-Based on the discussion above, it is recommended that relief be denied and that relief of this nature be evaluated only for specific components using appropriate
-technical justification presented by the licensee.
3.5.3.3 Request for Relief RI 12, Examination Ca.tegory B J, item B9.12 and Examination Categories C F 1 and C F 2, items C5.12, C5.22, C5.42, C5.52, C5.62 and C5.82. Examination of Class 1 and 2 Longitudinal Piping Wolds Code Requ/rsment-Saction XI, Table IWB 25001, Examination Category B J, item B9.12 requires surf ace and volumetric examinations of longitudinal piping welds in Class 1 piping i j
that is 4 inch nominal pipe size and larger in conjunction with exam! nation of the circumferential welds selected for examination, as defined in Figure IWB 2500 8. The length of lungitudinal vield required to be examined is at least one pipe diameter, but not more than 12 inches, from the circumferential weld intersection point.
/ Examination C,ategories C F-1 and C F 2, items C5.12, C5.22. C5.52, and C5.S2 require volumetric and surface examinations of longitudina' piping welds in Class 2 piping in conjunction with examination of circumferential wolds selected for examination, as defined in Figure IWC 2500 7. At least 2.5t of longitudinal weld is required to be examined. For items CM.42 and C5.82, a surface examinetion is required for longitudinal piping welds intersecting circumferential welds selected for examination, as defined in Figure IWC 2500 7. At least 2.5t of longitudinal weld is required to be examined.
L/censee's Code Re/le/ Request-Pursuant to 10 CFR 50.55a(a)(3)(1), the licensee proposed an alternative to performing the volumetric and/or surface examination for the length of longitudinal piping welds required to be examined in accordance with Tables IWB 2500 and IWC 2500.
Licensee's Basis-
" Specific relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.-
"The area of the longitudinal seam weld which is most susceptible to failure is that portion immediately adjacent to the circumferential weld. During the circumferential .
welding process, this area is most likely to undergo material changes, resulting in 3
+
flaw development and potential failure. This critical area is included in the required
, volume of material examined during the volumetric scanning of the circumferential weld."
- Licensee's Proposed Attemative-
"As an alternative to Code required volumetric examination and/or surface examination of Class 1 and 2 longitudinal pipe welds, CNS will perform the examinations in accordance with / SME Section XI Code Case N-524, " Alternative Examination Requirements for Long tudinal Pipe Welds in Class 1 and 2 Piping;Section XI Division 1."
f 49
/
Eva/ust/on-The licensee has proposed to implement the alternatives contained in Code Case N 524 for examination of Class 1 and 2 piping longitudinal welds. The licensee l
proposes to examine the potentially critical portions of the longitudinal welds (the portion that intersects the circumferential weld) in conjunction with examination of the circumferential welds.
When implementing the alternatives contained in Code Case N 524, longitudinal welds need not be examined beyond the examination zone of the associated circumferential -
! wold. When the longitudinal weld can be identified, only that portion of the longitudinal weld intersecting the circumferential weld is required to be examined for flaws parallel and transverse to the weld. Where the longitudinal weld cannot be identified,100% of the -
circumferential weld shall be examined for flaws parallel and transverse to the weld to ensure that the longitudinal /circumferential weld intersection is examined. Code Case N 524, when implemented in its entirety, leads to examination of the most critical area of the longitudinal weld, and thus provides an acceptable level of quality and safety (It should be noted that when implementing alternatives contained in Code Case N 524, requirements for examination of longitudinal welds contained in Table IWB 2500 are superseded.)
Conclus/on- An acceptable level of quality and safety is provided by the licensee's proposed alternative, use of Code Case N 524 for examinat!Sn of Class 1 and 2 piping longitudinal welds. Therefore,it is recommended that the use of Code Case N 524 be approved pursuant to 10 CFR 50.55ata)(3)(i). Use of Code Case N-524 should be authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time,if the licensee intends to continue to implement this code case, the licensee is to follow all provisions in Code Case N 524 with limitations issued in Regulatory Guide 1.147, if any.
3.5.3.4 Request for Relief RI 13, Examination of Coda Class Snubbers This request for relief is not in the scope of this review and is being evaluated in a separate report by the NRC.
3.5.3.5 Request for Relief RI 14 Revision 1 Use of Code Case N-509 for Selection and Examination of Class 1,2, and 3 Integrally Welded Attachn.ents Code Requirement-For Class 1, Examination Category B K-1, volumetric or surface examination, as applicable, is required for allintegrally-welded attachments exceeding 5/8 inch design thickness during the first and second intervals when implementing Program B.
For Class 2, Examination Category C C, surfact, examination is required for allintegrally-welded attachments exceeding 3/4 inch design thickness. For Class 3, Examination Categories D A, D-B, and D-C, surface examination is required for allintegrally-welded attachments corresponding to those component supports selected by IWF 2510(b).
Licensee's Code Relief Request-Pursuant to 10 CFR 50.55ata)(3)(i), the licensee proposed an alternative to implement Code Case N 509 for selection and examination of Class 1,2, and 3 integrally welded attachments.
50
1 =
i ~ Licensee's Basis-- _
'
- Specific relief is requested on the basis that the proposed attemative would
- provide an acceptable level of quality and_ safety. Code Case N 509, "Altemative
' Rules for the Selection and Examination of Integrally Welded Attachments," Section
- XI, Division 1, provides an altemative to the Tables of IWBIC/D 25001: for -
Integrally welded attachments.; The altamative requires a surface examination of-
<10% of the integrally welded attachments associated with the component supports
~
3 selected fur examination under IWF 2510. In addition, an examination is required whenever component' support member deformation is identified. This Code Case recognizes the results of over 20 years of inservice inspections and the-considerable attention that component supports have received through NRC bulletins."
l Licensee's troposed Attemative-
~ *CNS proposes to examine 10% of the integrally welded attachments in each Examination Category in each Code class in accordance with Code Case N 509 requirements."
Evaluation-In lieu of Code requirements for selection and examination of integral-
_ attachment welds, the licensee proposes to apply the attematives contained in Code Case N 509, Altemative Rules for the Selection and Examination of Class 1, 2; and 3 Integrolly Welded Attachments.;ln addition, CNS proposes to examine 10% of the integrally welded attachments in each Examination Category in each Code class in accordance with Code - 1 Case N 509 requirements. Considering that the Code often specifies sampling to assure
^ that service related degradatic, is not occurring, it is logical to extend the sampling process to welded integral attachments. With a sample of a minimum of 10% of all integral attachment welds in Code Class 1,2, and 3 systems, the INEL staff believes that degradation, if occurring, will be detected. Therefore, the INEL staff believes that the use of the altamatives contained in Code Case N 509. with a minimum 10% selection of all integrally welded attachments in each Code class, will provide an acceptable level of
-quality and safety.' . -
Conclusion-The licensee has' proposes to examine 10% of the integrally welded attachments in each Examination Category in each Code class in accordance with Code
' Case N-509 requirements. The INEL staff believes that the licensee's proposed altemative
- will provide an acceptable level of quality and safety. Therefore, pursuant to
,g _10 CFR 50.553(a)(3)(i), it is recommended that the licensee's proposed attemative be authorized. Use of alternatives contained in' Code Case N 509, augmented by the )
' licensee's proposed 10% sample of all nonexempt Class 1,2, and 3 integral attachments, should be authorized for the current interval or until such time as the code case is published in a future revision of Regulatory Guic'e 1.147. At that time,if the licensee intends to continue to implement this code case, the licensee should follow all provisions in Code Case N-509, with limitations issued in Regulatory Guide 1.147, if any.
51
3.5.3.6 Request for Relief RI 17, Class 1 and Class 2 Integrally Welded Shear Lugs on Piping Code Requirement-Table IWB 25001, Examination Category B K 1, items B10.10, B10.20, B10.30, and B10.40 require a volumetric or surface examination, as applicable, for integrally-welded attachments exceeding 5/8 inch design thickness. IWC 25001, Examination Category C C, item C3.70 requires a surface examination of 100% of the length of allintegral attachment welds exceeding 3/4 inch design thickness as defined by .
Figure IWC 2500 5.
L/censee's Code Relie/ Request-Relief is requested from performing the volumetric and -
! surface examinations to the extent required by the Code for the following Class 1 and 2 integral attachment welds: FWB BK1-8, FWC BK18, MSA BK16, PSA BK1 19, RR BK1-4A, RHB-BK1 16, RHA CE1-2, RSA CC 25, SDS CE121.
Licensee's Sosis for Requesting Relief-
" Specific relief is requested on the basis that the proposed attemative would provide and acceptable level of quality and safety.
"Certain of the integrally welded attachments on class 1 and 2 pipe supports are shear lugs adjacent to a pipe clamp or restraint. The shear lugs on horizontal piping runs prevent movement along the axis of the pipe. The shear lugs on vertical piping runs transfer load from the pipe to the support in the downward direction. Shear lugs are typically welded on the two sides orthogonal to the support by a groove plus a fillet weld as shown in Figure IWB 250015 or IWC 2500 5(a). Sometimes the shear lug is attached by a fillet all around as shown in Figure IWC-2500 5(b).
In order to examine 100% of the surface for 1/2" on either side of the weld, the pipe clamp or restraint must be disassembled. The Code does not usually require a component to be disassembled solely for oxamination. Disassembly may require considerable time, the erection of an altemate support and, depending on the location, may result in significant exposure. Examining the accessible portions of the attachment weld location, may result in significant exposure. Examining the accessible portions of the attachment weld without removing the clamp will cover approximately 80% of the required surface, willinclude the hiohest stressed regions of the attachment weld, and is sufficient to detect service induced flaws in the attachment welds."
Licensee's Proposed Altemative-
"In lieu of performing the Code required examinations, CNS proposes to examine integrally welded attachments in accordance with applicable Code requirements to the maximum extent practicable without removal of the clamp. The applicable NDE data record will describe in detail the extent of the limitation and will be available for review, if indications are detected adjacent to the intervening piping clamp, the clamp will be removed for further evaluation."
Evaluation-The Code requires a 100% surface examination of the subject welded attachments. The licensee has requested generic relief from removal cf clamps associated 52 l
l g7 "with piping integral attachments. The licensee is implementing Code Case N 509, which provides an altomative to the examination of 100%'of Class 1,2, and 3 integral-c L attachments and reduces the number of integral attachment welds being examined. The - -
INEL staff believes that the proposed attemative,in combination with the Code Case,is unacceptable. Other utilities have removed clamps for access to 100% of integral attachment welds, unless relief was sp*cifically authorized.
ConcAssion-The licensee's prcposed altamative in combination with Code Case N 509, is considered unacceptabl0. Because of the substantial reduction in integral attachment weld examinations allowed by this Code Case, all selected integral attachmera welds should be o - examined to the extent possible.- Therefore, it is recommended that relief be denied.
3.5.3.7 Request for Relief RI 24, IWB 2412(a) and IWB 2420(a), Succesolve Examination Requirements and The Sequence of Examination fot Class 1 Bolting i Code Regu/remont-Table IWB 2412(a) states that the required examinations in each examination category shall be completed during each successive inspection interval in accordance with Table IWB 24121. lWB 2420(a) states that the squence of component examinations established during the first inspection interval shall be repeated during each successive inspection interval, to the extent practical.
Licensee's Code Relief Request-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee requests relief from the successive examination and aequence of examination requirements of the Code for the reactor pressure vessel botting, uts, washers, and threads in flange.
Licensee's Bes/s for Requesting Relief-
"The reactor vessel closure head nuts and closure washers are (emoved each outage for disassembly _of the vessel. The studs and bushings are normally left in 7 - place. . During tne two previous inspecticn intervals, the studs, nuts bushings, washers, and the threads in flange were each divided into three groups. One group of each item was for examination during each period of the interval. Every period,-
an entry would be made into the reactor cavity to examine the scheduled threads in the iange. Every period, equipment would be brought up to the refueling floor to clean the nuts and washers scheduled for examiration. The same philosophy was
- applied to the closure studs'and bushings, except that the old bushings were-inspected while the studs were removed. The old axamination schedule appears to
.- have been based on an incorrect interpretation of tie Code Code Interpretation XI-86-74, clarifies that schedule requirements are t i be applied by examination category; not by item number.
"In order to more effectively use plant resources, and reduce perconnel exposure,- CNS would like to reschedule some of the Category 5-G 1 examinations. Rescheduling,
[ however, would conflict with the requirements of IWB 2412(a) since approximately 40% of the B G 1 examinations would be performed during the first period,20%
during the_ second period, and the balsnce during the third period. Rescheduling would ;
also conflict with the requirements of IWB-2420(a) because on third of the closure nuts '
and washers would be reexamined after more than ten years of service since the 53 Q3 *
'?
previous examination, and a third of the threads in the flango and'the bushings would
. be reexamined after less than ten years of service since the previous examination. The
_ studs would_be re examined at the Code required frequency."
Licensee's Proposed Attemet/ve-
"In lieu of maintaining the Code required successive exa'nination schedule, CNS proposes to re examine the threads in the reactor vessel flange and the bushings during the first period; the closure head studs during the second period, and the ,
closure washers and nuts during the third period."
Evaluet/on-The Code requires that the successive examination and sequence of .
examination be maintained during subsequent intervals. In addition, the Code requires that a sample of examination areas be examined each period in accordance with Table IWB-
_24121. The licensee has noted that the Code schedu;ing criteria is based on Examination Category._ However, the intent of the scheduling philosophy of items within an Examination Category is further defined in the 1991 Addenda of ASME Section XI.
lWB 2112(a)(3) states: "If there are less than three items to be examined in an examination _ Category, the items may be examined in any two periods, or in any one period if there is only one item ,in lieu of the percentage requirements of Table IWB 24121, Because there are greater than three bolts associated with the closure head, a schedule of examinations as required by Table IWB 24121 must be maintained. As such, it is.
recommended that relieV be denied.
Conclusion-The licensee's proposed alternative to scheduling examinations on Examinatiori Category B G 1, reactor pressure vessel stud, nuts, and threads in flange, is not supported by the sampling philosophy of the Code. Therefore,it is recommended that relief be denied.
3.5.3.8 Request for Relief PR-03, Revisior: 1, IWA 5250(a)(2), Corrective Action Resulting from Leakage at Bolted Connections Code Requirement-lWA 5220(a)(2) requires that the source of leakages detected during . . .e the, conduct of a system pressure test shall be located and evaluated by the Owner for corfective action. When the leakage is at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100, t
i licensee's Code Re/le/ Request-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee l
proposed an alternative to the ASME Section XI requirements for removal of botting at leaking connections for VT-3 visual examinc.in.
Licensee's Besis-
"In the event of a bolted connection leak detected during the conduct of a system l
pressure test current ASME Section XI Code requirements specify that all bolting j must be removed for the purpose of a VT-3 visual examination and evaluation. This would require placing the component or piping system out of service which could result in a plant shutdown, a delay of plant startup or, for continued operation, a 54
.1
. reduction in' plant safety, Additionally, removel'of all bolting for examination serves no practicable purpose if the bolting is fabricated of a material which is not i susceptible' to corrosion'due to contact with the leaking medium.: The following proposed alternative provides an acceptable level of quality and safety equivalent to
. that provided by applicable Code requirements."
Ucensee's Proposed Memative-c 1"If leakage occurs at a bolted connection ~ during the performance of a system pressure test, an engineering evaluation'shall be performed to determine if the
- associated bolting is susceptible to corrosion which could result in further
. degradation and increased leakage. This evaluation shall address'at a minimum:
1)- type and location of leakage
- 2) historical 1:akage .
3)- bolting matorials of the leaking component
- 4) visual evidence of corrosion '
- 5) history of the bolt material degradation due to corrosion in similar environments "If the engineering evaluation indicates that the bolting materialis not susceptible to corrosion, then bolt removal for visual examination and further evaluation shall not_be required. However, termination of leakage shall be addressed at the next
, available opportunity.
"If the evaluation determines the need for a VT-3 visual examination of the bolting,-
one bolt closest to the source of leakage shall be removed, and in lieu of performing .
the Code required VT-3 visual examination, the botting will be VT-1 visually
- examined per IWA 2211(a) and evaluated in accordance IWB 3517.1. If the removed bo!t has evidence of degradation, all remaining bolting shall be removed
.and VT-1 examined and evaluated accordingly. All examinations and evaluations shall be traceable to the VT-2 documentation originally detecting the leakage and applicable records will be maintained per IWA 6000."
N-Evaluation-The'1989 Edition of thn Code requires that, when leakage occurs at bolted i connections, all bolting be removed for VT-3 visual examination in lieu of this requirement, the licensee has propond to evaluate the botting to determine its susceptibility to corrosion, if the bolting is susceptible to corrosion or the initial evaluatien
, . indicates the need for a more in-depth evaluation, the bolt closest to the source of leakage will be removed, VT 1 examined, and evaluated in accordance with IWA-3100(a).
.- The INE'. naff believes that the licensee's proposed attemative is based on sound engineering judgement and that this attemative to the Code-required removal of botting at ,
a joint when leakage occurs should provide an acceptable level of quality and safety.
. Conclus/on-It is reasonable to conclude that the licensee's proposed alternative, to
- evaluate the botting at a leaking connection, would detect degradation of bolting, if present. Therefore,it is recommended that the proposed alternative be authorized pursuant'to 10 CFR 50.55a(a)(3)(i).
55 a__.._._._ . _. _
U
- 4. CONCLUSIONS Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee has determined that certain inseivice examinations cannot be performed to the extent required by Section XI of the ASME Code, in these cases, the licensee has demonstrated that specific Section XI requirements are impractical. Therefore,it is recommended that for Requests for F elief RI 05, RI 15, RI 18, RI 20 (Revision 1), RI 21, and RI 22, relief be granted as requeste1. Granting relief will not endanger life, property, or the common defense and security, and is otherwise in the -
public interest, Di ving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Pursuant to 10 CFR 50.55ata)(3),it is concluded that for Requests for Relief RI-02 (Revisloa 1), RI 03, RI 07, RI-OS, RI 10, RI 12, RI 14, RI 16, PR-01, PR 02, PR-03 (Revision 1), PR 04,' PR 05, PR-06, PR 07, PR 08, and PR 09, (a) the licensee's proposed alternatives will 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, in these cases,it is recommended that the proposed alternative be authorized.
For Requests for Relief RI 06 (Revision 1), RI-09, RI 11, RI 17, and RI 24, it is concluded that the licensee has not provided sufficient justification to support the determination that the Code requirement is impractical, and that requiring the licensee to comply with the Code requirement would result in hardship. Therefore, in these cases,it is recommended that relief be denied.
The licensee withdrew Requests for Relief RI-01 and RI 23.
Requests for Relief RI 13 and Rl 19 are not in the scope of this review and are being
, evaluated in separate reports by the NRC.
This technical evaluation has not identified any practical method by which the licensee can meet all the specific inservice inspection requirements of Section XI of the AbME Code for the existing Cooper Nuclear Station. Compliance with all the Section XI examination requirements would necessitate redesign of a significant number of plant systems, procurement of replacement components, installation of the new components, and baseline examination of 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 certain .
provisions of Section XI of the ASME Code that h6ke been determined to be impractical.
The licensee should continue to monitor the development of new or improved -
examination techniques. As improvements in these areas are achieved, the licensee should incorporate these techniques in the ISI program plan.
Based on the review of the Cooper Nuclear Station, Third 10-YearIntervalinservice Inspection Program Plan, Revision 1, the licensee's responses to the NRC's requests 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 Code, 56
~
-no deviations from regulatory requirements or commitments were identified, except as noted for Requests for Relief RI 06 (Revision 1), RI 09, RI 11, RI 17, and RI 24.
..p fn 2
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4
)
4 3
1 -
57
__ . . _ . _ . _ . - m__._.._....._.. . _ . . _ . . _. . _ _. _.
~
'a .. _
- 5. REFERENCES
-1. Cod 6 of Federal Regulations, Title 10, Part 50.
9
- 2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1,1989 Edition (except that the extent of examination of Class 1 piping welds has been determined by the 1974 Edition with Addenda through Summer 1975 as permitted by 10 CFR 50.55a(b)). .
- 3. Letter dated October 18,1995, John H. Mueller (NPPD) to Document Control Desk, containing the Cooper Nuclear Station, Third 10-Year Intervalinservice Inspection Program, Revision O.
- 4. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for 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.
- 5. Letter to G. R. Horn (NPPD') from J. R. Hall (US NRC) dated February 8,1996, containing NRC Request for Additional Information.
6 Letter dated April 11,1995, John H. Mueller (NPPD) to Document Control Desk, containing the response to the NRC request for additionalinformation on the Cooper Nuclear Station. Third 10 Year Intervalinservice Inspection Program, Revision O.
7 ' Letter dated April 11,1995, John H. Mueller (NPPD) to Document Control Desk, containing the Cooper Nuclear Station. Third 10-Year Intervalinservice Inspection Program, Revision 1.
- 8. Letter to G. R. Horn (NPPD) from D. L. Wigginton (US NRC) dated June 3,1996, containing NRC Request for Additional information.
- 9. Letter' dated August 5,1995 Philip D. Graham (NPPD) to Document Control De?'
containing the response to the NRC request for additionalinformation on the Cooper Nuclear Station, Third 10-Year Intervalinservice Inspection Program, Revision 1.
- 10. Letter to G. R. Horn (NPPD) from J. R. Hall (US NRC) dated October 24,1996, -
containing NRC Request for Additional information.
- 11. Letter dated December 31,1996, Philip D. Graham (NPPD) to Document Control Desk, containing the response to the NRC request for additional information on the Cooper Nuclear Station, Third 10-Year IntervalInservice Inspection Program, Revision 1.
4 58 4
1
3
- 12. Letter dated February 7,1997, Philip D. Graham (NPPD) to Document Control Desk,
. containing clarification on the Cooper Nucleas Station, Third 10 Year Intervalinservice Inspection Program, Revision 1.
- 13.- Regulatory Guide 1.147, Revision 10, inservice Inspection Code Case Acceptability, ASME Section XI Division 1. July 1993.
- 14.- HUREG 0619, BWR Feedwater No:zle and Control Rod Drive Retum Line Nozzle Cracking, Revision 1. July 1980.
- 15. NRC Generic Letter 88 01, Supplement 1, NRC Position on Interpranular Stress Corrosion Cracking (IGSCC)In BWR Austenitic Stainless SteelPiping, February 4, 1992.
- 16. S.H. Bush and R.R. Maccary, " Development of in Service liispection Safety Philosophy for U.S.A. Nuclear Power Plants," ASME 1971.
I i
e 59
4.
N:;C e;ny 333 U $ NUCLiaR RiGUL af oAY COMMil51oN sia;a g,.gg, fe$.,u h 33" '
= a:. BIBLIOGRAPHlO DATA SHEET
.s ,w o s P ~* ,o se =
1 taa as; semi INEL-95/0348 Technical Evaluation Report on the Third .0-Year 3 oin auca,.a ,,3.io Interval Inservice Inspection Program Plan: vcs . , ....
Nebraska Public Power' District March 1997
- Coo)er Nuclear Station ois ea cna~, s.cso j Doccet Number 50 298 JCN-J2229 (TWA-Al2) l , ,ywon g, e ivPe op aecar Technical M. T. Anderson, E. J. Teige, K. W. Hall ut aico Cov i a n o .. ~ - c,...
. ,i a.o a ,~c casamzation ** vi as o acoa t ss ", ~.c. ,-- o-+. . o"-- a- u s a- a -
- c-~~ ~ ~ e ~~ s <~~'- ~~
- - -.a Lockheed Martin Idaho Technologies Company .
P.O. Box 1625 Idaho Falls, ID 83415 2209
.-.c s, .,--, -- aac o-- o -., u s ..~ . . , c.-
i s,o~sgoacamu non - savi aso meon ess m acm w .
Civil Engineering and Geosciences Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 t0 $yPPL aMcNTany Note $
- 11. ABst a A;T IN0 =,,es ., wu, This report presents the results of the evaluation of the Cooper Nuclear Station, Third 10-Year interval inservice Inspection (ISI) Program Plan, Revision 0, submitted April 11, 1996, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements that the Iicensee has determined to be impractical. . The Cooper Nuclear Station, Third 10-Year Interval inservice Inspection (ISI) Program Plan, is evaluated in Section 2 of this report. The ISI Program Plan i evaluated for (a) compliance with the ap ,1, (b) acceptability of examination sample, (c)propriate correctness edition of the/ addenda application ofofSectior.
system or component examination exclusion criteria, and (d) compliance with ISI-related commitments
, identified during previous Nuclear Regulatory Commission (NRC) reviews. The requests for relief are evaluated in Section 3 of this report.
u s , La os c e s:a e v oas me . e ------.- ~ ~~, ~ ~~~ > " ^ ' ' * ~ " " ' " "
Unlimited
, . ,, w. . . u. w . .c. n a a.,,
Unclassified n . ,,,
Unclassified 16 NUM0ta OF PaGEs 16 PR Cl nac somw 32,i249'
- n. - _ . . . -. . .
Page 1 of 4 IC00PER NUCLEAR STATION Third 10-Year ISI Interval i
TABt.E I
,'i
SUMMARY
OF RELIEF REQUESTS S
o -
" Rellef Relief Regelred . , ,. y. og Request . System or Exam . Itse . -Vohame or Area to be - _
Enseined . -feethode Licenses Preposed Alternative ?StatesJ-Nember Cet - Category L No.'
GA NA NA Withdrawn RI-G1 Withdrawn MA RA Appendix ! Class 1 and 2 Cmponents Ultrasonics Use Entsting Calibration Authertred RI-02 Calibration NA and Ill Blocks Rev. ! Blocks f
B3.100 Norrie Inner Radius Sections Volumetric Surface examination of the- Autherfred
,RI-03 Reactor B-D l Pressure Vessel norrie inner radius section Blat
- RI-04 Blank Tubesheet-to-Shell Velds Volumetric Parform a VT-1 visual Granted RI-05 Residual Heat C-A .C1.30 Removal Heat examination l
Exchanger 3 Reactor 8-A Bl.11 Circiseferenttal Shell Welds Volumetric Develop and Ispienent en RPT Denied i RI-05 Pressure Vessel B1.12 Longitudinal Shell Welds examtr.stlan plan j Rev. 1.
81.21 Head Ciretssferential Weld N-491 4
i B1.22 Merldtonal Welds RI-07 Reactor B-H 88.10 Support Skirt veld Surface Perform a surfaca examination A> h rtred 1 Pressure Vessel of the accessible A-8 area and an ultrasonic examination of area A-B-C-D RI-08 Class 1 and 2 Generic NA Expansion Criteria for Weids NA Perform sample expansions as Authertred Piping Letter 88- required by Generic Lettee 01, NUREG- 88-01 and WUREG-0313 0313, Rev. 2 RI-09 Examination Appendix VII MA Examination Personnel NA Utiltre examination personnel Denied Personnel Qualiftest1ons quaitfled in .c ,it . e with IVA-2300 RI-lO Class 1, 2, 3 IVX-4000 NA Repairs and Replacements liA Apply the requirements of the Authortred Pfptng and IWX-7000 1999 Audende Components
Page 2 of 4 l COOPER NUCLEAR STATION Third 10-Year ISI Interval TABLE 1 l
SUMMARY
OF RELIEF REQUESTS
+
Rollef :
Relief JRequired( =f e Request ( !
Request Systen or Exas . Itse Vohme or Area to be , s .
" 5tatese Emmetned . Method- Licensee Prepaaed Alternettve
~
1hm6er . r_ _ _ ___ _ t . Category No. ,
NA Evaluate nature of flaw and Denied RI-11 Class 1 and 2 IWB-2420 NA Successive Exmalnation Reqctrements perform successive Yessels IWC-2420 exmainations based on that Pressure Vessel Class 1 end 2 piping Vohanetric laplement Code Case N-524 Authortred RI-12 Class 1 and 2 8-J. BE.12 Pipe C-F-1 C-F-2 C5.12, longitudinal welds and Surface
' Longitudinal C5.22 Welds C5.42 C5.52 C5.62 C5.82 NA NA NA NA Referred to RI-13 Snubbers NA NRC B-K-1, C-C. Class 1,. 2 and 3 integrally Volenetric or Implement Code Case E-509 Authertred l RI-14 Class 1, 2, and 510.10 D-A, and D-B B10.20, welded attactyents surface with a sintense 10% sample of
} 3 Piping all nonemenyt intsgral B10.30, 610.40, attactment welds j
C3.10, l* C3.20 C3.30, C3.40, DI.10, 31.20, 01.30, Pl.40 8-0 814.10 Cantrol lied Housing D-tve Surface Examine 501 of eight Granted RI-15 Reactor Pressure Vessel Welds peripheral CRD lower housing welds Reactor B-G-1 86.10 Closure Head Nuts surface rerform a VT-1 visual Auttertred RI-16 Pressure Vessel examination on the surfaces of all closure heed nuts
I Page 3 of 4 COOPER NUCLEAR STATION
, Third 10-Year ISI Interval TABLE 1 SUNNARY OF RELIEF REQUESTS l Relief
- Segnest C
' ' Systne or has Itna - ' Volume or Area to be -r -Required' . , . .
Request Licensee Preposed Altervistfwe Stat W
; Category . No.. : bester-j . Method -
Num6er Component ,
Integrally Vzided Shear Lugs Volonetric or Examine to the extent Dented RI-17 Class 1, 2, 3 8-K-1 B10.10 B10.20 St,? face practicable without clamp Piping C-C removal B10.30 B10.40 C3.10 C3.20 C3.30 C3.40 C3.30 Integrally welded Surface Perforin a VT-2 Granted RI-Its Class 2 Punps C-C attachments to Residual Heat
! Removal (RMt) Pgs IA,18, IC, and 10
{ Referred to i RI-19 Reactor B-D B3.90 Feedwater Nozzles M4A, N4C, MA NA NAC Pressure Vessel N4D B5.130 Weld RVD-BF-14 Surface Examine to the extent Granted
- 21-20, Class 1 Piping B-F i Rev. 1 practicable Reactor B-D B3.90 Nozzle-to-Shell Welds
- Voltanetric Examine to the extent Granted RI-21 Pressure Vessel B3.100 MIA, B pract fcable N2A-H, J K N3A-D M4A&C W4B&D NSA,B N6A 8 N7 N8A,8 N9 RI-22 Class 1 Piping B-J B9.31 Branch Pipe Connection Welds Voltanetric Examine to the extent Granted FWA-BJ-81, RAS-BJ-10, RBS- and Surface practicable BJ-6A RI-23 Class 2 Piping C-F-2 C5.51 Welds RWA-CF-10 RWA-CF-ll, NA NA Ytthdrawn and RWA-CF-12
i . , ".
- } , ~ '
,1 1 '
-t COOPER N!! CLEAR STATION Page 4 of 4[
l Third 10-Year ISI Interval ~
TABLE I SUNNARY OF RELIEF REQUESTS
' RelM :
aeller - - - I11 W> -
'" liegeest 5
Request .Systee or! Exam Itas _ Volume or Area to bel..
- Required . . . '. ' -
Category: IIo. :Exmained C Method " Licensee L ,- 2 Alternettve J5tatuni Nisaber rwt NA S*secessive and Se wence of Continue Examine flange and tueshings Dealed RI-24 Class 1, 2 & 3 IWB-2412(a)
IWB-7120(b) Bolting Examle.ations established 1st period, studs 2nd period, Sequence and weshers and nuts 3rd L
Tests perled.
PR-01 Class 1. 2 and IWA-4700(a) NA Repairs and Replacements YT-2 Code Case N-416-1 includtr2 a Authortred 3 Repair and surface examination of the Replacements root pass for Class 3 systems PR-02 Class 1 B-P . 815.10 Pres:ure Retalning Boundary YT-2 Close MS Line Dral. Valve. Authorized Pressure for System Leakage Tetts HPCI and RCIC VT-2 Retalning Boundary PR-03, Class 1, 2, and IVA-5220(a) NA Corrective Action Resulting Rem)ve all Engineering Evaluation Authertred l Rev. 1 3 & (b) from Leakage at Bolted bolts for VT-j 1 Connections a Visual Exam PR-04 Reactor B-P B15.50 RPV Head Flange Seal Leak YT-2 VT-2 of Leak Detection Line Authortred Pressure Ve--*1 ~
815.51 Detection System with the reactor cavity flooded PR-05 Class 2 Piping C-H C7.30 Containment Venetration Hydrostatic Code Case N-522 6t Peak [Authertred thru ' Piping Test Containment Dest M Pressure -
C7.80 and a Precedure fer Locating Leaks i
PR-06 *1 ass 3 7: ping D-A D1.10 Burled Pipe VT-2 Verify Flow with Authortred l Instris=ntation PR-07 Class Piping 8-P NA Pressure Testing Boundaries Hydrostatic Code Case N-498-1 Autherfred D-A Test. YT-2 I PR-08 Class 2 Piping D-A 01.li Alternate Testing for MSRV VT-2 Plant Surveillance Au*.hortrod l Discharge Piping PR-09 Class 2 Piping C-H C7.40 Pressure Testing HPCI and Hydrostatic Tested with Class 1 System Authortred RCIC Discharge Piping lest VT-2 m