L-96-026, Tech Evaluation Rept on Second 10 Year Inservice Insp Program Plan:Wolf Creek Nuclear Operating Corp
| ML20198Q320 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/31/1997 |
| From: | Mary Anderson, Beth Brown, Feige E IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20198Q266 | List: |
| References | |
| CON-FIN-J-2229 INEL-96-0264, INEL-96-0264-R01, INEL-96-264, INEL-96-264-R1, NUDOCS 9711120174 | |
| Download: ML20198Q320 (50) | |
Text
lt INEL 96/0264 Revision 1
, July 1997 l.l
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L j_ g idaho National iI Engineering t ,boratory Technical Evaluation Report on the .;
'l Second 10-year Interval inservice '
I inspection Program Plan:
, Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Siation, Docket Number 50-482 2
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M. T. Anderson l B. W. Brown
} E. J. Feige
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! A. M. Porter -
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l INEL 96/0264 !
Revision 1 Technical Evaluation Report on the Second 10-year intervalinser .co inspection Program Plan:
Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Docket Number 50 482 l
M. T. Anderson B. W. Brown E. J. Feige K. W. Ha!l
- j. A. M. Porter Published July 1997 Idaho National Engineering & Environmental Laboratory Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Idaho Operations Office 3
Contract DE AC07 94lD13223 JCN No. J2229 (Task Order A05)
ABSTRACT This report presents the results of the evaluation of the Wolf Creek Generating Station, -
Second 10 YearIntervalInservice Inspection Program Plan, submitted August 30,1995, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the liesnsee has determined to be impractical. The Wolf Creek Generating Station, Second 10-YearIntervalInservice Inspection Program Plan, is evaluated in Section 2 of this report. The Inservice Inspection (ISI) Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.
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This work was funded under:
U.S. Nuclear Regulatory Commission JCN No. J2229, (Task Order A05)
' Technical Assistance in Support of the NRC Inservice Inspection Program li O
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SUMMARY
The licensee, Wolf Creek Nuclear Operating Corporation, has prepared the Wolf Creek Generating Station, Second 10-YearIntervalInservice Inspection Program Plan to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section XI. The second 10-year interval began September 3,1995 and ends September 2,2005.
The information in the Wolf Creek Geaerating Station, Second 10-YearIntervalinservice Inspet: tion Program Plan, submitted August 30,1995, was reviewed. Included in the review were the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. As a result of this review, a request for additional information (RAl) was prepared describing the information and/or clarification required from the licensee in order to complete the review. A conference call was held on April 24,1996, to clarify the information requested. The licensee provided the requested information in the Response to Request for Additional Information, Second 10 year Inservice Inspection Interval, submitted l May 3,1996.
Additional RAls were transmitted to the licensee on October 10,1996, November 25,1996, and March 4,1997, to obtain additional information and/or clarifications that would permit the staff to evaluate the licensee's requests for relief with Code and Regulatory requirements. The licensee's responses to these requests were received on December 9,1996, January 23,1997, and May 6,1997, respectively.
Based on the review of the Wolf Creek Generating Station, Secono so YearInterval inservice Inspection Program Plan 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 requiremerna or commitments were identified in the Wolf Creek Generating Station Second 10-YearIntervalinservice Inspection Program Plan, except as noted in the evaluations of Requests for Relief 12R-09 and 12R-16.
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CONTENTS A B ST RA CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 11 S U M MA RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . lil
- 1. INTRODUCTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . . , . . . . . . . . 3 2.1 Documents Evaluated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 Compliance with Code Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -
2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.2 Acceptab;lity of the Examination Sample . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.3 Conclusions . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
- 3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1.1 Reactor Pressure Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1.1.1 Request for Relier 12R 03, Examination Category B-A, item B1.30, Reactor Vessel Shell-to-Flange Weld . . . , . . . . . . . . .6 3.1.1.2 Request for Relief 12R-04, Examination Category B-A, item B1.21, Reactor Vessel Closure Head Circumferential Weld . . 8 3.1.1.3 Request for Relief 12R.10, Examination Category B-G 1, item B6.10, Reactor Vessel Closure Head Nuts . . . . . . . . . . . . . . . 9 3.1.2 Pressurizer ....................................................11 3.1.2.1 Request for Relief 12R-06, Examination Category B-D, item B3.110, Pressurizer Nozzle-to-Vessel Weld . . . . . . . . . . . . . 11 3.1.2.2 Request for Relief 12R-07, Examination Category B-D, item B3.110, Pressurizer Surge Nozzle-to-Bottom Head Weld . . . . 12 3.1.2.3 Request for Relief 12R-08, Examination Category B-D, item B3.120, Pressurizer Surge Nozzie Inner Radius Section . . . . . 13 3.1.2.4 Request for Relief 12R-09, Examination Category B-F, item B5.40, Pressurizer Dissimilar Metal Nozzler to-Safe End Welds. .............................................. 15 3.1.3 Heat Exchanger and Steam Generators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.1.3.1 Request fer Relief 12R-05, Examination Category B-B,'
- ltem B2.40, Steam Generator Tuuesheet-to-Channel Head Weld.................................................16
- ' 3.1.4 Piping Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 -
3.1.4.1 Request for Relief 12R-18, Exa > nation Category B-J, item B9.31, Reactor Coolant System Branch Conne': tion We ld s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.1.5 Pump Pressure Boundary (No requests for relief) . . . . . . . . . . . . . . . . . . . . . . 19 3.1.6 Valve Pressure Boundary (No requests for relief) ...... .. .. ........ 19 3.1.7 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 iv i_,....a_. '
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3.1.7.1 Request for Relief 12R-14, Examination Category F A, item F1.40, Examination of Reactor Vessel Supportr . . . . . . . . . . . . . . . , , . . . . . . . . . 19 3.2 C lass 2 Compor.vnts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.1 Pressure Vessels (No requests for relief) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.2 P i ping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.2.1 Request for Relief 12R-12. Examination Category C-F-1, items C5.12, C5.22, and C5.42, Class 2 Longitudinal Piping
_ Welds in Austenitic Piping Systems . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.3 Pumps (No roquests for relief) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
' 3.2.4 Valves (No roquests for relief) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.2.5 General (No requests for relief) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.3 Class 3 Compo,ents (No requests for relief) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.4 Pres sure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.4.1 Class 1 System Pressure Tests (No requests for relief) . . . . . . . . . . . . . . . . . 23 3.4.2 Class 2 System Pressure Tests . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . 23 3.4.2.1 Request for Relief 12R-13, Table IWC-5250-1, Examination Category C-H, items C7.30, C7.40, C7.70, and C7.80, Pressure Testing of Class 2 Components at Containment Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.4.3 Class 3 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 3.4.3.1 Request for Relief 12R-17, Examination Category D-B, item D2.10, System Pressure Testing of Class 3 Diesel Generator Subsystems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 3.4.4 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .29 3.4.4.1 Request for Relief 12R-02, IWA-5250(a)(2), System Pressure Test Corrective Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 3.4.4.2 Request for Relief 12R-01, IWA-5242(a), System Pressure Tests for Insulated Components in Class 1 Borated Systems . . . . . 31 3.4.4.3 Request for Relief 12R-19, Request for Authorization to Use ASME Code Case N-498-1, Attema#ve Rules /br 10 Year System Hydrosta6c Tes6ng tbr Class 1, 2, and 3 Systems, Sec60n XI, Divisi0n 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.5 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... . . 36 ...............
3.5.1 Ultrasonic Examination Techniques . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , . . . . 36 3.5.1.1 Request for Relief 12R 16, Qualification of Nondestructin Examination Personnel for Ultrasonic Examination . . . . . . . . . . . . . 36 3.5.2 Exempted Components (No requests for relief) . . . . . . . . . . . . . . . . . . . . . . . . 37
- 3. 5.3 Other . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 3.5.3.1 Request for Relief 12R-11, IWB-2420(b,c) and IWC-2420(b,c),
Successive inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
. 3.5.3.2 Request for Relief 12R 20, Use of Code Case N-509. Attema#ve Rules tbr tte Selec60n and Examinadon of Class 1, 2, and 3 Integrally-Welded Attachments, Sec00n XI, Division 1. . . . . . . . . . . 38 3.5.3.3 Request for Relief 12R-15, lWF-5300(a) Snubber Examination . . . . 39
- 4. CONCLUSION . . . .. .. ....... ............ ............ ...... . 40
- 5. R E F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . . .42. . . . . .
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. c TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
. WOLF CREEK NUCLEAR OPERATING CORPORATION, WOLF CREEK GENERATING STATION, DOCKET NUMBER 50-482
- 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 ASME Code caction XI, Rules forInservice Inspection of Nuclear Power Plant Components (Reference 2), to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section i
of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of the Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limit & lions and modifications listed therein, and subject to Nuclear Regulatory Commission (NRC) approval. The licensee, Wolf Creek Nuclear -
Operating Corporation, prepared the Wolf Creek Generating Station, Second 10-YearInterval inservice inspection Program Plan (Reference 3) to meet the requirements of the 1989 Edition of the ASME Code,Section XI. The second 10-year interval began September 3,1995, and ends September 2,2005.
As required by 10 CFR 50.55a(g)(5), if the licensee determines that certein 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 attemative requirements that are determined to be authorized by law, will not endanger lifo, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on tha facility.
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.- o Altematively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the licensee's determination that either (1) the proposed altomatives 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 sitematives may l>s used when authorized by the NRC.
- The information in the Wolf Creek Generating Stabon, Second 10-YearIntervalInservice Inspeckon Program Plan, submitted August 30,1995, 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 inservice inspection (ISI) program plan was performed using Standard Review Plans, NUREG-0800 (Reference 4), Section 5.2.4, " Reactor Ceolant Boundary Inservice Inspections and Testing," and Section 6.6, " Inservice inspection of Class 2 and 3 Components."
In a letter dated March 6,1996 (Reference 5), the NRC requested additional information that was required to complete the review of the ISI Program Plan. A conference call was held at the request of the licensee on April 24,1996, to clarify the required information. Ths requested information was provided by the licensee, Wolf Creek Nuclear Operating Corporation, in the Response to Request Ibr AdditionalIn/brmation, Second 10-year Inservice Inspection Interval, dated May 3,1996 (Referetice 6). In this rasponse, Wolf Creek Nuclear Operating Corporation, withdrew Request for Relief 12R-11, revised Request for Relief 12R-17, and submitted new Requests for Relief 12R-19 and 12R-20.
Additional infommation was requested by the NRC in letters dated October 10,1996 (Reference 7), November 25,1996 (Reference 8), and March 4,1997 (Reference 9). These additional RAls were sent to the licensee to obtain additional information and/or clarifications in order to evaluate the licensee's requests for relief with respect to the Code and Regulatory requirements. The requested information was provided by the licenses in submittals dated December 9,1996 (Reference 10), January 23,1997 (Reference 11), and May 6,1997 (Reference 12), respectively.
The Wolf Creek Generating Station, Second 10-YearIntervalInservice Inspeckon Program Plan is evaluated in Section 2 of this report. The ISI Program Plan is ev1luated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination -
exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous rev'mws.
The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI,1989 Edition, Specific inservice test programs for pumps and valves are being evaluated in other reports.
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- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consisted of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements an,. any previous license conditions pertinent to ISI activities. This section describes the su'amittals reviewed and the results of the review.
2.1 Documents Evaluated Review has been completed on the following information from the licensee:
(a) Wolf Creek Generating Station, Second 10-YearIntervalInservice Inspection Program Plan -
(Reference 3);
(b) Corrections to inservice Inspection Program Ibr the Second 10-yearInterval, dated September 20,1995 (Reference 13);
(c) Response to Request for AdditionalInformation, Second 10-yearInservice Inspection Interval, dated May 3,1996 (Reference 6);
I (d) Response to Request for AdditionalInformation, Second 10-year Inservice Inspection Interval, dated December 9,1996 (Reference 10); and (e) Response to Request for AdditionalInformation, Second 10-yearInservice Inspection Interval, dated January 23,1997 (Reference 11L (f) Response to Request for AdditionalInformation, Second 10-yearInservice Inspection Interval, dated May 6,1997 (Reference 12).
2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions inservice inspection program plans are based on the Codo editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of September 3,1995, the Code applicable to the second 10-year interval ISI program is the 1989 Edition. As stated in Section 1 of this report, the licensee has prepared the Wolf Creek Generating Station, Second 10-YearIntervalinservice Inspection Program Plan to meet the requirements of 1989 Edition of the Code.
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 applied to systems and components as attematives to Code requirements. ASME Code Cases that have been found suitable for use by the NRC are listed in the current revision of Regulatory Guide 1.147, inservice inspection Code Case Acceptability, with conditions for use, as applicable. These Code Cases must be implemented in their entirety
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and the licensee may adopt them by providing written notification to the NRC, Attematively, 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 has requested to implement the attematives to Code requirements contained in the following Code Cases:
Code Request for Case Relief Title N-498-1 12R-19 Altemative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems N-509 12R-20 Altemative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally-Welded Attachments N 522 12R-13 Pressure Testing of Containment Penetration Piping The licensee has stated that Code Cate N-416-1, Altemative Pressure Test Requirement for Welded Repairs or Installation of Replacement Items by Welding, Class 1, 2 and 3, will be adopted with its incorporation into Revision 12 of Regulatory Guide 1.147. If a repair or l replacement requires a hydrostatic test prior to Revision 12 of Regulatory Guide 1.147 being issued, a request for relief to use the subject Code Case will be required.
2.2.2 Acceptability of the Examination Sample inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1,2, and 3 components and their supports using sampling schedules described in '
Section XI of the ASME Code and 10 CFR 50.55a(b). Sample size and weld selection have been implemented in accordance with the Code 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, lWC-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 Wolf Creek Generating Station, Second 10-YearIntervalInservice Inspection Program Plan, and appear to be correct.
2.2.4 Augmented Exam! nation Commitments in addition to the requirements in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:
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(a) Reactor vessel examinations in accordance with the requirements of NRC Regulatory Guide
' 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, Revision 1 (Reference 14);
(b) Volumetric examination of the reactor coolant pump flywheel high stress areas every 3 years, as well as volumetric and surface examinations with the flywheel removed at 10-year
, intervals, satisfying NRC Regulatory Guide 1.14, Reactor Coolant Pump Flywheellntegrity (Reference 15);
(c) Volametric examination of circumferential eM longitudinal pipe welds within the "no break zone' associated with high energy piping in containment penetration areas in accordance with Branch Technical Position MEB 31, High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, and; (d) Examination of the reactor vessel closure studs in accordance with Regulatory Guide 1.65, Materials and Inspectrons for Reactor Vessel Closure Studs (Reference 16).
2.3 Conclusions Based on the review of the documents listed above, no deviations from regulatory requirements or commitments were identified in the Wolf Creek Generating Station, Second 10-YearIntervalInservice Inspection Program Plan, i
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- 3. EVALUATION Cc RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the second 10-year inspection interval are evaluated in the following sections.
3.1 Class 1 Components 3.1.1 Reactor Prersure Vessel 3.1.1.1 Reqmst ;or Relief 12R 03, Examination Category B.A, item B1.30, Reactor Vessel SNil-O-Flange Weld Code f.or,uirement-Examination Category B-A, item B1.30 requires 100% volumetric exami te.lon of the reactor vessel shell-to-flange weld as defined in Figure IWB-2500-4 Licensee's Code Rollef Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination of the reactor vessel shell-to-flange weld.
l Licensee's Basis for Requesting Relief (as stated}-
" Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested on the basis that compliance with me specified requirements is impractical. Conformance with the inservice inspection requirements would necessitate a design modification to the Reactor Pressure Vessel Flange that would allow 100% examination coverage of the subject weld.
"Due to the vessel flange taper above the subject weld, (inner and outer diameter configuration),100% examination coverage can not be achieved. Figure 1 represents a drawing (not to scale) of the Flange to Upper Reactor Vessel Shell Weld and the applicable beam directions / angles are listed with the appropriate examination coverage.
"The manual examination portion from the flange surface has not been performed for the Second 10-Year inspection Interval but greater than 90% coverage was attained for the first interval. The amount of coverage for weld metal (WM) interrogation is shown as a separate value to identify the amount of coverage in the volume where a crack would have a higher probability of presence. The limitations described above were included in a request for relief for the first ten year interval at WCGS, and relief was subsequently granted.
"A significant portion of the weld required volume was examined which should provide for
- detection of significant pattems of degradation. The limited Code examination of the subject weld in conjunction with full Code volumetric examination of other similar welds in the Reactor Pressure Vessel and performance of the periedic visual examination VT-2 provides an acceptable verification of the structurclintegrity of this pressure retaining weld."
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k The licensee provided the fo"owing information in their response to the NRC's RAl.
Automated Examination Coverage Scan Angle Clockwise Direction Counterclockwise Direction Breakdown for Parallel Scans 45' 51.3 % 51.3 %
60' 51.3 % 51.3 %
70' 65.2 % 65.2 %
Breakdown for Perpendicular Scans 45' - 48.9% 64.5% -
60* 76.8 % 42.4%
70* 19.1 % 78 %
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- Manual examination from the flange face was 90% complete for the first interval."-
. Licensee's Proposed Alternative Examination (as stated}-
"None; the Code required volumetric examination will be completed to the maximum extent practical."
Evaluation-The Code requires 100% volumetric examination of the reactor vessel shell-to-flange weld. : However, the flange taper precludes scanning for 100% coverage of the required examination area, thus making the Code-required voharnetric examination to the extent required by Code impractical. To obtain complete volumetric coverage, modifications or replacement of .
the component with one of a design providing for complete coverage would be required.-
Imposition of this requirement would cause a considerable burden on the licensee, The licensee proposed no additional examinations. However, based on the percentage of
. volumetric coverage that has been obtained from the shell side and the 90% coverage obtained from the flange face, it is reasonable to conclude that degradation, if present, will be detected.
Therefore, reasonable assurance of continued structural integrity is provided.
Conclusion-The flange taper makes the Code-required volumetric examination to the extent required by Code impractical; Based on the significant amount of weld coverage obtained,
- reasonable assurance of structural intsgrity is provided. Therefore, pursuant to 10 CFR 50.55a(g)(6)(l), it is recommended that relief be granted .
7
3.1.1.2 Request for Rollef 12R 04, Examination Category B A, item B1.21, Reactor Vessel Closure Head Circumferential Weld Code Requirement-Examination Category B A, item B1.21 requires 100% volumetric examination of reactor vessel closure head circumferential welds as defined in Figure IWB 2500-3.
Licensee's Code Rollef Roguest-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination of reactor vessel closure head Weld CH-103101, Licensee's Basis for Requesting Relief (as stated}-
" Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the inservice inspection requirements would necessitate a design modification to the Reactor Pressure Vessel Closure Head tu remove obstructions that preclude 100% examination coverage of the subject weld.
"Due to the cooling duct ring limiting the required sc n path and three lifting lugs obstructing the portions of the weld examination volume, the Code required examination cannot be 100% completed. The weld metal was 100% inspected in one direction by the 45 degree and the 60 degree angle beam, but was limited in the other direction due to the cooling ring. This resulted in 16.8% of the weld metal not being examined by the 45 degree angle beam and 47.3% not being examined by the 60 degree angle beam. '
Obstruction by three lifting lugs results in 5.3% of the required examination volume not being examined.
"The Reactor Vessel was designed and fabricated in accordance with the stringent i
quailty controls of ASME Section lil; subsequent volumetric and surface examinations as
- well as pressure testing was performed on this weld with acceptable results. The weld l was examined during Preservice inspections (PSI) and during the first interval inservice j Inspection (ISI) with no irregularities found. The probability of a flaw occurring and not being detected by the examinations previously performed is small. The limitations described above were included in a request for relief for the first ten year interval at WCGS, and relief was subsequently granted.
l~
l "A significant portion cf the weld required volume can be examined which should provide for detection of significant pattems of degradation. Therefore, reasonable assurance of the continued inservice structural integrity of the subject weld is achieved without performing a o complete Code examination."
Two figures depicting the limitations were included in the response to the requests for additional information dated May 3,1996, and January 23,1997.
8
- - - - - - . - - . ~ . . . . .
e-
Licensee % Proposed Alternative Examination (as stated}-
"None; the Code required volumetric examination will be completed to the maximum extent practical."
Evaluatfor>-The Code roouires volumetric examination of the reactor vessel closure head
" weld. However, the location of the cooling duct ring and three lifting lugs restrict scanning making the Oode-required 100% volumetric examination impractical. To obtain complete -
volumetric coverage, modifications or replacement of the component with one of a design
, providing for complete coverage would be required. Imposition of this requirement would cause a considerable burden on the licensee.
The licensee proposed to perform the examination to the extent possible. Based on the significant percent of volumetric coverage that has been obtained, it is reasonable to conclude that degradation, if present, would be detected. Therefore, reasonable assurance of continued structuralintegrity is provided.
Conclusion-The component configuration makes the Code-required volumetric examination impractical. Based on the significant percent of weld volume that was examined, reasonable assurance of structural integrity is provided. Therefore, it is recommended that relief be I granted, pursuant to 10 CFR 50.55a(g)(6)(i).
i 3.1.1.3 Request for Relief 12R 10, Examination Category B-G 1, item B6.10, Reactor Vessel Closure Head Nuts .!
Code Requiremont-Examination Category B-G-1, item B6.10 requires 100% surface examination of the reactor vessel closure head nuts.
Licensee % Code Rollef Request- Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed an attomative to performing 100% surface examination of the reactor vessel closure head nuts.
Licensee's Basis for Requesting Relief (as stated}-
" Table IWB-2500-1 of 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 Requirements / Figure Numbar* and ' Acceptance Standard'. These provisions were still in the course of preparation.
~
"The 1989 Edition of ASME Section XI, Category B-G-1 employs a VT-1 visual examination for nuts associated with Heat Exchangers, Piping, Pumpc, and Valves (item Numbers B6.140, B6.170, B6.200, and B6.230, respectively). These Category B-G-1
_ requirements also provide an Acceptance Standard, IWB-3517, for the VT-1 examinations. Accordingly, these rules are deemed by WCGS as an acceptable and complete sets of rules to assure the integrity of reactor vessel closure nuts.
9 m .
" Based on the above, WCNOC 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 % Proposed Alternative Examinadon (as statedh--
"As an alternate examination, WCGS will perform a VT-1 visual examination of the
- surface of all reactor closure head nuts, utilizing the acceptance criteria of IWB-3517, as delineated in the 1989 Edition of ASME Section XI."
, Evaluation-The licensee has proposed to perform a VT 1 visual examination of reactor pressure vessel (RPV) closure head nuts in lieu of the Code-required surface examination. All items in Examination Category B-G-1 except the reactor pressure vessel closure head nuts and
- the closure studs (when removed) require VT-1 visual examinations and/or volumetric examination (as applicable).
Typical 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 (i.e., boric acid attack). The Code requires a surface examination for closure head nuts. Surface examination procedures are typically qualified for the detection of linear flaws (cracks) and have acceptance criteria specifying only rejectable linear flaw lengths. Acceptance criteria are not provided in the 1989 Edition of the Code, item B6.10, as they were in the course of preparation when the Code was published. Without clearly defined acceptance criteria, conditions that require corrective measures may not be adequately addressed. The 1989 Addenda of Section XI addresses these problems by changing the requirement for the subject !
reactor pressure vessel closure head nuts from surface to VT-1 visual examination and providing appropriate acceptance criteria.
Article IWB-3000, Acceptance Standards, IWB 3517.1, VisualExamina#on, VT-1, describes conditions that require corrective action prior to continued service for botting and associated ,
nuts. One of these requirements is to compare crack-like flaws to the flaw standards of IWB-3515 for acceptance. Because the VT-1 visual examination acceptance criteria include evaluation of crack-like indications and other conditions requiring corrective action, such as deformed or sheared threads, localized corrosion, deformation of part, and other degradation mechanisms, it can be conc 8uded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure head nut. As a result, the INE' staff believes that VT-1 visual examination provides an acceptable level of quality and NW ty.
. Conclualon-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 visual examination on reactor pressure vessel closure head nuts, it is concluded that an ~l acceptable level of quality and safety will be provided by the proposed attemative. Therefere, it is recommended that the proposed attemative, VT-1 visual examination, be authorized pursuant to 10 CFR 50.55a(a)(3)(1).
10
_a
3.1.2 Pressurizer 3.1.2.1 Request for Rollef 12R 06, Examination Category B D, item B3.110, Pressurizer Nozzle to Vessel Weld Code Requi.+. ment-Exammation Category B-D, item B3.110 requires 100% volumetric
, examination of pressurizer nozzle-to-vessel welds as defined in Figure IWB-2500-7.
Licensee's Code Rollef Regte Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing 100% volumetric examination of pressurizer nozzle-to vessel Welds TBB03-10B-C-W and TBB03108-D-W.
Licensee % Basis for Requesdng Relief (as stated}- '
" Pursuant to 10 CFR 50.55a(g)(5)(lii), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the inservice inspection requirements would necessitate a design modification to the Pressurizer
- Nozzle Welds which would allow 100% examination coverage of the subject weld.
" Relief is requested from the Code requirement to volumetrically examine 100% of the subject welds due to the Pressurizer Nozzle weld taper and base metal taper near the nozzle causing transducer to lift off, As a result of this transducer lift off, a portion of the weld required volume (WRV) neceives no shear wave examination when using 45* and 60* angle beams. This was observed when examining Nozzle to Vessel Welds TBB03108-C-W and TBB03108-D-W for the first ten year interval, which relief had I
been granted. For both welds,15% of the WRV could not be scanned using a 45' angle beam and 11.5% of the WRV could not be scanned using a 60* angle beam. These same percentages are expected, as a maximum, for the second ten year interval.
" Strict ASME Section lli quality controls were used when designing, fabricating, and installing these welds, in addition, these welds were examined using ultrasonic equipment, showing only geometrical indications. The probability of a flaw occurring only in one of these areas not being examined is extremely small.
"A significant portion of the weld required volume can be examined which should provide for detection of significant pattems of degradation. Therefore, reasonabie assurance of the -
continued inservice structural integrity of the subject weld is achieved without performing a complete Code examination."
Licensee > Proposed Alternative Examination (as stated}-
"None; the Code required volumetric examination will be completed to the maximum extent practical."
r Eve JatiorHThe Cade requires volumetric examination of pressurizer nozzle-to-vessel welds.
However, the tapers of the weld and base metal preclude achieving 100% coverage of nozzle =
to-vessel Welds TBB03-108-C-W and TBB03-108-D-W. These tapers cause the transducer to
" lift off", making the Code-required 100% volumetric axamination impractical. To obtain 11
- . a complete volumetric coverage, modifications or replacement of the component with MS of a design providing for complete coverage would be required. Imposition of this requirement wou
- d cause a considerable burden on the licensee.
i The licensee proposed to perform these examinations to the extent possible. Based on the significant amount of volumetric coverage that has been obtained (> 85% of each scan), it is
, reasonable to conclude that degradation, if present, will be detected. Therefore, reasonable i assurance of continued inservice structural integrity is provided.
Conclusion-The component configurstloa restriction makes the Code-required volumetric examination impractical. Based on the significant amount of weld coverage obtainable, reasonabir annurance of continued structural integnty is provided. Therefore, it is recommended that relief be granted, pur:;uant to 10 CFR 50.55a(g)(6)(i),
1 3.1.2.2 Request for Rollef 12R 07, Examination Category B D, item B3.110, Pressurizer Surge Nozzle to Bottom Head Weld Code Requirement-Examination Category B D, item B3.110 requires 100% volumetric examination of all r assurizer nozzle to-bottom head welds as defined in Figure IWB 2500 7(b).
Licensee's Code Rollef Request-The licensee requested relief from 100% volumetric examination of pressurizer surge nozzle to-bottom head Weld TBB0310A W.
Licensee's Basis for Roquesting Relief (as stated)-
" Pursuant to 10 CFR 50.55a(g)(5)(ill), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the inservice inspection requirements would necessitate a design modification to the Pressurizer Surge Nozzle and Pressurizer Heater to remove obstructions that preclude 100%
0xamination coverage of the subject weld.
"The Pressurizer shell configuration and the proximity of the heaters results in 35% of the weld required volume not being examined with 0 degree or with the parallel angle beam scan (See Figure 1)' in addition, the v, eld metal can only examined in the one direction and the adjacent base metal was not fully examined in one directbn with two beam angles (92% complete with 60 degree angle beam and 88.4% complete with 45 degree angle beam, See Figure 2)'.
" Strict ASME Section til quality controls were used when designing, fabricating, and installing this weld, in addition, this weld was volumetrically examined (PSI as well as current ISI), with no irregularities found. The probability of a flaw occurring only in one of these areas not being examined is extremely small. The limitations described above were included in a request for relief for the first ten year Interval at WCGS, and relief was subsequently granted.
- a. Figures 1 and 2 are not included in this report.
12
"A significant portion of the weld required volume can be examined which should provide for detection of significant pattems of degradation. Therefore, reasonable assurance of the continued inservice structural integnty of the subject weld is achieved without performing a complete Code examination."
Licensee's Proposed Alternative Examination (as stated}-
, "None; the Code requ! red volumetric examination will be completed to the maximum extent practical.'
Evaluation-The Code requires volumetric examination of the pressurizer nozzle-to-bottom head weld. However, the geometric configuration of the nozzle to head area precludes 100%
coverage of nozzle to-bottom head Weld TBB0310A W and makes the Code-required volumetric examination impiactical. .'o obtain complete volumetric coverage, modifications or replacement of the component with one of a design providing for complete coverage would be required. Imposition of this requirement would cause a considerable burden on the licensee.
The licensee proposed to perform the examination to the extent possible. The weld and adjacent base metal on the head side were examined in three of the four Code-required directions. This represents a significant portion of the Code-required examination volume and degradation, if present, should be detected. Therefore, reasonable assurance of continued structuralintegnty is provided.
Conclusion-The component configuration restriction makes the Code required volumetric examination impractical. Based on the extent of volumetric coverage achievable, reasonable assursnee of structuralintegrity is provided. Therefore,it is recommended that relief be granted, pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.2.3 Request for Rollef12R 08, Examination Category B D, item B3.*,.20, Pressurizer Surge Nozzle Inner Radius Section Code Requirement-Examination Category B-D, item B3.120 requires 100% volumetric examination of all pressurizer nozzle inner radius sections es defined in Figure IWB-2500-7(b).
Licensee's Code Relief Request-Pursuant to 10 CrR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination of pressurizer surge nozzle inner radius section TBB03-10A IR.
Licensee's Basis for Requesting Relief (ss stated}-
" Pursuant to 10 CFR 50.55a(g)(5)(ill), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the inservice inspection requirements would necessitate a design modification to the Pressurizer Surge Nozzle and Pressurizer Heater to remove obstructions that preclude 100%
examination coverage of the subject weld.
13
- Relief is requested from the Code requirement to volumetrically examine 100% of the subject welds due to the interferences by the Pressurizer heater penetrations and the nozzle configuration. As a result,70% of the required weld volume cannot be inspected.
See Figure 1.6
" Strict ASME Section 111 quality controls were used when designing, fabricating, and
- installing this weld. In addition, this weld was volumetrically examined (PSI as well as current ISI), with no irregularities found. The probability of a flaw occurring only in one of these areas not being examined is extremely small. The limitations described above
, were included in a request for relief for the first ten year interval at WCGS, and relief was subsequently granted.
"A portion of the weld required volume can be examined. In cddition, other pressurizer nozzles are receiving examinations. Therefore, reasonable assurance of continued inservice structuralintegrity of the subject weld is achieved without performing a complete Code examination."
Licensee's Proposed Alternative Examination (as stated}-
- None; the Code required volumetric examination will be completed to the maximum extent practical."
Evaluation-The Code requires volumet ic examination of the pressurizer nozzle inside radius sections. However, volumetric examination of the surge nozzle inner radius is restricted by heater penetrations and the nozzle's geometric configuration. These obstructions make the volumetric examination impractical to perform to the extent required by the Code. To meet the Code requirements, the surge nozzle and adjacent obstructions would have to be modified to allow access for examination imposition of this requirernent would create a considerable burden on the licensee.
Approximately 30% of the pressurizer surge nozzle inside radiu0 %ction can be examined, in addition, other pressurizer nozzl3 inner radius sections are receiving complete volumetric examination. Therefore, significant pattems of degradation should be detected by the examinations that are being performed, and reasonable assurance of the structuralintegrity is provided.
Conclusion-The pressuri' tr heater penetrations and the nozzle configuration design make the Code-required 100% volumetric examination of the subject pressurizer surge nozzle inner radius section impractical to complete. Based on the portion of the surge nozzle inner radius that can be examined, in corl unction with the Code examination of other prer.ririzer nozzle inner radius sections, reasonable assurance of structural integrity !s provided. Therefore, it is recommended that rclief be granted, pursuant to 10 CFR 50.55a(g)(6)(i).
- b. Figures are not included in this report.
14
. _ _ . . . . . . . . . _ _ +
3.1.2.4 Request for Rollef 12R 09, Examination Category B F. Item B5.40, Pressurizer Dissimilar Metal Nozzle to Safe End Welds Code Requirement-Examination Category B F, item B5,40 require 5100% volumetric and surface examination of the pressurizer dissimilar metal nozzle to safe end welds as defined in Figure IWB 2500-8.
Licensee's Code Rollef Request-Pursuant to 10 CFR 50.55a(a)(3)(li) the licensee requested relief from 100% volumetric examination of the following pressurizer nozzle-to safe end welds:
Weld Number Weld Description Percent Not Examined TBB03-4 W Relief Nozzle-to-Safe End 20% - 60' axial scan 45% 45' axial scan T2903-3 A W Safety Nozzle-to Safe End 50% 60' axial scan 35% 45' axial scan TBB031 W Surge Nozzle-to-Safe End 15% 60' axial scan 40% 45' axial scan TBB03 3 B W Safety Nozzle-to Safe End 55% 60' axial seca 40% 45' axial scan TBB03 2 W Spray Nozzle-to-Safe End 10% - 60' axial scan 40% - 45' axial scan TBB03-3-C W Safety Nozzle to-Safe End 20% - 60' axial scan 40% 45' axial scan Licensee's Basis for Requesting Relief (as stated}-
I
" Relief is requested from the Code requirement to volumetrically examine 100% of the subject welds due to surface undulations which restrict search unit inovement and due to the metal structure of inconel buttering which inhibits ultrasonic shear wave transmission.
"The data presented in the previous table was obtained during Interval 1 examination; however, WCNOC anticipates that use of improved examination techniques and equipment may capture more of tha required weld volume but still not attain the 90%
- coverage as allowed by Code Case N-460. The limitations described above were Included in a request for relief for the first ten year Intervil at WCGS, and relief was subsequently granted.
" Strict ASME Section lli quality controls were used when designing, fabricating, and installing this weld. In addition, this weld was volumetrically examined (PSI as well as current ISI), with no irregularities found. The probability of a flaw occurring only in one of these areas not being examined is extremely small. Future indications of s:gnificant size will be found by the examination of the weld at lt is now.
15
_ = _ _ _ __
2
-+'4 . wee N-fyh g 4 e.
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- 4' Mbi WSM%-* - *""
N
" Based on the above information, reasonable assurance of the continued inservice structural integrity of the subject welds is achieved without performing a complete Code examination.
" Compliance with the applicable Code requirements can only be accomplished by redesigning and refabricating the Pressurizer. WCNOC deems this course of action a
, hardship without a compensating increase in the level of quality and safety."
Licensee's Proposed Alternative Examination (as statedH "None; the Code required volumetric examination will be completed to the maximum extent practical' Evaluation-The Code requires volumetric and surface examination of the pressurizer dissimilar metal nozzle to safe end welds. However, the configuration of the subject nozzle-to-safe end welds precludes achieving 100% volumetric examination.
The licensee anticipatrs that the improved examination techniques that will be used during the second interval will improve coverage of the subject welds. Therefore, the INEEL staff recommends that the licensee resubmit for relief (if necessary) after performing these examinations. Further, the INEEL staff recontmends that the IMensee review NRC Information Notice No. 90 30, Ultrasonic Inspection Techniques for Dissimilar Metal Welds for guidance.
Conclusion-Since increased coverage will be obtained this interval due to enhanced examination techniques, it is ecommended that the licensee's attemative be denied at this time.
The licensee should resul; nut after completion of these skaminations if complete coverage
(>90%) is not obtained.
3.1.3 Heat Exchanger and Steam Generators 3.1.3.1 Request for Reilef 12R 05, Examination Category B 8, item B2.40, Steam Generator Tubesheet to Channel Head Wold Code Requiremont-Examination Category 0 B, item B2.40 requires 100% volumetric examination of the steam generator tubesheet to-channel head weld as defined in Figure IWB 2500-6.
Licensee's Code Rollef Requew . . _ ant to 10 CFR 50.55a(g)(5)(lii), the licensee requested relief from 100% volumetric examination of steam generator tubesheet to-channel head Weld EBB 01D-SEAM-1 W.
Licensee's Basis forRequesting Rollef(as stated >
" Pursuant to 10 CFR 50.55a(g)(5)(ill), relief is requested on the basis that compliance with the specified requirements is impractical. Confntmance with the inservice inspection requireme% would necessitate a design modification to the Steam Generator and associated supports to remove obstructions that preclude 100%
examination coverage of the subject weld.
16
_ __ =
=wwg,e' _eNg tga$.gS s.-4.@-e @h C MtN8 S JF F 4 h a M'*'*"W' C '"'
"*N--*M.'# # "
V
i . .
1 i
I !
i i "Due to the four (4) support legs obstructing 22.4% of the volumetric examination with, at !
j least an additional 9.3% and 33.4% o' the volumetric examination obstructed by the
- component design, while using the 45' and 60' angle beam, respectively, the Code required examination cannot be completed.
(
j "The Steam Generator was designed and fabricated in accordance wrth the stringent .
, quality controls of ASME Section lil; subsequent Preservice and first interval inservice i ultrasonic scans at 0,45, and 60 degrees revealed no indications. The probability of a i !
flew occurring and not being detected by the examination previously performed is s, mil. >
j ,
The limitations described above were included in a request for relief for the first ten year j interval at WCGS, and relief was subsequently granted.
i "A significant portion of the weld required volume can be examined which should provide for
! detection of significant pattoms of degradation. Therefore, reasonable assurance of the
! continued inservice structural integrity of the subject weld is achieved without performing a ,
! complete Code examination."
j Two figures depicting the limitations were included in the response to the request for additional
! information dated May 3,1996, i
- Licensee's Propoac*t Alternative Examination (as statedh-l "None; the Code required volumetric examination will be comp!sted to the maximum 4
extent practical?
- i j' Evaluation-The Code requires volumetric examination of the steam generator tubesheet to-4 l
channel head weld. However, the obstruction caused by the four support legs and steam generator design geometry make the Code required 100% volumetric examination of
- tubesheet to-channel head Wold EBB 01D SEAM 1 W, impractical to complete. To perform a
- complete volumetric examination, modifications or replacement of the component with one of a j design providir,g for complete coverage would be required, imposition of this requirement would
- cause a considerable burden on the licensee.
- l. The licensee proposed to examine the subject weld to the extent possible. Based on the
! significant amount of volumetric coverage that is obtainable, it is reasonable to conclude that a
! pattom of degradation, if present, will be detected. Therefore, reasonable assurarx:e of i continued structural integrity is provided, i
- ,-
- Conclualon-The component configuration and support restrictions make the Code-required j- ._ volumetric examination impractical. Based on the significant amount of weld coverage l obtainable, reasonable assurance of structural integrity is provided. ; Therefore, et is recommenced that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
I i
- 17
=_
L
_.. . . _ _ _ - . - . _ _ - . . - . - _ _ ~ _-.
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3.1.4 Piping Pressure Boundary !
3.1.4.1 Request for Relief 12R 18, Examination Category B.J. Item St.31, Reactor Coolant System Branch Connection Wolds !'
Code Requirement--Examination Category B J, item 89.31 requires 100% volumetric and surface examination of piping branch connection welds as defined in Figure IWB-2500 9. ;
Licensee's Ccde Rollef Moguest-Pursuant to 10 CFR 50.55a(g)(5)(ill), the licensee requested relief from 100% volumetric examination of piping branch connedion Welds BB-01 S1017, BB-01 S302 3, and BB-01 S402 3.
" Licensee *s Basis for Roguenting Relief (as stateefh-
" Pursuant to 10 CFR 50.55a(g)(5)(ill), relief is requested on the bcsis that compliance with the specified requirements is impractical. Confom)ance with the inservice inspection requirements would necessitate a design modification to the Reactor Coolant j' System loop piping to allow 100% examination of the subject weld.
I i
j 1
" Volumetric examination of these welds is limited to the pipe side only due to component j
geometry (pipe branch connection geometry) and metallurgical properties (contrifically j i
{ sic) cast stainless steel). Because of the coarse grain material and high attenuative l i
nature of the materials, it is necessary to use a refracted longitudinal sound wave to l
'- achieve the best ultrasonic response. This type of wave cannot be extended to provide to be,am path direction coverage, i
" Strict ASME Section. lllq' uality controls were used when designing, fabricating, and installing these welds, in addition, these welds were examined using ultrasonic equipment to the fullest extent possible, including examination of 100% of the volume in j= two beam path directions for reflectors transverse to the weld seem, with no irregularities i
- identified. This fact, in conjunction with the surface examination results and RCS visual i
examinations (VT-2) following each refueling outage, provides confidence that the welds
- are structurally sound and that the limited exam does not compromise the health and safety of the publie *
[-
l- The licensee proved the following information in their response to the NRC's RAl.-
4
}?, "These are Reactor Coolant System branch connection components that allowed only a
[ single sided examination (pipe side only). Coverage of 100% was achieved from the
- one side and of both directions on the parallel scan (comoosite coverage of 75%).
- l. However, because of material properties as discussed in the relief request, tto i examiners were unable to accompibh an extended beam path to examine from the
- opposing direction.'
y 1
i F 18
,-6-% K<r .T 1mA- n. Si dSON &' 94" M 918$} mgwam. M Q+ptOIw;&quum-Qt%MQM' O Y Q b& - _ _ . , _A- Vm n& -'M<4g "eumb a 4 4+- y :.,-q q-, s-
8 0 Licensee's Proposed Alternative Examination (as stated)-
"None; the Code required volumetric examination will be completed to the maximum extent practical."
Evaluation-The Code requires volumetric and surface examinations of the subject piping
- branch connection welds. However, the configuration and metallurgical properties preclude achieving 100% coverage of pipe branch connection Welds BB 01 S1017, BB-01 S302 3, and BB-01.S402 3.
4 The pipe branch connection geometry makes the Code-required 100% volumetric examination imprav.ical. To obtain complete volumetric coverage, modification or replacement of the branch connections with connections of a design providing for complete coverage would be required. Impcsition of this requirement would cause a considcrable burden on the licensee.
The licensee proposed to examine the subject welds to the extent possible. Extended beam paths were considered and found to be impractical due to the metallurgical properties of centrifugally cast stainless steel. However, based on the significant amount of the volume that will be examined (75%), it is reasonable to conclude that degradation, if present, will be detected. Therefore, reasonable assurance of continued structuralintegrity will be provided.
Conclusion-The component configuration restriction makes the Code required volumetric examination impractical. Based on the significant amount of weld coverage obtainable, reasonable assurance of structural integrity is provided. Therefore, it is recommended that relief be granted, pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.5 Pump Pressure Boundary (No requests for relief)
, 3.1.6 Valve Pressure Boundary (No requests for relief) 3.1,7 General 3.1.7.1 Request for Relief 12R.14, Examination Category F A, item F1.40, Examination of Reactor Vessel Supports Code Requirement-Code Case N-491, Tabla-25001, Examination Category F-A, item F1.40 requires 100% VT 3 visual examination of all Clast 1,2, and 3 supports other than piping
. supports.
Licensee's Code Rollef Request-Pursuant to 10 CFR 50.55a(g)(5)(lii), the licensee requested relief from performing 100% of the Code-required VT-3 visual examinations for the reactor vessel supports.
Licensee's Basis for Requesting Relief (as stated}-
" Pursuant to 10 CFR 50.55a(g)(5)(lii), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the inservice 19
. . _ ~ _ _ . . _ _ _ . _ _ _ _ . . . _ _ . .
. . i i
inspection requirements would necessn.W d design modification to the Roacr Pressure Vessel Supports and associated insulation /walkplate, to allow 100% visual i examination of the subject weld. . .
"The WCGS Reactor Vessel is supported by two cold leg noules and two hot leg ;
- noules. There is a support assembly at each of these noules that consists of a noule
, _wold build up, shoe plate, air cooled box, and steel support structure embedded in the 1 primary shield wall. Figure 1' depicts these support assemblies. As shown in the fgure, i only thS noule wold build up and shoe plate are completely accessible for a visual VT 3 i
, examination. The majority of the air cooled box and the entire steel support structure ers - '
located beneath a steel walk plate and only the top of the air cooled box is directly accessible. An additional 20 to 30 percent of the air cooled box and a very small _
percentage of the steel support structure would be made accessible if the steel walk plate and insulation were removed.
"The Reactor Vessel supports are located in a confined space below the refueling pool permanent seal ring. The area can only be accessed inrough four seal ring hatches. In addition to difficult access, the radiu lon in the area is between 1.5 to 2.0 man-rom per
- hour. It is estimated that the removal and re installation of the walk plate and insulation in this confined space, combined with the visual VT 3 examination, would result in an
- exposure of approximately 06 man-rom. Removal of the walk plate and insulation under these conditions to increase the examination of the air cooled box by approximately 20 ;
! to 30 percent and a very small percentage of the steel support structure is considered
- . impractical without a commensurate increase in quality or safety. Based on this, relief is requested from the visual VT-3 examination of the air cooled box and steel support
- structure that is obstructed by the walk plate and insulation. The limitations described ,
[ . above were included in a request for relief for the first ten year interval at WCGS, and j relief was subsequently granted.
L "A visual examination VT 3, with the walk plate and insulation installed, shall be performed to i' the extent practical on the accessible NF portions of the Reactor Vessel support assemblies i to satisfy the requirements of Code Case N-491, Table 25001, item No. F1.40. If L conditions are discovered during this limited visual examination VT 3 that do not meet the acceptance standards of N-491, .3400, the walk plate or insulation will, if necessary, be removev 10 neet the evaluation requirements of N-491,-3112.2 or 3112.3."
- Licensee % Proposect Altemative Examination (as statedf--
l . "None; the Code required visual examination VT-3 will be completed to tha rnaximum extent practical."
l Evaluation-Code Case N-491 requires a VT-3 visual examination of the reactor vessel supports. _However, due to access restrictions, the support design, and high local radiation
- levels, the licensee proposed to perform a limited visual examination.
i.
]
- c. Figure is not included as part of this evaluation.
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The support assembly at each of tTe reactor vessel nonles consists of nonle weld build up, shoe plate, air-cooled box, and steel support structure. The steel support structure is embedded in the primary shield wall, making the VT 3 visual examination of the reactor vessel support impractical to perform to the extent required by the Code.
The licensee's proposed attemative, to perform a VT 3 visual examination of the entire nonle weld build up and shoe plate and approximately 30% of the air-cooled box, will provide reasonable assurance of continued structuralintegrity.
, ConcluslorHDue to the design of the reactor vessel supports, it is impractical to perform the VT 3 visual examination to the extent required by the Code. The licensee's attemative will provide reasonable assurance of structural integrity; therefore, it is recommended that relief be granted pursuarn to 10 CFR 50.55a(g)(6)(l).
3.2 Class 2 Components 3.2.1 Pressure Vessels (No requests for rollef) 3.2.2 Piping 3.2.2.1 Request for Rollef 12R 12 Examination Category C F.1, items C5.12, C5.22, and C5.42, Class 2 Longitudinal Piping Welds in Austenitic Piping Systems Code Requirement-Examination Category C-F 1, items C5.12, and C5.22, require surface and volumetric of examination Class 2 longitudinal piping welds. Item C5.42 requires surface examination of Class 2 longitudinal piping welds. The examination volume / surface area includes 2.5t at the intersection with circumferential welds required to be examined.
Licensee's Code Rollef Request-Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed '
an attemative to performing 100% of the Code-required surface and volumetric examinations on Class 2 longitudinal piping welds in austenitic piping systems to the extent required by Code.
Licensee's Basis for Requesting Relief (as stated}--
- Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed attemative would provide an acceptable level of quality and safety.
" Based on the following discussion, the performance of surface and volumetric examination on longitudinal piping welds has a negligible compensating effect on the quality or safety of Class 2 pipir'g. In addition, there is little if any, technical benefit associated with the performance of these examinations, and may result in a substantial man-rem exposure and cost.
"1) Throughout the nuclear industry, there has been no evidence of rejectable service induced flaws being attributed to longitudinal piping welds.
21
"2) During the first inservice inspection intervsl at WCGS, no inservice flaws were detected in longitudinal piping welds.
"3) There are distinct differences between the processes used in the manufacturing of longitudinal and circumferential wolds which enhance the integrity of longitudinal welds. First, longitudinal welds are typically manufactured under controlled shop
- conditions whereas circumferential welds are produced in the field under less than Ideal conditions. Secondly, longitudinal wolds usually undergo heat treatment in the shop which improves their material properties and relieves the residual stresses
, created by welding. Finally, shop manufacturing inspections can be performed ;
under more favorable conditions which further increase the confidence level of the longitudinalweld quality.
- "4) During field installation of hing, the ends of the longitudinal welds may be affected during welding of the intersecting circumferential field welds. This small area falls -
within the circumferential weld inspection boundaries. Therefore, the ends of the longitudinal welds will still be subject to examination.
"5) From industry wide standpoint, there has been no evidence of longitudinal weld defects compromising safety at nuclear generating facilities.
l l
"6) No significant loading conditions or known material degradation mechanisms have become evident to date which specifically relete to lor gitudinal seam welds in nuclear plant piping.
"7) There is a significant accumulation of man-rom exposure and cost associated with the !nspection of Class 2 longitudinal piping welds.
"8) The proposed attemative examinations provide an acceptable level of quality and safety without causing undue hardship or difficulties."
Licensee > Proposed Alternative Examination (as statedF-
" Surface and volumetric examinations shall be performed, as applicable, on the length of the longitudinal weld that is normally examined during examination of the intersecting circumferential weld (s). The volumetric examination at the intersection of circumferential and longitudir.al welds will include both transverse and parallel scans within the length of the longitudinal weld the t falls within the circumferential wvld examination boundary."
Evaluettork -The licensee requested relief from performing the surface and volumetric examinations, to the extent required by Code, of the longitudinal welds in Class 2 piping. The
- licensee proposes to examine the potentially critical portions of the longitudinal welds (the portions that intersect circumferential welds) in conjunction with examination of the circumferential welds.
The licensee's altomative is based on the position that, due to fabrication controls and lack of susceptibility to the conditions that lead to failure, longitudinal welds are unlikely to fall. The potentially critical portions of the longitudinal welds are the portions that intersect circumferential 22
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l Welds; these regions will be examined in conjunction with the circumferential welds. With this attemative, the most critical area of the long tudinal weld is examined, thus providing an acceptable level of quality and safety. Based on the ouality of longitudinal welds and the extent of examinations performed, this provides an acceptable level of quality and safety.
Conclusion-An acceptable level of quality and safety is provided by the licensee's proposed
' attemative. Therefore, it is recommended that the licensee's altemative be approved pursuant to 10 CFR 50.55a(a)(3)(l).
, 3.2.3 Pumps (No requests for relief) 3.2.4 Valves (No requests for relief) 3.2.6 General (No requests for relief) 3.3 Class 3 Components (No requests for relief) 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests (No requests for relief) 3.4.2 Class 2 System Pressure Tests 3.4.2,1 Request for Rollef 12R 13, Table IWC-62501, Examination Category C H, items C7,30, C7.40, C7,70, and C7.80, Pressure Testing of Class 2 Components at Containment Penetrations Code Requirement-Examination Category C-H, items C7.30, C7.40, C7.70, and C7.80, in conjunction with Code Case N-490-1, require system leakage testing of Class 2 piping and valves once each inspectioa period.
Licensee's Code Reflef Request-Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed an attemative to performing the Code-require.1 systein teakage test for the following Code Class 2 piping and valves at containment penetrations where the balance of the system is outside the scope of Section XI.
e 23
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Table 1 Line Number Penetration Description BB 103-HCB 1" P-62 M 12BB02 BL-028-HCB 3" P 25 M-12BLO1 BM-053-HBB 3" P 78 M-12BM01 EC-067 HCB-6" P 53 M 12ECO2 EC 072 HCB-6" P 54 M 12ECO2 EC-081 HCB 3" P 55 M 12ECO2 EM-071 BCB-3/4" P 92 M 12EM01 GP 003 HBB 1" P 51 M 12GP01 l GP-005-HBB 1" P 51 M 12GP01 GS-025 HBB-6" P-65 M 12GS01 GT 007 HBB 36" V 160 M 12GT01 s
GT-004 HBB-36" V 161 M 12GT01 GT-029-HBG 18" V-161 M 12GT01 GT-034 HBB 18" V-160 M-12GT01 GT-033 HBB 1R" V 160 M-12GT01 GT-030-HBb la" V 161 M 12GT01 HB-015-HUB 3" P 26 M-12HB01 HB-025-HBB-3/4" P-44 M-12HB01 HD-015-HBB 2" P-43 M 12HD01 KA 244-HCB-1%" P 30 M 12KA01 KA 259-HCB-1%" P-30 M 12KA01 KA-051-HBB-4" P 63 M-12KA02 KA 261-HBB 1" P-63 M 12KA02 KA 732-HBB 1" N/A M 12KA05 KA 733-H681" N/A M 12KA05 KB-001 HCB 2" P 98 M-12KA05 KC 560 HBB-4" P-67 M-12KCO2 LF-842 HCB-6" P-32 M 12LF09 SJ-002 BCB 1" P-64 M 12SJ01 SJ-003 ECB-1" P-95 M 12SJ01 24
, i 1
i Licensee's Basis for Requesting Relief (as stated >
"Specifically WCNOC request relief from the requirement to perform a pressure test in accordance with ASME Section XI, Table IWC-25001. Examination Category C-H on the Code Class 2 lines listed in Table 1.
"The lines listed in Table 1 are portions of non safety related piping systems that penetrate the primary reactor containment. At each containment penetration, the process pipe is classified Code Class 2 and provided with isolation valves that are either locked shut during normal operation, capable of automatic closure, or capable of remote ,
[' closure to support the :ontainment safety function. These components preform no othe.-
4 safety function.
"The primary reactor containment integrity, including all containment penetrations, is periodically verified by performing leakage tests in accordance with a 10 CFR 50, Appendix J. Each of the Code Class 2 lines listed in Table 1 and their associated l lsolation valves are terted during an Appendix J, Type A, B, or C leakage test at a pressure not less than 48.1 psig.
" Performance of these Appervix J leak tests verifies the integrity of the subject Code
!- Class 2 lines at each respective penetration. The performance of ASME Section XI,
} Examination Category C-H pressure tests on these same lines will provide little, if any, ;
j_
additional verification of primary reactor containment integrity and are thus redundant.
- "Por the preceding information, WCNOC request relief from the ASME Section XI requirements for pressure testing these Code Class 2 containment penetration components on the basis that Proposed Allemate provisions provide an adequate level j_ of quality and safety." 4 Licensee's Proposed Alternative Examination (as stated > '
"WCNOC shall perform 10 CFR 50, Appendix J leakage tests on the primary reactor containment penetration lines listed above, and on their associated valves as required -
by 10 CFR 50, etc. WCGS Technical Specification 3/4.6."
- In the December 9,1996 response to the NRC RAl, the licensee provided the following additionalinformation:
- - "(1) When Section XI pressure testing credit is to be taken for Appendx J, Type C tests, the peak calculated containment pressure as defined by the Technical Specifications will be used as the test pressure.
"2) When Appendix J Type C testing of the containment penetration listed in Table I of 12R 13, and on the associated valves,' results in identification of a 0 (zero) leak rate, ASME Section XI pressure testing requirements are satisfied. However, when a leak rate is present and air or gas is used as the testing medium, the test procMure will include methods for detection and location of through-wall leaks when credit for ASME Section XI pressure test requirements is desired."
25
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Evaluation-The system leakage test required by Examination Category C H pro / ides periodic verification of the leak-tight integrity of Class 2 piping systems or segments orece every 40 months. Pipe segments from non-Code class systems that penetrate containment are
- designed and examined as Class 2 pipe to protect the integrity of containment. The Appendix J pressure testing provides periodic vocification of the leak tight integnty of the primary reactor containment, and of systems and components that penetrate containment. The Appendix J test
, frequency provides assurance that the containment pressure boundary is being maintained at an accuptable level while monitoring for deterioration of seals, valves, and piping.
The Class 2 containment isolation valves (CIW) and connecting pipe segments must withstand the peak calculated containment intomal pressure related to the maximum design containment pressure. The containment penetration piping is classified as Class 2 because it is - -
part of the containment pressure boundary, and because containment integrity is the only safety-related function performed by this piping. Therefore, it is logical to test the penetration piping portion of the associated system to the Appendix J criteria. The INEEL staff finds that the pressure retaining integrity of the CIVs and connecting piping, and their associated safety functions, may 1:e verified with an Appendix J, Type C test if it is conducted at the peak calculated containment pressure.
l IWC-5210(b) requires that where air or gas is used as a testing medium, the test procedure shall include methods for detodion and location of through-wall leaks in system components.
Because an Appendix J, Type C test most likely uses air as a testing medium, the licensee's test procedure should meet the above requirement for the CIVs and pipe segments between the CIVs. The licensee has committed to this requirement.
Conclualon-The INEEL staff concludes that compliance v/ith Appendix J would provide an acceptable level of quality and safety for the subject Class 2 piping that penetrates containment, where the balance of the piping system is non-Code class. Thorofore, it is recommended that the proposed attomative 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 Rollef 12R 17 Examination Category D 8, item D2.10, System Pressure Testing of Class 3 Diesel Generator Subsystems Code Requiremont-Examination Category D B, item D2.10 requires system leakage testing '
of Class 3 piping and valves once each inspection period and hydrostatic testing once each
. Interval.
Licensee's Code Mohef Request-Pursuant to 10 CFR 50.55a(s)(3)(i), the licensee proposed an attemative to performing the Code required system leakage test for the Code Class 3 diesel generator subsystems.
26
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Licensee's Basis for Requesting Relief (as stated}--
"The focus of the Technical Specification Surveillance Requirements !s slightly different; l 4.8.1.1.2 concentrates on component operability and 4.0.5 (inservice inspection) concentrates on component pressure boundary integrity. Because successful Standby Emergency Diesel Generator operability testing requires the associated subsystems such as the Starting Air, Jacket Cooling Water, intake and Exhaust Air, and Fuel Rack Supply Air to maintain pressure boundary integrity and therefore deemed to provide an equivalent level of quality and safety to that of ASME Section Xi inspections.
- "The repeatability of subsystem instrumentation (pressure and temperature) recorded during testing provides supporting data for the ' indirect verification of component integrity ' In addition, operations personnel specifically trained in the design and testing of the Standby Emergency Diesel Generators are aware of the necessity to maintain pressure boundary integrity for certain components and also of the necessity to maintain adequate flow characteristics for open ended components which provide intake air and process exhaust air. Although not a specific step in the Surveillance Procedure, visual .
venfication of component pret,sure boundary integrity and adequate flow in an understood responsibility of the operations personnel performing Standby Emergency Diesel Generator operability testing. If evidence of leakage or inadequate flow is identified during testing, written notification is forwarded to plant maintenance for corrective actions or repairs and follow-up confirmatory testing is performed.
"The following paragraphs provide specific procedural actions which support the use of the attemative operability testing in lieu of ASME Section XI System Pressure Testing and VT-2 Visual examination.
Startino Air Subsystem "Per WCGS Surveillance Procedures STS KJ-00SA & B which are performed monthly, the Standby Emergency Diesel Generators are tested for operability. As part of these procedures, the pressure of the two starting air tanks for each Standby Emergency Diesel Generator is recorded to assure the associated discharge valves are properly performing their function (thereby satisfying Inservice Testing for valves). The satisfactory completion of this test also provides a positive indication that the pressure boundary integrity of the Starting Air Subsystem is intact. In addition, WCGS Surveillance Procedures STS KJ-002A & B, which are also performed monthly, include steps to verify starting air tank pressure at 2,5,10 and 15 minute points during the subsystem valve testing. A pressure drop of 20 psig maximum (from 600 psig, approximately 3.34%)is allowable to satisfactorily complete the test. This data also provides a positive indication that pressure boundary integrity is being maintained for ti,e Starting Air Subsystem. Based on the monthly frequency and data collected during these attemative test, WCNOC considers that testing performed to satisfy the Technical Specification Surveillance Requirements is an acceptable attemative to Section XI System Pressure Testing.
27
. ... - .. . - . . - . - . - - . - - - . - - - . -~ -
1 Jacket Water Coolina Subsystem "Similar to the Starting Air Subsystem, Jacket Water Cooling prer,sure and temperature data is recorded every 30 minutes as part of Standby Emergency Diesel Generator testing in accordance with Surveillance Procedures STS KJ-005A & B. Normal values i for this data are provided within the procedures as well as limits (s) for the recorded
- values which provide a means to assess the data recorded. Again, this data provides a positive indication that pressure boundary integrity is being maintained. Based on the monthly frequency and data collected during these altemative tests, WCNOC considers
, that testing performed to satisfy the Technical Specificatinn Surveillancs Requirements
, -is an acceptable attemative to Section XI System Pressure Testing.
Air and Fuel Rack Sunolv Air Subsystem "The air manifold temperature and pressure dets associated with the Air and Fuel Rock !
Supply Air Subsystem is also recorded every 30 minutes as part of Standby Emergency Diesel Generator testing in accordance with Surveillance Procedures STS KJ-005A & B.
Normal values for this data are provided within the procedures as well as limit (s) for the recorded values which provide a means to assess the data recorded. Again, this data provided a positive indication that pressure boundary integrity is being maintained.
i Based on the monthly frequency and data collected during these attemative test, WCNOC considers that testing performed to satisfy the Technical Specification l Surveillance Nguirements is an acceptable attomative to Section XI System Pressure Testing, ,
intake and ExbW Air Subsystem
" Adequate flow of the intake and Exhaust Air Subsystem is demonstrated by successfully opersiing the Standby Emergency Diesel Generators during testing.
Specifically, the intake air vacuum data in inches of we.or is recorded every 30 minutes as part of Standby Emergency Diesel Generator testin l in accordance with Survt,illance ,
Procedures STS KJ-005A & B.. This data provides a positive indication that proper flow is being maintained and in tum proper exhaust flow is also being maintained. WCNOC .
1 i
considers the testing perfoimed to satisfy the Technical Specificatio,1 Surveillance Requirements is an acceptable attemative to Section XI System Pressure Testing.
"Per Surveillance Requirement 4.8.1.1.2.g, each Standby Emergency Diesel 15
, subjected to an inspection in accordance with procedures prepared per the manufacturer's recommendation. The inspections performed per this procedure provide
, additional assurance that the components within the Starting Air, Jacket Cooling Water,
, Intake and Exhaust Air, and Fuel Rack Supply Air subsystems demonstrate pressure -
l boundary integrity and the ability to provide adequate flow for satisfactory Standby
- Emergency Diesel Generator Operation.
- " Based on the information prove, id d WCNOC request relief from the Section XI
- requirements to perform system pressure testing on the Code Class 3 Stand 5y
- Emergency Ciesel Generator subsystems listed above on the basis that Technical t 28
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Specification Surveillance Refluirements of 4.8.1.1.2 provide an acceptable level of quality and safety.'
l i
Licensee's Proposed Alternative Examination (as stated >- l "WCNOC will imploment the operability tetting of Technical Specification Surveillance 1 Requirements 4.8.1.1.2."
Evaluation-The Code requires system leakage testing of Class 3 piping and valvet once f l , . each inspection period, in lieu of the Code-required testing, the licensee has proposed to i perform leak or flow testing each month on the emergency diesel generator subsystems.
I Performing leak and/or flow testing provides an indirect verification of system integrity. Each
- l i
system receives these tests every thirty days, whid is a more rigorous testing seedule than .
that required by the Code. Each test has pressure or flow indicators that are monitored during f the test. Each of these indicators has an associated maximum / minimum allowable value, which t i will alert an operator if a leak exists. The cause of the unallowable indication is then located and
- repaired if necessary. The INEEL staff believes that the licensee's proposed attemative to the j
Code required periodic pressure tests will provide an acceptable level of quality and safety, i 4
Conclusion-It is reasonable to conclude that the licensee's proposed attemative will detect !
leakage, if present, providing an acceptable level of quality and safety. Therefore, it is i
recommended that the proposed altemative be authorized pursuant to 10 CFR 50.55a(s)(3)(l).
l
! 3.4.4 General ,
3.4.4.1 Request for Relief 12R 02, IWA 8260(a)(2), System Pressure Test Corrective i Measures l Code Requiremens-lWA-5250(a)(2) states that if leakage occurs at a bolted connection l during a system pressute test, then all bolting must be removed and a VT 3 visus'l examination
- performed to detect corrosion.'
- Licensee's Code Rollef Rege- ' Pursuant to 10 CFR 50.55a(s)(3)(i) the licensee proposed an altomative to performing the Code-required removal and VT 3 visual examination of bolting if
- leakage occurs during a system pressure test of Class 1,2, and 3 systems.
4 l.. Licensee's Basis for Roguesting Rollef(as steted)~
3
- Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed i- attemative would provide an acceptable level of quality and safety.
! _" Removal of pressure retaining botting at mechanical connections for visual, VT-3
- examination and subsequent evaluation in locations where leakage has been identified i is not always the most prudent course of action to determine the acceptability of the j_ bolting. The Code requ!rement to remove, examine and evaluate botting in this situation '
j does not allow the Owner to consider other factors which may indicate the acceptability 29-k
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of mechanicaljoint botting. WCNOC considers this requirement to be unneosssarily prescriptive and restrictive.
- 0ther factors which should be considered when evaluating botting acceptability when leakage has been identified at a mechanical joint include, but should not be limited to:
joint bolting materials servloe , age of joint bolting mat i ler a s, location of the leakage, history of leakage at the joint, evidence of corrosion with the joint assembled and corrosiveness of process fluid.
- 'ASME Section XI is written to primarily address examinations and testing during periods of plant or system shutdown. No guidance is given to adoress comF,onents that are examined or tested while the plant or system is in service. However, mar y Code Class 3 and a few Codt. Class 2 systems are pressure tested, including VT 2 visually examined, utilizing the -
- inservios test
- requirements of IWA 5000.
" Performance of the test while the system is inservice may identify leakage at a bolted connection that, upo<1 evaluation, may conclude that the joint's structural ir.tegrity and pressure retaining ability is not challenged. It would not be prudent to negatively impact a j-safety system's availability by removing the system from service to address a leak that does not challenge the system's ability to perform its safety function.
"In addition, a situation frequently encountered at censsc,;al nucisar plants such as WCGS is the complete replacement of botting materials (studs, bolts, nuts, washers, etc) at medanicaljoints during plant outages. Whcn the associated system process piping is pressurized during plant start up, leakage is identified at these joints. The root cause of this leakage is most ohen due to thermal expansion of the piping and botting materials ai the joint and subsequent process fluid seepage at the joint gasket. Proper retorquing of the joint botting, in most cases, stops the leakage. Removal of any of the joint botting to evaluate for corrosion would be unwarranted in this situation due to new conditien of the botting materials. ASME Section XI interpreta' ion XI 192-01 has recognized this situation as one which the requirements of IWA 5250(a)(2) do not apply, t
"WCNOC proposes the following attemative methodology to the requirements of
- IWA-5250(a)(2) which will provide an equivalent level of quality and safety when evaluating leakage and bolting material acceptability at Class 1, 2, and c ' olted connections."
, Licensee % Propoaed Alternative Examinadon (as statedF-
"When leakage is identified at bolted connections by visual, VT 2 examination during system pressure testing, an evaluation will be performed to determine the susceptibility of the botting to corrosion and assess the potential for failure. The following factors, may be considered, as applicable, but not limited to, when evaluating the acceptability of the botting:
"1) Bolted materials "2) Corrosiveness of the process fluid
,_ "3) Leakage location 30
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"4) Leakage history at connection or other system components "5) - Visual evidence of corrosion at connection (connection assembled)
- 6) Servios age of botting materials
'When the pressure test is performed with the system in service or required by the Technical Specifications to be operable, and the botting is susceptible to corrosion, the evaluation shall address the connedions's structural integrity until the next component / system outage of sufficient duration, if the evaluation concludes that the system can perform its safety related function, removal of the bolt closest to the leakage and VT 3 visual examination of the bolt
, will be performed when the system or component is taken out of service for a sufficient duration for accomplishment of other system maintenance activities.
- For botting that is susceptible to corrosion, and when the initial evaluation indicates that the connection cannot conclusively perform its safety function until the next componer,t/ system outage of sufficient duration, the bolt closest to the source of leakage will be removed, receive a VT 3 visual examination, and be evaluated in accordance with IWA 3100(a)."
l Evaluation-In accordance with the 1989 Edition of the Code, when leakage occurs at bolted connections, all botting is required to be removed for VT 3 visual examination. In lieu of the -
Code required removal of botting to perform a VT-3 visual examination, the licensee has proposed to perform an evaluation of the bolted connection to determine the susceptibility of the botting to corrosion and the potential for failure.
This attemative allows the licensee to utilize a systematic approach and sound engineering judgement; provided as a minimum, all six evaluation factors listed in the licensee's proposed altomative are considered. Futhermore, if the in tial evaluation indicates the need for a more :
In<lepth evaluation, the bolt closest to the source of leakage shall be removed, VT 3 examined,-
and evaluated in accordance with lWA 3100(a). With this commitment, the licensee's altomative to tne Code-requirad removal of botting at a joint when leakage occurs will provide an acceptable level of quality and safety, as the integrity of the joint will be maintained.
Conclualon-The licensee's proposed altemative, to use a systematic approach and sound engineering judgement, will provide an acceptable level of quality and safety. Therefore, it is recommended that the proposed att?mative be authorized pursuant to 10 CFR 50.55a(a)(3)(i),
considering that if the initial evaluation indicates the need for a more in-depth evaluation, the bolt closest to the source of leakage is removed, VT 3 examined, and evaluated in accordance with IWA 3100(a).
3.4.4.2 Request for Rollef 12R 01, lWA4242(a), System Pressure Tests for Insulated Components in Class 1 Borated Systems
. 1 Code Requiremont--tWA 5242(a) states that for systems borated for the purpose of-controlling reactivity, insulation shall be removed from pressure-retaining bolted connections for a direct VT 2 visual examination.
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Licensee's Code Rollef Request-Pursuant to 10 CFR 50.5Sa(a)(3)(i), the licensee proposed an attemative to the Code-required removal of insulation for VT 2 visual examinations of bolted connections in Class i borated tystems, Licensee's Basis for Request.'ng Rollef(as stated)- -
- Pursuant to 10 CFR 50.5Sa(a)(3)(l), relief is requested on the basis that the proposed
- attemative would provide an acceptable level of quality and safety. Specifically, relief is requested from the requirement to remove insulation at bolted connections for visual, VT 2 examination duiing system pressure testing for the following reasons:
i "1) Code Class 1 and 2 systems borated for the purpose of controlling reactivity are extensive and large systems extend into multiple areas and elevations. Scaffolding will be required to access many of the bolted connections. In addition, many of the bolted connections are located in difficult to access areas and in medium to high radiation areas. Insulation removal combined with scaffolding requirements will increase the financial cost, personnel exposure, and generation of radwaste associated with performance of visual VT 2 examinations.
"2) The visual VT 2 examination of Class 1 systems, primarily the Reactor Coolant l System (RCS) piping and components, is performed between plant mode 3 and 2 i
ascending. As required by IWB 5221, the RCS is at a normal operating pressure of 2235 psig. Between modes 3 an,4.1 ascending, the temperature is apprraimately 557'F. Performance of a visual VT 2 examination, installation of insulation, and disassembly of scaffolding at bolted connections under these operating conditions is a personnel safety hazard. The visual VT 2 examination is a critical path activity and normally has a duration of six to eight hours. Since the majority of Class 1 piping is inside the containment building bio-shield wall, insulation installation and disassembly of scaffolding will add to the outage duration. Critical path cost is currently estimated at $216,000 per day.
"The following WCNOC botting examination commitments and material control programs in conjunction with the proposed attemate provision provide an acceptable level of safety and quality for bolted connections in systems borated for the purpose of controlling reactivity.
"A) In response to NRC Generic Letter 88-05,' Boric Acid Corros.on Of Carbon Steel Reactor Pressure Boundary Components in PWR Plants', Wolf Creek Generating
, Stetion (WCGS) has established a program to inspect all boric acid leaks discovered in the containment building and to evaluate the impact of those leaks on carbon steel or low alloy steel components. All evidence of leaks, including boric acid crystals or residue, is inspected and evalueted regardless of whether the leak was discovered at power or during an outage. Based on the evaluation, appropriate corrective actions are initiated to prevent reoccurrence of the leak and to repair, if necessary, any oegraded materials or components.
"B) In addition to the nondestructive examinations requised by ASME Section XI, WCGS bes committed to the bolting examination requirements of NRC 32 0
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4
Bulletin 82 02,' Degradation Of Threaded Fasteners in The Reactor Coolant Pressure Boundary Of PWR Plants'. In accordance with this Bulletin, at least two nondestructive examination techniques (e.g., ultrasonic, liquid penetrant, magnetic particle, or visual VT 1) are performed on bolted connections of the following components' Steam Generator primary manways, Pressurizer primary manway, l
Pressurizer safety valves, and a total of 22 Reactor Coolant System isolation !
, valves that are greater than 6" NPS. As a minimum, two nondestructive examination techniques are used whenever the bolted connection of one of the subject components is dissembled for maintenance or other inspection. These '
,_ additional examinations ensure that degradation mechanisms such as Stress Corrosion Cracking or corrosion do not go undetected in bolted connect'ons critical to reactor safety.
"C) The only carbon steel and low alloy preswre boundary components at the WCGS that are in systems borated for the purpose of controlling reactivity are clad with stainless steel. Specifically, these clad components are the Reactor Vessel, Steam .
Generators (primary side), and Pressurizer. All other pressure boundary piping and components in borated systems that are within inservice inspection boundarie.s are constructed of stainless steel. There is substantial information, such as EPRI NP-5679, attesting to the resistance of stainless stools to boric acid corrosion. To ensure that degradation mechanisms in stainless steels are mitigated, WCNOC maintains a program for controlling materials (insulation, thread lubricant, boron, etc.) that may come in contact with safety related components, including botting.
This program ensures that impurities are not present in concentrations that would promote development of Stress Corrosion Cracking in stainless steel bolted connections.*
Licensee > Proposed Altemative Examination (as stateit) --
" Pressure retaining bolted connections in Class 1 components subject to the pressure test at the completion of each refueling outage, will be VT-2 visually ex6 mined with the insulation removed. However, it is not required to have the system pressurized. Any evidence of -
leakage will be evaluated in accordance with IWA 5250.
"A VT-2 visual examination of pressure retaining bolted connections in Class 1 components will be performed during the system pressure test at the completinn of each refueling outage, without removing the insulation, after satisfying the requirements of IWA-5213, Test Condidon Holding Time." ^
Evaluation-Paragraph IWA 5242(a) requires the removal of allinsulation from pressure-retaining bolted connections in systems borated for tta purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. The licensee has committed to a 4-hour hold time at test conditions (pressure and temperature) prior to the VT-2 visual examination without removal of insulation, and during the refueling outage,- the connection will not be required to be at operating conditions for performance of the VT 2 visual examination with the insulation removed. An/ evidence of leakage will be evaluated in
~
accordance with IWA 5250. Requiring removal and installation of insulation, and disassernbly of scaffolding at Class 1 bolted connections when the plant is at operating conditions for the sole 33
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- purpcss of the VT 2 visual examination is a personnel safety hazard. Based on the review of the licensee's basis for relief and proposed attemative.11 has been determined that the licensee's approach to the Code-required insulation removal is acceptable.
Conclusion-The INEEL staff believes that the proposed attemative provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed
, attemative be authorized pursuant to 10 CFR 50.55a(a)(3)(1).
3.4.4.3 Request for Relief 12R.19, Request for Authorization to Use ASME Code Case N 4981, Altemative Rules for 10 Year System Hydrostatic Testing for Class 1, 4, and 3 Systems,Section XI, Division 1 Code Requirement-Exam: nation Categories B-P, C H, D A, D 8, and D-C require that a system hydrostatic test be performed in accordance with IWA 5000 once each 10 year interval.
Licensee's Code Relief Request-Pursuant to 10 CFR 50.55a(s)(3)(i), the licensee requested approval to use Code Case N 4981 in lieu of performing the Code-required hydrostatic pressure test. .
Licensee's Basis for Requesting Rollef(as stated}--
" Relief is being requested to allow the use of attemative requirements for system hydrostatic testing of Class 1,2 and 3 systems as detailed in Code Case N 4981. Th3 basis for this requast is as follows:
"1) The ASMd Section XI Working Group on Pressure Testing concluded that no additional benefit would be gained by conducting the existing Class 1,2, and 3 system hydrostatic tests versus performing the 40 month pressure tests at nominal operating pressure.
"2) Licensees incur considerable time, radiation dose, and monetary resources carrying out hydrostatic test requirements.
"3) The NRC staff has recognized. through approval for use of Code Case N-498-1 in WCNOC's first inspection interval, that '... compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty for the licensees without a compensating increase in the level of quality and safety.'
"4) The attemate rules of Code Case N-498-1 provide an acceptable level of quality and safety."
Licensee's Proposed Alternative ExarrInstion (as stated}--
"The requirements of Code Case N-498-1 may be used as an attemative to the system hydrostatic testing requiremtents."
34
. o EvaluarlorHThe Code requires the performance of a system hydrostatic test once per interval in accordance with the requirements of IWA 5000 for Class 1,2, and 3 pressure-retaining syctems, in lieu of the Code-required hydrostatic testing requirements, the licensee has requested authorization to use Code Case N-4981, Attemative Rules for 10-Year System Hydrostthe 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 structuralintegrity of the system but rather an enhanced leakage test.8 Hydrostatic testing only subjects the piping components to a small increase in pressure over the design pressure; therefore, piping dead
, weight, thermal expansion, and seismic loads present far greater challenges to the structural integrity 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, 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--4n most cases leaks are being found when the system is at normal operating pressure.
Code Case N-498, Attemative 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, Revision N498-1 specifies requirements identical to those for Class 2 components (requirements for Class 1 and 2 systems in N-498-1 are unchanged from N 498). In liec of hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-4981 requiri.: s VT 2 visual examination at nominal operating pressure 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 the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. 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 failures 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; therefore, Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms.
System hydrostatic testing entails considerable time, radiation dose, and financial resources.
The safety assurance provided by the enhanced leakage gair'ed from a slight increase in system pressure during a hydrostatic test may be offset or negated by the necessity to gag or remove Code safety and/or relief valves (placing the system, and thus the plant, in an off-
- d. S. H. Bush and R. R. Maccary, Development of In-Service Inspection Safety Philosophy for U.S.A. Nuclear Power Plants, ASME,1971
- e. Documented in Licensee Evert Reports and the Nuclear Plant Reliabinty Data System databases.
35
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a-
. . o normal state), erect temporary supports in steam lines, and expend resources to set up testing with special equipment and gages. Therefore, performance of system hydrostatic testing at pressures greater than operating pressure represents a considerable burden without a compensating increase in quality and safety.
Conclusion.- Giving cont'deration to the minimal amount of increased assurance provided by
, the elevated pressure of a hydrostatic test versus the pressure for the system leakage test, and the excessive burden associated with performing the hydrostatic test, the INEEL 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 qutlity and safety. Therefore, it is recommended that the use of Code Case N-496-1 for Code Class 1,2, and 3 systems be authorized for the second interval pursuant to 10 CFR 50.55a(a)(3)(ii), until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.
3.5 General 3.5.1 Ultrasonic Examination Techniques 3.5.1.1 Request for Rollef 12R 18, Qualification of Nondestructive Examination Personnel for Ultrasonic Examination Code Requiremens-lWA 2311(b) requires that the training, qualification, and certification of ultrasonic examination personnel comply with the requirements specified in Appendix Vll.
Licensee's Code Rollef Roguest-Relief is requested from implementation of Appendix Vil until the performance demonstration requirements of Appendix Vill are fully implemented.
Licensee's Basis 96r Requesting Rollet (as stated}-
"WCNOC requests relief from implementation of Appendix Vil until the performance demonstration requirements of Appendix Vill are fully implemented, implementation of Appendix Vil prior to full implementation of Appendix Vill is considered impractical and without a compensating increase in quality and safety,
" 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 C
practices in the requirements for qualification of ultrasonic axamination personnel. New training programs trust be developed and taught by trained instructors, employer's written practims must be completely rewritten, examination question banks must be developed, flew specimens containing actual or simulated flaws must be acquired, and performance demonstrations (practical examinations) must be completed.
" Implementation of Appendix Vil will require a substantialindustry effort. #though work is_ progressing towards compliance with Appendix Vil, full implementation has not yet been achieved. Since Appendix Vil provides for use of specimens prepared for 36
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utrasonic performance demonstrations per Appendix Vill, many NDS vendors are developing these two programs concurrently in order to avoid duplicated effort. Though currently not required, the nuclear industry anticipates that the Appendix Vill l performance demonstration requirements will be mandated by a backfit ruling in the Federal Register, in anticipa!!on of this ruling, the Performance Demonstration Initiative (PDI) Committee is currently leading an industry wide effort to implement Appendix Vill. j The tentative completion dates for pipe weld performance demonstrations and reactor
- l vessel performance demonstrations are January of 1996, and January of 1997, l
respectively. l "The WCNOC intends to implement Appendix Vil when the performance demonstrations of Appendix Vill are mandated by a back fit ruling in the Federal Reglster."
Licensee's Proposed Alternative Examination (as stated}--
"WCGS shall utilize ultrasonic exr.mination personnel qualified in accordance with the requirements of IWA 2300, except for IWA 2311(b). The additional Appenuk v'il training, qualification, and certification requirements referenced in IWA 2311(b) shall be fully implemented when the performance demonstrations of Appendix Vill are mandated by a ruling in the Federal Register."
EvaluatlorHAppendix Vil was incorporated into the Code in 1988 to enhance the qualifications of uttrasonic examiners. This appendix places controls on a wide variety of classroom and laboratory training. Although Appendices Vil and Vill are both designed to improve flaw detection confidence, their concurrent implementation is not necessary.
The INEEL staff believes that the licensee has been given sufficient time to develop an Appendix Vil program. Appendix Vil has been generally adopted through-out the industry.
Although Appendix Vill will further iniprove flaw detection confidence, its implementation is not required in conjunction with Appendix Vll. An Appendix Vil program willincrease quality and safety and is not considered impractical.
ConclusforHSufficient technical justification has not been provided an:f, therefore, it is recommended that the licensee's request be denied.
3.5.2 Exempted Components (No requests for relief) 3.5.3 Other 3.5,3.1 Request for Rollef 12R.11, IWB 2420(b,c) and IWC.2420(b,c), Successive inspections In the May 3,1996, response to the NRC's RAl, the licensee withdrew Request for Relief 12R 11, stating that 'a number of issues still require resolution.'
37
1 1
3.6.3.2 Request for Rollef 12R 20, Use of Code Case N 509, Altomative Rules for the Selection and Examination of CIsss 1, 2, and 3 Integrally-Welded Attachments,Section XI, Olvision 1 Code Regulromont-Tables IWB/C/D-2500-1, Examinat.'on Categories B-H, B-K-1, C-C, D A, D 8, and D-C require volumetric or surface examination of 100% of the non-exempt integrally.
, welded attachments.
Non-mandatory Code cases may be used for ISI after general acceptance by the NRC staff and incorporation into Regulatory Guide 1.147. Fursuant to 10 CFR 50.55a, Code cases not incorporated into Regulatory Guide 1.147 may be used provided specific NRC authorization is ootained.
(.lconsee's Code Rollef Request-Pursuant to 10 CFR 50.55a(s)(3)(i), the licensee requested authorization to use Code Case N-509, Attemate Rules tbr the Selection and Examination of Class 1, 2, and 3 Integrelly Welded Atta:hments,Section XI, Division 1.
Licensee's Basis for Requesting Rollef(as stated}--
- 1) During the first inservice inspection interval at the Wolf Creek Generating Station, no inservice flaws were detected in integrally welded attachments which would affect safety or compromise the integrity of the plant.
"2) Within the commercial nuclear power industry, failures of integral attachmerts have been very rare and have not affected plant safety. When failures or inservice defects are found in integral attechments, they are usually associated with a support which has been damaged during operation. Therefore, flawed or broken integral attachments are typically detected during the investigation of damaged supports rather during scheduled inservice inspections. One feature of Code Case N 509 is to focus the examination of integral attachments on instances where the deformation of the associated supports is identified. This requirement will increase the likelihood of locating damaged integral attachments and thereby increase the level of quality and safety provided by those attemative rules, as compared to the rules of the 1989 Edition of Section XI.
"3) There is a significant amount of man-rem exposure and cost associated with the scheduled inspection of Class 1,2 and 3 integral attachments.
. "4) Unlike ASME Section XI,1989 Edition, the attemate selection critoria of Code Case N-509 does not impose a minimum thickness requirement for the inspection of an integral attachment. Therefore, a greater population of integral attachments will be available for inspection because selection will not be limited to those above an arbitrary thickness, This provision improves the quality and safety level established by these examinations.
"5) The attemate rules of Code Case N 509 provide an acceptable level of qual?y and safety."
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! In a response to the NRC's RAl, the licensee stated that 10% of the total population of non-
) exempt integral attachments will be examined using Code Case N 509.
I Licensee's Proposed Alternative Examination (as statedh--
'The requirements of Code Case N 509 will ba used to select ano examine integrally ,
welded attachments (10% of the total population of non-exempt integral attachments will '
!, be examined using Code Case N 509).* ;
i Evaluatfor>-The licensee has proposed, as an attomative to the Code requirements, to apply
- l. the requirements of Code Case N-509 for the examination of integrally-welded attachments on j Class 1,2, and 3 piping and components. Code Case N 509 provides attemative sampling ,
j requirements for the examination of Class 1,2, and 3 integral attachments. ;
l' sview of this Code Case indicates that there is an ambiguity in the notes of the i emination tables that would allow the selection of a 10% sample of the integrally welded l sttachments from the percentage of component supports selected for examination under the '
- rules of the Code (specifically, Subsection lWF of the 1990 Addenda). This could potentially ,
reduce the examination sample to an insignificant amount, or to no integral attachments at all.
l The INEEL staff believes that Code Case N 509 should be augmented to ensure that this does '
not occur. Considering that most of the Code examination requirements are based on sampling l to ensure the detection of service induced degradation, extending the sampling philosophy to i the integral attachment welds will provide an equivalent level of quality and safety for those welds. Therefore, it is concluded that the use of Code Case N-509 provides an acceptable level
} of quality and safety when including the licensees proposal to examine a minimum of 10% of the total number of all nonexempt Class 1,2, and 3 piping, pump, and valve integral attachments.
l Conclualon-The INEEL staff believes that the proposed altemative provides an acceptable
) level of quality and safety with the commitment stated above. The licensee has committed to
- examination of 10% of the total number of integral attachments in all Class 1,2, and 3 piping, i pumps, and valves. Therefore, it is recommended that the licensee's proposed attemative be >
l authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of this Code Case, with the above j commitment, should be authorized for the second interval at Wolf Creek Generating Station, or
- until the Code Case is approved for general use by reference in Regulatory Guide 1.147.' After that time, the licensee must follow the conditke.5, if any, specified in the regulatory guide.
3 t
! 3.5.3.3 Request for Relief 12R 15, IWF4300(a) Snubber Examination
- ,- This request for relief is considered part of the Inservice Testing Program (IST) and is,-
therefore, not included in this evaluatiors. The Snubber Testing Program will be evaluated by
)-
the Mechanical Engineering Branch of NRC.
j- 39
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.-_.brL: 271:2::2:::2 ::~~ ::.L: TT::: L:LL::. ::j ::-. .
r-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
l_ 4. CONCLUSION Pursuant to 10 CFR 50.54a(g)(6)(i), it has been determined that certain inservice exarninations cannot be performed to the extent required by Section XI of the ASME Code, in the casef of Requests for Relief 12R-03,12R-04,12R4)S,12R-06,12R-07,12R-08,12R 14, and 12R-18, the I:censee has demonstrated that specific Section XI requirements are impractical; it is recommended that relief be granted as requested. The granting of relief wil: not endanger life, property, or th9 common defense anc' security and is otherwise in the public interest, giving due consideration to the burden up; I the licensee that could result if the requirements were Imposed on the facility.
Pursuant to 10 CFR 50.55a(a)(3)(i), it is concluded that for Requests for Relief 12R * -
12R-17 the licensee's proposed altematives will provide an acceptable level af qu a -
,c in lieu of the Code-required examinations. In these cases, it is recommended th Sc .m attematives be authorized. The propowd attematives for Requests for Relief li .
' Lw, 12R 12,12R 13, and 12R-20 are recommended to be authorized only if the licenst stisfios the commitments stated in the request for relief evaluations above.
Pursuant to 10 CFR 50.55a(a)(3)(ll), it is concluded that for Request for Relief 12R-19 the licensee bra demonstrated that specific Section XI requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In this case, it is recommended tiwt the preposed attemative be authorized.
Request for Relief 12R-11 was withdrawn by the licensee, and deleted from the ISI Program Plan, by letter dated May 3,1996, in rurponse to the NRC's request for add;tional information.
E or Requests for Relief 12R-09 and 12R-16, it l's cancluded that the licensee has not provided sufficient justification to support the determistion that the Code requirements are
- impractical or that compliance with the Code requirement would result in hardship. Therefore, in that,3 ctses it is recorrmended that relief De denied. ,
Request for Relief 12R-15 was not evaluated in this repo t.
This technical evaluation has not identified any prsctical method by which the licensee can meet all the specific inservice inspection requiremerits of Section XI of the ASME Code for the j
existing Wolf Creek Generating Station. Compliance with all of the Section XI examination requirements would necessitate redesign of a significent number of plant systems, procuremont of replacement components, installation of the new components, and performance of baseline examinations for these components. P en 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 have been determined to be impractical.
The licensee should continue to monitor the developw.9 of new or improved examination techniques. As improvements in these areas are achieved, the licensee should incorporate inese techniques irc:o the ISI program plan examination requirements.
40
[
A Bas;d on the review of tne Wolf Creek Generating Station, Second 10-YearInterval inservice inspection Pr'. 7 tam Plan, and the recor6mendations 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 identified, except those noted in the evaluation of Requests for Relief 12R 09 and 12R 16 .
4 r
1 41
..o
- 5. REFERENCES i- 1. Code of Federal Regulations, Title 10, Part 50.
l
- 2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI,
- Rules Ibr Inservice Inspection of NuclearPower Plant Components," Division 1,1989
, Edition.
L
\ 3. Wolf Creek Generato,$ Station, Second 10 YearIntervalinservice Inspeckon Program
. Plan, submitted August 4,1995.
- 4. NUREG-0800, Standarri Review Plans, 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.
Letty, dated March 6,1996, J. C. Stons (NRC) to N. S. Cams (WCNOC)) containing request for additional informatbn on the Wolf Creek Generating Station, Second 10-Year Interval Inservice Inspeckon Program Plan.
6.
Letter, dated May 3,1996, R. A. Muench (WCNOC) to Document Control Desk (NRC) containing the Response to Request for Additional information, S ,cond 10-Year Inservice inspection Interval.
Letter, dated October 10,1996, J. C. Stone (NRC) to N. S. Cams (WCNOC) containing request for additional information on the Wolf Creek Generefog Station, Second 10 Year Interval Inservice Inspeckon Program Plan reques% for relief.
8.
Letter, dated November 25,1996, J. C. Stone (NRC) te N. S. Cams (WCNOC) containing request for additional information on the Wolf Creel Gertrating Station, Second 10 Year Interval inservice Inspeckon Program Plan requests for relief.
9 Letter, dated March 4,1997, J. C. Stone (NRC) to N. S. Cams (WCNOC) containing request for additional information on the Wolf Creek Generating Station, Second 10 Year Interval I.nervice Inspechon Program Plan requests for relief.
- 10. Letter, dated December 9,1996, N. S. Cams (WCNOC) to Document Comsl Desk (NRC) containing the Response to Request for Additional Information, Second 10-Year Interval Inservice inspection Program Plan and Associated Requests for Relief.
- 11. Letter, dated January 23,1997 O. L. Maynard (WCNOC) to Document Control Desk
' j (NRC) containing the Response to Request for Additional Information, Second 10-Year '
interval Ins 3rvice laspection Prograrr plan and Associated Requests for Relief.
- 12. . Letter, dated May 6,1997, O. L. Maynard (WCNOC) to Document Control Desk (NRC) containing the Response to Request for Additional information, Second 10-Year Interval inservice Inspection Program Plan and Associated Requests for Relief.
42
__._.__..._-_._._,i
. , , o i
- 13. Letter, dated September 20,1995, R. C, Hagan (WCNOC) to Document Control Desk (NRC) containing Corrections to inservice Inspection Program for the Seconti Ten Year interval.
- 14. NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During
, Preservice andInservice Examinations, Revis'on 1, February 1983. ,
15, NRC Regulatory Guide 1.14, Reactor Coolant Pump F!ywheellntegrity, Revision 1, August
. 1975.
- 16. NRC Regulatory Guide 1.65, .'Aaterial and Inspections for Reactor Vessel Closure Studs, dated October 1973.
4 e
i 43
ce- no. v s wucasa asswatorv cewsso 1. REPORT NuGER
-co nn BIBLIOGRAPHIC DATA SHEET U Z Z .7. @ '"'
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INEL.9613264
- 2. TITL( AND SUETITLt Technical Evaluation Report on the Second 10 Year 3. oAtt arront Pur 15Ht0 Interval inservice inspection Program Plan: " cath t**r Wolf Creek Nuclear Operating Corporation July 1997 Wolf Creek Generating Station 4 Flu m sRANT auMatu Docket Number 50482 - ' - <= ^
- an 5, '#UTHOR(5) 6. It'PE'0I~R'6 0RI' " ' '
Technical M. T. Anderson, 3. W. Brown, E. J. Feige, K. W. Hall, A. M. Pntter
- 7. PER100 C0vtRt0 ,
- 4. PERFORNING ORGANIZAT10m . NAME AND ADDRES5 w unc.s =m.o a.oss== = a=,.a.u.s.a --
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.e. . sun. .as m.dme es J INEEL/LMITCO P.O. Box 1625 Idaho Falls, ID 83415-2209 -
- 9. $PON10 RING ORGANIZ AT10N - NAME AND ADDR!ss m wt. im *a = % e == we. sinc o.= n.on = ane u.s. ini==. a ,v
- ETvTarM~sceer' ices Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washirwan o c 9nnnn
- 10. SUPPLEMENTARY NOTES
^
- 11. A8STRACT noo .i=i This report pros,ents the results cf the evaluation of the Wolf Creek Genera #ng Station, Second 10 Yearinterval mserwce inspeccon Psopram Plan, submitted August 30,1995 including the requests for relief from the Amencen Society of Mschanical Engineers Boiler and Pressure Vessel Code Section XI requirements that the licensee has determined to be impredical. Th? Wolf Creek Genera #ng Station, Second 10-Yearinfo.vallnservice s.,*W Program Plan, is evaluated in Section 2 of this report. The Inservice inspection (ISI) Prograr.1 Plan is evaluated for (a) complia*1ce with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion crfioria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this repoct.
- 12. .. KEY WOR 05/0ESCRIPT0R$ am . ww.* =r = *aw a. '=aa 13 .' AVAILABILITY STATEMINT Unlimited
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- 15. NUMBER OF PAGES
- 16. PRICE
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