ML17261A036

Request for Request 3-ISI-28 Regarding Third 10-Year Inservice Inspection Interval Examination Coverage of Inside Radius and Nozzle-to-Vessel Welds

ML17261A036
Person / Time
Site:	Browns Ferry
Issue date:	10/06/2017
From:	Undine Shoop Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Saba F, NRR/DORL/LPL2-2, 415-1447
References
CAC MF9257
Download: ML17261A036 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 6, 2017 Mr. Joseph W. Shea Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 3R-C Chattanooga, TN 37 402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNIT 3 - REQUEST FOR RELIEF 3-ISl-28 REGARDING THIRD 10-YEAR INSERVICE INSPECTION INTERVAL EXAMINATION COVERAGE OF INSIDE RADIUS AND NOZZLE-TO-VESSEL WELDS (CAC NO. MF9257)

Dear Mr. Shea:

By letter dated January 31, 2017, as supplemented by letter dated June 30, 2017, Tennessee Valley Authority (the licensee) submitted Requests for Relief 3-ISl-28 and 3-ISl-29, requesting relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," for Browns Ferry Nuclear Plant, Unit 3.

Specifically for 3-ISl-28, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), "ISi program update: Notification of impractical ISi Code requirements," the licensee requested relief from the "essentially 100 percent" volumetric examination coverage requirements of the ASME Code,Section XI, for the reactor pressure vessel nozzle-to-vessel full penetration welds and inside radius section on the basis that the compliance with the ASME Code requirement is impractical due to plant design. Request for Relief 3-ISl-29 was reviewed in a separate safety evaluation dated August 18, 2017.

Based on the Nuclear Regulatory Commission (NRC) staff review of the information submitted by the licensee, the staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest, given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Considering that further examination would require either extensive modifications or replacement of the components to remove limitations inherent to the nozzle-to-vessel weld design, the NRC staff determined that the examinations were performed to the maximum extent practical. Furthermore, the licensee was able to achieve coverage of the weld regions most susceptible to cracking, and no indications of service-induced degradation were found during these examinations or during any prior interval examinations. Operating experience with other plants has shown no significant signs of service-induced degradation; therefore, the staff determined that the examinations performed provide reasonable assurance that the structural integrity of the subject components will be maintained during continued operation. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the NRC staff grants relief to Tennessee Valley Authority for the examinations of the subject nozzle-to-vessel full penetration welds and the inside radius section,

J.Shea as requested in Request for Relief 3-ISl-28 for the Browns Ferry Nuclear Plant, Unit 3, third 10-year ISi interval, which began on November 19, 2012, and ended on January 31, 2016.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

If you have any questions, please contact the Project Manager, Farideh E. Saba, at 301-415-1447 or Farideh.Saba@nrc.gov.

Sincerely,

.~ d--r Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-296

Enclosure:

Safety Evaluation cc w/

Enclosure:

Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF 3-ISl-28 REGARDING EXAMINATION COVERAGE OF INSIDE RADIUS AND NOZZLE-TO-VESSEL WELDS TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT. UNIT 3 DOCKET NO. 50-296

1.0 INTRODUCTION

By letter dated January 31, 2017, as supplemented by letter dated June 30, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML17031A351 and ML17181A279, respectively), Tennessee Valley Authority (TVA, the licensee) submitted Requests for Relief 3-ISl-28 and 3-ISl-29, requesting relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),

Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," for Browns Ferry Nuclear Plant (Browns Ferry), Unit 3. Specifically, pursuant to Title 1O of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), the licensee requested relief from the "essentially 100 percent" volumetric examination coverage requirements of the ASME Code,Section XI, for the welds on the basis that the code requirement is impractical. Request for Relief 3-ISl-29 was reviewed in a separate safety evaluation dated August 18, 2017 (ADAMS Accession No. ML17135A146).

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements set forth in the ASME Code,Section XI to the extent practical, within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(a), 12 months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2).

The regulation in 10 CFR 50.55a(g)(5)(iii) states that licensees may determine that conformance with certain ASME Code requirements is impractical and that the licensee shall notify the Nuclear Regulatory Commission (NRC or the Commission) and submit information in support of the determination. Determination of impracticality in accordance with this section must be based on the demonstrated limitations experienced when attempting to comply with the code requirements during the inservice inspection (ISi) interval for which the request is being submitted. Requests for relief made in accordance with this section must be submitted to the Enclosure

NRC no later than 12 months after the expiration of the initial 120-month inspection interval or subsequent 120-month inspection interval for which relief is sought.

The regulation in 10 CFR 50.55a(g)(6)(i) states that the Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical.

The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. The licensee has requested relief from ASME Code requirements pursuant to 10 CFR 50.55a(g)(6)(i).

3.0 TECHNICAL EVALUATION

3.1 Applicable ASME Code Required The applicable code of record for the third Browns Ferry, Unit 3, 10-year ISi interval is the 2001 Edition with 2003 Addenda of the ASME Code,Section XI. Specifically, Table-2500-1, Category B-D, Item Nos. B3.90 and B3.100, require essentially 100 percent volumetric coverage of the subject welds and adjacent base material as depicted in Figure IWB-2500-7.

3.2 Components Affected The following table lists the affected Browns Ferry, Unit 3, ASME Section XI, Code Class 1 inside radius and nozzle-to-vessel full penetration welds that are addressed in the request for relief. These welds were examined during the second and third periods (spring of 2012 and spring of 2014 respectively) of the third 10-year interval.

Table 1 Code Category Item Number Description Weld Number B-D B3.90 Recirculation Outlet Nozzle N1B-NV B-D B3.90 Recirculation Nozzle N4A-NV B-D B3.90 Recirculation Nozzle N4B-NV B-D B3.90 Recirculation Nozzle N4C-NV B-D B3.90 Recirculation Nozzle N4D-NV B-D B3.90 Recirculation Nozzle N4E-NV B-D B3.90 Recirculation Nozzle N4F-NV B-D B3.90 Jet Pump Instrument Nozzle N8A-NV B-D B3.90 CRD Hydraulic Control Nozzle N9-NV Differential Pressure and Liquid B-D B3.90 N10-NV Control Nozzle Differential Pressure and Liquid B-D B3.100 N10-IR Control Nozzle 3.3 Licensee's Basis and Alternative The licensee stated that the ultrasonic testing (UT) examinations of the welds listed in Table 1 were performed to the maximum extent practical for the component design configuration of the reactor pressure vessel (RPV) nozzle-to-vessel welds and inner radius sections.

To support the impracticality in achieving full ASME Code-required coverage, the licensee stated:

In order to examine the weld in accordance with the ASME Code requirements, the RPV would require extensive design modifications. The physical arrangement of the nozzle-to-vessel welds precludes UT examination from the nozzle side. The limitations are inherent to the nozzle-to-vessel weld design. UT scanning from the inner nozzle surfaces is ineffective due to the weld location and the asymmetrical inside surface where the nozzle and vessel converge.

The blend radius of the weld restricts the scanning movement and/or transducer contact. The areas receiving little or no examination coverage are located toward the outside surface of the reactor vessel in the general area of the nozzle outside blend radius. Degradation, if present, at the inside surface or inner 15% volume would have been detected by the performed UT examinations.

The welds were examined with the latest techniques, procedures, equipment, and personnel qualified to the requirements of the Performance Demonstration Initiative (POI) Program, in accordance with the requirements of the 2001 Edition through 2003 Addenda of the ASME Code.

UT examination of the subject areas to the maximum extent practical for the design of the nozzle-to-vessel weld joints provides an acceptable level of quality and safety. Since the inner 15% thickness was fully examined for nozzle to vessel welds and 90% of the N10 inner radius was fully examined detectable degradation, if present, would have been identified by the performed UT examinations. As a result, reasonable assurance of operational readiness of the subject welds has been provided by the performed UT examinations.

Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), TVA requests that relief be granted for the BFN Unit 3 third ten-year inspection interval.

3.4 NRG Staff Evaluation Section XI of the ASME Code requires essentially 100 percent volumetric examination of all reactor vessel nozzle-to-vessel welds, adjacent base metals, and nozzle inner radius sections.

The licensee chose to implement Code Case N-460 which states, "When the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10%." Additionally, the licensee implemented Code Case N-613-1, which reduces the examination area of Category B-D nozzle-to-vessel welds to the weld plus half on either side of the weld. Table 1 of Regulatory Guide 1.147 lists both of these Code Cases as acceptable for application in the licensee'sSection XI in ISi programs. Pursuant to 10 CFR 50.55a(g)(4), the licensee attributed the inability to achieve the required examinations to the geometric configuration of the nozzle.

Radial and circumferential scans of the Category B-D, Item No. 83.90 nozzle-to-vessel welds were performed utilizing a 60° refracted longitudinal wave. The procedure for the radial exams is qualified to detect flaws throughout the entire thickness of the weld and adjacent base metal when scanning perpendicular to the weld. This same procedure, which is used for the circumferential exams, is qualified to detect flaws when scanning in the parallel and tangential

directions to the weld in the outer 85 percent thickness of the weld. Additional coverage in the circumferential direction was achieved using a separate procedure qualified to detect flaws when scanning in the parallel and tangential directions to the weld in the inner 15 percent thickness of the weld. The procedures were qualified through the industry's POI Program in accordance with AMSE Section XI, Appendix VIII, Supplements 4, 6, and 7. The table below lists the examination volume obtained with each scan and the total percentage of the ASME Code-required coverage obtained.

Weld Numbers Radial Outer 85% Inner 15% Total ASME Code examination Circumferential Circumferential Required coverage examination examination obtained N1B-NV 35.53% 0% 100% 25.27%

N4A-NV, N4D-NV 49.61% 0% 100% 32.31%

N4B-NV, N4C-NV, 49.75% 0% 100% 32.38%

N4E-NV, N4F-NV N8A-NV 95.74% 44.82% 100% 77.78%

N9-NV 40.92% 0% 100% 27.96%

N10-NV 93.70% 39.95% 97% 74.10%

The NRC staff reviewed the licensee's submittals for relief of these welds in the previous ISi interval, 3-ISl-18, to determine if there were any significant changes in examination coverage achieved. The coverage obtained for these same welds in the past interval ranged from 68 percent to 97percent coverage. Therefore, by e-mail dated June 8, 2017 (ADAMS Accession No. ML17163A050), the NRC staff requested that the licensee explain what caused the decrease in obtainable coverage and justify how the third ISi interval examinations provide an equivalent or greater standard of quality and safety.

The licensee, in its response dated June 30, 2017, explained that coverages in the second ISi interval were calculated in accordance with TVA Procedure N-GP-28, which was retired because it did not provide specific details for calculating nozzle-to-shell weld coverages. Prior to performing third ISi interval examinations, TVA developed Procedure N-GP-31, "Calculation of ASME Code Coverage for Section XI," in response to discovering inconsistencies in the examination coverage estimates generated under Procedure N-GP-28. All of the nozzle-to-shell weld coverage calculations performed in the third ISi interval were performed in accordance with the more conservative N-GP-31 methods.

In addition to the differences in calculation methods, the examination coverages achieved for the N1 B, N4A, N4B, N4C, N4F, and N9 nozzle-to-shell welds in the second and third intervals were differed due to the procedures that were utilized. The second ISi interval for these welds was performed according to ASME Section V, Articles 4 and 5, which used transducer and wedge combinations that were not ideal for obtaining meaningful data in the radial direction and were not optimized for proper orientation in the axial direction from the nozzle blend radius.

Since there was no industry standard for establishing what data could be credited as obtained coverage, the calculations allowed for credit of much of the outer 85 percent thickness. In the third ISi interval, these same welds were inspected with procedures qualified in accordance with ASME Section XI, Appendix VIII, Supplements 4,6, and 7, which used modeling to ensure that the maximum possible coverage was achieved using complex-curved transducers from the blend radius for the inner 15 percent thickness. Even though the reported coverages are lower for the third ISi interval due to more restrictive procedures, the techniques that were utilized

provided a higher quality of examination that focused on demonstrating coverage in the regions where flaws are likely to initiate.

Examination procedures and techniques used to examine nozzle-to-shell welds N4D, N4E, and N10 in the second ISi interval were similar to those applied during the third interval. Both processes used standard search units to perform tangential and outer 85 percent thickness circumferential exams. They also used specially designed search units to maximize circumferential examination of the inner 15 percent thickness. The higher claimed coverage of these welds in the second interval was based solely on the less conservative calculation methodologies, taking credit for areas that were not conducive to adequate transducer contact in the outer 85 percent thickness.

Full ASME Code-required examination coverage could not be obtained during the third ISi interval, primarily due to the proximity of the welds to the nozzles. Specifically, the curvature limited full radial examination and completely prevented circumferential examination of the outer 85 percent thickness of most of the welds. In addition to the nozzle curvature, additional geometric constraints precluded full examination of the weld and adjacent base metal. The crown configuration of circumferential weld C-3-4 limited radial examinations of welds N4A-NV, N4B-NV, N4C-NV, N4D-NV, N4E-NV, and N4F-NV for a distance of 28.69 degrees.

Additionally, the proximity to the N11A and N11 B nozzles further limited the examination of the N4A-NV and N4D-NV welds respectively for a distance of 10.38 degrees. A permanent insulation support ring and a permanent insulation support bracket limit radial coverage of the N9-NV and N10-NV welds for a distance of 92 degrees and 26 degrees, respectively. However, full examination of the inner 15 percent thickness of all of the welds was obtained utilizing the circumferential scans, and no recordable indications were observed during any of the exams.

In accordance with the POI-qualified procedure for Category B-D, Item No. B3.100 nozzle inside radii, the licensee conducted four scans of the N10-IR inside radius section utilizing 65- and 70-degree shear waves from the outer diameter vessel surface. Due to the configuration of the nozzle, scanning in one direction was limited to 50 percent of the weld length, while the other scans were performed in full without any limitations. Therefore, the licensee was able to obtain a total volumetric coverage of 87.5 percent. No recordable indications were found during these exams.

The NRC staff reviewed the licensee's submittal and determined that the examinations of the subject nozzle-to-vessel full penetration welds and the inside radius section were performed to the greatest possible extent, considering the geometric limitations precluding full examination.

Based on the licensee's assessment that obtaining essentially 100 percent exam volume coverage would require extensive RPV reconstruction, the NRC staff concluded that it is impractical for the licensee to comply with this ASME Code requirement, and its imposition would cause an unnecessary burden, without an increase in safety. Furthermore, considering that none of the subject welds lie within the beltline region, the exams performed covered the regions where cracking would most likely occur, no recordable indications were observed, and operating experience of similar welds at other plants has shown no signs of significant service-induced degradation. Therefore, the NRC staff concludes that there is reasonable assurance that the structural integrity of the welds will be maintained with the examination coverage achieved.

4.0 CONCLUSION

As set forth above, the staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest, given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Considering that further examination would require either extensive modifications or replacement of the components to remove limitations inherent to the nozzle-to-vessel weld design, the NRC staff determined that the examinations were performed to the maximum extent practical. Furthermore, the licensee was able to achieve coverage of the weld regions most susceptible to cracking, and no indications of service-induced degradation were found during these examinations or during any prior interval examinations. Operating experience with other plants has shown no significant signs of service-induced degradation; therefore, the staff determined that the examinations performed provide reasonable assurance that the structural integrity of the subject components will be maintained during continued operation. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i).

Therefore, the NRC staff grants relief to TVA for the examinations of the subject nozzle-to-vessel full penetration welds and the inside radius section, as requested in Request for Relief 3-ISl-28 for Browns Ferry, Unit 3, for the third 10-year ISi interval, which began on November 19, 2012, and ended on January 31, 2016.

All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor: Austin Young Date: October 6, 2017

J. Shea

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNIT 3 - REQUEST FOR RELIEF 3-ISl-28 REGARDING THIRD 10-YEAR INSERVICE INSPECTION INTERVAL EXAMINATION COVERAGE OF INSIDE RADIUS AND NOZZLE-TO-VESSEL WELDS (CAC NO. MF9257) DATED OCTOBER 6, 2017 DISTRIBUTION:

PUBLIC LPL2-2 R/F RidsACRS _MailCTR Resource RidsNrrDeEvib Resource RidsNrrDorlLpl2-2 Resource RidsNrrLABClayton Resource RidsNrrPMBrownsFerry Resource RidsRgn2MailCenter Resource AYoung, NRR TClark, OEDO ADAMS A ccess1on No.: ML17261A036 *b1y e-mar*1 OFFICE DORL/LPLI 1-2/PM DORL/LPLI 1-2/LA DE/EVIB/BC* DORL/LPLI 1-2/BC NAME FSaba BClayton DRudland UShoop (FSaba for)

(LRonewicz for)

DATE 09/25/2017 09/25/2017 08/31/2017 10/06/2017 OFFICIAL RECORD COPY