ML17069A283

From kanterella
Jump to navigation Jump to search

Relief Request No. IP3-ISI-RR-09, for Alternative to the Depth Sizing Qualification Requirement
ML17069A283
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 03/16/2017
From: James Danna
Plant Licensing Branch 1
To:
Entergy Nuclear Operations
Pickett D, NRR/DORL/LPL1, 415-1364
References
CAC MF8896
Download: ML17069A283 (9)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 16, 2017 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3- RELIEF REQUEST NO. IP3-ISl-RR-09 FOR ALTERNATIVE TO THE DEPTH SIZING QUALIFICATION REQUIREMENT (CAC NO. MF8896)

Dear Sir or Madam:

By letter dated December 2, 2016, as supplemented by letter dated January 26, 2017, Entergy Nuclear Operations, Inc. (the licensee), submitted Relief Request No. IP3-ISl-RR-09 to the U.S. Nuclear Regulatory Commission (NRC), requesting the use of an inspection procedure at Indian Point Nuclear Generating Unit No. 3 with a depth sizing error that is greater than the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and ASME Code Case N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1,"

for the fourth 10-year inservice-inspection (ISi) interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii),

the licensee requested relief from the depth sizing uncertainty qualification requirement for ultrasonic examinations conducted from the inside diameter of pipes (i.e., root-mean square (RMS) error not greater than 0.125 inches) contained in ASME Code Cases N-695 and N-696.

The licensee requested relief from the requirements for ISi items on the basis that the ASME Code requirement is impractical.

The NRC staff has concluded, as set forth in the enclosed safety evaluation, that using a vendor with a 0.189-inch RMS error for depth sizing for ASME Code Case N-695 and a 0.245-inch RMS error for depth sizing for ASME Code Case N-696, provides reasonable assurance of the structural integrity and leak-tightness in the subject welds. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff grants the licensee's Relief Request No. IP3-ISl-RR-09 at Indian Point Nuclear Generating Unit No. 3 for the fourth 10-year ISi interval, which began on July 21, 2009, and is currently scheduled to end on July 20, 2019. All other ASME Code,Section XI requirements for which relief has not been specifically requested remain applicable, including a third-party review by the Authorized Nuclear lnservice Inspector.

Vice President, Operations If you have any questions, please contact the Project Manager, Douglas Pickett, at 301-415-1364 or by e-mail at Douglas.Pickett@nrc.gov.

Sincerely, r~  :-J

""' I C/VVL-i.-_ v~~

Ja~es G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELIEF REQUEST NO. IP3-ISl-RR-09 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

1.0 INTRODUCTION

By letter dated December 2, 2016, as supplemented by letter dated January 26, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML16350A104 and ML17038A408, respectively), Entergy Nuclear Operations, Inc. (the licensee) submitted Relief Request No. IP3-ISl-RR-09 to the U.S. Nuclear Regulatory Commission (NRC or the Commission) requesting the use of an inspection procedure at Indian Point Nuclear Generating Unit No. 3 with a depth sizing error that is greater than the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and ASME Code Case N-696, "Qualification Requirements for Mandatory Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1," for the fourth 10-year inservice-inspection (ISi) interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii),

the licensee requested relief from the depth sizing uncertainty qualification requirement for ultrasonic examinations conducted from the inside diameter (ID) of pipes (i.e., root-mean square (RMS) error not greater than 0.125 inches) contained in ASME Code Cases N-695 and N-696.

The licensee requested relief from the requirements for ISi items on the basis that the ASME Code requirement is impractical.

2.0 REGULATORY EVALUATION

In its letter dated December 2, 2016, the licensee requested relief from the 0.125-inch RMS error depth sizing acceptance criteria contained in ASME Code Cases N-695 and N-696 pursuant to 10 CFR 50.55a(g)(5)(iii). The following regulations were considered in the NRC staff's review:

Enclosure

lnservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147).

  • 10 CFR 50.55a(g)(5)(iii) states, in part, that licensees may determine that conformance with certain Code requirements is impractical and that the licensee shall notify the Commission and submit information in support of the determination.
  • 10 CFR 50.55a(g)(6)(i) states, in part, that the Commission will evaluate determinations under paragraph (g)(5) of this section that Code requirements are impractical and that the Commission may grant such relief and may impose such alternative requirements as it determines are authorized by law and will not endanger life or property.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to grant, the relief requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 The Licensee's Relief Request Component Descriptions The welds covered by Relief Request No. IP3-ISl-RR-09 are four reactor vessel hot leg and four cold leg nozzle to safe-end dissimilar metal (OM) welds and the eight corresponding safe-end to pipe elbow austenitic steel welds.

Applicable Code Requirement

The code of record for the fourth 10-year ISi interval is ASME Code,Section XI, 2001 Edition with 2003 Addenda.

The hot and cold leg OM weld examination requirements are covered in ASME Code Case N-770-1, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1."

ASME Section XI, Code Case N-770-1, Item A-2, "Unmitigated Hot Leg Butt Welds," specifies that a visual inspection is performed every refueling outage, and a volumetric examination is performed every 5 years. ASME Section XI, Code Case N-770-1, Item B, "Unmitigated Cold Leg Butt Welds," specifies that a visual examination is performed once per interval, and a volumetric examination must be performed every second inspection period, not to exceed 7 years.

The safe-end to pipe elbow welds are covered under the risk-informed program. Risk-informed R-A requires volumetric examination of safe-end to pipe/elbow austenitic welds every 10 years.

The welds will be examined from the inner diameter, requiring the use of ASME Code

Cases N-695 and N-696, as ASME Code Section XI, 2001 Edition, with 2003 Addenda, has no rules for Appendix VIII qualifications for inner-diameter examinations.

Proposed Inspection The licensee proposes to use Code Cases N-695-1 and N-696-1, which allow a depth sizing RMS error of 0.25 inches instead of the 0.125 inches specified for depth sizing in Code Cases N-695 and N-696, and ASME Code,Section XI, Appendix VIII, Supplements 2 and 10.

The examination vendor contracted to perform the safe end examinations has demonstrated the ability to depth size indications in OM welds with an RMS error of 0.189 inches for the reactor pressure vessel (RPV) nozzle to safe-end OM welds (Appendix VIII, Supplement 10) and 0.245 inches RMS error for the safe-end to pipe austenitic welds. If the examination vendor demonstrates an improved depth sizing RMS error prior to the examination, improved RMS error will be used in any flaw sizing instead of the 0.189-inch and 0.245-inch RMS error.

If a reportable flaw is detected and determined to be ID surface connected during examination of the welds in accordance with this relief request, the licensee will provide a flaw evaluation, including the measured flaw size as determined by ultrasonic examination for NRC review.

Eddy current testing will be used to determine if flaws are surface connected. Additional data, including details of the surrounding ID surface contour in the region of the flaw and percentage of the examination area where ultrasonic testing (UT) probe lift-off is evident, if any, will be included.

In the event that any flaws requiring depth sizing are detected during examination of welds in accordance with this relief request, the following criteria shall be implemented:

  • Flaws detected and measured as less than 50 percent through-wall in depth shall be adjusted by adding a correction factor to the flaw depth such that the adjusted flaw depth is equal to the measured flaw depth +(applicable vendor RMS error-
0. 125 in.), prior to comparison to the applicable acceptance criteria;
  • For flaws detected and measured as 50 percent through-wall depth or greater, and to remain in service without mitigation or repair, the licensee will submit flaw evaluations for review and approval prior to reactor startup. The flaw evaluation will include:
1. Information concerning the mechanism that caused the flaw.
2. Information concerning the inside surface roughness and/or profile of the region surrounding the flaw in the examined piping weld.
3. Information concerning areas where UT probe lift-off is observed, if any.

Basis for the Request During the upcoming Spring 2017 refueling outage, the licensee will perform ultrasonic examination of the four hot leg safe-end to nozzle OM welds, and during the 2019 outage, the four cold leg safe-end to nozzle OM welds will be ultrasonically examined. These examinations will be performed from the ID of the weld utilizing remote inspection equipment. ASME Code Case N-695 will be used to establish the qualification requirements when performing OM weld inspections only, and ASME Code Case N-696 will be used to establish the qualification

requirements for the inspectors when performing coordinated inspections of both DM and austenitic stainless steel welds.

To date, the contracted vendor, who is qualified for detection and length sizing on these welds, has not met the RMS error requirement for depth sizing. The examination vendor has demonstrated ability to meet the depth sizing qualification requirement with an RMS error of 0.189 inches for the RPV nozzle to safe-end DM welds and 0.245 inches RMS error for the safe-end to pipe welds instead of the 0.125 inches required by the Code Case.

Duration of the Proposed Relief Relief is requested for the fourth 10-year interval, which began on July 21, 2009, and is currently scheduled to end on July 20, 2019.

3.2 NRC Staff Evaluation The licensee will use NRC-approved Code Case N-695 to satisfy the requirements of ASME Code,Section XI, Appendix VIII, Supplement 10, and ASME Code Case N-696 to satisfy the requirements of ASME Code,Section XI, Appendix VIII, Supplement 2. Code Cases N-695 and N-696 require that procedures used to inspect welds from the inside surface of the pipe be qualified by performance demonstration. The acceptance criterion in Code Cases N-695 and N-696 specify that the RMS error of the examination procedures shall not be greater than 0.125 inches. The licensee's inspection vendor was able to depth size with an RMS error of 0.189 inches. The licensee is requesting relief from the 0.125-inch depth sizing requirement in ASME Code Cases N-695 and N-696, in accordance with 10 CFR 50.55a(g)(5)(iii).

The NRC staff has confirmed that since 2002, the industry has not been able to satisfy the RMS error acceptance criterion of less than 0.125 inches when qualifying the volumetric examination inspection procedures performed from the inside surface of a pipe. Developing new technology capable of meeting the 0.125-inch RMS error, and qualifying the new technology to meet the requirements of ASME Code Cases N-695 and N-696, would be a burden on the licensee. The staff concludes that this repeated inability to qualify inside surface UT inspection techniques in accordance with ASME Code Cases N-695 and N-696 constitutes an impracticality, as described in 10 CFR 50.55a(g)(5)(iii).

To address the issue of increased potential for undersizing of flaws by inside surface UT inspection procedures that do not meet ASME Code Cases N-695 and N-696 acceptance criterion, in 2012, the NRC staff, in conjunction with personnel from the Performance Demonstration Initiative (POI), examined the proprietary UT examination data set compiled from all attempts to date to qualify inside surface UT inspection procedures to the acceptance criterion contained in ASME Code Cases N-695 and N-696. Based on this examination, the NRC staff concluded that:

(a) For flaw depths less than or equal to 50 percent pipe wall thickness, a flaw could be appropriately depth sized if a correction factor is added to the measured flaw depth such that the adjusted flaw depth is equal to the measured flaw depth plus the difference between the vendor procedure qualification RMS error and 0.125 inches.

(b) For flaw depths greater than 50 percent wall thickness, the variability of sizing errors is sufficiently large so that no single mathematic flaw size adjustment formula is sufficient to provide reasonable assurance of appropriate flaw depth sizing. As a

result, the NRC staff finds it necessary to evaluate the flaws that have depth greater than 50 percent through-wall on a case-by-case basis.

To provide reasonable assurance of the structural integrity of examined welds, the NRC staff determined that the following compensatory measures shall be applied to any inspection not meeting the 0.125-inch RMS error for depth sizing to address the measurement uncertainty in flaw depth sizing when examining welds from the inside surface:

(1) Examine the welds under consideration using a UT technique that is qualified for flaw detection and length sizing.

(2) For flaws with a measured depth of less than 50 percent of the wall thickness, the depth shall be adjusted by adding the measured flaw depth to the difference between the procedure qualification RMS error and 0.125 inches.

(3) For flaws with measured depth of greater than 50 percent of the wall thickness, either the degraded weld needs to be repaired in accordance with the ASME Code, or a flaw evaluation needs be submitted to the NRC staff for review and approval prior to reactor startup.

(4) In addition to information normally contained in flaw evaluations performed in accordance with the ASME Code,Section XI, IWB-3600, the submitted flaw evaluation shall include (a) information concerning the degradation mechanism that caused the crack, (b) information concerning the surface roughness and/or profile in the area of the examined pipe and/or weld, and (c) information concerning areas in which the UT probe may "lift off' from the surface of the pipe and/or weld.

(5) Perform eddy current examinations to confirm whether a flaw is connected to the inside surface of the pipe and/or weld.

The nozzle-to-safe-end and safe-end-to-pipe configurations differ from the POI mockups in two respects. The weld surfaces were machined smooth, and the stainless steel safe-ends and welds were cladded on both inner and outer surfaces. The NRC staff has no technical issues with the ID examinations of the wrought pipe to cast stainless steel elbow welds. As with the corrosion resistant clad, UT and followup ET inspections from the ID should be effective at detecting and length sizing flaws. Depth sizing is very challenging in cast stainless steels, but the proposed inspection procedure addresses this concern effectively.

The NRC staff concludes that the licensee's alternative is consistent with the compensatory measures discussed above, because (1) the licensee will add the correction factor to the crack tips, (2) the licensee will use eddy current testing to verify whether an embedded flaw is connected to the inside surface, and (3) the licensee will submit any flaw analysis for flaws greater than 50 percent through-wall to the NRC staff for review and approval prior to startup.

Therefore, the NRC staff determines that relief from the depth sizing RMS error acceptance criterion of ASME Code Case N-695 and using a vendor with a 0.189-inch RMS error for depth sizing and ASME Code Case N-696 with a 0.245-inch RMS error for depth sizing provides reasonable assurance of the structural integrity and leak tightness of the subject welds.

4.0 CONCLUSION

As set forth above, the NRC staff determines that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law, 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.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(5)(iii). Therefore, the staff grants the licensee's Relief Request No. IP3-ISl-RR-09 at Indian Point Nuclear Generating Unit No. 3 for the fourth 10-year ISi interval, which began on July 21, 2009, and is scheduled to end on July 20, 2019.

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

Principal Contributor: S. Cumblidge Date: March 16, 2017

Vice President, Operations

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3- RELIEF REQUEST IP3-ISl-RR-09 FOR ALTERNATIVE TO THE DEPTH SIZING QUALIFICATION REQUIREMENT (CAC NO. MF8896) DATED MARCH 16, 2017 DISTRIBUTION:

Public RidsNrrDorlLpl 1 RidsNrrPMlndianPoint LPL 1 Reading File RidsNrrLALRonewicz RidsACRS_MailCTR RidsRgn1 MailCenter TSetzer, R-1 JBowen, OEDO RidsNrrDeEpnb SCumblidge, NRR ADAMS A ccess1on Num b er: ML17069A283 *b1y e-ma1 OFFICE DORL/LPLl/PM DORL/LPLl/LA DE/EPNB/BC* DORL/LP LI/BC NAME DPickett LRonewicz DAiley JDanna DATE 03/13/2017 03/13/2017 03/09/2017 03/16/2017 OFFICIAL RECORD COPY

Retrieved from "https://kanterella.com/w/index.php?title=ML17069A283&oldid=2440027"