ONS-2017-072, Fifth Ten Year Lnservice Inspection Interval, Relief Request No. 17-0N-001; Alternative to Extend the Code Case N-770, Inspection Item B Examination Frequency

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Fifth Ten Year Lnservice Inspection Interval, Relief Request No. 17-0N-001; Alternative to Extend the Code Case N-770, Inspection Item B Examination Frequency
ML17279A108
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 10/03/2017
From: Grant H
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ONS-2017-072, RR 17-ON-001
Download: ML17279A108 (8)
v • d • e


Text

H. Todd Grant General Mqnager, Nuclear Engineering Oconee Nuclear Station Duke Energy ONO 1VP I 7800 Rochester Hwy Seneca, SC 29672 ONS-2017-072 0. 864.873.6767 f: 864.873.5791 Todd. Grant@duke-energy.com October 3, 2017 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Duke Energy Carolinas, LLC (Duke Energy)

Oconee Nuclear Station, Units 1 and 2.

Docket Numbers 50-269 and 50-270 Renewed License Numbers DPR-38 and DPR-47

Subject:

Fifth Ten Year lnservice Inspection Interval, Relief Request No. 17-0N-001; Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency Pursuant to 10 CFR 50.55a(z)(2), Duke Energy requests the NRC to grant relief from Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and the augmented inspections of ASME Code Case N-770 as prescribed by 10CFR50.55a(g)(6)(ii)(F). Relief is being sought due to the hardship, with no compensating increase in quality or safety, caused by performing an augmented inspection required by code case N-770 at a frequency of less than 10 years from the previously inspection.

In the absence of this relief being granted, a core barrel removal will be required for the Unit 1 Fall 2018 outage. Because of the significant amount of lead time needed to plan and procure services to remove the core barrel, Duke Energy's desire is that this Relief Request receive review and approval as soon as practical and at least by April 12, 2018.

Relief Request 17-0N-001 is provided as an enclosure to this letter.

If there are any questions or further information is needed you may contact David Haile at (864) 873-4742, Sincerely, H.~JfJrd-H. Todd Grant General Manager, Nuclear Engineering Oconee Nuclear Station

Enclosure:

Relief Request: 17-0N-001: Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency NRC FORM 366 (04-2017)

ONS-2017-072 October 3, 2017 Page 2 cc:

Ms. Catherine Haney, Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Ms. Audrey L. Klett, Project Manager (by electronic mail only)

U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-08B1A Rockville, MD 20852-2738 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Station

Enclosure to ONS-2017-072 Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1 and 2 Relief Request 17-0N-001 Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency

Oconee Nuclear Station, Units 1 and 2 Relief Request 17-0N-001 Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency

1. ASME Code Components Affected:

The Code Components included in the scope of this relief request are the nozzle-to-safe end Dissimilar Metal Butt Welds (DMBWs) for the Core Flood Nozzles on Oconee Nuclear Station's Units 1 and 2 (ONS1 & ONS2). This relief request addresses the examination frequency requirements of Code Case N-770-2 (Reference 8.1 ), Inspection Item B for the following welds.

Weld

Description:

ONS1: 1-RPV-WR53 - Reactor Vessel Core Flood Nozzle-to-Safe End, 14" DMBW ONS1: 1-RPV-WR53A- Reactor Vessel Core Flood Nozzle-to-Safe End, 14" DMBW ONS2: 2-RPV-WR53 - Reactor Vessel Core Flood Nozzle-to-Safe End, 14" DMBW ONS2: 2-RPV-WR53A- Reactor Vessel Core Flood Nozzle-to-Safe End, 14" DMBW

2. Applicable Code Edition and Addenda

ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition and 2008 Addenda.

Additionally 10 CFR 50.55a(g)(6)(ii)(F)(1) mandates the use of ASME Code Case N-770-2.

3. Applicable Code Requirement

Code Case N-770-2, Table 1, Item B requires volumetric examination of "Unmitigated butt welds at Cold Leg operating temperature ~ 525°F and < 580°F" every second inspection period not to exceed 7 years.

Note: This relief request will apply to future versions of Code Case N-770 that may be incorporated in 10 CFR 50.55a(g)(6)(ii)(F)(1) provided the Item B maximum examination frequency continues to be less than 1O years.

4. Reason for Request

Relief is being sought to allow the 7 year maximum examination frequency to be extended to a maximum of every 10 years to coincide with the core barrel removal for the Reactor Vessel weld inspections (Table IWB-2500-1, Category B-A exams).

Oconee's plant design limits access to the outer surface of the Core Flood Nozzle-to-Safe End DMBWs. Acceptable examination coverage of these welds can only be achieved when approached from the inner diameter of the weld. The outer surfaces of these welds are confined within an approximately 3 ft. wide Reactor Vessel annulus, and underneath the annulus shield blocks. The physical interferences and high dose rates associated with the weld location, prevent effective ultrasonic examinations from being performed from the outside diameter of the Core Flood Nozzles.

It is possible to access and perform an effective volumetric examination of the weld from the inner diameter, provided the reactor vessel core barrel is removed. However, core barrel removal is a hardship that produces increased plant risk and elevates the outage dose to workers.

The ONS1 and ONS2 welds had baseline volumetric examinations performed during their last respective Reactor Vessel 10 Year ISi in the fall of 2012, and fall of 2013 with essentially 100%

coverage. The baseline inspections were performed in accordance with ASME Section XI, Appendix VIII, Supplement 10 requirements and found no reportable circumferential or axial Page 1of5 ONS 2017 Enclosure

Oconee Nuclear Station, Units 1 and 2 Relief Request 17-0N-001 Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency indications (Note: The IWB, Item B5.10 inspections of these welds conducted over the past 40 years (the past 4 ISi intervals) have never found reportable circumferential or axial indications).

Based on the acceptable results from conservative fracture mechanics analyses for these welds and the hardship of removing the core barrel to perform this examination, there is no compensating increase *;

in quality or safety by limiting the frequency to every 7' years as opposed to every 10 years.

5. Proposed Alternative and Basis for Use:

Proposed Alternative: The proposed alternative is to extend the weld examination frequency required by Code Case N-770-2, Table 1, Item B for the ONS1 and ONS2 Core Flood nozzle DMBWs (listed in section 1 above) from "every second inspection period not to exceed 7 years" to "a maximum of 10 years from the previous examination." This alternative will allow the N-770 inspections of these welds to coincide with the 10 Year ISi Refueling Outages for the respective units, currently scheduled as follows:

ONS1 ONS2 Outage ID: 1EOC32 2EOC31 Date: Fall 2022 Fall 2023 Basis for Use: 10 CFR 50.55a(z) states the following:

"Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:

(1) The proposed alternative would provide an acceptable level of quality and safety, or (2) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety."

Duke Energy is seeking approval of the proposed alternative in accordance with 10 CFR 50.55a(z)(2) on the basis that the proposed alternative would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

Analytical results provide expectation of structural integrity and leak tightness of the weld for 10 years.

A review of the fabrication records for the ONS1 and ONS2 Core Flood Nozzles, supports that there were no weld repairs of Inside Diameter (RCS wetted) surface of the DMBWs during the weld installation. The Baseline Inspection of these welds, in 2012 and 2013 respectively, found no reportable indications.

Oconee specific fracture mechanics analyses demonstrate that, with the assumption that an undetected 10% through wall flaw is present and 10 years of Primary Water Stress Corrosion Cracking (PWSCC) crack growth, neither a circumferential nor axial flaw will exceed the ASME Code Section XI, IWB-3640 acceptance criteria of 75% through wall depth. These results substantiate that deferring the Core Flood nozzle DMBW examination frequency to 1O years has no adverse impact on the quality or safety of the weld as compared to the 7 year examination frequency. Thus, the 10 year Page 2 of 5 ONS 2017 Enclosure

Oconee Nuclear Station, Units 1 and 2 Relief Request 17-0N-001 Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency frequency extension provides reasonable assurance that the structural integrity and the leak tightness of the weld will be maintained.

The results of Oconee-specific residual stress analyses and crack growth analyses for circumferential

& axial cracking, including the maximum calculated and allowable flaw sizes for the Core Flood DMBWs are tabulated below. Note that the Allowable (Critical) flaw lengths are based on elastic-plastic fracture mechanics analyses.

Circumferential Flaws: Allowable vs. Analvtical Results Allowable *,

Aspect Ratio (I/a) Analytical results Flaw Length = 16.1" t*. NA Flaw length at 75% through wall = 4.23" Through Wall = 75% (a/t) 10:1 Time to Reach 75% wall = 11.0 years Axial Flaws* Allowable vs Analvtical Results Ii Allowable '.;t Aspect Ratio (I/a) Analytical results Flaw Length = 2. 7 4" .

" NA Flaw length at 75% through wall = 1.02"

., 2:1 Time to Reach 75% wall = 10.4 years '

','* ... Time to Reach 75% wall = 18.3 years Through Wall = 75% (a/t)  : Variable

[.'., Time to Reach 75% wall= 18.4 years I *. 1:1 Key aspects of the fracture mechanics calculations are summarized below:

1. No reportable flaws were found by the past volumetric Table IWB-2500-1 exams in the previous four lnservice Inspection (ISi) Intervals (40 years of service).
2. Results of the most recent volumetric examinations that were performed to Section XI, Appendix VIII, Supplement 10 requirements, achieved essentially

~ - ,_ ... ....... ,. ___ _ 100% coverage for both Circumferential

~-- ~-

and Axial flaws and resulted in no reportable PWSCC indications.

3. Core Flood nozzle weld fabrication sequence incorporated to model the DMBW residual stresses.
4. Two rigorous records reviews indicate that ONS1 & 2 had no Inside Diameter weld repairs
  • during fabrication.
5. A nominal reactor coolant temperature at the Core Flood nozzle was assumed to be 557 °F.
  • 6. The circumferential flaw analysis used EPRl's Materials Reliability Program (MRP)-287
  • (Reference 8.2) guidance, assumed an initial 10% through wall flaw, a 50% through wall repair during fabrication, maximum stress path and PWSCC crack growth rates in accordance with MRP-115 (Reference 8.3)guidance.
7. The axial flaw analysis assumed an initial 10% through wall flaw, no through wall repair during fabrication (matching the fabrication records), maximum stress path and PWSCC crack growth rates in accordance with MRP-115 guidance. Applied stress intensity factors were developed from a finite element fracture mechanics model for the axial flaw growth that resulted in* an aspect ratio that was not fixed, but varied as the crack grew through wall under the applied hoop Page 3 of 5 ONS 2017-72 *Enclosure

Oconee Nuclear Station, Units 1 and 2 Relief Request 17-0N-001 Alternative to extend the Code Case N-770, Inspection Item 8 Examination Frequency stress field. In addition, Handbook solutions for fixed aspect ratios of 1: 1 and 2: 1 were also evaluated.

Factors for the circumferentially oriented PWSCC flaw analysis included:

  • the modeling of the adjacent stainless steel safe end and weld,
  • the modeling of the weld residual stresses used the original fabrication sequence, including the hydro test and normal operating conditions,
  • Use of a conservative envelope for Core Flood nozzle external loads,
  • An assumed initial 10% through wall flaw depth, an initial 10:1 (length to depth) flaw aspect ratio and

Factors for the axially oriented PWSCC flaw analysis included:

  • The modeling of the adjacent stainless steel safe end and weld,
  • The modeling of the weld residual stresses used the original fabrication sequence, including the hydro test and normal operating conditions, *
  • Normal Operating Pressure,
  • An assumed initial 10% through wall flaw depth, with multiple flaw aspect ratios, and

Note that the analytical methodologies applied in this Relief Request are similar to the methodologies used for the previously approved Relief Request listed as "Precedents" in Section 7. However, residual stresses were based on ONS fabrication records which supports that no weld repairs were performed on the subject welds.

The initial assumed flaw size of 10% through wall was selected because the requirements of the ASME Section XI, Appendix VIII, Supplement 10 inspection qualification procedure, sets the minimum detectable flaw depth as being 10% of the wall thickness.

6. Duration of Proposed Alternative:

This request may be implemented for the remainder of the 5th Interval ISi plan and applies to ONS units 1 and 2. This relief will be re-submitted, if needed, for subsequent ISi intervals.

7. Precedents 7.1 Indian Point Unit 3 Docket Number 50-286, Relief Request IP3-ISl-RR-07 for Code Case N-770-1 Weld Inspection Frequency Extension, including NRC Safety Evaluation dated 8/4/2014, (ADAMS Accession Number ML14017A054).

7.2 Farley Nuclear Power Station, Units 1 and 2, Docket Numbers 50-348 & 50-364, Request for Relief FNP-ISl-ALT-15, version 1, Alternative to In-Service Inspection Regarding RPV Cold-Leg Nozzle Dissimilar Metal Welds, (ADAMS Accession Number ML14262A317).

Page 4 of 5 ONS 2017 Enclosure

Oconee Nuclear Station, Units 1 and 2 Relief Request 17-0N-001 Alternative to extend the Code Case N-770, Inspection Item B Examination Frequency

8. References 8.1 Code Case N-770-2, 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.

8.2 MRP-287, Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC)

Flaw Evaluation Guidance, December, 2010.

8.3 MRP-115, Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182 and 132 Welds, November 2004.

Page 5 of 5 ONS 2017-72 Enclosure m

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