ML12333A317

Response to Request for Additional Information for Time Limited Aging Analysis on Reactor Vessel Internals

ML12333A317
Person / Time
Site:	Oconee
Issue date:	11/16/2012
From:	Gillespie T P Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML12333A317 (13)

Text

bDiuke OEnergyT. PRESTON GILLESPIE, Jr.Vice President Oconee Nuclear Station Duke Energy ONOI VP / 7800 Rochester Hwy.Seneca, SC 29672 864-873-4478 10 CFR 50.90 864-873-4208 fax T. Gillespie@duke-energy.

corn November 16, 2012 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Station (ONS), Units 1, 2 and 3 Renewed Facility Operating Licenses Numbers DPR-38, 47 and 55;Docket Number 50-269, 50-270 and 50-287;Response to Request for Additional Information for Time Limited Aging Analysis on Reactor Vessel Internals On February 20, 2012, Duke Energy Carolinas, LLC (Duke Energy) submitted a Time Limited Aging Analysis (TLAA) to demonstrate that the reactor vessel internals will meet deformation limits at the end of the renewed operating license. The Nuclear Regulatory Commission (NRC)requested Duke Energy to provide additional information related to that submittal by electronic mail dated October 4, 2012. Duke Energy's response to that request is provided in the Enclosure 3.Information contained in Enclosure 3 is classified by AREVA NP as proprietary.

The appropriate affidavit from AREVA NP is provided in Enclosure 1 in accordance with the provisions of 10 CFR 2.390. A non proprietary version of Enclosure 3 has been provided in Enclosure 2.There are no regulatory commitments being made as a result of this response.

Inquiries on this submittal should be directed to Boyd Shingleton, ONS Regulatory Affairs Group, at (864) 873-4716.I declare under penalty of perjury that the foregoing is true and correct. Executed on November 16, 2012.Sincerely, T. Preston Gillespie, Jr., Vice President, Oconee Nuclear Station Enclosure 3 to this letter contain proprietary information.

Withhold From Public Disclosure Under 10 CFR 2.390.Upon removal of Enclosure 3, this letter is uncontrolled.

www. duke-energyom cor U. S.' Nuclear Regulatory Commission November 16, 2012 Page 2

Enclosures:

1. AREVA NP Affidavit 2. Response to NRC Request for Additional Information

-Non Proprietary

3. Response to NRC Request for Additional Information

-Proprietary

• U. S Nuclear Regulatory Commission November 16, 2012 Page 3 cc w/enclosures:

Mr. Victor McCree, Regional Administrator U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. John Boska, Project Manager (by electronic mail only)Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop O-8G9A Rockville, MD 20852 NRC Senior Resident Inspector Oconee Nuclear Station Susan E. Jenkins Manager, Infectious and Radioactive Waste Management, Bureau of Land and Waste Management Department of Health & Environmental Control 2600 Bull Street, Columbia, SC 29201 Enclosure 1 AREVA NP Affidavit for Enclosure 3

Encl6sure 1- AREVA NP Affidavit for Enclosure 3 November 16, 2012 Page 1 AFFIDAVIT COMMONWEALTH OF VIRGINIA) ss.CITY OF LYNCHBURG 1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information contained in the report ANP-3178(P), Revision 0, entitled "Response to NRC Request for Additional Information on Analysis for Oconee Units 1, 2, and 3 Vessel Internals, ME8436, ME8437, and ME8438," dated November 2012 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is Enclosure 1- AREVA NP Affidavit for Enclosure 3 November 16, 2012 Page 2 requested qualifies under I OCFR 2.390(a)(4) "Trade secret and commercial or financial information." 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

Enclbsure 1- AREVA NP Affidavit for Enclosure 3 November 16, 2012 Page 3 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this day of 2012.Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31115 Reg, # 110864 d KATHLEEN ANN SENNETT di Natiry Public COMMllfl4l0h of Virginia Ag110.64 MY Commission Expires Aug 31, 2016 Enclosure 2 Response to NRC Request for Additional Information Non Proprietary Encl6sure 2 Response to Request for Additional Information

-Non Proprietary November 16, 2012 Page 1 Introduction This enclosure has been redacted to remove AREVA proprietary information provided in Enclosure

3. AREVA Proprietary information removed from the RAI responses is shown as blank spaces within the square brackets, "[ ]". The NRC references section at the end of this enclosure applies to all three RAIs. The Duke Energy references section applies to all three RAI responses.

RAI 1 Section 3.2 of the report entitled "Update of Irradiation Embrittlement in BAW-10008 Part 1 Rev. 1," (Ref. 1) provided the neutron flux and neutron fluence for the core barrel flanges.Section 3.2 of the Reference 1 describes the assumptions used in determining the fluence but does not describe the methodology used to calculate the fluence.Describe the methodology used to determine the neutron fluence and address the consistency of the fluence methodology with the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Response to RAI 1 The methodology used to determine the neutron fluence was based on AREVA's NRC approved fluence analysis methodology, described in topical report BAW-2241 P-A. This fluence analysis methodology is consistent with the guidance of Regulatory Guide 1.190,"Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." In addition, a 60 year (54 effective full power years (EFPY)) fluence was calculated for the core support shield using the methodology described above. A fast fluence (E > 1.0 MeV) rate was calculated for one representative ONS unit and cycle. An 80% and 90% capacity factor for 20 and 40 years, respectively, was used to project the fluence value. This representative fast fluence (E > 1.0 MeV) was determined to be [ ] at the bottom of the core support shield. Fluence at the top of the core support shield would be less, due to being farther from the core.RAI 2 Appendix E to BA W-10008, Part 1, Rev. 1, "Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake," (Ref. 2) evaluates the adequacy of the ductility of the Oconee Nuclear Station (ONS) Units 1, 2, and 3 reactor vessel internals (RVI) under the combined loading resulting from a loss of coolant accident (LOCA) and seismic event. Section 3.4 of Reference 1 provides a justification for the use of uniform elongation data from slow strain rate tests (SSRT) which are performed at strain rates of around 10-7/second rather than data from standard tensile tests which are performed at 10-2/second to Enclosure 2 Response to Request for Additional Information

-Non Proprietary November 16, 2012 Page 2 10-4/second, to evaluate the ductility of the internals in the update of this analysis for 60 years.However, it is not clear how the strain rates from the different test methods relate to the strain rate that would occur in the postulated LOCA plus seismic event.What strain rates are assumed in the loadings evaluated in BAW-10008 Part 1, Rev. 1? If the strain rates for the test data are substantially different from the strain rates for the postulated event, discuss how the test data for uniform elongation are conservative for evaluating the ductility of the RVI during the postulated event.Response to RAI 2 In preparation of this RAI response, an error 1 was identified in the submitted document, "Update of Irradiation Embrittlement in BAW-10008 Part 1 Rev. 1," Reference

1. This document identified the locations of maximum stress intensity during a combined LOCA/seismic event as the upper and lower core barrel flanges, based on wording in Appendix E of BAW-1 0008, Part 1, Rev. 1. However, as documented in Section 3.2.3.2 of BAW-1 0008, Part 1, Rev. 1, the peak stress intensities of interest occur near the upper and lower core support shield flanges. This section also states that "Stress intensities for the core barrel are less severe than for the core support shield." The material for the ONS Units 1, 2, and 3 upper and lower core support shield flanges is solution annealed Type 304 stainless steel. The representative fluence for the bottom of the core support shield (i.e., the lower flange) is [ ] at 60 years (54 EFPY) for the ONS units. The fluence at the core support shield upper flange is less than this value due to its proximity to the core. The fluence at the core support shield lower flange is equivalent to approximately

[ ], using the light water reactor conversion factor of 1022 neutrons/cm 2 , E>1.0 MeV = 15 dpa, page 6 of Reference

2. At this dose, any decrease in ductility due to irradiation embrittlement is insignificant for Type 304 stainless steel, Figure 13(c)of Reference

2. Therefore, the irradiated tensile and slow strain rate tensile test data are not pertinent to the upper and lower core support shield flanges; instead, the effect of strain rate on ductility of unirradiated material is considered for the upper and lower core support shield flanges.As shown in Figure 2-1,of Reference 3 (figure from reference also provided below), the uniform elongation of unirradiated solution annealed Type 304 stainless steel at 600°F is seen to decrease slightly with increasing strain rate. However, even at the highest tested strain rates, at 600'F the uniform elongation is well above the 20% uniform elongation of irradiated material credited for 40 years in Appendix E of BAW-1 0008, Part 1, Rev. 1 and the 8.6% allowable strain specified in Appendix A of BAW-1 0008, Part 1, Rev. 1. Furthermore, as shown in Figure 2-2 of Reference 3 (figure from reference also provided below), it is also observed that yield strength increases with increasing strain rate at 600 0 F. Thus, in addition to having sufficient ductility at 60 years relative to the allowables of BAW-1 0008, Part 1, Revision 1, the upper and lower core support shield flanges will have greater resistance to plastic deformation at increased strain rates.Therefore, the original conclusions from Appendix E to BAW-1 0008, Part 1, Revision 1 concerning the acceptable ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime for the ONS units.1 This error and subsequent corrective actions are being addressed through the corrective action programs for Duke Energy and AREVA.

Encl6sure 2 Response to Request for Additional Information

-Non Proprietary November 16, 2012 Page 3 Figure 2-1: Effect of strain rate and temperature on the uniform elongation of Type 304 stainless steel, Reference 3 50.4 I, S=30 8 -~-. A 0..-**'--a.- a sod',£** i~ur~aI60d~.I ...~ ...i........I

-I ~ -~ ...a..,.3 S 10 lt o NOMINAL STRAIN RAl, SIC 1-Figure 2-2: Effect of strain rate and temperature on the 0.2 percent yield stress of Type 304 stainless steel, Reference 3 so oU0AF*10WPF 401 S: It CS 30 14MO0£IffiF 8 10 0..W ... I ..10-3 I0-2 10-1 alo 101 NOMINAL MiRAIN RAnI, SEC-Encl6sure 2 Response to Request for Additional Information

-Non Proprietary November 16, 2012 Page 4 RAI 3 Section 3.4 of Reference 1 indicates that the majority of the uniform elongation data for solution annealed Type 304 stainless steel (Type 304SA) presented in Reference I is from SSRT, which are performed at strain rates of around lOE-7/second, while standard tensile tests are performed at 1 OE-2/second to 1 OE-4/ second. Section 3.4 of the report references Figure 3-11 of Reference 1 with respect to the effect of strain rate on the uniform elongation of Type 304SA.Figure 3-11 provides uniform elongation data from Type 304SA irradiated in the EBR-11 test reactor as a function of strain rate for room temperature, 450 0 F, and 700'F. Section 3.4 of Reference 1 states that uniform elongation is seen to decrease moderately with decreasing strain rate at elevated temperatures of 450°F and 700°F [thus] the uniform elongation values from SSRT at -IOE-7/second are conservative compared to those obtained from conventional tensile tests at a strain rate typically ranging from lOE-4/second to 10E-2/second.

However, the staff notes that the uniform elongation data shown in Figure 3-11 are from material irradiated to approximately 1x10E23 neutrons per square centimeter (n/cm 2) (Energy > 0. 1 megaelectronvolt (Me V)), which is two orders of magnitude higher than the predicted end-of-life fluence for the limiting high-strain component for the ONS, Units 1, 2, and 3 RVI. Additionally, the strain rates of the tests depicted in Figure 3-11 range from 2xlOE-3/minute (3.3xlOE-5/second) to 2/minute (3.3xlOE-2/second), which do not overlap the strain rates of the SSRT.Provide furtherjustification for applying the data in Figure 3-11 of Reference 1 to demonstrate that the use of SSRT data to predict the uniform elongation of Type 304SA in the ONS Units 1, 2, and 3 RVI is conservative, given the higher neutron fluence of the EBR-11 materials and the different strain rate range of the testing.Response to RAI 3 As indicated in the response to RAI 2, the locations of maximum stress intensity are the core support shield upper and lower flanges. Any decrease in ductility due to irradiation embrittlement is insignificant for these components.

Therefore the irradiated test data described in RAI 3 are no longer applicable.

NRC RAI References

1. Update of Irradiation Embrittlement in BAW-1 0008, Part 1, Rev. 1, Areva Document No. 51-9038244, Duke Energy Calculation OSC-1 0237, Enclosure to letter from T. Preston Gillespie to NRC dated February 20, 2012;

Subject:

Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1,2, and 3, Docket Numbers 50-269, 50-270, and 50-287, License Renewal Commitment to Submit a Time Limiting Aging Analysis for the Reactor Vessel Internals to the NRC For Review. (ADAMS Accession No. ML12053A332).

2. AREVA NP Inc. Document BAW-1 0008, Part 1, Rev. 1, "Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake," June, 1970.

Encl6sure 2 Response to Request for Additional Information

-Non Proprietary November 16, 2012 Page 5 Duke Energy RAI Response References

1. "Oconee, Units 1, 2 and 3 -License Renewal Commitment to Submit a Time Limiting Aging Analysis for the Reactor Vessel Internals to the NRC for Review," NRC ADAMS Accession Number ML12053A332, February 20, 2012.2. NUREG/CR-7027, "Degradation of LWR Core Internal Materials due to Neutron Irradiation," December 2010, NRC ADAMS Accession Number ML1 02790482.3. "High Strain Rate Tensile Properties of AISI Type 304 Stainless Steel," J. M. Steichen, Journal of Engineering Materials and Technology, July 1973.