ML15028A493

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ANP-3281Q1NP, Rev. 0, Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals, Attacment 2
ML15028A493
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Issue date: 01/31/2015
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Attachment 2 to ICAN011506 AREVA document ANP-3281Q1 NP, "Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals,"

NON-PROPRIETARY

A AREVA ANP-3281QINP Response to NRC Request for Revision 0 Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals January 2015 AREVA Inc.

(c) 2015 AREVA Inc.

Copyright © 2015 AREVA Inc.

All Rights Reserved

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

AREVA Inc. ANP-3281Q1 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page ii Contents Page NRC BACKGROUND ................................................................................................... 1 RAI-1-CB Flange: ............................................................................................. 2 Statement of RAI ................................................................................... 2 Response to RAI-1-CB Flange .............................................................. 3 RAI-2-CSSUF: .................................................................................................. 5 Statement of RAI ................................................................................... 5 Response to RAI-2-CSSUF: ................................................................. 5 RAI-3-Neutron Fluence: .................................................................................... 7 Statement of RAI ................................................................................... 7 Response to RAI-3-Neutron fluence ..................................................... 8 REFERENCES .............................................................................................................. 19 List of Tables None.

List of Figures Figure 1 B&W PW R Internals General Arrangement ............................................... 16 Figure 2 Sketch of R-8 DORT Core Model (Not to Scale) ........................................ 17 Figure 3 Sketch of R-Z DORT Model (Not to Scale) ................................................. 18

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page iii Nomenclature (Ifapplicable)

Acronym Definition AMP Aging Management Program ANO-1 Arkansas Nuclear One, Unit 1 B&W Babcock & Wilcox BWOG B&W Owners Group CB Core Barrel CSSBF Core Support Shield Bottom Flange CSSUF Core Support Shield Upper Flange (NRC terminology) dpa Displacements per atom DORT Discrete ORdinate Transport code EOI Entergy Operations, Incorporated MeV Mega electron Volts ksi thousand pounds per square inch (kip/in 2)

LOCA Loss-of-Coolant Accident MRP Materials Reliability Program PWR Pressurized Water Reactor RAI Request for Additional Information RV Reactor Vessel RVI Reactor Vessel Internals SE Safety Evaluation

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Pa-ge 1 NRC BACKGROUND In a letter dated May 6, 2014 (1CAN051401, Reference 1), Entergy Operations Inc.,

(Entergy, the licensee) submitted an aging management program (AMP) for the reactor vessel internals (RVI) at Arkansas Nuclear One Unit 1 (ANO-1). This AMP was developed by the industry in MRP-227-A report, "Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines." MRP-227-A and the supporting reports were used as technical bases for developing ANO-1 AMP. The staff reviewed this report and issued a final safety evaluation (SE) on December 16, 2011. In the Submittal of May 6, 2014 (Reference 1), a Time-Limited Aging Analysis (TLAA) for the reactor vessel internals was included in which the licensee addressed the loss of ductility due to the exposure of neutron radiation in the RVI components at ANO-1, at 60 Years.

The staff reviewed the submittal and based on the review conducted thus far, the staff has developed request for additional information (RAI), Reference 2, as discussed below.

In Section 2.0 of its submittal dated May 6 2014 (Reference 1), Entergy Operation, Inc.

(EOI) stated that:

I

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for thp ANO-1 Rpqr*ctnr Vp..*.*.I IntprnpI-, Pnrip 2 This document provides the requested additional information on core barrel and core support shield flanges. Each RAI is stated individually with the corresponding response.

The statements of RAI and the above background are from Reference 2 with parentheticals added for reference identification where appropriate.

RAI-1-CB Flange:

Statement of RAI

[

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 3 b)[

Response to RAI-I-CB Flange Part A Appendix A of the 1970 RV internals topical report is provided For Information Only on a proprietary CD (3 copies) enclosed with this response.

Part B The maximum stress intensity for the core barrel top flange is reported in Table 1 of the 1970 RV internals topical report to be [ I under Load Case IV, which includes design loads, seismic loads, and large break (outlet and inlet pipe) loss of coolant accident (LOCA) loads. This stress intensity does not exceed the un-irradiated yield strength for solution annealed 304 stainless steel at 600°F (Appendix A, Figure A-2 and Appendix C, Table C-1 of the 1970 RV internals technical report).

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Pa-qe 4

[

Thus, the core barrel bottom flange will also remain below the un-irradiated yield strength of solution annealed 304 stainless steel.

The maximum fluence projected in ANO-1 specific fluence analysis for the core barrel flanges at 54 EFPY is [ ]. This fluence value was not available during the preparation of MRP-189 Revision 1, Reference 4. The yield strength for 304 stainless steel increases with irradiation, while the applied stress intensity will not change. Since the stress intensity at the core barrel flanges will remain below the un-irradiated yield strength, there is no plastic deformation and no impact due to irradiation induced change in ductility. Thus, as previously concluded in Reference 3, the core barrel will maintain its function during LOCA and earthquake conditions.

I[

I

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 5 RAI-2-CSSUF:

Statement of RAI I

Response to RAI-2-CSSUF:

Section 4.0 of Attachment 1 to Reference 1 states:

"As described in Section 2.0 of this document, the bottom of the core support shield (i.e., at the bottom flange) is the location of interest and the projected 54 EFPY fluence for this location at ANO-1 is [ ] ... In addition, it is noted that the fluence at the top of the core support shield (i.e., at the upper flange) would be less than at the bottom of the core support shield because of the increased distance from the core." [emphasis added] These components and configuration are depicted in Figure 1, which is on the cover of Reference 4. Sections 2.0 and 5.0 of Attachment 1 to Reference 1, also indicate that the location of highest stress intensity occurs at the core support shield bottom flange.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 6 The projected 54 EFPY neutron fluence (E > 1 MeV) for ANO-1 reactor vessel internals location of maximum stress intensity, the core support shield bottom flange, is [

] whereas the [

I I

] The core support shield assembly, bottom flange neutron fluence (E > 1 MeV) used for screening and categorization was based on Babcock & Wilcox Owners Group (BWOG) preliminary fluence and temperature analyses available at the time as indicated in the flowchart for screening and categorization, Figure 1-2 of Reference 4. This includes [

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Paae 7 Pane 7 RAI-3-Neutron Fluence:

Statement of RAI

[

I 2 Guidance in Section 1.2, "Core Neutron Source," of RG 1.190 advises representing the peripheral assemblies with a pin-wise source distribution. The basis document for this guidance considered "pin- wise source distributions in three different core/vessel configurations" to quantify the effect of neglecting a pin'-wise representation on the total estimated vessel fluence (M. Todosow and J.F. Carew, "Evaluation of Selected Approximations Used in Pressure Vessel Fluence Calculations," Transactions of the American Nuclear Society, Vol. 46, p. 658, June 1984). A similar study, specifically showing that a more explicit representation of internal fuel assemblies produces a negligible effect on the core barrel fluence estimate, is expected.

AREVA Inc. ANP-3281Q1 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Paqe 8 b) [

I c) [

Response to RAI-3-Neutron fluence PartA Core source representations discussed in Section 3.1.2 of BAW-2241P-A, Reference 6, are consistent with Regulatory Guide 1.190, Reference 5, and also adequate for calculating flux at a location closer to the core than the vessel as a pin-wise source distribution was used. A different source representation was not used.

The source representations used to calculate reactor vessel (E > 1 MeV) fluence are also used for core barrel and core support shield flange (E > 1 MeV) reactor vessel internals fluence, or anywhere else in the model. As indicated in Appendix D of BAW-2241P-A (Reference 6), in response to Set 1 - Question 6, "the multi-planar re sources and multi-channel rz sources are produced from the results of pin-by-pin, three-3 Refer to Regulatory Position 1 of RG 1.190, which states, in part, "The uncertainty of the fluence must be 20% (1a) or less [emphasis added] when the fluence is used to determine RTPTS and RTNDT for complying with 10 CFR 50.61 and Revision 2 of Regulatory Guide 1.99, "RadiationEmbrittlement of Reactor Vessel Materials," respectively. It should be recognized that this 20% uncertainty value has been included in the margin term for the RTPTS. Uncertainty of fluence for other applications should be determined using Regulatory Position 1.4 and included as an uncertainty allowance in the fluence estimate, as appropriate for the specific application."

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 9 dimensional, time-averaged source distributions. The three-dimensional source distributions come from explicit three-dimensional fuel-cycle calculations, such as those from the NEMO or PDQ codes. The calculations of the sources are produced during core-follow benchmarks of the code results to measured power densities."

Fluence values, including for reactor vessel internals components (in regions that directly surround the effective height of the active core) that are 1) determined with NRC-approved methods and 2) adherent to RG 1.190 guidance are considered acceptable as described in Section 3.2 of Reference 3. As described in the NRC safety evaluation (SE) for BAW-2241 P, Enclosure 1 of BAW-2241 P-A, Reference 6:

The approach used in BAW-2241-P is semi-analytic using the most recent fluence calculational methods and nuclear data sets. In the proposed methodology, the vessel fluence is determined by a transport calculation in which the core neutron source is explicitly represented and the neutron flux is propagqated from the core through the core barrel, the baffle, and the downcomer to the vessel (rather than by an extrapolation of the measurements). [emphasis added]

For ANO-1, subject reactor vessel internals locations, mesh points, were selected in support of the demonstration that reactor internals will have adequate ductility and meet deformation limits at the end of life, Applicant Action item 12 in BAW-2248-A, Reference 7. [

I

AREVA Inc. ANP-3281Q1 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 10 The fluence method used is NRC-approved, and adheres to RG 1.190, for regions that directly surround the effective height of the active core. It is based on a pin-by-pin source distribution and includes approximations for fluence determination. As such, a more explicit representation of internal fuel assemblies is not warranted for ANO-1 reactor internals ductility and deformation limit evaluations.

Part B Specific uncertainty was not determined for the fluence estimates at reactor vessel internals points of interest; best-estimate is considered appropriate for the application.

The fluence at the core support shield bottom flange is equivalent to approximately [

1, using the light water reactor conversion factor of 1022 neutrons/cm 2, E > 1 MeV = 15 dpa, page 6 of Reference 8. Relative to the effect of neutron irradiation on reactor vessel internals component ductility, dpa of this order of magnitude are not discernable on Figure 5-3 of ANP-3281P (Reference 9), Revision 1, which corresponds to Figure 13(c) of NUREG/CR-7027, Reference 8. The NUREG/CR-7027 figure shows ductility due to irradiation embrittlement as a function of dpa for Type 304 stainless steel. The variability of uniform elongation in the unirradiated 304 stainless steel is greater than the uncertainty in loss of uniform elongation due to neutron exposure of this order of magnitude. Therefore, the use of best-estimate fluence analyses to confirm the order of magnitude of fluence is appropriate for this application.

With respect to fluence uncertainty, the AREVA fluence methodology documented in BAW-2241P-A, Reference 6, was developed through a full-scale benchmark experiment that was performed at the Davis Besse reactor. The results of the benchmark experiment demonstrated the accuracy of a fluence analysis that employs the AREVA methodology is unbiased and has a precision within the RG 1.190, Reference5, suggested one (1) standard deviation (a) limit of 20% for reactor vessel items located in the reactor vessel beltline, i.e., surrounding the effective active height of the fuel/core.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Paqe 11 The most recent synthesis results from fluence transport calculations based on r6 and rZ DORT cases were used to determine neutron fluence at the ANO-1 core support shield bottom flange, as Well as the core barrel upper and lower flanges in support of evaluations summarized in ANP-3281P (Reference 9), Revision 1. The combined (synthesized) results provide calculations of the three-dimensional fast neutron fluence rate (time-averaged flux) for the selected reactor vessel internals points of interest. The synthesized fluence rates (fluxes) were applied to 60 years at conservative 90% overall load capacity factor or 54 effective full power years (EFPY) to determine end-of-life fluence for those locations.

Although reactor vessel internals fluence results are best-estimate as described above, AREVA has assessed the uncertainty of fluence estimates associated with the reactor internals in the beltline region. The fluence-related beltline region is defined by upper and lower radial planes that are adjacent to the active fuel and extend radially to the vessel cavity - biological shield structure. The upper and lower planes are [

] above and below the active fuel height.

The fluence-related reactor vessel beltline region extends throughout the reactor internals, from the lower plane to the upper one. It has uncertainties associated with those in BAW-2241P-A, Reference 6. The fluence modeling is unbiased with an estimated standard deviation (uncertainty in calculated results) of 7.0 percent. [

The uncertainty associated with calculated fluence values throughout the fluence-related beltline is based on two mathematical statistics principles as described in BAW-2241 P-A, Reference 6. The first is that unbiased measured data must be available throughout the reactor internal structures, and the cavity between the vessel and biological shield structure. Benchmarks of the calculated values to the data must be performed to identify any biases in the calculational methodology.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 12 If the calculated values can be shown to have no biases, then the root mean square of the deviations. in the benchmark comparison of calculated values to the data must be evaluated. Deviations which are truly random have no functional relationship to any part of the methodology. Therefore, the root mean square standard deviation determined with core data must be statistically consistent with the standard deviation determined 'with reactor internals' data and likewise with cavity data. The consistency between root mean square benchmark deviations from various locations within and outside of the reactor, and the standard deviations from benchmarks of other reactor systems ensures that the calculational methodology has the same standard deviation independent of location. Thus, all calculated values located within the fluence-related beltline, such as the core barrel flanges, are considered to have a standard deviation that is less than 7%.

For reactor vessel internals locations above or below the active height of the fuel, including the core supportshield bottom flange, fluence results are best-estimate due to the lack of surveillance capsules or cavity dosimetry data. As described above, this best-estimate value is considered appropriate for the particular application of confirming the order of magnitude for the fluence.

Part C The flux synthesis methods employed for the ANO-1 fluence calculation are based on the methods described in BAW-2241 P-A, Reference 6. The methods provide a reliable estimate for regions that are both (a) above the core's active fuel, as well as (b) regions that are far from the fuel-baffle - plate surface. The accuracy and precision of the methods was validated by a full-scale benchmark experiment performed at the Davis Besse reactor. The results of the benchmark verified that the methodology is unbiased and has a standard deviation less than 20%, as noted in the part B discussion above.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 13 The benchmark experiment results demonstrated that the fluence rate from neutrons leaving the fuel - baffle-plate surface [

As discussed further below, the geometrical modeling of the benchmark experiment is that used for the ANO-1 analysis supporting reactor vessel internals time-limited aging analysis (TLAA); the only updates reflect the advancement in technological and computational capabilities. [

The base methodology described in BAW-2241P-A was followed for the fluence calculations. However, improvements in the application of the methodology were also used for the ANO-1 reactor vessel internals fluence calculations. These include use of the BUGLE-96 cross section library, more detail in the geometrical and quadrature modeling of the reactor system, and more detail representing greater accuracy of the the neutron source.

Model changes unrelated to the ANO-1 reactor vessel internals fluence calculation have utilized the advances in computer storage and speed that have occurred since the release of BAW-2241P-A. There have been improvements to both the R-0 (re) model and the R-Z (rZ) model; with the changes more pronounced in the R-Z model. The detail and number of components in the system have increased as has the number and uniformity of the mesh intervals to incorporate the added detail in a single model.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 14 Figure 1 provides a sketch of the general arrangement of the reactor vessel internals.

This can be compared to the sketches of the re and rZ models shown in Figure 2 and Figure 3, respectively. These second two figures are updates of Figures 3-2 and 3-3 in BAW-22411P-A with the requested location discussed in ANP-3281P (Reference 9),

Revision 1, identified.

The primary difference in the re model, Figure 2, from BAW-2241 P-A Figure 3-2 is (1) the increase in the number of mesh intervals to represent the core configuration that is located within the core-baffle liner, and (2) additional detail in the cavity for evaluating the excore dosimeters.

Like the re model, the rZ model, represented by Figure 3, now encompasses the reactor system in a single model as opposed to the three sub-models referenced in BAW-2241P-A. In addition, the base model has been extended to span part of the inlet plenum below the lower grid-plate (at the elevation of the Dutchman weld) up to the vessel head flange mating surface. [

] These extended regions have been added to allow analyses of locations outside the traditional beltline region of the previous fluence calculations for extended life considerations. [

The locations discussed in ANP-3281 P (Reference 9), Revision 1, relative to neutron fluence include the core barrel flanges, high-fluence locations inside the beltline region with radial planes of the model as described above, and the core support shield bottom flange, high-stress location above the effective height of the active core. Their locations are illustrated in Figures 2 and 3. [

I

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 15

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for thA ANO-1 Reactor Vessel Internals PaOa 16 Figure 1 B&W PWR Internals General Arrangement Support Shield 4 E I 43 Barrel E

Be5 (Note: same coWonent item awe rtated or damift)

From cover of MRP-1 89, Rev. 1, Reference 4.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Paqe 17 Figure 2 Sketch of R-0 DORT Core Model (Not to Scale)

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Page 18 Figure 3 Sketch of R-Z DORT Model (Not to Scale)

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Paqe 19 REFERENCES

1. Entergy letter 1CAN051401, "Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years,"( ADAMS Accession # ML14126A816 for non-proprietary version), May 6, 2014.
2. U.S. NRC Request for Additional Information, "OFFICIAL USE ONLY -

SENSITIVE PROPRIETARY - ANO, Unit 1 - Time-Limited Aging-Analysis Regarding Reactor Vessel Internals - TAC NO. MF42013," (ADAMS Accession # ML14349A007 for non-proprietary version), December 12, 2014

3. U.S. NRC Safety Evaluation, "Oconee Nuclear Station, Units 1, 2, and 3 -

Approval of Time-Limited Aging Analysis for Reactor Vessel Internals,"

(ADAMS Accession #ML13045A489), February 19, 2013.

4. MaterialsReliability Program:Screening, Categorization,and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev. 1).

EPRI, Palo Alto, CA: 2009. 1018292. (ADAMS Accession Number ML092250189).

5. U.S. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001 (Reviewed May 2013).
6. Letter from James F. Malley to NRC Document Control Desk, "Submittal of BAW-2241 P, Revision 2, Fluence and Uncertainty Methodologies,"

June 2, 2003, (ADAMS Accession Number ML031550365. 4 ,

4 Revision 0 of BAW-22411P-A is for Babcock & Wilcox (B&W) reactor designs and includes the NRC Safety Evaluation Report (SER) applicable to ANO-1. Revisions 1 and 2 of BAW-2241 P-A also contain the associated SERs and increase applicability to include Boiling Water Reactors (BWRs) and Westinghouse or Combustion Engineering (CE) reactors, respectively.

AREVA Inc. ANP-3281Q1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals Paqe 20

a. BAW-2241 NP-A, Revision 2, "Fluence and Uncertainty Methodologies," April 30, 2006 (ADAMS Accession Number ML073310660).
7. B&WOG Topical Report 43-2248-00, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," BAW-2248A, Revision 0, April 2000 (ADAMS Accession Number ML003708443), includes NRC Safety Evaluation Report (ADAMS Accession Number ML993490303).
8. U.S. NRC NUREG/CR-7027, "Degradation of LWR Core Internal Materials Due to Neutron Irradiation," December 2010 (ADAMS Accession Number ML102790482).
9. AREVA Topical Report ANP-3281 P, Revision 1, "Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years," March 25, 2014.

to ICAN011506 Affidavit

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle Elliott. I am Manager, Product Licensing, for AREVA Inc.

(AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteda.
3. I am familiar with the AREVA information contained in document ANP-3281 Q1 P, Revision 0, titled "Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals," dated January 2015 and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. 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 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

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA'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.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA, would be helpful.to.competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d) and 6(e) above.

7. In accordance with AREVA's policies governing the protection and control of

.information, proprietary information contained in this Document has been made available, on a limited .basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this day of j0LN/v Vy*' ,2015.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129 SHERRY L.MOFADEN NOtWy PuMlPc CommronwIalt of Virginli 7079129 My Commission Expires Oct 31.2018

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