ML12311A029
| ML12311A029 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/01/2012 |
| From: | Price J Dominion, Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME7881, 12-578 | |
| Download: ML12311A029 (19) | |
Text
Dominion Nuclear Connecticut, Inc.
F 5000 Dominion Boulevard, Glen Allen, VA 23060 lDominioEn Web Address: www.dom.com Proprietary Information - Withhold Under 10 CFR 2.390 November 1,2012 U. S. Nuclear Regulatory Commission Serial No.12-578 Attention: Document Control Desk NSSLIWDC RO Washington, DC 20555 Docket No.
50-336 License No.
DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES (TAC NO. ME7881)
By letter dated January 25, 2012, Dominion Nuclear Connecticut, Inc. (DNC) submitted the Millstone Power Station Unit 2 (MPS2) 30-day Report for Emergency Core Cooling System Model Changes. The report documents changes to, or errors discovered in, an acceptable Small Break Loss of Coolant Accident (SBLOCA) evaluation model application for the emergency core cooling system for MPS2, in accordance with 10 CFR 50.46(a)(3)(ii). By letter dated August 28, 2012, the NRC transmitted a request for additional information (RAI) to DNC related to the report. DNC agreed to respond to the RAI by November 2, 2012. provides DNC's response to the NRC's RAI and contains proprietary AREVA information extracted from Reference 2.
The non-proprietary version of the response to the NRC's RAI is provided in Attachment 2. It is respectfully requested that be withheld from public disclosure in accordance with 10 CFR 2.390. The AREVA application for withholding and affidavit is provided in Attachment 3.
DNC understands that Supplement 1 of Reference 1, AREVA S-RELAP5 Based PWR SBLOCA Evaluation Model, is currently with the NRC for review and approval. Within one year of NRC approval of Supplement 1 of Reference 1, DNC plans to submit a reanalysis of the SBLOCA event using Supplement 1 of Reference 1.
If you have any questions regarding this submittal, please contact William D. Bartron at (860) 444-4301.
Sincerely, J.
Price Vice President - Nuclear Engineering Attachment I contains information that is being withheld from public
/o 0ob disclosure under 10 CFR 2.390. Upon separation from the attachment, this letter is decontrolled.
Serial No.12-578 Docket No. 50-336 Page 2 of 2
References:
- 1. AREVA Report EMF-2328(P)(A), Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 2001.
- 2. AREVA Report ANP-3171(P), Revision 0, "Millstone Unit 2 RAI Responses Regarding the 2011 10CFR50.46 Report," October 2012.
Attachments:
- 1. Response to Request for Additional Information Regarding 30-Day Report for Emergency Core Cooling System Model Changes (Proprietary)
- 2. Response to Request for Additional Information Regarding 30-Day Report for Emergency Core Cooling System Model Changes (Non-Proprietary)
- 3. AREVA application for withholding and affidavit Commitments made in this letter: None cc:
U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd.
Suite 100 King of Prussia, PA 19406-2713 J. S. Kim Project Manager - Millstone Power Station U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08 C2A Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station
Serial No.12-578 Docket No. 50-336 ATTACHMENT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES (NON-PROPRIETARY)
DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 1 of 12 In a letter dated January 25, 2012, Dominion Nuclear Connecticut, Inc. (DNC) submitted the Millstone Power Station Unit 2 (MPS2) 30-day Report for Emergency Core Cooling System Model Changes. The report documents changes to or errors discovered in an acceptable loss of coolant accident evaluation model application for the emergency core cooling system for MPS2, in accordance with 10 CFR 50.46(a)(3)(ii).
In a letter dated August 28, 2012, the NRC transmitted a request for additional information (RAI) to DNC related to the report. This attachment provides DNC's response to the NRC's RAI. The response to RAIs 3, 4, 5, 6, 8, 11, and 12 contain information extracted from a proprietary AREVA report.
The affidavit for withholding for the AREVA report is provided in Attachment 3.
RAI 1
The cover letter states:
"DNC [Dominion Nuclear Connecticut, Inc.] considers the requirement for reconsideration for reanalysis specified in 10 CFR [Title 10 of the Code of Federal Regulations] 50.46(a)(3)(ii) to be satisfied with the submission of this amendment."
10 CFR 50.46(a)(3)(ii) states: "...If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 50.46 requirements..."
DNC needs to provide a schedule for reanalysis or take other action to show compliance with Section 50.46. The peak clad temperature (PCT) for small break loss-of-coolant accident (SBLOCA) at Millstone has changed by an absolute value of 314°F since the analysis was performed in January, 2002. Simply reporting the changes and errors in the methodology does not satisfy the intent of the regulation.
DNC Response DNC employed the option in 10 CFR 50.46(a)(3)(ii) of "taking other action as may be needed to show compliance with 50.46 requirements...." Specifically, DNC had an analysis of the limiting 0.08 ft2 break performed that explicitly included all SBLOCA Evaluation Model (EM) PCT rackup items (errors and changes) in a single corrected case. This case produced the 1801°F SBLOCA PCT reported in DNC's January 25, 2012 letter. In addition, break sizes of 0.06, 0.07 and 0.10 ft2 around the limiting break size of 0.08 ft2 were analyzed incorporating PCT rackup changes and errors.
The results from the four break sizes are included in the response to RAI 8. The 0.08 ft2 break continues to be limiting.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 2 of 12 Because the SBLOCA event was analyzed incorporating all reported changes and errors, the AOR limiting break size of 0.08 ft2 was confirmed to be limiting, and 399°F margin to the PCT limit of 2200'F exists for the limiting case, compliance with 10 CFR 50.46 requirements was demonstrated and no further action was deemed necessary.
As a result of "taking other action" to show compliance with 10 CFR 50.46, DNC did not propose a reanalysis schedule.
RAI 2
Please provide the estimated effects of the S-RELAP5 kinetics and heat conduction model changes separately. By reporting the delta PCT of multiple errors bundled as one the staff can not completely understand the changes and therefore can not appropriately assess their safety significance.
DNC Response The error in the S-RELAP5 point kinetics model resulted in a +4'F increase in PCT.
The programming error in the heat conduction solution resulted in a 0°F change in PCT.
RAI 3
Please explain how the Sleicher-Rouse correlation in S-RELAP5 was different from the formulation of the correlation used in other industry codes.
DNC Response The Sleicher-Rouse heat transfer correlation is used in S-RELAP5 for predicting convective heat transfer to single-phase vapor.
The concern is related to the form of the equation for calculating the exponent of the temperature ratio correction term. The Sleicher-Rouse correlation is derived from Reference 2 equation number (12).
The S-RELAP5 form is Nub = 5 + 0.012*Rebs 8 3 * (Prb +0.29)*(Tw/Tb)n with (Form A) n = - Iogjo(T,'Tb)114 + 0.3 The alternative form used in other industry codes is
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 3 of 12 Nub = 5 + 0.012*RebD° 83 * (Prb +0.29)*(Tw/Tb)n with (Form B) n = -[1ogjo(TvT/Fb)]1/4 + 0.3 The alternative form is more consistent with other heat transfer correlations, like correlation (6-168) of Reference 3, and expected physical trends.
RAI 4
Please provide the 'alternative formulation' of the Sleicher-Rouse correlation that is now used and show the 'closer agreement' to other industry codes.
DNC Response The alternative form used in other industry codes is Nub = 5 + 0.012*Reb0 83 * (Prb +0.29)*(Tw/Tb)n with n = - [Ioglo(Tv-Tb)'114 + 0.3 Equation (8) in Reference 2:
(Form B)
Nub =Nuo*(Tw/Tb)n with n = - a*/og1o(T/-/b) - 0.36 Equation (12) in Reference 2:
Nub = 5 + 0.012*Reb° 83 * (Prb +0.29)*(Tw/Tb)n with (Form B) n = - [1oglo(Tv--b)]14 + 0.3
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 4 of 12 Figure 4-1 shows the comparison between the Form A, Form B, and the Petukhov formulation and it demonstrates that Form B is the correct form to be used to implement the temperature degradation factor (n).
Figure 4-1: Heat Transfer Enhancement Factor Trend
RAI 5
Please provide the initial conditions and results of the sample problems that were performed to predict a PCT increase of 830F for SBLOCA at MPS2 as a result of the Sleicher-Rouse modification.
DNC Response To assess the effect of the alternate form of the Sleicher-Rouse correlation, a new version of S-RELAP5 with the corrected formulation was prepared. The current AOR
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 5 of 12 for MPS2 SBLOCA presented in final safety analysis report (FSAR) Section 14.6.5.2 was evaluated using this new version.
The evaluation is performed specifically for MPS2 and the impact was not evaluated with a sample problem. The original initial conditions for the AOR provided in Table 14.6.5.2-3 of the MPS2 FSAR were maintained in the evaluation of the Sleicher-Rouse correction. The 0.06, 0.07, 0.08 and 0.1 ft2 were analyzed with the new version of the code containing the correction for the Sleicher-Rouse correlation.
The new code version contains the correction of the Sleicher-Rouse correlation in addition to the other changes reported in DNC's January 25, 2012 30-day report letter. The PCT value of 1801'F for the 0.08 ft2 break size represents the impact of the changes reported in that letter, including the Sleicher-Rouse correlation correction.
RAI 6
Please explain why 'RODEX2 Thermal Conductivity Degradation' is listed in the SBLOCA PCT rack-up in Attachment 2 but is not discussed in Attachment 1. Also, please provide justification to support the estimation of its effect on PCT to be O°F.
DNC Response The RODEX2 thermal conductivity degradation PCT assessment of 0°F was included in the Attachment 2 PCT rack-up of DNC's January 25, 2012 letter. Since the assessment determined that thermal conductivity degradation did not have any impact on the calculated PCT of the SBLOCA, no discussion of that impact was included in of the January 25, 2012 letter.
The RODEX2 code is used to determine the initial core and hot pin stored energy for SBLOCA evaluations.
Small breaks evolve through a pump coastdown and natural circulation phase to a loop draining phase followed by a boil-down and refill phase. The pump coastdown phase lasts approximately 100 seconds. For most of this phase, a single or two-phase forced circulation exists within the RCS which prevents a cladding temperature excursion and acts to remove the initial energy of the fuel and deposits it in the steam generators or the containment. In either case, the energy content of the fuel has been reduced to that required to transport decay heat out of the fuel by the end of the pump coastdown phase. Thus, the peak cladding temperatures, which occur later in the transient, depend on decay heat versus heat transfer and are not coupled to the initial stored energy within the fuel.
Consequently, adjustments to the initial fuel temperature, to address the thermal conductivity degradation, have no significant effect on the SBLOCA cladding temperatures or the local oxidation. For this reason, the RODEX2 thermal conductivity degradation was classified as a 00F impact on SBLOCA PCT.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 6 of 12
RAI 7
Please specify whether any of the errors listed in the table in Attachment 2 are bundles of more than one error or change and, if so, provide the values of the individual delta PCTs.
DNC Response Other than the RELAP5 kinetics and heat conduction model error discussed in response to RAI 2 above, none of the other changes and errors provided in Attachment 2 to DNC's January 25, 2012 letter are bundles of more than one error or change.
RAI 8
The Millstone Unit 2 SBLOCA spectrum consisted of breaks with the following PCTs Break Size, ft**2 PCT OF 0.05 1616 0.06 1782 0.08 1941 0.10 1808 0.12 1701 0.15 1525 Since the initial analysis using this break spectrum was performed, staff review activities have indicated that a more refined break spectrum is necessary to provide assurance that the limiting break sizes, locations, and other properties have been analyzed, and that the most severe postulated LOCA has been calculated. Changes in the break area by as little as 0.005 ft**2 can produce increases in PCT from 100 - 200 OF. Typically, the limiting small break is that maximum break size that just precludes SIT actuation and suggests a break between 0.06 and 0.08 ft**2 could be more limiting since SIT actuation occurs at the time of the PCT for the 0.08 ft**2 break. Also, there is the potential for the worst break to be between 0.08 and 0.1 ft**2. The possibility for more limiting break sizes to be identified when a coarse spectrum is used is documented, for example, see Section 5.3 of ADAMS Accession MLl 10390263. The treatment of this issue as an error in the application of an ECCS evaluation model, and its associated safety significance for plants who's PCTs exceed 1700 degrees Fahrenheit, is also documented in ML062000027. As such, the staff requests that additional breaks be analyzed to produce an estimate of the effect of a potential error in the break spectrum analysis, or to show that the limiting break has been identified in the spectrum analysis.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 7 of 12 The following information is quoted from the Federal Register, Volume 53, Page 35996, concerning the significance of ECCS evaluation model error reports.
[Paragraph (a)(3) of section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected. The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation model...
Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model... More timely reporting is required for significant errors or changes...the final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements.
Based on the above, the NRC staff has determined that it is necessary to perform confirmatory calculations to ensure that the MPS2 ECCS evaluation remains adequate with regard to demonstrating compliance with the 10 CFR 50.46(b) acceptance criteria.
Please also provide the following data:
- a. limiting top peaked axial power distribution. Please provide a table of peaking factor vs. axial elevation.
- b. a table of minimum HPSI flow into the four loops vs. pressure.
- c. a curve of reactivity vs. moderator density for the most positive MTC.
DNC Response While performing the evaluation for the Sleicher-Rouse formulation correction, DNC performed a calculation to evaluate the PCT for an additional break size (0.07 ft2) that was not in the AOR break spectrum.
The break sizes currently analyzed for the Sleicher-Rouse are listed below with the SIT (Safety Injection Tank) Injection time compared to the time PCT occurs:
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 8 of 12 SBLOCA Break Size Evaluation Break Size Analysis of Record PCT (°F) with SIT Injection (ft)
Sleicher Rouse Timing (ft2)
(Correction 0.06 17821 16192 No Injection 0.07 17392 After PCT 0.08 19411 18012
-PCT 0.10 18081 16962 Before PCT (1) For the AOR the number of the loop seals that clear for each break size, is indicated in the FSAR Table 14.6.5.2-5. For break sizes of 0.06 and 0.08 ft2 two loop seals clear. For the 0.10 ft2 break all four loop seals clear.
(2) For Sleicher-Rouse correction, two loop seals clear in the evaluation of 0.06, 0.07, 0.08 ft2 break sizes. For the 0.10 ft2 break size all four loop seals clear. This isconsistent with the results of the AOR.
The PCT reported above is the temperature obtained with the Sleicher-Rouse evaluation and the errors and changes that are listed in the SBLOCA PCT margin utilization table provided in Attachment 2 of DNC's January 25, 2012 letter. The table above shows the trend of the SIT Injection time versus the time PCT occurs. For the smaller break sizes, the SIT does not inject during the transient. For the 0.07ft2 break size, the PCT occurs before the SIT injection, while for the 0.08ft2, identified as the limiting break size, the PCT and SIT injection occur approximately at the same time.
For larger break sizes, the SIT injection occurs before the PCT is reached and the increase in the PCT is terminated by the safety injection. This trend shows that for the 0.08 ft2 break size, the increase in the PCT is terminated because of the SIT.injection, unlike the smaller break sizes for which the PCT reduction is independent of the SIT injection. The trending of this break spectrum shows that PCTs are increasing as the break size increases to 0.08 ft2 and as the break size decreases to 0.08 ft2. Further, the rate of increase in PCT is about the same regardless of the path. This is the typical behavior of a SBLOCA transient and indicates that the limiting break size will be very close to a break size of 0.08 ft2.
In addition, variance of the temperature will be small such that the analysis performed is more than sufficient to demonstrate that the cladding temperature and oxidation are kept well below the 10 CFR 50.46 acceptance criteria.
The break sizes investigated, with the 0.01 ft2 increment, are sufficient to analyze the impact of the key mitigating systems (SIT Injection and high pressure safety injection) for the SBLOCA event.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 9 of 12 8a.
The table peaking factor vs. axial elevation is provided below:
The table below represents the limiting EOC axial power profile from the core design results; the shape is further modified for the analysis by adjusting the axial shape to the Fq/FAh limit (the values are reported in the third column of the following table). This approach is consistent with the approved methodology of Reference 1.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 10 of 12 8b.
The minimum High Pressure Safety Injection (HPSI) flows vs. pressure used in the analysis of record is provided in MPS2 FSAR Table 14.6.5.2-3 and is presented below:
RCS Pressure HPSI Injection (psia) per Loop (gpm) 200 136 300 128 500 112 700 94 900 73 1000 60 1050 53 1100 44 1150 31 1190 15 1204 0
8c.
I This approach is consistent with the Reference 1 approved methodology at the time of the analysis.
RAI 9
Also, was charging flow credited in the analysis? If so, please provide the mninimum charging flow for one pump operating injecting into the RCS.
DNC Response As stated in MPS2 FSAR Section 14.6.5.2.5.3 "Plant Description and Summary of Analysis Parameters," no charging pump flow was credited in the SBLOCA analysis.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 11 of 12
RAI 10
Please also provide an analysis of severed injection line.
DNC Response DNC has not previously performed an analysis of a severed ECCS injection line for MPS2.
This break size is at the high end of the SBLOCA spectrum and has been qualitatively judged to be a non-limiting break size in the past.
This historic qualitative judgment has been quantitatively demonstrated for the Calvert Cliffs Nuclear Power Plant (CCNPP) as described in Section 5.3.2.2 of the Safety Evaluation Report included in Reference 4. CCNPP is a plant of similar CE-NSSS design to the MPS2 plant. As indicated in Section 5.3.2.2 of the Safety Evaluation Report included in Reference 4, the CCNPP analysis of the severed injection line included the bounding assumption of crediting only safety injection tank flow.
The analysis did not credit HPSI pump flow. With this conservative assumption, the results of the severed injection line were bounded by the limiting SBLOCA break size.
RAI 11
Please provide the SIT and RWST temperatures used in the SBLOCA analyses.
DNC Response In the SBLOCA Analysis of Record, the SIT is initialized to the pressure of 214.7 psia, which is the minimum Tech Spec value, and the temperature of 106.8°F from representative plant data. The Refueling Water Storage Tank (RWST) temperature is set to 140 0F, which bounds the maximum administrative limit for the RWST temperature. The use of the nominal SIT temperature based on representative plant data is in accordance with the SBLOCA methodology of Reference 1 and analysis guidelines.
Serial No.12-578 Docket No. 50-336 RAI Response to 30-Day ECCS Model Changes Report, Page 12 of 12
RAI 12
Please discuss whether the hot leg nozzle gaps or core barrel leakage paths were modeling in the RELAP5 nodalization.
DNC Response t
]
This approach is consistent with the approved methodology of Reference 1.
REFERENCES:
- 1.
PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, EMF-2328(P)(A)
Revision 0, March 2001.
- 2.
C. A. Sleicher and M. W. Rouse, "A Convenient Correlation for Heat Transfer to Constant and Variable Property Fluids in Turbulent Pipe Flow," International Journal of Heat and Mass Transfer, 18, pp. 677-683, 1975.
- 3.
USNRC Document, "TRACE V5.0 Theory Manual - Field Equations, Solution Methods, and Physical Models", ADAMS Accession Number ML120060218.
- 4.
USNRC Letter Dated February 18, 2011, "Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Amendment RE: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (TAC Nos. ME2831 and ME2832)," ADAMS Accession Number ML110390224.
- 5.
Millstone Power Station Unit 2 Final Safety Analysis Report, Revision 30, September 2012.
Serial No.12-578 Docket No. 50-336 ATTACHMENT 3 AREVA APPLICATION FOR WITHHOLDING AND AFFIDAVIT DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2
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-3171(P), Revision 0, entitled "Millstone Unit 2 RAI Responses Regarding the 20111 10 CFR 50.46 Report," dated October 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
requested qualifies under 10 CFR 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.
- 9.
The foregoing statements are true and correct to the best of my knowledge, information, and belief.
SUBSCRIB me his day of 2012.
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