NRC-13-0057, Response to Request for Additional Information Regarding the Proposed License Amendment to Relocate the Pressure and Temperature Curves to a Pressure and Temperature Limits Report

Response to Request for Additional Information Regarding the Proposed License Amendment to Relocate the Pressure and Temperature Curves to a Pressure and Temperature Limits Report

ML13291A363
Person / Time
Site:	Fermi
Issue date:	10/17/2013
From:	Conner J DTE Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML13291A389	List: NRC-13-0057, Response to Request for Additional Information Regarding the Proposed License Amendment to Relocate the Pressure and Temperature Curves to a Pressure and Temperature Limits Report
References
NRC-13-0057, TAC MF0446
Download: ML13291A363 (29)

Text

J Todd Coner Sie Vice President DTE Energy Company 6400 N. Dbde Highway, -Newport, MI 48166 Tel: 734.5864849 Fax: 734.5865295 Email: connerj@cteenergycoin DTE Energy' Proprietary Information - Withhold Under 10 CFR 2.390 10 CFR 50.90 October 17, 2013 NRC-13-0057 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) DTE Electric Company Letter to NRC, "Proposed License Amendment to Relocate Pressure and Temperature Curves to a Pressure Temperature Limits Report," NRC-12-0037, dated December 21, 2012 (ADAMS Accession No. ML13004A134)

Subject:

Response to Request for Additional Information regarding the Proposed License Amendment to Relocate the Pressure and Temperature Curves to a Pressure and Temperature Limits Report In Reference 2, DTE Electric Company (DTE) submitted a license amendment request for Fermi 2 to relocate the Pressure and Temperature (P/T) Curves from the Technical Specifications to a Pressure and Temperature Limits Report (PTLR). In an email dated September 10, 2013, from Mr. Mahesh Chawla of the NRC to Mr. Alan Hassoun of DTE (TAC No. MF0446, ADAMS Accession No. ML13260A372), the NRC staff requested additional information to complete the review. Enclosure 1 of this letter provides DTE's response to the NRC staff request.

DTE requests withholding Enclosure 1 from public disclosure in accordance with 10 CFR 2.390. Enclosure 1 contains information that is considered proprietary by General Electric - Hitachi (GEH). A non-proprietary version of Enclosure 1 is contains Proprietary Information - Withhold Under 10 CFR 2.390.

When separated from Enclosure 1, this document is decontrolled.

USNRC NRC-13-0057 Page 2 provided in Enclosure 2 and an affidavit supporting the request for withholding from public disclosure is provided in Enclosure 3. Enclosure 4 provides a copy of the GEH affidavit originally provided in Enclosure 6 of Reference 2. The affidavit in Enclosure 4 was not separated from other proprietary documents when previously transmitted in Reference 2.

No new commitments are being made in this submittal.

Should you have any questions or require additional information, please contact Mr.

Zackary W. Rad of my staff at (734) 586-5076.

Sincerely, : to GEH Letter 318327-4, "GEH Responses to PT Curve RAIs 1-5"- PROPRIETARY : to GEH Letter 318327-4, "GEH Responses to PT Curve RAIs 1-5" - NONPROPRIETARY :

GE-Hitachi Nuclear Energy Americas LLC Affidavit for of 318327-4 :

GEH affidavit for Enclosure 1 of GEH letter, 1-2RXPKK-13, "GEH Responses to GGNS PT Curve RAIs" cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission

USNRC NRC-13-0057 Page 3 I, Mathew S. Caragher, do hereby affirm that the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Mathew S. Caragher Director, Nuclear Engineering On this 17* day of October, 2013 before me personally appeared Mathew S.

Caragher, being first duly sworn and says that he executed the foregoing as his free act and deed.

Notary Public SHARON S. MARSHALL NOTARY PUBLIC, STATE OF MI COUNTY OF MONROE MY COMMISSION EXPIRES Jun 14, 2019 ACTING IN COUNTY OF f-( c to NRC-13-0057 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Response to Request for Additional Information - NONPROPRIETARY to GEH Letter 318327-4, "GEH Responses to PT Curve RAIs 1-5" - NONPROPRIETARY

318327-4 GEH Responses to PT Curve RAIs 1-5 Non-Proprietary Information-Class I (Public)

Non-Proprietary Notice This is a non-proprietary version of Enclosure 1 of 318327-4 which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( )).

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 2 of 16 NRC RAI 1 The final two columns of Tables B-5 and B-6 of Enclosure 5 to the submittal refer to shift and ART values. The values contained in these columns represent the 1/4T shift and ART values.

This is not consistent with the column titles. The staff request that the column titles be revised to correctly identify the information contained below them.

Response

The headings of the columns in Tables B-5 and B-6 are based on the format in the NRC-approved GEH PT Curve LTR (References 1 and 2) and similar to previously approved PTLR License Amendment Requests. The format of the tables indicates that the values contained in these columns correspond to the 1/4T fluence values presented for each entry in the table.

References

1.

GEH Nuclear Energy, NEDC-33178P-A, Revision 1, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, Report for BWR Owners Group, Sunol, California, June 2009 (GEH Proprietary).

2.

Safety Evaluation Report, Thomas B. Blount (NRC) to Doug Coleman (BWROG), Final Safety Evaluation for Boiling Water Reactors Owners Group Licensing Topical Report NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (TAC No. MD2693), April 27, 2009.

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 3 of 16 NRC RAI 2 Provide a basis for the difference in the margin terms between the Tables B-5 and B-6 of and why this is acceptable.

Response

The basis for the difference in the margin terms in Tables B-5 and B-6 is due, in part, to the effective fluence associated with 24 and 32 EFPY. For many of the Fermi 2 materials, the margin term is dependent on the RTNDT. This is consistent with Position 1.1 of Regulatory Guide 1.99, Revision 2, which provides the methodology for determining adjusted reference temperature (ART). The final paragraph of this section states that (standard deviation for RTNDT) is 17°F for plates and 28°F for welds, but that need not exceed 0.5*RTNDT. The Fermi 2 ART calculation has incorporated the use of 0.5*RTNDT for all materials, where applicable.

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 4 of 16 NRC RAI 3 Clarify precisely how the N16 water level nozzle was accounted for in the PTLR including relevant inputs and calculations.

Response

A separate PT curve was developed for the N16 water level instrumentation (WLI) nozzle as documented in Appendix J of the Reference 1. ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) were used to determine KI for calculating the WLI PT curve. Sample calculations and relevant inputs for the WLI nozzle for Curve A and Curve B are provided in the response to RAI 4, along with sample calculations and relevant inputs for the limiting beltline material other than the WLI nozzle.

Reference

1.

GEH Nuclear Energy, NEDC-33178P-A, Revision 1, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, Report for BWR Owners Group, Sunol, California, June 2009 (GEH Proprietary).

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 5 of 16 NRC RAI 4 Clarify precisely how the beltline limiting portions of the curve were generated including relevant inputs and calculations.

Response

The following response applies to RAIs 3 and 4. The Fermi 2 WLI nozzles that are located in the beltline region are of a partial penetration design, similar to that shown in Figure 1 of Appendix J of Reference 1. The methodology of Appendix J was used to calculate the PT curves to represent the WLI component. The following sample calculations and relevant inputs represent 32 EFPY. Please note that the results presented herein may differ slightly from those in the PTLR; this is due to hand-calculation versus Excel rounding.

Beltline (excluding WLI nozzle) Pressure Test (Curve A) at 1050 psig for 32 EFPY The limiting beltline material is the bounding component for Curve A; therefore, a sample calculation for this material, not including the WLI nozzle is provided for 1050 psig.

The limiting ART applied to the beltline PT curves is 82°F for the lower shell axial weld, which is also in the Integrated Surveillance Program (heat CE-2 (WM)(13253,12008)). This ART conservatively uses a shift of 121°F (in lieu of the 120°F shift shown in Table B-6 of the PTLR);

this is identified in Table B-7 of the PTLR. In addition, an adjustment of 5°F was included to assure protection of the existing flaw in the axial weld of Shell Ring 2 as detailed in Section 3.5 of the PTLR; this is applicable to the Pressure Test curve only (Curve A), and not Core Not Critical (Curve B).

Pressure is calculated to include hydrostatic pressure for a full vessel:

P

=

1050 psig + (H - B)*0.0361 psi/inch (H=vessel height; B=elevation of bottom of active fuel)

=

1050 + (861.6 - 216.3)

• 0.0361

=

1073.3 psig Pressure Stress:

=

PR/t (P=pressure; R=vessel radius; t=vessel thickness)

=

1.073

• 127 / 6.125

=

22.25 ksi Mm

=

0.926t

=

0.9266.125

=

2.29 to 318327-4 Non-Proprietary Information-Class I (Public)

Page 6 of 16 The stress intensity factor, KIt, is calculated as described in Section 4.3.2.2.4 of Reference 1, except that G is 20°F/hr instead of 100°F/hr.

Mt

=

0.2942, from ASME Appendix G, Figure G-2214-2 T

=

GC2 / 2 G = coolant heatup/cooldown rate of 20°F/hr C = minimum vessel thickness including clad = 6.125+0.3125=6.4375=0.5365 ft

= thermal diffusivity at 550°F = 0.354 ft2/hr

=

(20*(0.5365)2) / (2*0.354)

=

8.13°F KIt

=

Mt

• T

=

0.2942

• 8.13

=

2.39 KIm

=

• Mm

=

22.25

• 2.29

=

50.96 T-RTNDT

=

ln[(1.5*KIm + KIt - 33.2)/20.734]/0.02

=

ln[(1.5*50.96 + 2.39 - 33.2)/20.734]/0.02

=

39.44°F T is calculated by adding the ART:

T

=

39.44 + 82

=

121.4°F for P = 1050 psig at 32 EFPY This temperature represents the limiting point on Curve A and is cited as 121.5°F in Table 2 of the PTLR.

Beltline (excluding WLI nozzle) Core Not Critical (Curve B) at 1250 psig for 32 EFPY The WLI nozzle is the bounding component between 580 psig and 1220 psig; therefore, a sample calculation for the limiting beltline material, not including the WLI nozzle, is provided for 1250 psig in order to present a point visible on the PT curve.

As discussed above and in Section 3.5 and Table B-7 of the PTLR the limiting ART applied to the beltline Curve B is 77°F for the lower shell axial weld, which is also in the Integrated Surveillance Program (heat CE-2 (WM)(13253,12008)).

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 7 of 16 The T term is calculated as shown above for the Pressure Test case, but the temperature rate change is 100°F/hr instead of 20°F/hr. Therefore, T equals 40.65°F.

P

=

1250 psig + (H - B)*0.0361 psi/inch (H=vessel height; B=elevation of bottom of active fuel)

=

1250 + (861.6 - 216.3)

• 0.0361

=

1273.3 psig Pressure Stress:

=

PR/t (P=pressure; R=vessel radius; t=vessel thickness)

=

1.273

• 127 / 6.125

=

26.4 ksi KIm

=

• Mm

=

26.4

• 2.29

=

60.46 KIt

=

Mt

• T (for the 100°F/hr case)

=

0.2942

• 40.65

=

11.96 T-RTNDT

=

ln[(2.0*KIm + KIt - 33.2)/20.734]/0.02

=

ln[(2.0*60.46 + 11.96 - 33.2)/20.734]/0.02

=

78.51°F T is calculated by adding the ART:

T

=

78.51 + 77

=

155.5°F for P = 1250 psig at 32 EFPY This temperature represents the limiting point on Curve B and is cited in Table 2 as 155.6°F.

Pressure Test (Curve A) at 1050 psig KI for the discontinuity is determined considering the KI obtained from Table 7 of Appendix J for hydrotest. For 1050 psig, this KI is ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) as follows:

KI = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

T-RTNDT is calculated in the following manner:

T-RTNDT = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

T-RTNDT = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The ART is added to T-RTNDT to obtain the required T:

T = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

for P = 1050 psig at 32 EFPY This temperature is not obvious from the PT curve as the WLI curve for the entire pressure range from 0 to 1400 psig is bounded by the limiting beltline weld material requirement. The limiting beltline weld temperature of 121.5°F at 1050 psig is shown in Table 2.

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 8 of 16 Core Not Critical (Curve B) for the WLI Nozzle at 1050 psig The WLI nozzle is the bounding component between 580 psig and 1220 psig; therefore, a sample calculation is provided for 1050 psig. As discussed above and in Table B-6 of the PTLR, the limiting ART applied for the WLI is 64°F.

KI for the discontinuity is determined considering the KI obtained from Table 5 of Appendix J.

((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The transient used for the WLI nozzle, defined in Appendix J, is used in determination of KI.

The total stress for the WLI exceeds the yield stress; therefore, the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according the following equation based on the assumptions and recommendation of WRC Bulletin 175 (Reference 2) as shown in Equation 4-7 of Reference 1.

R = [ys - pm + ((total - ys)/30)]/(total - pm)

Applied to the WLI:

R = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The total KI is therefore:

KI = [Safety Factor(pressure)

• KI(pressure)] + [KI(thermal) * (Transient T)

• R]

KI = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

T-RTNDT is calculated in the following manner:

T-RTNDT = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

The ART is added to T-RTNDT to obtain the required T:

T = ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ))

for P = 1050 psig at 32 EFPY This temperature represents the limiting point on Curve B and is cited in Table 2 as 149.3°F.

References

1.

GEH Nuclear Energy, NEDC-33178P-A, Revision 1, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, Report for BWR Owners Group, Sunol, California, June 2009 (GEH Proprietary).

2.

Welding Research Council (WRC) Bulletin 175, PVRC Recommendations on Toughness Requirements for Ferritic Materials, August 1972.

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 9 of 16 NRC RAI 5 10 CFR Part 50, Appendix G, Paragraph IV.A states that, the pressure-retaining components of the reactor coolant pressure boundary [RCPB] that are made of ferritic materials must meet the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code

[ASME Code],Section III, supplemented by the additional requirements set forth in[paragraph IV.A.2, Pressure-Temperature (P-T) Limits and Minimum Temperature Requirements"]

Therefore, 10 CFR Part 50, Appendix G requires that P-T limits be developed for the ferritic materials in the reactor vessel (RV) beltline (neutron fluence 1 x 1017 n/cm2, E > 1 MeV), as well as ferritic materials not in the RV beltline (neutron fluence < 1 x 1017 n/cm2, E > 1 MeV).

Further, 10 CFR Part 50, Appendix G requires that all RCPB components must meet the ASME Code,Section III requirements. The relevant ASME Code,Section III requirement that will affect the P-T limits is the lowest service temperature requirement for all RCPB components specified in Section III, NB-2332(b).

The P-T limit calculations for ferritic RCPB components that are not RV beltline shell materials may define P-T curves that are more limiting than those calculated for the RV beltline shell materials due to the following:

• RV nozzles, penetrations, and other discontinuities have complex geometries that may exhibit significantly higher stresses than those for the RV beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperature (RTNDT) for these components is not as high as that of RV beltline shell materials that have simpler geometries.

• Ferritic RCPB components that are not part of the RV may have initial RTNDT values, which may define a more restrictive lowest operating temperature in the P -T limits than those for the RV beltline shell materials.

Consequently, please describe how the P-T limit curves submitted for Fermi 2 and the methodology used to develop these curves, considered all RV materials (beltline and nonbeltline) and the lowest service temperature of all ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G.

Response

The methods described in Topical Report NEDC-33178P-A, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (Reference 1) were used to develop the pressure-temperature (PT) limits for Fermi 2. The Reference 1 report describes methods for compliance with the requirements of 10 CFR 50 Appendix G (Reference 2), ASME Code Section XI, Appendix G (Reference 3), and Welding Research Council (WRC) Bulletin 175 (Reference 4). The NRC has reviewed the methods described in Reference 1 and approved the report by issuance of a Safety Evaluation Report (SER) dated April 27, 2009 (Reference 5).

The Fermi 2 P-T limits were generated consistent with the requirements of 10CFR50 Appendix G and Regulatory Guide 1.99 (Reference 6) as defined in Reference 1. The reference to 318327-4 Non-Proprietary Information-Class I (Public)

Page 10 of 16 temperature for nil-ductility transition (RTNDT) values of the reactor vessel beltline and non-beltline plates, welds, forgings, and bolting were determined using the NRC-approved GE/BWROG methodology defined in NEDC-32399-P (Reference 7).

Reference 1 is applicable to Fermi 2 and much of the BWR fleet. Tables 4-4 and 4-5 of Reference 1 provide a list of all non-beltline vessel RCPB components included in the P-T limit evaluation. In accordance with the definition of materials exposed to sufficient neutron radiation damage (Reference 2), the topical report identifies all materials in the vessel beltline region that experience end of license fluence greater than or equal to 1.0e17 n/cm2. This typically extends the beltline region considered to some distance both above and below active fuel. Appendix E of Reference 1 demonstrates the method used to identify these materials, and the Adjusted Reference Temperature (ART) table in Section 4 of Reference 1 includes the evaluation of all materials exposed to fluence greater than or equal to 1.0e17 n/cm2.

The GEH topical report (Reference 1) describes the methodology employed to develop bounding PT limit curves for three (3) regions of the vessel, shown below. Two (2) PT curves were developed to represent all vessel non-beltline discontinuities; beltline discontinuities are evaluated in accordance with ASME Code requirements.

• The non-beltline upper vessel curve is based upon the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), the

((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), and its associated transients. The ((° ° ° ° ° ° ° ° ° ° ° )) transient for the upper vessel is defined in Figure 4-3 of the topical report. The ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) in the upper vessel curve (see Table 4-4 of Reference 1). The upper vessel curve is

((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) of Reference 1; therefore, it is assured that the upper vessel curve bounds the requirements for each of these components.

• The non-beltline bottom head curve is based upon the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° )). All components identified in Table 4-5 of Reference 1 are included in the evaluation. The ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) of Reference 1; therefore, it is assured that the bottom head curve bounds the requirements for each of these components. ((° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), the ((° ° ° ° ° ° ° ° ° ° ° )) transients applied for the bottom head curve are ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) conditions as defined in Figure 4-2 of Reference 1.

• The beltline curve considers all vessel materials adjacent to the reactor core plus all materials above and below this region that will be exposed to fluence greater than or equal to 1.0e17 n/cm2 at end of license.

Each non-beltline discontinuity was evaluated and is considered to be bounded by either or both the upper vessel and bottom head curves. T-RTNDT was determined for each discontinuity using finite element analysis (FEA) methods, conservatively based upon ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) is accepted by the ASME Code. Specific examples showing the method of assuring that the most limiting discontinuity is considered in the development of each curve are to 318327-4 Non-Proprietary Information-Class I (Public)

Page 11 of 16 provided in Section 4.3.2 of Reference 1. The ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )), due to the consideration of loading, transients, and the dimensions of the vessel and nozzles, are more limiting relative to stress than any of the Class 1 ferritic branch piping in the RCPB. None of the RCPB large bore piping has been modified, repaired, or replaced since the start of commercial operation, and the wall thickness of all RCPB piping and welds is less than two inches. Small bore modifications (e.g., socket welded connections) were performed following ASME Section XI and Section III requirements, using materials consistent with original installation.

The GEH methodology for P-T curve development (Reference 1) contains a number of

((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) considerations. Detailed stress analyses of ((° ° ° ° ° ° ° ° )) non-beltline components were performed specifically for the purpose of fracture toughness. The ((° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) designs as defined in Section 4.3.2 of Reference 1. The analyses considered mechanical loading and anticipated thermal transients that ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° )) In addition, ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) in this evaluation. The results of these transients experienced by and evaluated for the PT curve evaluation are more severe for the vessel and nozzles than those anticipated for the attached RCPB piping and equipment.

ASME Code Section XI, Appendix G, Article G-3000, paragraph G-3100 states that for materials used for piping, pumps, and valves for which impact tests are required (NB-2311), the tests and acceptance standards of Section III, Division 1 are considered to be adequate to prevent non-ductile failure under the loadings and with the defect sizes encountered under normal, upset, and testing conditions. Level C and Level D Service Limits should be evaluated on an individual case basis (G-2300). As described in Section 4.3 of Reference 1, ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° )) in the development of all non-beltline PT limits.

The following excerpt from the Fermi 2 UFSAR (Reference 8), Section 5.2.4.2.1, provides information regarding Fermi 2 compliance with 10 CFR 50 Appendix G.

Versions of 10 CFR 50, Appendix G, prior to the 1983 Edition had specific requirements for the preparation and testing of all reactor coolant pressure boundary materials. In lieu of these specific requirements, the present version of Appendix G requires that for a reactor vessel which was constructed in conformance with an ASME Code Section III earlier than the Summer 1972 Addenda of the 1971 Edition, the fracture toughness data and data analyses must be supplemented in a manner approved by the Director, Office of Nuclear Reactor Regulation, to demonstrate equivalence with the present fracture requirements of Appendix G. The Fermi 2 reactor vessel was constructed in compliance with an ASME Code earlier than the summer 1972 Addenda of the 1971 Edition. The NRC has stated in Supplement 1 to NUREG 0798, the Fermi 2 Safety Evaluation Report, to 318327-4 Non-Proprietary Information-Class I (Public)

Page 12 of 16 that the alternative methods proposed by Fermi 2 to demonstrate compliance with Appendix G has been reviewed, evaluated, and found to provide the safety margin required by Appendix G. Accordingly, Fermi 2 has supplied sufficient information to demonstrate equivalency with the fracture toughness requirements of the present version of 10 CFR 50, Appendix G, (1983 as amended November 1986 and October 1988).

With respect to the concern regarding irradiation effects on RCPB piping, a qualitative fluence assessment was performed. With a 32 EFPY peak surface ID fluence of 9.68e17 n/cm2 as the maximum fluence of concern, accrual of fluence greater than 1.0e17 n/cm2 outside the vessel for 32 EFPY is not expected, based on historical calculations of flux vs. vessel thickness.

The General Design Criteria (GDC) in effect at the time that Fermi 2 was fabricated are discussed in Section 3.1 of the Fermi 2 UFSAR (Reference 8). Fermi 2 complies with the intent of the GDC for Nuclear Power Plants, 10 CFR 50, Appendix A with respect to fracture toughness and the RCPB. The NRC requirement regarding the GDC is that the plant application provide assurance that its principal design criteria encompass all those facility design features required in the interest of public health and safety. Criteria 14, 30, 31, and 32 specify requirements with respect to the RCPB. These sections of the Fermi 2 UFSAR are provided for clarification.

UFSAR Section 3.1.2.2.5 Criterion 14, Reactor Coolant Pressure Boundary, states:

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Criterion 14 conformance is demonstrated as follows:

The piping and equipment pressure parts, which extend through the outer isolation valve(s) but which are within the RCPB, are designed, fabricated, erected, and tested to provide a high degree of integrity throughout the plant lifetime. Systems and components within the RCPB are classified in Section 3.2 as Code Group A. The design requirements, codes, and standards applied to this Code Group ensure a quality product in keeping with the safety functions to be performed.

To minimize the possibility of brittle fracture within the RCPB, the fracture or notch properties and the operating temperature of ferritic materials are controlled to ensure adequate toughness when the system is pressurized to more than 20 percent of the design pressure. Subsection 5.2.4 describes the methods used to control toughness properties. Materials are to be impact tested in accordance with ASME Boiler and Pressure Vessel (B&PV) Code Section III, 1971. The fracture toughness temperature requirements of the RCPB materials also apply for the RCPB piping which penetrates the containment, up to and including the outermost isolation valve.

Piping and equipment pressure parts of the RCPB are assembled and erected by welding unless applicable codes permit flanged or screwed joints. The welding procedures used are designed to to 318327-4 Non-Proprietary Information-Class I (Public)

Page 13 of 16 produce welds of complete fusion and free of unacceptable defects. All welding procedures, welders, and welding machine operators are qualified in accordance with the requirements of Section IX of the ASME B&PV Code for the materials to be welded.

Qualification records, including the results of procedure and performance qualification tests and identification symbols assigned to each welder, are maintained.

Subsection 5.2.3 contains the detailed material and examination requirements for the piping and equipment of the RCPB prior to and after its assembly and erection. Leakage testing and surveillance are accomplished as described in the evaluation against Criterion 30.

The design, fabrication, erection, and testing of the RCPB ensures an extremely low probability of failure or abnormal leakage, thus satisfying the requirements of Criterion 14.

UFSAR Section 3.1.2.4.1 Criterion 30, Quality of Reactor Coolant Pressure Boundary, states:

Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Criterion 30 conformance is demonstrated as follows:

By using conservative design practices and detailed quality control procedures, the pressure-retaining components of the RCPB are designed and fabricated to retain their integrity during normal and postulated accident conditions. Accordingly, components that comprise the RCPB are designed, fabricated, erected, and tested in accordance with the recognized industry codes and standards listed in Sections 5.2, 5.5, and 5.5. Further, product and process quality planning is provided as described in Chapter 17 to ensure conformance with the applicable codes and standards and to retain appropriate documented evidence verifying compliance. Because the subject matter of this criterion deals with aspects of the RCPB, further discussion on this subject is treated in the response to Criterion 14, Reactor Coolant Pressure Boundary.

Means are provided for detecting reactor coolant leakage. The leak detection system consists of sensors and instruments to detect, annunciate, and in some cases, isolate the RCPB from potentially hazardous leaks before predetermined limits are exceeded. Small leaks are detected by temperature and pressure changes, increased condensate flow from the primary containment cooling system, increased frequency of sump pump operation, and measurement of fission product concentration. In addition to these, large leaks are detected by changes in flow rates in process lines and reactor water level. The allowable leakage rates are based on the predicted and experimentally determined behavior of cracks in pipes, the ability to make up coolant system leakage, the normally expected background leakage due to equipment design, and the detection capability of the various sensors and instruments. The total leakage rate limit is established so that, in the absence of normal AC power associated with a loss of feedwater supply, makeup capabilities are provided by the RCIC system. While the leak detection system provides to 318327-4 Non-Proprietary Information-Class I (Public)

Page 14 of 16 protection from small leaks, the ECCS network provides protection for the complete range of discharges from ruptured pipes. Thus, protection is provided for the full spectrum of possible discharges. The RCPB and the leak detection system are designed to meet the requirements of Criterion 30.

UFSAR Section 3.1.2.4.2 Criterion 31, Fracture Prevention of Reactor Coolant Pressure Boundary, states:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady-state, and transient stresses, and (4) size of flaws.

Criterion 31 conformance is demonstrated as follows:

Brittle fracture control of pressure-retaining ferritic materials is provided to ensure protection against nonductile fracture. To minimize the possibility of brittle fracture failure of the RPV, it is designed to meet the requirements of ASME B&PV Code Section III, 1968 Edition through Summer 1969 addenda, which considers material properties; steady-state and transient stresses; and the size of flaws, and conforms very closely with Appendix G, which was added in the Summer 1972 Addenda (see Section 5.2 for a discussion of the degree of conformance).

The nil ductility transition (NDT) temperature is defined as the temperature below which ferritic steel fails in a brittle rather than ductile manner. The RTNDT temperature increases as a function of neutron exposure at integrated neutron exposures greater than 1.0 x 1017 n/cm2 with neutrons of energies in excess of 1 MeV. Since the material RTNDT temperature dictates the minimum operating temperature at which the reactor vessel can be pressurized, it is desirable for the NDT temperature to be low.

The reactor assembly design provides an annular space from the outermost fuel assemblies to the inner surface of the RPV that serves to attenuate the fast neutron flux incident upon the reactor vessel wall. This annular volume contains the core shroud, jet pump assemblies, and reactor coolant. Assuming plant operation at rated power, availability of 80 percent, and a plant operating life of 40 years, the maximum fast neutron fluence at the inner surface of the RPV is calculated to be 9.68 x 1017 n/cm2 (fast neutron fluence consists of neutrons having energies greater than 1 MeV) as detailed in Table 4-3.2. The end-of-life (EOL) RTNDT temperature as calculated from the EOL fluence and chemical composition indicates a substantial margin against the occurrence of brittle fracture. For hydrostatic test, the RPV will not be pressurized until the RPV temperature exceeds the RTNDT by at least 60°F.

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 15 of 16 The RCPB piping, pumps, and valves are designed, maintained, and tested such that adequate assurance is provided that the boundary will behave in a nonbrittle manner throughout the life of the plant.

UFSAR Section 3.1.2.4.3 Criterion 32, Inspection of Reactor Coolant Pressure Boundary, states:

Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Criterion 32 conformance is demonstrated as follows:

The RPV design and engineering effort includes provisions for inservice inspection. Removable plugs in the sacrificial shield and/or removable panels in the insulation provide access for examination of the vessel and its appurtenances. In addition, all of the remaining portion of the RCPB is provided with removable insulation. Inspection of the RCPB is in accordance with the ASME B&PV Code Section XI. The Inservice Inspection Plan, access provisions, and areas of restricted access are defined in Section 5.2.

Reactor pressure vessel material surveillance samples are located within the RPV to enable periodic monitoring of material properties with exposure. The program includes specimens of the base metal, heat-affected zone metal, and weld material. The samples are placed near the core midplane to obtain maximum exposures. Tests include tensile and impact testing. The test program is in accordance with ASTM E185-73 and the appropriate requirements of 10 CFR 50, Appendixes G and H. Subsequent to developing this surveillance program, the BWRVIP developed an integrated surveillance program (ISP) which replaces the Fermi specific surveillance program. This program is described in Section 5.2.4.4.3.

The plant testing and inspection programs ensure that the requirements of Criterion 32 will be met.

to 318327-4 Non-Proprietary Information-Class I (Public)

Page 16 of 16 The Fermi 2 UFSAR, Section 3.2, also contains the classification of the RCPB components.

Tables 3.2-1, 3.2-2, and 3.2-3 define the classification of each component and the minimum code requirements for the quality group classification.

In summary and as previously detailed, the ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° °

° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) are more limiting relative to stress than any of the Class 1 ferritic branch piping in the RCPB. To address irradiation effects on RCPB piping, a qualitative assessment was performed; the accrual of fluence greater than 1.0e17 n/cm2 ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )) based on historical calculations of flux vs. ((° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° ° )). Additionally, the WLI nozzles located in the beltline region along with the RPV beltline and non-beltline materials were considered in the development of the PT curves.

As discussed in the UFSAR sections cited above, the GDC explanations define the manner in which the RCPB was designed and constructed to ensure a high degree of integrity with adequate toughness throughout the plant life. The RCPB components were designed and fabricated, and are maintained and tested such that adequate assurance is provided that the boundary will behave in a nonbrittle manner throughout the life of the plant.

References

1. GEH Nuclear Energy, NEDC-33178P-A, Revision 1, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, Report for BWR Owners Group, Sunol, California, June 2009.

2. 10CFR50 Appendix G, Fracture Toughness Requirements, December 1995.

3. ASME Code Section XI Appendix G, Fracture Toughness Criteria for Protection Against Failure, 2004 Edition.

4. WRC Bulletin 175, PVRC Recommendations on Toughness Requirements for Ferritic Materials, August 1972.

5. Safety Evaluation Report, Thomas B. Blount (NRC) to Doug Coleman (BWROG), Final Safety Evaluation for Boiling Water Reactors Owners Group Licensing Topical Report NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (TAC No. MD2693), April 27, 2009.

6. USNRC Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988.

7. GE Nuclear Energy, Basis for GE RTNDT Estimation Method, NEDC-32399-P, BWR Owners Group, San Jose, CA, September 1994; Letter from B. Sheron to RA Pinelli, Safety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation Method, September 1994, USNRC, December 16, 1994.

8. Fermi 2 UFSAR, Revision 18, October 2012.

to NRC-13-0057 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318327-4

318327-4 GEH Affidavit for Enclosure 1

GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318327-4 Page 1 of 3 AFFIDAVIT I, Peter M. Yandow, state as follows:

(1) I am the Vice President, Nuclear Plant Projects/Services Licensing, Regulatory Affairs, of GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosure 1 of GEH letter, 318327-4, Fermi PT Curve Responses to RAIs 1-5, dated September 27, 2013. The GEH proprietary information in Enclosure 1, which is entitled GEH Responses to PT Curve RAIs 1-5, is identified by a dotted underline inside double square brackets. ((This sentence is an example.{3})) In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C.

Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a.

Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b.

Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c.

Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d.

Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my

GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318327-4 Page 2 of 3 knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a need to know basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains the detailed GEH methodology for pressure-temperature curve analysis for the GEH Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the pressure-temperature curves were achieved at a significant cost to GEH.

The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their

GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318327-4 Page 3 of 3 own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 27th day of September 2013.

Vice President, Nuclear Plant Projects/Services Licensing, Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd Wilmington, NC 28401 Peter.Yandow@ge.com to NRC-13-0057 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 GEH affidavit for Enclosure 1 of GEH letter, 1-2RXPKK-13, "GEH Responses to GGNS PT Curve RAIs"

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Edward D. Schrull, PE, state as follows:

(1) I am the Vice President, Regulatory Affairs, Services Licensing, of GE-Hitachi Nuclear Energy Americas LLC ("GEH"), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosure 1 of GEH letter, 1-2RXPKK-13, "GEH Responses to GGNS PT Curve RAIs," dated October 19, 2012. The GEH proprietary information in Enclosure 1, which is entitled "GEH Responses to GGNS PT Curve RAIs," is identified by a dotted underline inside double square brackets. ((is sentence is an example.3)) In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), -and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from. disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a.

Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b.

Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c.

Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d.

Information that discloses trade secret and/or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my Affidavit for Enclosure 1 of 1-2RXPKK-13 Page 1 of 3

GE-Hitachi Nuclear Energy Americas LLC knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary and/or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains the detailed GEH methodology for pressure-temperature curve analysis for the GEH Boiling Water Reactor (BWiR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the pressure-temperature curves were achieved at a significant cost to GEH.

The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering., analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their Affidavit for Enclosure 1 of 1-2RXPKK-13 Page 2 of 3

GE-Hitachi Nuclear Energy Americas LLC own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 19th day of October 2012.

Edward D. Schrull, PE Vice President, Regulatory Affairs Services Licensing GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd.

Wilmington, NC 28401 Affidavit for Enclosure 1 of 1-2RXPKK-13 Page 3 of 3