NLS2019040, 3-EN-DC-147, Rev. 5C1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (Efpy). (Non-proprietary)

3-EN-DC-147, Rev. 5C1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (Efpy). (Non-proprietary)

ML19267A111
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Issue date:	08/07/2019
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ML19267A115	List: ML19267A110 NLS2019040, 3-EN-DC-147, Rev. 5C1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (Efpy). (Non-proprietary)
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NLS2019040 3-EN-DC-147, Rev. 5C1
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NLS2019040 Page 1 of 79 Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY) (Non-Proprietary)

Cooper Nuclear Station Docket No. 50-298, DPR-46

NUCLEAR QUALITY RELATED 3-EN-DC-147 I REV. 5C1

~ Entergy MANAGEMENT INFORMATIONAL USE PAGE 1 of 33 MANUAL Enaineerina ReDorts ATTACHMENT 9.1 ENGINEERING REPORT COVER SHEET Engineering Report No. __

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Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY)

( Non-Proprietary)

Engineering Report Type: (3)

New D Revision cgj Cancelled O

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Superseded by:

Revision 1: Removed "Proprietary Brackets" from Table 4 values for Cu and Ni content.

Revision 2: Update information of BWRVIP ISP surveillance capsule data for 120° coupon.

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Section 1.0 2.0 3.0 4.0 5.0 6.0 Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Table 1 Table 2 Table 3 Table 4 Appendix A Appendix B Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 2 of 33 Table of Contents Purpose Applicability Methodology Operating Limits Discussion References CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54 EFPY CNS P-T Curve B (Normal Operation - Core Not Critical) for 54 EFPY CNS P-T Curve C (Normal Operation - Core Critical) for 54 EFPY Cooper Feedwater Nozzle Finite Element Model [ 19]

Cooper Core Differential Pressure Nozzle Finite Element Model CNS Pressure Test (Curve A) P-T Curves for 54 EFPY CNS Core Not Critical (Curve B) P-T Curves for 54 EFPY CNS Core Critical (Curve C) P-T Curves for 54 EFPY CNS ART Calculations for 54 EFPY Page 3

3 4

5 6

13 17 18 19 20 21 22 25 28 31 Cooper Reactor Vessel Materials Surveillance Program 32 BWRVIP-135, Revision 3: BWR Vessel and Internals Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations, Technical Report No.

3002003144, December 2014 (Pages 1 - 45)

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 3 of 33 1.0 Purpose The purpose of the Cooper Nuclear Station (CNS) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing;

2. RCS Heatup and Cooldown rates;

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision 1-A, contained within BWROG-TP-11-022-A, Revision 1 [l], and 0900876.401, Revision 0-A, contained within BWROG-TP-11 -023-A, Revision O [2].

2.0 Applicability This report is applicable to the CNS RPV for up to 54 Effective Full-Power Years (EFPY).

The following CNS Technical Specifications (TS) are affected by the information contained in this report:

TS RCS Pressure and Temperature (P-T) Limits TS Surveillance Requirements 3.0 Methodology Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 4 of 33 The limits in this report were derived as follows:

The methodology used is in accordance with Reference [ 1] and Reference [2],

incorporating the NRC Safety Evaluations in References (3] and [ 4], respectively.

The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [5], using the RAMA computer code, as documented in Reference [6].

The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (7], as documented in Reference [8].

The pressure and temperature limits were calculated in accordance with Reference [ 1],

"Pressure - Temperature Limits Report Methodology for Boiling Water Reactors," June 2013, as documented in NPPD Calculation NEDC 07-048, Reference [9].

This revision of the pressure and temperature limits is to incorporate the following changes:

Revision O - Initial Issue of PTLR Revision 1 - Update pressure and temperature limits for 54 EFPY.

Revision 2 - Update information on BWRVIP ISP surveillance capsule data for CNS representative materials.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPY, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 ( I OJ, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

r Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 5 of 33 Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumptions modifications in the UFSAR, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 54 EFPY for Cooper Nuclear Station, as documented in Reference [9]. The C S P-T curves for 54 EFPY are provided in Figures 1 through 3, and a tabulation of the curves is included in Tables 1 through 3. The adjusted reference temperature (ART) tables for the C S vessel beltline materials are shown in Table 4 for 54 EFPY (Reference

[8]). The resulting P-T curves are based on the geometry, design and materials information for the CNS vessel with the following conditions:

Heatup and Cooldown rate limit during Hydrostatic Class 1 Leak Testing (Figure 1:

Curve A): ~ 25°F/hour 1 [9].

1 Interpreted a the temperature change in any I-hour period is less than or equal to 25°F.

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 6 of 33 ormal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): :S l 00°F/hour2 [9).

RPV bottom head coolant temperature to RPV coolant temperature ~T limit during Recirculation Pump startup:~ 145°F.

Recirculation loop coolant temperature to RPV coolant temperature ~ T limit during Recirculation Pump startup: ~ 50°F.

RPV flange and adjacent shell temperature limit 2: 70°F [9].

To address the NRC condition regarding lowest service temperature in Reference [3, Section 4.0], the minimum temperature is set to 70°F for Curves A and B, which bounds RTNoT,max and the C S shutdown margin analysis, and 80°F for Curve C, which is equal to RTNDT,max + 60°F.

These values are consistent with the minimum temperature limits approved for use by the NRC in Reference [ 11].

The composite P-T curves are extended below O psig to -14. 7 psig based on the evaluation documented in Reference [12], which demonstrates that the P-T curves are applicable to negative gauge pressures. A pressure of -14.7 psig bounds the maximum expected vacuum pressure as well as externally applied pressures the RPV may experience. Since the P-T curve calculation methods used do not specifically apply to negative values of pressure, the tabulated results start at O psig. However, the minimum analyzed RPV pressure is -14.7 psig 5.0 Discussion The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [7] provides the methods for 2 Interpreted as the temperature change in any I-hour period is less than or equal to 100°F.

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 7 of 33 determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the CNS vessel plate, weld, and forging materials [8]. This evaluation included the results of two surveillance capsules for the representative plate material and three surveillance capsules for the representative weld material. The Cu and i values were used with Table 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds. The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings. However, the fitted CF for the limiting plate (which is based on credible surveillance data) in the CNS vessel bounds the RG 1.99 CF. Therefore, the fitted CF is used for the limiting beltline plate.

The peak RPV ID fluence value of 2.23 x 1018 n/cm2 at 54 EFPY used in the P-T curve evaluation were obtained from Reference [ 6] and are calculated in accordance with RG 1.190

[5]. These fluence values apply to the limiting beltline lower intermediate shell plate (Heat No.

C2307-2). The fluence values for the lower intermediate shell plate are based upon an attenuation factor of 0. 72 for a postulated l/4T flaw. As a result, the l/4T fluence for 54 EFPY for the limiting lower intermediate shell plate is 1.62 x 1018 n/cm2 for C S.

The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. The water level instrument (WLI) nozzle is located in the lower-intermediate shell beltline plates [9]. The nozzle material is not ferritic, however the effect of the penetration on the adjacent shell is considered according to the methodology in Reference [2]. The RPV ID fluence value of 5.44 x 1017 n/cm2 at 54 EFPY used in the P-T curve evaluation of the WLI nozzle was obtained from Reference [6]

and is calculated in accordance with RG 1.190 [ 5]. This fluence value applies to the limiting WLI nozzle location (Heat No. EV-26067). The fluence value for the WLI nozzle location is

Cooper uclear Station PTLR ER 2016-042, Rev 2 Page 8 of 33 based upon an attenuation factor of 0.72 for a postulated 1/4T flaw. As a result, the l/4T fluence for 54 EFPY for the limiting WLI nozzle location is 3.94 x 10 17 n/cm2 for CNS. There are no additional forged or partial penetration nozzles in the extended beltline.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the l/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location ( outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stresses at the l/4T location are assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, and for a given pressure, the coolant saturation temperature is well above the P-T curve limiting temperature. Consequently, the material toughness at a given pressure would exceed the allowable toughness.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cool down temperature rate of::: 100°F /hour for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level A/B RPV thermal transients. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of::: 25°F/hour must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 9 of 33 The initial RT oT, the chemistry (weight-percent copper and nickel), and ART at the l/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 10 17 n/cm2 for E > lMeV) are shown in Table 4 for 54 EFPY [8]. Initial RT OT values were reported in the ART calculation in C S Amendment 120 [ 13].

Per Reference [8] and in accordance with Appendix A of Reference [ 1], the C S representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [ 14]. The representative heat of the plate material (C2307-2) in the ISP is the same as the lower intermediate shell plate material in the vessel beltline region of CNS. For plate heat C2307-2, since the scatter in the fitted results is less than 1-sigma ( l 7°F), the margin term ( crc. = 17°F) is cut in half for the plate material when calculating the ART. The representative heat of the weld material (20291) in the ISP is not the same as the limiting weld material in the vessel beltline region of CNS. Therefore, CFs from the tables in RG 1.99 were used in the determination of the ART values for all CNS beltline materials except for plate heat C2307-2.

Additionally, the most recent BWRVIP ISP representative weld and plate surveillance material data for CNS have been reviewed. The results of testing of the CNS 120° capsule were published in 2018 in EPRI Letter 2018-064 [28]. The impact of the new surveillance data and updated reactor vessel fluence projections on the 54 EFPY P-T limit curves has been evaluated in Reference [29]. As documented in that evaluation, if the updated CF and fluence projections for plate heat C2307-2 were considered, the 54 EFPY limiting ART value would decrease relative to the limiting ART value in Table 4 (developed in Reference [8]). Therefore, the limiting ART value in Table 4 and the P-T curves in this PTLR remain conservative for 54 EFPY.

Consequently, the P-T curves and ART presented in thi PTLR have not been revised at this time to incorporate the new surveillance data from Reference [28]

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 10 of 33 The only computer code used in the determination of the C S P-T curves was the ANSYS finite element computer program:

ANSYS, Revision 5.3 [ 15] for the feedwater (FW) nozzle (non-beltline) pressure and thermal down shock stresses.

Mechanical and Prep Post, Release 11.0 (Service Pack 1) [ 16] for the development of the generic WLI nozzle stress intensity factors in [2].

Mechanical APDL and Prep Post, Release 12.1 [ 17] for the FW nozzle (non-beltline) thermal ramp stresses and the core differential pressure (DP) nozzle (bottom head) pressure stress distribution.

A SYS finite element analyses were used to develop the stress distributions through the FW, WLI, and core DP nozzles, and these stress distributions were used in the determination of the stress intensity factors for these nozzles (2, 18, 19, 20]. At the time that each of the analyses above was performed, the A SYS program was controlled under the vendor's 10 CFR 50 Appendix B (21] Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 (22] was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems.

The plant-specific CNS FW nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients [ 18, 19].

Detailed information regarding the analysis can be found in References [ 18] and [ 19]. The following inputs were used as input to the finite element analysis:

With respect to operating conditions, stress distributions were developed for two bounding thermal transients. A thermal shock, which represents the maximum thermal shock for the FW nozzle during normal and upset operating conditions [ 18], and a thermal ramp were analyzed (19]. Potential leakage past the primary and secondary

Cooper uclear Station PTLR ER 2016-042, Rev 2 Page 11 of 33 thermal sleeves is considered in the heat transfer calculations. The thermal down shock of 450°F, which is associated with the turbine roll transient during startup, produces the highest tensile stresses at the l/4T location. Because operation is along the saturation curve, these stresses are scaled to reflect the worst-case step change due to the available temperature difference. It is recognized that at low temperatures, the available temperature difference is insignificant and could potentially result in a near zero stress distribution. Therefore, a minimum stress distribution is calculated based on the thermal ramp of 100°F/hour, which is associated with the shutdown transient. Therefore, the combination of the thermal down shock and thermal ramp stresses represent the bounding stresses in the FW nozzle associated with I 00°F /hour heatup/cooldown limits associated with the P-T curves for the upper vessel FW nozzle region.

Heat transfer coefficients were given in the CNS FW nozzle design basis stress report and are a function of FW temperature and flow rate. Bounding, or larger, convection coefficients were used in the present P-T curve analysis [ 18, 19]. Therefore, the heat transfer coefficients used in the analysis bound the actual operating conditions in the FW nozzle at CNS.

A two-dimensional finite element model of the FW nozzle was constructed (Figure 4 ).

The pressure stresses are multiplied by a factor of 2.5 to account for the 3-D effects [18].

Material properties were taken at 350°F, which is approximately the average temperature for the shutdown transient, from the 1989 ASME Code [23]. The use of temperature independent material properties is consistent with original design basis documents. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.

The plant-specific CNS core DP nozzle analysis was performed to determine a through-wall pressure stress distribution [20]. Detailed information regarding the analysis can be found in Reference [20]. The following inputs were used as input to the finite element analysis:

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 12 of 33 o thermal transients were analyzed as part of the plant-specific core DP nozzle evaluation. Thermal stresses were addressed generically as specified in [ 1] with the use of a stress concentration factor of 3.0 to account for the discontinuity in the bottom head.

A two-dimensional finite element model of the core DP nozzle was constructed (Figure 5). Material properties were taken at 325°F from the vessel stress report (20]. The use of temperature independent material properties is consistent with original design basis documents. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.

6.0 References Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 13 of 33

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, June 2013.

2. BWROG-TP-11 -023-A, Revision 0, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations, May 2013.

3. U.S. NRC Letter to BWROG dated May 16, 2013, "Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, 'Pressure-Temperature Limits Report Methodology for Boiling Water Reactors"' (TAC NO. ME7649, MLl 3277 A557).

4. U.S. NRC Letter to BWROG dated March 14, 2013, "Final Safety Evaluation for Boiling Water Reactor Owners" Group Topical Report BWROG-TP-11-023, Revision 0, November 2011, 'Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations"' (TAC NO. ME7650, MLl 3183AO 17)

5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 2001.

6. Cooper Nuclear Station Calculation NEDC 07-032, Revision 4, "CNS Review of TransWare Calculations NPP-FLU-003-R-002, Revision 0, NPP-FLU-003-R-004, and NPP-FLU-003-R-005, Reactor Pressure Vessel Fluence Evaluation", August 2019, that incorporated TransWare Enterprises Report No. NPP-FLU-003-R-005, Revision 0, "Non-Proprietary Version of Cooper Nuclear Station Reactor Pressure Vessel Fluence Evaluation," January 2011.

7. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May 1988.

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 14of 33

8. Cooper Nuclear Station Calculation NEDC 07-045, Revision 4, August 2019, "Review of SIA Calculation 1100445.301, Proprietary and Non-Proprietary Versions, ~RTNoT and ART Evaluation," dated July 2010.

9. Cooper Nuclear Station Calculation, NEDC 07-048, Revision 8, August 2019, "Review of SIA Calculation 14004 73.302 Cooper Updated P-T Curve Calculation for 54 EFPY",

dated December 2015.

10. U.S. Code of Federal Regulations, Title 10, Part 50, Section 59, "Changes, tests and experiments," August 29, 2017.

11. Cooper uclear Station Amendment 245 as approved by the NRC on February 22, 2013.

(ML l 3032A526)

12. Cooper Nuclear Station Calculation NEDC 16-024, Revision 0, September 2016, "Review of SIA Calculation 11004 73.30 l Cooper Vacuum Assessment", Revision 0 dated December 2015.

13. Cooper Nuclear Station Amendment 120 as approved by the NRC on April 26, 1988.

(ML02 l 3 604 24)

14. BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. SI File No. BWRVIP-135P. EPRI PROPRIETARY INFORMATION.

15. ANSYS, Revision 5.3, ANSYS Inc., October 1996.

16. ANSYS Mechanical and PrepPost, Release 11.0 (w/ Service Pack 1), ANSYS, Inc.,

August 2007.

17. ANSYS Mechanical APDL and PrepPost, Release 12.1 x64, ANSYS, Inc.,

ovember 2009.

18. Cooper Nuclear Station Calculation No. NEDC99-020, "Review of Structural Integrity Report SIR-99-069 and Calculations No. NPPD-13Q-301, NPPD-13Q-302, NPPD-13-Q-

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 15 of 33 303," specifically Structural Integrity Associates Calculation No. NPPD-13Q-302, Revision 2, "Feedwater Nozzle Stress Analysis," June 1999.

19. Cooper Nuclear Station Calculation No. NEDC99-020, Structural Integrity Associates Calculation No. 1100445.302, Revision 0, "Finite Element Stress Analysis of Cooper RPV Feedwater Nozzle," June 2011.

20. Cooper Nuclear Station Calculation, NEDC 16-025, "Review of SIA Calculation 1100445.304 Core Differential Pressure Nozzle Finite Element Model and Stress Analysis" dated July 2011.

21. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, "Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants".

22. U.S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, "License Qualification for Performing Safety Analyses", June 24, 1999.

23. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 1989 Edition.

24. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," August 29, 201 7.

25. Letter NLS2002104 dated December 31, 2002, "License Amendment Request to Adopt an Integrated Reactor Vessel Material Surveillance Program, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46", from M.T. Coyle (NPPD) to U.S.

uclear Regulatory Commission, ADAMS Accession No. ML030080070, SI File No.

1400473.202.

26. Cooper Nuclear Station Amendment 201 as approved by the NRC on October 23, 2003.

(ML033090607)

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 16 of 33

27. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012.

1025144. EPRI PROPRIETARY INFORMATION.

28. EPRI Letter 2018-064, "Advance Notification of New BWRVIP Integrated Surveillance Program (ISP) Data Applicable to Cooper," June 12, 2018. EPRI PROPRIETARY INFORMATION.

29. SI Report No. 1800664.401, Revision 0, "Evaluation of the Effects of BWRVIP ISP Capsule Test Results on the Cooper Pressure-Temperature Curves," December 19, 2018.

30. BWRVIP-318NP: BWR Vessel and Internals Project, Testing and Evaluation of the Cooper 120° Surveillance Capsule. EPRI, Palo Alto, CA: 2018. 3002013102.

31. Cooper Nuclear Station Amendment 256, Cooper Nuclear Station as approved by the NRC on July 25,2016 (ML16158A022)

32. U. S. Nuclear Regulatory Commission, Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits", January 31, 1996.

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 17 of 33 Figure 1: CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54 EFPY 1300 1200 1100 1000 -

Curve A - Pressure Test, Composite Curves

-- Beltline Bottom Head Non-Beltline Overall

, I

/ I

, I I

I I

I -

I I

/

I T --,-----

1 I

900 I

- * ---, -- I aii 800 -

I ','

70°F, 814 psig _

..e:

]

~ 700 0...

V 600 C:

70°F, 426 psig 300 -

I 70°F, 313 psig 200 100 0

-100 -

0 50 ll0°F, 633 psig I 110°F, 313 psig 100 150 200 Safe Operating Region Minimum RPV Pressure = -14. 7 psig Minimum Bolt-Up Temperature > 70°F 250 Minimum Reactor Vessel Metal Temperature (°F) 300

QO "ij;

.9:

a:;

QI >

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QI a:

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.E QI

i QI 1£.

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 18 of 33 Figure 2: CNS P-T Curve B (Normal Operation - Core Not Critical) for 54 EFPY Curve B - Core Not Critical, Composite Curves

-- Beltline Bottom Head Non-Beltline Overall 1300

• r I

I I

1200 ----------t--- L I

I I

I I

1100 -

T - T I

I I

1000 -

I I

I I

900 J..--- I I

II 800 1.. "

I I

I 700 I

I I

600 I

I I

70°F, 499 psig I

I I

~

Safe 500 -

.J.

r -

124°F, 140°F, 503 psig I

Operating I

432 psig Region I

400 t

I I 70°F, 313 psig I :_ _

140°F, 313 psig 300 ------,-

I 200 70°F, 184 psig I

100 -

+-

Minimum RPV Pressure= -14.7 psig 0

~

Minimum Bolt-Up Temperature > 70°F

-100

... L 0

50 100 150 200 250 300 Minimum Reactor Vessel Metal Temperature {°F)

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 19 of 33 Figure 3: CNS P-T Curve C (Normal Operation - Core Critical) for 54 EFPY 1300 1200 1100 1000 900 tic 800

'iii

~

ai 700 QI >

0...

I.I re, QI 600 -

c::

C:... *e 500 QI

i QI C:

400 300 200 100 -

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Curve C - Core Critical, Composite Curves

-- Beltline --- Bottom Head Non-Beltline I

I I

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I I

I I

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I rl I

I,1,+

I f

I I

I-I I

I I

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~-------,-

99°F, 313 psig

/ '~--~

I 1

164°F, I

I I

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I 432 psig I

l I

I t

Overall Safe Operating Region 80°F, 332 psig I,,

-I

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• I 80°F, 212 psig I : /

J 1so°F, 313 psig 145°F, 313 psig I

~ -

--~

800F, 126 psig I

~

50 100 150 200 Minimum RPV Pressure= -14.7 psig Minimum Core Critical Temperature> 80°F 250 Minimum Reactor Vessel Metal Temperature {°F}

300

ELEMENTS MA.T NOM Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 20 of 33 Figure 4: Cooper Feedwater Nozzle Finite Element Model [19]

J\\N' "*

APR 20 2011 15 :18 :48 Plill NO.

1

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 21 of 33 Figure 5: Cooper Core Differential Pressure Nozzle Finite Element Model (20]

J\\N s Core DP Nozzle, Pre ssure Stress

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 22 of 33 Table 1: CNS Pressure Test (Curve A) P-T Curves for 54 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF psi 70.0 0.0 70.0 426.0 80.6 466.6 89.3 507.2 96.8 547.7 103.2 588.3 109.0 628.9 120.9 678.1 130.6 727.2 138.7 776.4 145.6 825.5 151.7 874.7 157.2 923.9 162.1 973.0 166.5 1022.2 170.6 1071.4 174.4 1120.5 178.0 1169.7 181.2 1218.9 184.3 1268.0 187.2 1317.2

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 23 of 33 Table 1: CNS Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF psi 70.0 0.0 70.0 312.6 110.0 312.6 110.0 1563.0

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 24 of 33 Table 1: CNS Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF psi 70.0 0.0 70.0 814.0 74.8 864.0 79.2 913.9 83.3 963.8 87.0 1013.8 90.5 1063.7 93.8 1113.6 96.8 1163.5 99.7 1213.5 102.4 1263.4 105.0 1313.3

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 25 of 33 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 54 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure OF psi 70.0 0.0 70.0 184.3 86.2 233.8 98.5 283.3 108.3 332.8 116.5 382.3 123.6 431.7 135.5 480.9 145.1 530.1 153.2 579.3 160.1 628.5 166.2 677.7 171.6 726.8 176.5 776.0 181.0 825.2 185.1 874.4 188.9 923.6 192.4 972.8 195.7 1022.0 198.8 1071.1 201.7 1120.3 204.4 1169.5 207.0 1218.7 209.5 1267.9 211.9 1317.1

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 26 of 33 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure OF psi 70.0 0.0 70.0 312.6 140.0 312.6 140.0 1563.0

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 27 of 33 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure OF psi 70.0 0.0 70.0 498.6 76.2 547.0 81.6 595.4 86.6 643.7 91.1 692.1 95.2 740.5 99.0 788.9 102.5 837.3 105.8 885.7 108.9 934.0 111.8 982.4 114.6 1030.8 117.2 1079.2 119.7 1127.6 122.1 1175.9 124.3 1224.3 126.5 1272.7 128.6 1321.1

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 28 of 33 Table 3: CNS Core Critical (Curve C) P-T Curves for 54 EFPY Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure OF psi 80.0 0.0 80.0 126.2 104.0 169.8 120.2 213.5 132.4 257.1 142.2 300.8 150.4 344.4 157.4 388.1 163.6 431.7 175.5 480.9 185.1 530.1 193.2 579.3 200.1 628.5 206.2 677.7 211.6 726.8 216.5 776.0 221.0 825.2 225.1 874.4 228.9 923.6 232.4 972.8 235.7 1022.0 238.8 1071.1 241.7 1120.3 244.4 1169.5 247.0 1218.7 249.5 1267.9 251.9 1317.1

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 29 of 33 Table 3: CNS Core Critical (Curve C) P-T Curves for 54 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure OF psi 80.0 0.0 80.0 211.8 87.3 245.4 93.5 279.0 98.8 312.6 180.0 312.6 180.0 1563.0

Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 30 of 33 Table 3: CNS Core Critical (Curve C) P-T Curves for 54 EFPY ( continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure OF psi 80.0 0.0 80.0 331.9 90.9 381.1 99.8 430.4 107.4 479.6 113.9 528.9 119.7 578.1 124.9 627.4 129.7 676.6 134.0 725.8 137.9 775.1 141.6 824.3 145.0 873.6 148.2 922.8 151.2 972.1 154.1 1021.3 156.8 1070.6 159.3 1119.8 161.7 1169.0 164.1 1218.3 166.3 1267.5 168.4 1316.8

Table 4: CNS ART Calculations for 54 EFPY Beltline ID Code No.

Heat No.

Flux Type Initial Cu Ni RTNO'l' l'FI lwto/o l

/wt%)

Lower Shell Plate G-2803-1 C2274-l 14.0 0.20 0.68 Lower Shell Plate G-2803-2 C2307-l 00 0.21 0.73 Plates Lower Shell Plate G-2803-3 C2274-2

-8.0 0.20 0.68 Lower Int. Shell Plate G-2802-1 C233 I-2 10.0 0.16 0.62 Lower Int. Shell Plate G-2802-2 C2307-2

-20.0 0.21 0.76 Lower Int. Shell Plate G-2801-7 C2407-l

-10.0 0.13 0.65 Lower Shell Axial Welds 2-233A 12420 LINDE 1092

-50.0 0.270 1.035 Lower Shell Axial Welds 2-2338 12420 LINDE 1092

-50.0 0.270 1.035 Lower Shell Axial Welds 2-233C 12420 LINDE 1092

-50.0 0.270 1.035 Welds Lower Int. Shell Axial Welds I-233A 2 7204/12008 LINDE 1092

-50.0 0.219 0.996 Lower Int. Shell Axial Welds l-233B 27204/12008 LINDE 1092

-50.0 0.219 0.996 Lower Int. Shell Axial Welds l-233C 27204/ 12008 LINDE 1092

-50.0 0.219 0.996 Lower/Lower Int. Shell Circ Weld 1-240 21935 LINDE 1092

-50.0 0.183 0.704 Nozzles Nozzle N-l 6A G-2822 EV-26067

-10.0 0.13 0.65 Nozzle N-16B G-2822 EV-26067 10.0 0.16 0.62 Fluence Data Beltline ID Code No.

Heat No.

Wa ll Thickness (in.)

Fluence Attenuation at ID Full 1/41 (11/cm'l e *O,l *h Lower Shell Plate G-2803-1 C2274-1 6.375 1.59 l.75E+l8 0.68 Lower Shell Plate G-2803-2 C2307-I 6.375 1.59 1.75E+18 0.68 Plates Lower Shell Plate G-2803-3 C2274-2 6.375 1.59 l.75E+l8 0.68 Lower Int. Shell Plate G-2802-1 C233 l-2 5.375 1.34 2.23E+ 18 0.72 Lower Int. Shell Plate G-2802-2 C2307-2 5.375 1.34 2.23E+!8 0.72 Lower Int. Shell Plate G-2801-7 C2407-I 5.375 1.34 2.23E+ 18 0.72 Lower Shell Axial Welds 2-233A 12420 6.375 1.59 1.72E+ l8 0.68 Lower Shell Axial Welds 2-233B 12420 6.375 1.59 l.72E+ l8 0.68 Lower Shell Axial Welds 2-233C 12420 6.375 1.59 1.72E+ l8 0.68 Welds Lower Int. Shell Axial Welds l-233A 27204/12008 5.375 1.34 1.26E+l8 0 72 Lower Int. Shell Axial Welds 1-233 8 27204/12008 5.375 1.34 l.26E+ 18 0 72 Lower Int. Shell Axial Welds l-233C 27204/12008 5.375 1.34 I.26E+ 18 0.72 Lower/Lower Int. Shell Circ Weld 1-240 2 1935 5.375 1.34 l.75E+18 0 72 Nozzles Nozzle N-I 6A G-2822 EV-26067 5.375 1.34 5.44E+ 17 0 72 Nozzle N-1 68 G-2822 EV-26067 5.375 1.34 5.44E+ 17 0.72 CF

("Fl 153.0 162.8 153.0

((..

'JJ

{[-

IEJjJ 92.3 254.4 254.4 254.4 23 1.1 23 1.1 23 1.1 172.2 92.3 118.5 Fluence at 1/41 (n/cm')

l.19E+ l8 1.19E+l8

l. 19E+ l8 1.62E+ l8 l.62E+l8 l.62E+ 18 l.17E+ l 8 l.1 7E+l8 l.17E+l8 9.13E+17 9.13E+ l7 9.13E+17 1.27E+ 18 3.94E+ l7 3.94E+17 Cooper Nuclear Station PTLR ER 20 16-042, Rev 2 Page 31 of 33 dRTNoT Margin Terms Total Ma l1!in

!"Fl C'6 (°FI 171 l°FI l'FI 69.3 17.0 0.0 34.0 73.8 17.0 00 34.0 69.3 17.0 00 34.0 77.7 8.5 00 17.0 134.2 8.5 0.0 17.0 47.9 17.0 0.0 34.0 114.4 28.0 00 56.0 114.4 28.0 0.0 56.0 114.4 28.0 0.0 56.0 92.2 28.0 0.0 56.0 92.2 28.0 0.0 56.0 92.2 28.0 0.0 56.0 80.2 280 0.0 56.0 23.7 8.3 00 16.5 30.4 10.6 0.0 21.2 Fluence Factor, FF f(O,?N

• O.IU log f) 0.453 0.453 0.453 0.520 0.520 0.520 0.450 0.450 0.450 0.399 0.399 0.399 0.466 0.257 0.257 ART l'FI 117.3 107.8 95.3 104.7 131.2 71.9 120.4 120.4 120.4 98.2 98.2 98.2 86.2 37.4 70.8

Appendix A Cooper Nuclear Station PTLR ER 2016-042, Rev 2 Page 32 of 33 COOPER REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [24], two surveillance capsules were removed from the CNS reactor vessel in 1985 at 6.8 EFPY and 1991 at 11.2 EFPY [25, Attachment 3]. The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region.

CNS is currently committed to use the BWRV1P ISP, and has made a licensing commitment to use the ISP for CNS during the period of extended operation. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by NRC. Nebraska Public Power District committed to use the ISP in place of its existing surveillance programs in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated October 31, 2003 [26].

Additionally, CNS served as a host plant for three of the nine surveillance capsules irradiated as part of the Supplemental Surveillance Program; the SSP-A, SSP-8, and SSP-C capsules were removed from CNS and tested in 2003 [27] The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. CNS continues to be a host plant under the ISP. One additional standby Cooper capsule is currently scheduled to be removed and tested under the ISP during the license renewal period in approximately 2029 at 40 EFPY [27].

~

~ Entergy ATTACHMENT 9.3 SHEET 1 OF 1 NPPD Engineering Re ort Number Quality Related:

Comment Section/ Page No.

Number NIA NIA Verified/Reviewed 8 :

Site/De rtment:

NUCLEAR MANAGEMENT MANUAL No Review Comment None QUALITY REL.Al eo 3-EN-DC-147 I REV. SC1 INFORMATIONAL Use PAGE 330F 33 TECHNICAL REVIEW COMMENTS AND RESOLUTION FORM Engineering Report Technical Review Comments and Resolutions Form Rev.

2

Title:

Cooper Nuclear Station Pressure and Temperature Limits Report PTLR for 54 Effective Full-Power Years EFPY Non-Pro rieta Special Notes or Instructions:

Response/Resolution NIA Resolved 8 Date:

Reviewer's Acee t Initials NIA

r=~r.::::11 1 ELECTRIC POWER

~I-le;;;;;.

RESEARCH INSTITUTE Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary)

Page 1 of 45 2014 TECHNICAL REPORT BWRVIP-135, Revision 3: BWR Vessel and Internals Project Integrated Surveillance Program {ISP) Data Source Book and Plant Evaluations WARNING:

Please read the Export Control Agreement on the back cover.

NOTICE: This reporl conloins proprielory information that,s the inlellecluol property of EPRI. Accordingly, it is ovoiloble only under license from EPRI and may nol be reproduced or disclosed, wholly or in port, by any licensee lo any other person or organization.

BWRVIP-135, Revision 3: BWR Vessel and Internals Project Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary)

Page 2 of 45 Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations 3002003144 Technical Report, December 2014 EPRI Project Manager R. Carter All or a portion of the requirements of the EPRI Nuclear Quality Assurance Program apply to this product.

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Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary)

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Cooper Representative Surveillance Materials Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary)

Page 4 of 45 The ISP Representative Surveillance Materials for the Cooper vessel target weld and plates are shown in the following table.

Table 2-22 Target Vessel Materials and ISP Representative Materials for Cooper Target Vessel Materials ISP Representative Materials Weld 27204/12008 20291 Plate C2307-2 C2307-2 Summary of Available Surveillance Data: Plate The representative plate material C2307-2 is contained in the following ISP capsules:

Cooper Capsules Specific surveillance data related to plate heat C2307-2 are summarized in Appendix A-2.

Two capsules containing this plate heat have been tested. The Charpy V-notch surveillance results are as follows:

Table 2-23 T30 Shift Results for Plate Heat C2307-2 Capsule Cu Ni Fluence

,Hao(oF)

(wt%)

(wt%)

(10 17 n/cm 2

, E > 1 MeV)

Cooper 30° 2.4 52.2 0.21 0.76 Cooper 300° 2.8 52.2 The results given in Appendix A-2 show a fitted chemistry factor (CF) of { {

cEJ} }, as compared to a value of 164.6°F from the chemistry tables in Reg. Guide 1.99, Rev. 2. The maximum scatter in the fitted data is { {

(El which is well within the 1-sigma value of l 7°F for plates as given in Reg. Guide 1.99, Rev. 2. Conclusions and Recommendations Because the representative plate material is the same heat number as the target plate in the Cooper vessel, and because there are two irradiated data sets for this plate that fall within the 1-sigma scatter band, the ISP surveillance data in Appendix A-2 should be used to determine the projected ART value for the target vessel plate. Recommended guidelines for use of ISP surveillance data are provided in Section 3 of this Data Source Book.

Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 5 of 45 An archival plate heat from the Cooper vessel, Plate Heat C233 l-2, was included in the Supplemental Surveillance Program (SSP), and irradiated data from SSP Capsules D, G, E, I, A, and Bare provided in Appendix A-19. The credible surveillance data should be considered when a revised ART is calculated for vessel heat C233 l-2. Summary of Available Surveillance Data: Weld The representative weld material 20291 is contained in the following ISP capsules: Cooper Capsules SSP Capsule C Specific surveillance data related to weld heat 20291 are presented in Appendix B-2 and the results are summarized below. Three capsules containing weld heat 20291 have been tested. The Charpy V-notch surveillance results are as follows: Table 2-24 T30 Shift Results for Weld Heat 20291 Capsule Cu Ni Fluence AT30(0F) (wt%) (wt%) (10 11 n/cm 2 , E > 1 MeV) Cooper 30° 2.4 60.9 Cooper 300° 0.23 0.75 2.8 63.8 SSPC 3.29 73.0 The results given in Appendix B-2 show a fitted chemistry factor (CF) of { { fEJ} }, as compared to a value of 194.5°F from the chemistry tables in Reg. Guide 1.99, Rev. 2. The maximum scatter in the fitted data is well within the I-sigma value of 28°F for welds as given in Reg. Guide 1.99, Rev. 2. Conclusions and Recommendations Because the representative weld material is not the same heat number as the target weld in the Cooper vessel, the utility should use the chemistry factor from the Regulatory Guide 1.99, Rev. 2 tables to determine the projected ART value for the target vessel weld. Cooper surveillance weld heat 20291 is not in the Cooper vessel beltline. Recommended guidelines for evaluation of ISP surveillance data are provided in Section 3 of this Data Source Book.

A-2 Plate Heat: C2307-2 Summary of Available Charpy V-Notch Test Data Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 6 of45 The available Charpy V-notch test data sets for plate heat C2307-2 are listed in Table A-2-1. The source documents for the data are provided, and the capsule designations and fluence values are also provided for irradiated data sets. Table A-2-1 ISP Capsules Containing Plate Heat C2307-2 Capsule Fluence (E> 1 MeV, 10 11 n/cm2) Reference Unirradiated Baseline Data Reference A-2-1 Cooper 300° 2.8 Cooper 30° 2.4' Reference A-2-2 1 From Reference [A-2-1], which updated and superseded the fluence provided by Reference [A-2-2] for this capsule. The CVN test data for each set taken from the references noted above are presented in Tables A-2-7 through A-2-9. The BWRVIP ISP uses the hyperbolic tangent (tanh) function as a statistical curve-fit tool to model the transition temperature toughness data. Tanh curve plots for each data set have been generated using CVGRAPH, Version 5 [A-2-3] and the plots are provided in Figures A-2-1 through A-2-3. Best Estimate Chemistry Table A-2-2 details the best estimate average chemistry values for plate heat C2307-2 surveillance material. Chemical compositions are presented in weight percent. If there are multiple measurements on a single specimen, those are first averaged to yield a single value for that specimen, and then the different specimens are averaged to determine the heat best estimate. Table A-2-2 Best Estimate Chemistry of Available Data Sets for Plate Heat C2307-2 Cu (wt%) Ni (wt%) p (wt%) s (wt%) Si (wt%) Specimen ID Source 0.21 0.73 0.010 0.014 0.20 Plate CMTR Reference A-2-2 and A-2-4 0.22 0.77 0.007 J64 Reference A-2-2 and A-2-4 0.22 0.78 0.006 J6L 0.21 0.76 0.011 J63 Reference A-2-1 0.21 0.75 0.011 J6M 0.21 0.76 0.009 0.014 0.20 +Best Estimate Average Calculation of Chemistry Factor (CF): The Chemistry Factor (CF) associated with the best estimate chemistry, as determined from U.S. NRC Regulatory Guide 1.99, Revision 2 [A-2-5], Table 2 (base metal), is: CF(C2307-2) = 164.6°F

Effects of Irradiation Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 7 of 45 The radiation induced transition temperature shifts for heat C2307-2 are shown in Table A-2-3. The T30_[30 ft-lb Transition Temperature], T50 [50 ft-lb Transition Temperature], and T35mil [35 mil Lateral Expansion Temperature] have been determined for each Charpy data set, and each irradiated set is compared to the baseline (unirradiated) index temperatures. The change in Upper Shelf Energy (USE) is also shown. The unirradiated and irradiated values are taken from the CV GRAPH fits presented at the back of this sub-appendix ( only CVN energy fits are presented). Comparison of Actual vs. Predicted Embrittlement A predicted shift in the 30 ft-lb transition temperature (~T30) is calculated for each irradiated data set using the Reg. Guide 1.99, Rev. 2, Regulatory Position 1.1 method. Table A-2-4 compares the predicted shift with the measured ~T30 (°F) taken from Table A-2-3. Comparison of Actual vs. Predicted Decrease in USE Table A-2-5 compares the actual percent decrease in upper shelf energy (USE) to the predicted decrease. The predicted decrease is estimated from USNRC Regulatory Guide 1.99, Rev. 2, Figure 2; the measured percent decrease is calculated from the values presented in Table A-2-3. Credibility of Surveillance Data The credibility of the surveillance data is determined according to the guidance of Regulatory Guide 1.99, Rev. 2 and IO CFR 50.61, as supplemented by the NRC staff [A-2-6]. The following evaluation is based on the available surveillance data for irradiated plate heat C2307-2. The applicability of this evaluation to a particular BWR plant must be confirmed on a plant-by-plant basis to verify there are no plant-specific exceptions to the following evaluation.

Table A-2-3 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties of Plate Heat C2307-2 T30, 30 ft-lb T50, 50 ft-lb T3smu* 35 mil Lateral Transition Transition Expansion Material Identity Capsule Temperature Temperature Temperature ID Unirrad lrrad AT3o Unirrad lrrad ATso Unirrad lrrad AT3sm;1 (OF) (OF) (OF) (OF) (OF) (OF) (OF) (OF) (OF) 30° -43.0 9.2 52.2 -12.8 47.7 60.q_ -27.5 7.0 34.5 CPR C2307-2 300° -43.0 9.2 52.2 -12.8 43.9 56.7 -27.5 33.0 60.5 Table A-2-4 Comparison of Actual Versus Predicted Embrittlement for Plate Heat C2307-2 RG 1.99 Rev. 2 Capsule Material Fluence Measured Shift' Predicted Shift 2 Identity (x10 18 n/cm 2 ) OF OF CPR 30° Plate Heat C2307-2 in Cooper 0.24 52.2 31.7 CPR 300° Plate Heat C2307-2 in Cooper 0.28 52.2 34.7 Notes:

1.

See Table A-2-3, ti. T,,. Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page B of 45 CVN Upper Shelf Energy (USE) Unirrad lrrad Change (ft-lb) (ft-lb) (ft-lb) 132.6 124.9 -7.7 132.6 125.8 -6.8 RG 1.99 Rev. 2 Predicted Shift+Margin 2 3 OF 63.3 68.7

2.

Predicted shift= CF x FF, where CF is a Chemistry Factor taken from tables from USN RC Reg. Guide 1.99, Rev. 2, based on each material's Cu/Ni content, and FF is Fluence Factor, 102** 0 1010, where f = fluence (10" n/cm', E > 1.0 MeV).

3.

Margin= 2-.J(cr,' + cr1,, 2 ), where cr, = the standard deviation on initial RT""' (which is taken to be 0°F), and cr1,, is the standard deviation on !!.RT""' (28°F for welds and 17°F for base materials, except that cr1,, need not exceed 0.50 times the mean value of !!.RT "0,). Thus, margin is defined as 34 °F for plate materials and 56°F for weld materials, or margin equals shift (whichever is less), per Reg. Guide 1.99, Rev. 2.

Table A-2-5 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 9 of 45 Comparison of Actual Versus Predicted Percent Decrease in Upper Shelf Energy (USE) for Plate Heat C2307-2 Measured RG 1.99 Rev. 2 Capsule Material Fluence Cu Content Decrease in Predicted Identity (x10 10 n/cm 2 ) (wt%) USE 1 (%) Decrease in USE 2 (%) CPR 30° Plate Heat C2307-2 in Cooper 0.24 0.21 5.8 12.4 CPR 300° Plate Heat C2307-2 in Cooper 0.28 0.21 5.1 12.9 Notes:

1. See Table A-2-3, (Change in USE)/(Unirradiated USE).

2. Calculated using equations in Regulatory Guide 1.162 [A-2-7] that accurately model the Charpy upper shelf energy decrease curves in Regulatory Guide 1.99, Revision 2.

Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 10 of 45 Per Regulatory Guide 1.99, Revision 2 and 10 CFR 50.61, there are 5 criteria for the credibility assessment. Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement. In order to satisfy this criterion, the representative surveillance material heat number must match the material in the vessel. Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously. Plots of Charpy energy versus temperature for the unirradiated and irradiated condition are presented in this sub-appendix. Based on engineering judgment, the scatter in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy. Hence, this criterion is met. Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ~RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than l 7°F for plates. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice that value. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [A-2-.8] For plate material C2307-2, there are 2 surveillance capsule data sets currently available. The functional form of the least squares fit method as described in Regulatory Position 2.1 is utilized to determine a best-fit line for this data and to determine if the scatter of these ~T NDT values about this line is less than l 7°F for plates. Figure A-2-4presents the best-fit line as described in Regulatory Position 2.1 utilizing the shift prediction routine from CVGRAPH, Version 5.0.2. The scatter of ~TNDT values about the functional form of the best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A-2-6. Table A-2-6 Best Fit Evaluation for Surveillance Plate Heat C2307-2 Measured Best Fit Scatter <17° F Material Fitted Capsule FF ARTNDT ARTNDT of ARTNaT (Base Metal) CF (°F) <28° F (30 ft-lb) (°F) (OF) (OF) (Weld metal) { (E)}} 30° 0.192 52.20 (( (E)}} (( (El}} Ye s C2307-2 { (E)}} 300° 0.211 52.20 (( (E)}} (( (E)}} Ye s Table A-2-6 indicates that the scatter is within acceptable range for credible surveillance data. Therefore, plate heat C2307-2 meets this criterion. Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within+/ - 25°F.

~---------------------------------------~ ----- Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 11 of 45 BWRVIP-78 [A-2-9] established the similarity of BWR plant environments in the BWR fleet. The annulus between the wall and the core shroud in the region of the surveillance capsules contains a mix of water returning from the core and feedwater. Depending on feedwater temperature, this annulus region is between 525°F and 535°F. This location of specimens with respect to the reactor vessel beltline is designed so that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperature will not differ by more than 25°F. Any plant-specific exceptions to this generic analysis should be evaluated. Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material. Few ISP capsules contain correlation monitor material. Generally, this criterion is not applicable. For plate heat C2307-2, these criteria are satisfied (or not applicable). The surveillance data are nominally credible because the scatter criterion is met. Prior to application of the data, a plant should verify that no plant-specific exceptions to these criteria exist. Table A-2-7 Unirradiated Charpy V-Notch Results for Surveillance Plate C2307-2 (LT) Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear EP4 -100 8 6 9 EPE -60 10 11 14 EPP -50 21 20 16 ET4 -40 41 35 28 EPL -40 33 29 25 EPK -30 44.5 37 30 EPJ -20 45.5 40 30 EUK 20 72.5 60 42 ETE 60 108 79 75 EU5 100 114 83 87 EUA 150 132 92 100 EUB 200 133 88 100

Table A-2-8 Charpy V-Notch Results for C2307-2 (LT) in CPR 30° Capsule Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear EUD -20 13.5 22.0 10 ETK 0 27.5 31.0 10 EPM 10 32.5 39.0 10 EPC 20 38.5 42.0 15 EPA 40 45.0 49.0 30 ETT 60 55.0 49.0 40 EUC 80 73.0 64.0 50 EU1 120 86.5 64.0 85 EP7 160 112.0 88.0 80 EP3 200 117.7 78.0 90 EU6 300 121.7 93.0 90 ETB 400 125.3 95.0 100 Table A-2-9 Charpy V-Notch Results for C2307-2 (LT) in CPR 300° Capsule Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear EP1 -20 18 12 11 EU? 0 24 18 27 EP6 20 42.5 33 25 EPD 60 53.5 46 30 EUJ 100 91 68 63 ETD 150 111 83 83 EU3 200 119 88 100 EU4 300 125 76 100 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 12 of 45

Tanh Curve Fits of CVN Test Data for Plate Heat C2307-2 PLATE HEAT C2307-2 (CPR) Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/19/2002 08:55 AM Page I Coefficients of Curve l A= 67,57 B = 65.07 C = 79.1 TO= 9.01 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=l32.6 Lower ShelfEnergy=2.5(Fixed) Temp@30 ft-lbs=-43.0 Deg F Temp@50 ft-lbs=-12.8 Deg F Plant: Cooper Material: SA53381 Heat: C2307-2 Orientation: LT Capsule: UNIRRA Fluence: 0.0 n/cm"2 300.--~-~- __... 1 -~- --1 -------,---- -- : ~ :: t 11-_ -l ~~~-----~--1--~-j ~ I I I I g t 1 1 1

u.

I i I i I l'" 0 r--r T ~r- ---:--*__L_ ___ I z I I I I I ~ *:r~_T_~r __ ~- I --+-~~---1_ _ I i I I 0 -1====:l:=-::-:::_-::_, __ o, __ (._ ____ J_.~-- ~---J __, __ J ___, ---

• 300

-200 Temperature - 100. 00 - 60. 00 - 50. 00 - 40. 00 - 40. 00 - 3 0. 00 - 20. 00

20. 00

60. 00 I 00. 00 15 0. 00 Figure A-2-1

-100 0 100 200 300 Temperature in Deg F Charpy V-Notch Data lnputCVN Computed CVN

8. 00 IO. 2 8 IO. 00

21. 85 2 I. 00 26.40 4 I. 00 3 I. 7 3 3 3. 00 3 I. 7 3

44. 5 0 3 7. 8 6

45. 50 44.73

72. 50

76. 56 I 08. 00 I 04. 5 4 I 14. 00 12 0. 79 132. 00 129. 0 6 Charpy Energy Data for Plate C2307-2 (LT) Unirradiated 400 500 600 Differential

- 2. 2 8 - I I. 8 5 - 5. 4 0 9.27

l. 2 7

6. 64

. 77 - 4. 0 6 3.46 - 6. 79

2. 94

Temperature 200. 00 Figure A-2-1 PLATE HEAT C2307-2 (CPR) Page 2 Plant: Cooper Material: SA533Bl Heat: C2307-2 Orientation: LT Capsule: UNIRRA Fluence: 0.0 n/cmA2 Charpy V-Notch Data lnputCVN 13 3. 00 Correlation Coefficient=.992 Computed CVN 13 l. 61 Charpy Energy Data for Plate C2307-2 (LT) Unirradiated (Continued) Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Pa e 14 of45 Differential

l. 3 9

Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) IRRADIATED PLATE HEAT C2307-2 (CPR-30) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/22/2003 08:24 PM Page 1 Coefficients of Curve I A= 63.68 B = 61.18 C = 98.32 TO= 70.05 D = O.OOE+oO Equation is A+ B * [Tanh((T-To)/(C+DT))) Upper ShelfEnergy=l24.9 Lower ShelfEnergy=2.5(Fixed) Ternp@30 ft-lbF9.2 Deg F Ternp@SO ft-lbs=47. 7 Deg F Plant: COOPER Material: SA533Bl Heat: C2307-2 Orientation: LT Capsule: 30 DEG Fluence: 2.4E+ 17 n/crn"2 aoo r-*-r*-- t I _ T ____ - / I i .1 I 1 l I 250 1 I I ---1-- --** --- ----1-- 1---1 ~ 200 +-----~--!--~-- -r-------L--~-----i o j I I f I 1 i*** ! I I +- f ___ ! --! ~ j I I r; 100 r-* --<-------+- r-T--j j I I 50 t --+-----+----! _____ l ___ ] l I I O 1==::!==:::t=:...._-L~---L-----t--

---

L~- L--- ___ J

-300 Temperature - 20. 00 . 00 I 0. 00 20.00 40.00

60. 00

80. 00 120. 0 0 160. 00 Figure A-2-2

-200 -100 0 100 200 300 Temperature in Deg F Charpy V-Notch Data lnputCVN

13. 50 27.50 32.50 3 8. 50

45. 00 55.00

73. 00

86. 50 I 12. 00 Computed CVN l 9. 3 9 26.22 3 o. 3 6

34. 97 4 5. 54 57.45

69. 85 92.34 I 07. 94 Charpy Energy Data for Plate C2307-2 (LT) in CPR 30° Capsule 400 500 600 Differential

• 5. 8 9 I. 28

2. 14

3. 5 3

• . 54

- 2. 45

3. 1 5

- 5. 84 4.06 1

Temperature 200. 00 300. 00 400.00 Figure A-2-2 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Pa e 16 of 45 IRRADIATED PLATE HEAT C2307-2 (CPR-30) Page 2 Plant: COOPER Material: SA533Bl Heat: C2307-2 Orientation: LT Capsule: 30 DEG Fluence: 2.4E+ 17 n/cm"2 Cbarpy V-Notcb Data InputCVN 11 7. 70 121. 70 125. 3 0 Correlation Coefficient=.997 Computed CVN 116. 7 3 123. 7 3 124. 71 Differential . 97 - 2. 0 3 . 59 Charpy Energy Data for Plate C2307-2 (LT) in CPR 30° Capsule (Continued)

Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) IRRADIATED PLATE HEAT C2307-2 (CPR-300) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/22/2003 08:26 PM Page I Coefficients of Curve I A = 64.13 B = 61.63 C = 88.9 TO = 64.63 D = O.OOE+OO Equation is A+ B "'[Tanh((T*To)/(C+DT))] Upper ShelfEnergy=l25.8 Lower ShelfEnergy=2.5(Fixed) Temp@30 ft-lbs=9.2 Deg F Temp@50 ft-lbs=43.9 Deg F Plant: COOPER Material: SA533B I Heat: C2307-2 Orientation: LT Capsule: 300 DE Fluence: 2.SE+ 17 n/cm"2 31JO I I I -r- -- 1 -i--r 250;

• -i--t--+-- l)~-1-1

~ 200' I I ----+---+----r---t---*--: ! \\ I I I I -+' e,1sot--r--t--*--r--- 1 - --1 l I I I ~ 100 1* II

1.

--t-~ I I I i -----r-r*---t-1---: O 4===1=='.:+:::__._-J. ___ J __ __ L_ 1 _._J_ __ J -300 Temperature -20.00 . 00 20.00 60.00 I 00. 00 150. 00 200. 00 300. 00 Figure A-2-3 -200 -100 0 100 200 300 Temperature in Deg F Cbarpy V-Notch Data InputCVN I 8. 00 24.00 42.50

53. 50

91. 00 111.00 119. 00 125. 00 Correlation Coefficient =.995 Computed CVN

18. 48 25.84 3 5. 55 60.92 87.43 110. 0 I 120. 16 125. 15 Charpy Energy Data for Plate C2307-2 (LT) in CPR 300° Capsule 400 500 600 Differential

-.48 - I. 84 6.95 -7.42

3. 5 7

. 99 - I. 16 -. I 5 4

(( Figure A-2-4 Fitted Surveillance Results for Plate Heat C2307-2 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 18 of45

References Appendix 8, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 19 of45 A-2-1. GE Nuclear Energy, "Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis," GE-NE-523-159-1292, February 1993. A-2-2. "Cooper Nuclear Station Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis," T.A. Caine, B.J. Branlund, and S. Ranganath, General Electric, MDE-103-0986, DRF B13-01389, May 1987. A-2-3. CVGRAPH, Hyperbolic Tangent Curve Fitting Program, Developed by ATI Consulting, Version 5.0.2, Revision 1, 3/26/02. A-2-4. Letter from G.R. Horn (NPPD) to USNRC, "Response to Generic Letter 92-01, Revision 1, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-44," Nebraska Public Power District, NSD920629, dated July 1, 1992. A-2-5. "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988. A-2-6. K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. A-2-7. "Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels," USNRC Regulatory Guide 1.162, February 1996. A-2-8. ASTM E-185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, July 1982. A-2-9. BWR Vessel and Internals Project: BWR Integrated Surveillance Program Plan (BWRVIP-78). EPRI, Palo Alto, CA and BWRVIP: 1999. TR-114228.

A-19 Plate Heat: C2331-2 Summary of Available Charpy V-Notch Test Data Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 20 of 45 The available Charpy V-notch test data sets for plate heat C2331-2 are listed in Table A-19-1. The source documents for the data are provided, and the capsule designation and fluence values are also provided for irradiated data sets. Table A-19-1 ISP Capsules Containing Plate Heat C2331-2 Capsule Fluence (E> 1 MeV, 10 17 n/cm 2 ) Reference Unirradiated Baseline Data Reference A-19-1 SSPD 10.118 Reference A-19-2 SSPG 18.487 SSP E 17.192 Reference A-19-3 SSPI 27.085 SSPA 3.82 Reference A-19-12 SSP B 4.79 Reference A-19-12 The CVN test data for each set taken from the references noted above are presented in Tables A-19-7 through A-19-13. The BWRVIP ISP uses the hyperbolic tangent (tanh) function as a statistical curve-fit tool to model the transition temperature toughness data. Tanh curve plots for each data set have been generated using CVGRAPH, Version 5 [A-19-4] and the plots are provided in Figures A-19-1 through A-19-7. Best Estimate Chemistry Table A-19-2 details the best estimate average chemistry values for plate heat C2331-2 surveillance material. Chemical compositions are presented in weight percent. If there are multiple measurements on a single specimen, those are first averaged to yield a single value for that specimen, and then the different specimens are averaged to determine the heat best estimate.

Table A-19-2 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 21 of 45 Best Estimate Chemistry of Available Data Sets for Plate Heat C2331-2 Cu (wt%) Ni (wt%) P(wt%) s (wt%) Si (wt%) Specimen ID Source 0.17 0.58 0.01 0.017 0.22 CMTR Reference A-19-5 0.15 0.69 0.022 0.023 0.25 SSP 0.15 0.64 0.012 SSP Reference A-19-1 0.15 0.665 0.017 0.023 0.25 SSP Average 0.16 0.62 0.014 0.020 0.24 +Best Estimate Average Calculation of Chemistry Factor (CF): The Chemistry Factor (CF) associated with the best estimate chemistry, as determined from U.S. NRC Regulatory Guide 1.99, Revision 2 [A-19-6], Table 2 (base metal), is: CF(C2331-2) = 118.5°F Effects of Irradiation The radiation induced transition temperature shifts for heat C2331-2 are shown in Table A-19-3. The T30 [30 ft-lb Transition Temperature], T50 [50 ft-lb Transition Temperature], and T35mil [35 mil Lateral Expansion Temperature] have been determined for each Charpy data set, and each irradiated set is compared to the baseline (unirradiated) index temperatures. The change in Upper Shelf Energy (USE) is also shown. The unirradiated and irradiated values are taken from the CVGRAPH fits presented at the end of this sub-appendix (only CVN energy fits are presented). Comparison of Actual vs. Predicted Embrittlement A predicted shift in the 30 ft-lb transition temperature (i1T30) is calculated for each irradiated data set using the Reg. Guide 1.99, Rev. 2, Regulatory Position 1.1 method. Table A-19-4 compares the predicted shift with the measured i1T30 (°F) taken from Table A-19-3. Decrease in USE Table A-19-5 shows the percent decrease in upper shelf energy (USE). The measured percent decrease is calculated from the values presented in Table A-19-3.

Table A-19-3 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 22 of 45 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties of Plate Heat C2331-2 T30, 30 ft-lb T50, 50 ft-lb Tasm;i, 35 mil Lateral CVN Upper Shelf Energy Transition Transition Material Capsule Temperature Temperature Expansion Temperature (USE) Identity ID Unirrad lrrad ATao Unirrad lrrad ATso Unirrad lrrad ATasm11 Unirrad lrrad Change (OF) (OF) (OF) (OF) (OF) (OF) (OF) (OF) (OF) (ft-lb) (ft-lb) (ft-lb) SSP D -13.3 48.7 62.0 30.1 92.8 62.7 34.1 86.3 52.2 100.0 89.3 -10.7 SSP G -13.3 78.7 92.0 30.1 127.2 97.1 34.1 118.2 84.1 100.0 81.6 -18.4 CPR C2331-2 SSP E -13.3 62.8 76.1 30.1 105.8 75.7 34.1 124.2 90.1 100.0 82.3 -17.7 SSPI -13.3 80.4 93.7 30.1 128.8 98.7 34.1 128.3 94.2 100.0 80.3 -19.7 SSPA -13.3 28.2 41.5 30.1 77.9 47.8 34.1 44.4 10.3 100.0 91.0 -9.0 SSP B -13.3 21.4 34.7 30.1 62.5 32.4 34.1 39.2 5.1 100.0 97.7 -2.3 Table A-19-4 Comparison of Actual Versus Predicted Embrittlement for Plate Heat C2331-2 RG 1.99 Rev. 2 RG 1.99 Rev. 2 Capsule Material Fluence Fluence Factor Measured Shift 1 Predicted Shift' Predicted Identity (x10 18 n/cm 2 ) OF OF Shift+Margin 2 3 OF SSP D 1.0118 0.419 62.0 49.7 83.7 SSPG 1.8487 0.551 92.0 65.3 99.3 SSP E Plate Heat C2331-2 1.7192 0.534 76.1 63.3 97.3 SSPI from Cooper 2.7085 0.644 93.7 76.3 110.3 SSPA 0.382 0.252 41.5 29.9 59.8 SSP B 0.479 0.286 34.7 33.9 67.8 Notes:

1. See Table A-19-3,,H30 *

2. Predicted shift= CF x FF, where CF is a Chemistry Factor taken from tables from USN RC Reg. Guide 1.99, Rev. 2, based on each material's Cu/Ni content, and FF is Fluence Factor, 1'"'*0*

1"" ', where f = fluence (1 O" n/cm', E > 1.0 MeV).

3. Margin= 2../(cr,' + 0 6 2), where er,= the standard deviation on initial RT NOT (which is taken to be 0°F), and era is the standard deviation on LiRT NoT (28°F for welds and 17°F for base materials, except that er6 need not exceed 0.50 times the mean value of t,,RTN0T). Thus, margin is defined as 34°F for plate materials and 56°F for weld materials, or margin equals shift (whichever is less), per Reg. Guide 1.99, Rev. 2.

Table A-19-5 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 23 of 45 Comparison of Actual Versus Predicted Percent Decrease in Upper Shelf Energy (USE) for Plate Heat C2331-2 Measured RG 1.99 Rev. 2 Fluence Cu Content Predicted Capsule Identity Material (x10 18 n/cm 2 ) (wt%) Decrease in Decrease in USE 1 (%) USE 2 (%) SSP D 1.0118 0.16 10.7 14.5 SSP G 1.8487 0.16 18.4 16.8 SSP E 1.7192 0.16 17.7 16.5 Plate Heat C2331-2 in Cooper SSPI 2.7085 0.16 19.7 18.3 SSPA 0.382 0.16 9.0 11.5 SSP B 0.479 0.16 2.3 12.2 Notes:

1. See Table A-19-3, (Change in USE)/(Unirradiated USE).

2. Calculated using equations in Regulatory Guide 1.162 [A-19-7] that accurately model the Charpy upper shelf energy decrease curves in Regulatory Guide 1.99, Revision 2.

Credibility of Surveillance Data Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 24 of 45 The credibility of the surveillance data is determined according to the guidance of Regulatory Guide 1.99, Rev. 2 and 10 CFR 50.61, as supplemented by the NRC staff [A-19-8]. The following evaluation is based on the available surveillance data for irradiated plate heat C2331-2. The applicability of this evaluation to a particular BWR plant must be confirmed on a plant-by-plant basis to verify there are no plant-specific exceptions to the following evaluation. Per Regulatory Guide 1.99, Revision 2 and 10 CFR 50.61, there are 5 criteria for the credibility assessment. Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement. In order to satisfy this criterion, the representative surveillance material heat number must match the material in the vessel. Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously. Plots of Charpy energy versus temperature for the unirradiated and irradiated condition are presented in Figures A-19-1 through A-19-7. Based on engineering judgment, the scatter in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy. Hence, this criterion is met. Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LiRTNoT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than l 7°F for plates. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice that value. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [A-19-9]. For plate material C2331-2, there are 6 surveillance capsule data sets currently available. The functional form of the least squares fit method as described in Regulatory Position 2.1 is utilized to determine a best-fit line for this data and to determine if the scatter of these LiRTNoT values about this line is less than l7°F for plates. Figure A-19-8 presents the best-fit line as described in Regulatory Position 2.1 utilizing the shift prediction routine from CVGRAPH, Version 5.0.2. The scatter of LiRTNoT values about the functional form of the best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A-19-6.

Table A-19-6 Best Fit Evaluation for Surveillance Plate Heat C2331-2 Measured Material Fitted Capsule FF ART NOT CF (°F) (30 ft-lb) (°F) SSP D 0.419 62.0 SSP G 0.551 92.0 SSP E 0.534 76.1 C2331-2 { (E)}} SSPI 0.644 93.7 SSPA 0.252 41.5 SSP B 0.286 34.7 Best Fit ART NOT (OF) (( (El} (( (El}} (( (El}} (( (E)}} (( (El}} (( (E)}} Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 25 of 45 Scatter of <17°F (Base Metal) ART NOT <28°F (OF) (Weld metal) (( (El}} Yes (( (El}} Yes (( (El}} Yes (( (El}} Yes (( (El}} Yes (( (El}} Yes Table A-19-6 indicates that the scatter is within acceptable range for credible surveillance data. Therefore, plate heat C233 l-2 meets this criterion. Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within+/ - 25°F. BWRVIP-78 [A-19-11] established the similarity of BWR plant environments in the BWR fleet. The annulus between the wall and the core shroud in the region of the surveillance capsules contains a mix of water returning from the core and feedwater. Depending on feedwater temperature, this annulus region is between 525°F and 535°F. This location of specimens with respect to the reactor vessel beltline is designed so that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperature will not differ by more than 25°F. Any plant-specific exceptions to this generic analysis should be evaluated. Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material. Few ISP capsules contain correlation monitor material. Generally, this criterion is not applicable. For plate heat C233 l-2, these criteria are satisfied (or not applicable). The surveillance data are nominally credible because the scatter criterion is met. Prior to application of the data, a plant should verify that no plant-specific exceptions to these criteria exist.

Table A-19-7 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 26 of 45 Unirradiated Charpy V-Notch Results for Surveillance Plate C2331-2 (TL) Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear SSP 1 -80 12.0 5.0 3 SSP2 -60 15.5 5.0 0 SSP3 -40 24.5 12.5 19 SSP4 -20 20.0 13.0 16 SSP5 -20 31.5 20.0 20 SSP6 0 43.5 28.5 23 SSP7 20 46.0 29.5 30 SSP8 40 52.5 32.5 49 SSP9 60 53.5 37.0 47 SSP10 60 49.5 37.0 44 SSP 11 80 91.5 , 67.5 87 SSP12 100 86.0 63.0 89 SSP13 180 97.0 70.0 100 SSP14 300 97.0 73.0 100 SSP15 400 106.0 73.5 100 Table A-19-8 Charpy V-Notch Results for C2331-2 (TL) in SSP Capsule D Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear 1 0 9 8 5 2 25 24.5 21 10 3 50 28 19 15 4 75 50.5 39 20 5 100 49 37 40 6 150 68 51 90 7 200 83.75 57 100 8 250 92.5 65 100 9 300 94 71 100 10 400 87 71 100

Table A-19-9 Charpy V-Notch Results for C2331-2 (TL) in SSP Capsule G Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear 1 25 18.5 15 5 2 75 33.5 24 20 3 100 35.5 27 10 4 125 41.5 35 35 5 140 51.5 37 40 6 150 63.5 47 70 7 200 76 57 100 8 250 84.5 58 100 9 300 83 70 100 10 400 83 65 100 Table A-19-10 Charpy V-Notch Results for C2331-2 (TL) in SSP Capsule E Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear EP130C 0 10.5 1 0 EP130E 40 28.0 12 15 EP130A 70 27.5 17 30 EP130H 100 49.0 27 55 EP130D 125 54.5 32 50 EP130F 150 68.5 43 90 EP130B 200 80.0 55 100 EP130G 225 82.0 54 100 EP1301 250 81.5 61 100 EP130J 300 85.5 59 100 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 27 of 45

Table A-19-11 Charpy V-Notch Results for C2331-2 {TL) in SSP Capsule I Spec ID Temp {°F) CVN (ft-lb) LE (mils) %Shear IP130B 0 8.5 2.0 0 IP130J 30 20.0 8.0 5 IP130A 70 27.5 14.0 20 IP130H 100 30.0 18.0 30 IP130G 125 50.5 37.0 55 IP130C 150 60.0 46.0 65 IP130D 200 71.5 54.0 85 IP1301 250 77.5 59.0 100 IP130E 300 83.0 69.0 100 IP130F 400 80.5 64.0 100 Table A-19-12 Charpy V-Notch Results for C2331-2 (TL) in SSP Capsule A Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear AP1-30-10 -40.36 10.07 10.5 8.3 AP1-30-8 -20.56 15.78 16.0 11.9 AP1-30-7 19.94 30.17 28.5 21 AP1-30-9 19.94 33.14 30.5 20.7 AP1-30-1 67.64 39.22 39.0 26.9 AP1-30-2 110.84 57.99 52.0 47.4 AP1-30-3 160.70 85.95 73.0 99 AP1-30-4 250.88 88.94 77.0 100 AP1-30-5 300.74 90.17 73.0 100 AP1-30-6 399.56 99.00 76.5 100 Appendix 8, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 28 of 45

Table A-19-13 Charpy V-Notch Results for C2331-2 (TL) in SSP Capsule B Spec ID Temp (°F) CVN (ft-lb) LE (mils) %Shear BP1-30-8 -20.20 10.03 10.0 9.3 BP1-30-10 0.32 27.15 26.0 16.6 BP1-30-7 20.48 35.30 31.0 19 BP1-30-1 68.00 46.36 39.5 36.2 BP1-30-9 89.60 60.70 55.5 37.7 BP1-30-2 120.74 82.25 66.0 73.8 BP1-30-3 180.32 90.96 72.0 100 BP1-30-4 249.44 100.08 81.0 100 BP1-30-5 299.66 99.12 77.0 100 BP1-30-6 400.28 100.70 74.5 100 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 29 of 45

Tanh Curve Fits of CVN Test Data for Plate Heat C2331-2 UNIRRADIATED PLATE HEAT C2331-2 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 30 of 45 CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 06/23/2003 03:05 PM Page I Coefficients of Curve I A= 51.25 B = 48.75 C = 98.29 TO= 32.52 D = O.OOE+oO Equation is A+ B * [Tanh((T-To)/(C+D1))] Upper ShelfEnergy=IOO.O(Fixed) Lower ShelfEnergy=2.5(Fixed) Temp@30 ft-Jbs=-13.3 Deg F Temp@50 ft-lbs=JO. I Deg F Plant: Cooper Material: SA533BI Heat: C2331-2 Orientation: TL Capsule: UNIRRA Fluence: 0 n/cm"2 300~--~ 250---+--T+ -1 +- ,.., r -+ -H. e, 150 ------+--!----+---*-1---+--1-- GI C ~ II l l () 100 4----+----- ----* -- ---+--rd:=---,~-+---+---, I I ! 50 +---+---"-----+~'-W---'----+-----'-----1--------i 0 J,,..,,=+/-=:::::J=--~..L- ' -300 -200 -100 0 100 200 300 Temperature in Deg F Charpy V-Notch Data Temperature lnputCVN Computed CVN -80.00

12. 00 I l. 4 7

- 60. 00

15. 50

15. 3 8

- 40. 00 24.50 20.64 - 20. 00 20.00 27.43 - 20. 00 3 l. 5 0 27.43 . 00

43. 5 0 3 5. 68 20.00 46.00 45.07

40. 00

52. 5 0

54. 95 60.00

53. 50

64. 53 Figure A-19-1 Cooper Unirradiated Plate Heat C2331-2 Charpy Energy Plot 400 500 600 Differential

. 5 3 . 12

3. 8 6

- 7. 4 3

4. 07

7. 82

. 93 - 2. 45 - I I. 03

Temperature

60. 00

80. 00 l 00. 00 18 0. 00 300.00 400. 00 Figure A-19-1 UNlRRADIATED PLATE HEAT C2331-2 Page 2 Plant: Cooper Material: SA533Bl Heat: C233I-2 Orientation: TL Capsule: UNIRRA Fluence: 0 n/cm"2 Charpy V-Notch Data InputCVN

49. 5 0

91. 5 0 8 6. 00

97. 00

97. 00 l 06. 00 Correlation Coefficient ~. 969 Computed CVN

64. 5 3

73. 12

80. 29 95.38

99. 5 8 99.94 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary)

Differential - 15. 0 3 I 8. 3 8

5. 71 I. 62

- 2. 5 8

6. 06 Cooper Unirradiated Plate Heat C2331-2 Charpy Energy Plot (Continued)

Plate Heat C2331-2 in SSP-D Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 32 of 45 CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/01/2006 05:51 PM Pagel Coefficients of Curve I A= 45.9 B = 43.4 C = 92.06 TO= 84.06 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+D1))] Upper Shelf Energy=89.3(Fixed) Lower Shelf Energy=2.5(Fixed) Temp@JO ft-lbs=48.7 Deg F Temp@50 ft-lbs=92.8 Deg F Plant: Oyster Creek Material: SA533B 1 Heat: C233 l-2 Orientation: TL Capsule: SSP-D Fluence: I.Ol 18E+18 n/cm"2 300 i-* ---r- -i --- - *1 *--*: ----- r--- ___ J_ ---* 1-*-----+-- i i T i 250 }-----.1 ** -/-- I --- *1 /-- 2 200 L I i i "T -*-.1 0 ~ i!;j 150 i---- cii f C ~ w z (i 100 50 +----+- __ _L_ ___ *----- , --- ---- -+ ---/-- --- 1 -----

.1 ---

/ --- ---t-: - ___ I_ ---- ------ '------~ t + -: + _;..i -~ 0 -=-l_--_---nf,-____ -_*_-;--: _* _____,-1 I / I i ~--***1 I 01====-----,,*-*="*-=-~--~-::::::_

• • -* f 400

-300 Temperature . 0 0

25. 00 5 0. 0 0 7 5. 0 0 I 00. 00 150. 00 200. 00 250. 00 300. 00 Figure A-19-2

-200 -100 0 100 200 300 Temperature in Deg F Charpy V-Notch Data lnputCVN Computed CVN

9. 00

14. 5 4

24. 50

21. 34 2 8. 0 0

30. 54

50. 50 4 I. 64 49.00 5 3. 3 4 6 8. 0 0

72. 5 8 8 3. 7 5

82. 8 3 9 2. 5 0 8 7. 0 0 94.00 8 8. 5 I Cooper Irradiated Plate Heat C2331-2 (SSP-D) Charpy Energy Plot 500 600 Differential

- 5. 5 4

3. I 6

- 2. 54

8. 8 6

- 4. 34 - 4. 5 8 .92

5. 5 0 5.49

Temperature 400. 00 Figure A-19-2 Plate Heat C2331-2 in SSP-D Page 2 Plant: Oyster Creek Material: SA533B I Heat: C233 l-2 Orientation: TL Capsule: SSP-D Fluence: 1.0 I I 8E+l 8 n/cm"2 Charpy V-Notch Data InputCVN 8 7. 0 0 Correlation Coefficient=.987 Computed CVN

89. 21 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary)

Pa e 33 of 45 Differential - 2. 2 I Cooper Irradiated Plate Heat C2331-2 (SSP-D) Charpy Energy Plot (Continued) l

Plate Heat C2331-2 in SSP-G Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Pa e 34 of45 CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/01/2006 05:54 PM Page 1 Coefficients of Curve 1 A = 42.05 B = 39.55 C = 93.59 TO= 108.08 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+Dn)J Upper Shelf Energy=8 l.6(Fixed) Lower ShelfEnergy=2.5(Fixed) Temp@30 ft-lbs=78.7 Deg F Temp@50 ft-lbs=!27.2 Deg F Plant: Oyster Creek Material: SA533Bl Heat: C233 l-2 Orientation: TL Capsule: SSP-G Fluence: I. 8487E+ 18 n/cm"2 Joo -r -

• -; -- -~

r---*---- *- ~-- 1 1 I I I I i -+----.!-- i -~ -_ _J_ *--*- ---** I 1---1 U) t I =!il 200 ~: -------* o I ~ e> 150 m C w z t, 100 ___ I __ _

---

*--**-*-. --*-y-***

Q i ! 0

• ----+---+

i i

• • i I

r 0,

• • • --L I

---. _. ---*+---+-*--~-* -300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential

25. 0 0

18. 5 0

13. 9 6

4. 5 4 7 5. 0 0 3 3. 50 2 8. 63

4. 8 7 l 00. 00 3 5. 5 0 3 8. 64

- 3. 14 1 25. 00 4 I. 5 0

49. 12

- 7. 62 140.00

51. 50 5 5. 04

- 3. 5 4 l 50. 00 6 3. 5 0 5 8. 6 7

4. 83 200. 00

76. 00 71.87

4. l 3 250. 00

84. 5 0

77. 96

6. 54 300. 00 8 3. 00 8 0. 3 l 2.69 Figure A-19-3 Cooper Irradiated Plate Heat C2331-2 (SSP-G) Charpy Energy Plot

Temperature 400. 00 Figure A-19-3 Plate Heat C2331-2 in SSP-G Page 2 Plant: Oyster Creek Material: SA533B 1 Heat: C233 l-2 Orientation: TL Capsule: SSP-G Fluence: l.8487E+l8 n/cm"2 Charpy V-Notch Data lnputCVN 8 3. 00 Correlation Coefficient~.982 Computed CVN 8 I. 45 Appendix 8, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 35 of 45 Differential

l. 5 5 Cooper Irradiated Plate Heat C2331-2 (SSP-G) Charpy Energy Plot (Continued)

Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Pa e 36 of 45 IRRADIATED PLATE HEAT C2331-2 (SSP-E) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 06/23/2003 03:05 PM Page 1 Coefficients of Curve 1 A= 42.4 B = 39.9 C = 83.62 TO= 89.63 D = O.OOE+oO Equation is A+ B * (Tanh((T-To)/(C+DT))] Upper ShelfEnergy=82.3(Fixed) Lower ShelfEnergy=2.5(Fixed) Temp@JO ft-lbs=62.8 Deg F Temp@SO ft-lbs=105.8 Deg F . Plant: Cooper Material: SA533Bl Heat: C233 !-2 Orientation: TL Capsule: SSP-E Fluence: l.7192E+I8 n/cm"2 300.--~---r-~~~~~~~~~--.~~~~~~--.--~--,, I 250 +----f----1--~-+---'----+-----'-----,-----'----i I ~ 200 +---+---+---+------!----'-----'-----+---~ ~ +-----+-! I I I ~1~ I j I I t> 100 1 - f ! : ---+----t------- 1 I 50+---+---+---+-----~-----!f-----'----------1 0 I o -l===l===t==::::::_i_ __ i.___~-L~~-i ---_L_ -300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature lnputCVN Computed CVN Differential . 00 l O. 50 l 0. 87 -.37

40. 00 2 8. 00 2 I. 16

6. 84

70. 00

27. 50 3 3. 20

- 5. 70 I 00. 00 49.00 4 7. 32 I. 68 125. 00

54. 50 5 8. 34

- 3. 84 15 0. 00

68. 50

67. 06 I. 44 200.00

80. 00

76. 98

3. 02 225.00

82. 00 79.29

2. 71 250.00

81. 50

80. 61

. 89 Figure A-19-4 Cooper Irradiated Plate Heat C2331-2 (SSP-E) Charpy Energy Plot

Temperature 300. 00 Figure A-19-4 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) 7 f4 IRRADIATED PLATE HEAT C2331-2 (SSP-E) Page 2 Plant: Cooper Material: SA533Bl Heat: C2331-2 Orientation: TL Capsule: SSP-E Fl_uence: 1.7192E+I8 n/cm"2 Charpy V-Notch Data lnputCVN

85. 50 Correlation Coefficient=.991 Computed CVN 8 I. 7 8 Differential

3. 72 Cooper Irradiated Plate Heat C2331-2 (SSP-E) Charpy Energy Plot (Continued)

IRRADIATED PLATE HEAT C2331-2 (SSP-I) Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Pa e 38 of45 CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 06/23/2003 03:05 PM Page 1 '"i" 200 t 0 if Coefficients of Curve I A= 41.4 B = 38.9 C = 91.86 TO = 108.11 D = O.OOE+oO Equation is A+ B "[Tanh((T-To)/(C+DT))] Upper ShelfEnergy=80.3(Fixed) Lower ShelfEnergy=2.5(Fixed) Ternp@30 ft-lbs=80.4 Deg F Ternp@50 ft-lbs=l 28.8 Deg F Plant: Cooper Material: SA533Bl Heat: C2331-2 Orientation: TL Capsule: SSP-I Fluence: 2. 7085E+ 18 n/crn"2 f 150 I ~ f i I ! tS 100 1 --*--~--4---~~--+-----+ i ~-¥----Ir---+----, I 50 +----'---.----+----1--9'---+----+----+-----+-----j 0 -F=======::;==:::=-+-~--l---+----l----+---1--~~ -300 -200 Temperature . 00 3 0. 00 70.00 100. 00 125. 00 15 0. 00 200. 00 250.00 300.00 Figure A-19-5 -100 0 100 200 300 Temperature in Deg F Charpy V-Notch Data lnputCVN Computed CVN 8.50 9.25 20.00

14. 51

27. 50 2 6. 13 30.00 37.97 50.50 48.47 60.00 5 8. 00

71. 5 0 7 I. 03

77. 50

76. 91

83. 00 7 9. 13 Cooper Irradiated Plate Heat C2331-2 (SSP-1) Charpy Energy Plot 400 500 600 Differential

-. 7 5 5.49 I. 3 7 - 7. 97 2.03 2.00 . 4 7 . 59

3. 87

Temperature 400. 00 Figure A-19-5 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) f4 IRRADIATED PLATE HEAT C2331-2 (SSP-1) Page 2 Plant: Cooper Material: SA533Bl Heat: C2331-2 Orientation: TL Capsule: SSP-I Fluence: 2.7085E+l8 nJcmA2 Charpy V-Notch Data InputCVN

80. 50 Correlation Coefficient=.992 Computed CVN

80. 17 Differential

. 3 3 Cooper Irradiated Plate Heat C2331-2 (SSP-1) Charpy Energy Plot (Continued)

Plate Heat C2331-2 in SSP Capsule A Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 40 of 45 CV GRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 06/07/2006 12: 13 PM Page I 300 250 iJ 200 "T 0 ~ 150 Q) C: w z () 100 50 Coefficients of Curve I A=* 46.75 B = 44.25 C = 105.31 TO= 70.12 D = O.OOHOO Equation is A+ B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=91.0(Fixed) Lower ShelfEnergy=2.5(Fixcd) Temp@30 rt-lbF28.2 Deg F Temp@50 ft-lbs=77.9 Deg F Plant: COOPER Material: SA533Bl Heat: C2331-2 Orientation: Tl. Capsule: SSP-A Fluence: N/A n/cmA2 I o* 0 0 0 0 1--,,-....,..,,----,-...,--:-::-~ -300 -200 -100 0 100 200 300 400 Temperature in Deg F Charpy V-Notch Data Temperature lnputCVN Computed CVN . 4 0. 3 6 IO. 0 7

12. I 7

- 20. 56 I 5. 7 8 I 5. 92 19.94 3 0. I 7 2 7. I J I 9. 94 3 3. 1 4 2 7. I 3 6 7. 6 4 3 9. 22 4 5. 7 I I IO. 8 4 5 7. 99 6 3. 06 I 6 0. 7 0 8 5. 9 5 7 7. 5 6 250. 88 8 8. 94 8 8. 23 300. 74 9 u. I 7

89. 91 Figure A-19-6 Cooper Irradiated Plate Heat C2331-2 (SSP-A) Charpy Energy Plot 500 600 Differential

- 2. I 0 -. 14

3. 0 4

6. 0 I

- 6. 4 9 - 5. 0 7

8. 3 9

. 71 .26

Tl!mpcramre 399. 56 Figure A-19-6 Plate Heat C2331-2 in SSP Capsule A Page 2 Plant: COOPER Material: SA533Bl Heat: C2331-2 Orientation: TL Capsule: SSP-A Fluence: NIA n/cm"2 Charpy V-Notch Data Input CVN 9 9. 00 Correlation Coefficient -.9&9 Computed CVN 9 0. 8 3 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 41 of45 Differential

8. I 7 Cooper Irradiated Plate Heat C2331-2 (SSP-A) Charpy Energy Plot (Continued)

Plate Heat C2331-2 in SSP Capsule B Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 42 of 45 CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 06/07/2006 12: l 7 PM Page I 300 250 il 200... 0 if ~ 150 Q) C w z (j 100 50 !* 0 -300 Temperature

• 20. 20

.32 2 0. 4 8 6 8. 00 8 9. 6 0 I 2 0. 7 4 180. 3 2 249. 44 299. 66 Figure A-19-7 Coefficients of Curve I A= 50.1 B = 47.6 C = 91.63 TO= 62.66 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+DD)l Upper Shelf Encrgy=97. 7(Fixed) Lower Shelf Energy=2.5(Fixed) Temp@30 ft-lhs=2 I.4 Deg F Temp@50 ft-lbs=62.5 Deg F Plant: COOPER Material: SA533BI Heat: C2331-2 Orientation: TL Capsule: SSP-8 Fluence: NIA n/cm"2 -200 -100 f :a C!l

a.

0 a ! I 0 100 0 200 300 Temperature in Deg F Charpy V-Notch Data JnputCVN Computed CVN IO. 0 3 I 5. 90 2 7. I 5 2 I. 9 3 J 5. 3 0 2 9. 6 1 46.36

52. 8 7

60. 7 0 6 3. 70

82. 25 7 6. 79

90. 9 6

90. 9 2 I 00. 0 8 9 6. I I

99. t 2

97. I 6 400 Cooper Irradiated Plate Heat C2331-2 (SSP-B) Charpy Energy Plot 500 600 Differential

- 5. 8 7

5. 2 2

5. 69

- 6. 5 I - 3. 00 5.46 . 04

3. 97

1. 96

T ernperature 400. 28 Figure A-19-7 Plate Heat C2331-2 in SSP Capsule B Page 2 Plant: COOPER Material: SA533B I Heat: C233 1-2 Orientation: TL Capsule: SSP-B Fluence: NIA n/cm"2 Charpy V-Notch Data lnputCVN IO O. 70 Correlation Coefficient =.991 Computed CVN g 7. 64 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 43 of 45 Differential

3. 06 Cooper Irradiated Plate Heat C2331-2 (SSP-B) Charpy Energy Plot (Continued)

(( Figure A-19-8 Fitted Surveillance Results for Plate Heat C2331-2 Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page 44 of 45 (El}}

References Appendix B, ER 2016-042 Cooper PTLR (Non-Proprietary) Page45 of 45 A-19-1. "Progress Report on Phase 2 of the BWR Owners' Group Supplemental Surveillance Program," T.A. Caine, S. Ranganath, and S.J. Stark, GE Nuclear Energy, GE-NE-523-99-0792, January 1992. A-19-2. BWRVIP-87NP, Revision 1: BWR Vessel and Internals Project Testing and Evaluation of BWR Supplemental Surveillance Program Capsules D, G, and H. EPRI, Palo Alto and BWRVIP: 2010. 1021553. A-19-3. BWRVIP-l J JNP, Revision 1: BWR Vessel and Internals Project, Testing and Evaluation of BWR Supplemental Surveillance Program Capsules E, F and I. EPRI, Palo Alto, CA: 2010. 1021554. A-19-4. CV GRAPH, Hyperbolic Tangent Curve Fitting Program, Developed by A TI Consulting, Version 5.0.2, Revision 1, 3/26/02. A-19-5. General Electric, "Cooper Nuclear Station Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis," T.A. Caine, BJ. Branlund, and S. Ranganath, MDE-103-0986, May 1987. A-19-6. "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988. A-19-7. "Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels," USNRC Regulatory Guide 1.162, February 1996. A-19-8. K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. A-19-9. ASTM E-185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, July 1982. A-19-10. BWR Vessel and Internals Project: BWR Integrated Surveillance Program Plan (BWRVIP-78). EPRI, Palo Alto, CA and BWRVIP: 1999. TR-114228. A-19-11. Not used. A-19-12. BWRVIP-169NP: BWR Vessel and Internals Project, Testing and Evaluation of BWR Supplemental Surveillance Program (SSP) Capsules A, B, and C. EPRI, Palo Alto, CA: 2010. 1021556.

NLS2019040 Page 1 of 4 Affidavit for Proprietary Information Contained in the Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY) Cooper Nuclear Station Docket No. 50-298, DPR-46

fl'=:;;'~~H I ELECTRIC POWER ~u-u~9 RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietary Information Included In: NPPD, Cooper Nuclear Station, Pressure and Temperature Limits Report (PTLR) for 54 Effective Full Power Years (EFPY), ER 2016-041, Rev 2 I, Neil Wilmshurst, being duly sworn, depose and state as follows: I am the Vice President and Chief Nuclear Officer at Electric Power Research Institute, Inc. whose principal office is located at 3420 Hillview Avenue, Palo Alto, California ("EPRI") and I have been specifically delegated responsibility for the above-listed Report which contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Proprietary Information on behalf of EPRI. EPRI Proprietary Information is identified in the above referenced report with text inside double brackets. Examples of such identification is as follows: ((This sentence is an example(El}} Tables containing EPRI Proprietary Information are identified with double brackets before and after the object. In each case the superscript notation (El refers to this affidavit and all the bases included below, which provide the reasons for the proprietary determination. EPRI requests that the Proprietary Information be withheld from the public on the following bases: Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information (see e.g. 10 C.F.R. §2.390(a)(4))::

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i I I (State of North Carolina) (County of Mecklenburg) (or affirmed) before me on this £!!.day of , 20Jj by _.L.J,!w..-U~rEffe.~"4,JJ~-------' proved to me on the basis of satisf ory evidence to be the Signature &za/i ;( ~o.1.141 (Seal) My Commission Expires rl.y of t1fd ,20~}}