ML17018A152

Pressure and Temperature Limits Report, Revision 1

ML17018A152
Person / Time
Site:	Cooper
Issue date:	11/01/2016
From:	McClure T Entergy Corp, Nebraska Public Power District (NPPD)
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17018A177	List: ML17018A151 ML17018A152 ML17018A153
References
ER 2016-042, Rev 1
Download: ML17018A152 (32)

Text

NLS2017003 Enclosure 2 Page 1 of77 Enclosure 2 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 Attachment 9.2 EC 16-046, Rev 1 Page 1 of 76

• NUCLEAR QUALITY RELATED 3-EN-DC-147 I REV. 5C1 MANAGEMENT INFORMATIONAL USE PAGE 1of31 MANUAL Engineering Reports ATTACHMENT

9.1 ENGINEERING

REPORT COVER SHEET Engineering Report No. __ 2_0_1_6-_0_4_2

__ Rev 1 Page of 31 Engineering Report Cover Sheet Engineering Report Title: Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY) (Non-Proprietary)

Engineering Report Type: (3) New 0 Revision Cancelled 0 Superseded 0 Superseded by: Revision 1: Removed "Proprietary Brackets" from Table 4 values for Cu and Ni content. (6) ECR No. NIA EC No. 16-46 (4) Report Origin: CNS 0 Vendor Vendor Document (5) Quality-Related: Yes 0 No Prepared by: Tim McClure/ Date: /(-t Responsible Engineer (Print Name/Sign)

Date:

Reviewed by: NI A Date:

Date: 12/; !f6 a I Attachment 9.2 EC 16-046, Rev 1 Page 2 of76 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 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 2 of31 Table of Contents Purpose Applicability Methodology Operating Limits Discussion References CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54EFPY CNS P-T Curve B (Normal Operation

-Core Not Critical) for 54EFPY CNS P-T Curve C (Normal Operation

-Core Critical) for 54EFPY Cooper Feedwater Nozzle Finite Element Model [19] Cooper Core Differential Nozzle Finite Element Model [20] 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 3 3 4 5 6 12 16 17 18 19 20 21 24 27 30 Appendix A Cooper Reactor Vessel Materials Surveillance Program 31 Appendix B BWRVIP-135, Revision 3: BWR Vessel and Internals Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations, Technical Report No. 3002003144, December 2014 (Non-Proprietary, Pages 1 -45)

• Attachment 9.2 EC 16-046, Rev 1 Page 3 of 76 1.0 Purpose Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 3of31 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 [1], and 0900876.401, Revision 0-A, contained within BWROG-TP-11-023-A, Revision 0 [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 Attachment 9.2 EC 16-046, Rev 1 Page 4 of76 3.0 Methodology The limits in this report were derived as follows: Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 4 of31 1. The methodology used is in accordance with Reference

[1] and Reference

[2], incorporating the NRC Safety Evaluations in References

[3] and [4], respectively.

2. 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]. 3. 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]. 4. 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]. 5. This revision of the pressure and temperature limits is to incorporate the following changes:

• Update pressure and temperature limits for 54 EFPY. Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 [10], provided the above methodologies are utilized.

The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumption modifications in the UFSAR, cannot Attachment 9.2 EC 16-046, Rev 1 Page 5 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 5 of31 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 CNS P-T curves for 54 EFPY are provided in Figures 1through3, and a tabulation of the curves is included in Tables 1 through 3. The adjusted reference temperature (ART) tables for the CNS 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): :5 25°F/hour 1 [9].

• Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B -non-nuclear heating, and Figure 3: Curve C -nuclear heating):

'.S 100°F/hour 2 [9]. 1 Interpreted as the temperature change in any 1-hour period is less than or equal to 25°F. 2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100°F.

Attachment 9.2 EC 16-046, Rev 1 Page 6 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 6 of31

• RPV bottom head coolant temperature to RPV coolant temperature L'l T limit during Recirculation Pump startup:::;;;

145°F.

• Recirculation loop coolant temperature to RPV coolant temperature L'lT limit during Recirculation Pump startup:::;;;

50°F.

• RPV flange and adjacent shell temperature 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 RT NDT,max and the CNS 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 0 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 0 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 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 Attachment 9.2 EC 16-046, Rev 1 Page 7 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 7 of31 surveillance capsules for the representative plate material and three surveillance capsules for the representative weld material.

The Cu and Ni 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 10 18 n/cm 2 at 54 EFPYused 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 1/4T flaw. As a result, the 1/4T fluence for 54 EFPY for the limiting lower intermediate shell plate is 1.62 x 10 18 n/cm 2 for CNS. 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 10 17 n/cm 2 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 based upon an attenuation factor of0.72 for a postulated 1/4T flaw. As a result, the 1/4T fluence for 54 EFPY for the limiting WLI nozzle location is 3.94 x 10 17 n/cm 2 for CNS. There are no* additional forged or partial penetration nozzles in the extended beltline.

Attachment 9.2 EC 16-046, Rev 1 Page 8 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 8of31 The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T 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 1/4T 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 cooldown 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 ifthe pressure test heatup/cooldown rate limits cannot be maintained.

The initial RTNoT, the chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 10 17 n/cm 2 for E > lMe V) are shown in Table 4 for 54 EFPY [8]. Initial RT NDT values were reported in the ART calculation in CNS Amendment 120 [ 13].

Attachment 9.2 EC 16-046, Rev 1 Page 9 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 9 of31 Per Reference

[8] and in accordance with Appendix A of Reference

[1], the CNS 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 ( cr = l 7°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 RGl.99 were used in the determination of the ART values for all CNS beltline materials except for plate heat C2307-2. The only computer code used in the determination of the CNS 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 PrepPost, Release 11.0 (Service Pack 1) [ 16] for the development Of the generic WLI nozzle stress intensity factors in [2].

• Mechanical APDL and PrepPost, Release 12.l [17] for the FW nozzle (non-beltline) thermal ramp stresses and the core differential pressure (DP) nozzle (bottom head) pressure stress distribution.

ANSYS 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 ANSYS 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 Attachment 9.2 EC 16-046, Rev 1 Page 10 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 10 of31 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 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 1/4T 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 100°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.

Attachment 9.2 EC 16-046, Rev 1 Page 11 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 11of31

• 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 detemi.ine 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:

• No 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 m 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.

Attachment 9.2 EC 16-046, Rev 1 Page 12 of 76 6.0 References Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 12 of31 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 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, ML13277A557).

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, ML13183A017)

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 3, "CNS Review of Trans Ware Calculations NPP-FLU-003-R-002, Revision 0, NPP-FLU-003-R-004, and NPP-FLU-003-R-005, Reactor Pressure Vessel Fluence Evaluation", April 2013, that incorporated Trans Ware Enterprises Report No. NPP-FLU-003-R-005, Revision 0, 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.

Attachment 9.2 EC 16-046, Rev 1 Page 13 of 76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 13 of31 8. Cooper Nuclear Station Calculation NEDC 07-045, Revision 3, September 2016, "Review of SIA Calculation 1100445.301, Proprietary and Non-Proprietary Versions, NDT and ART Evaluation," dated July 2010. 9. Cooper Nuclear Station Calculation, NEDC 07-048, Revision 7, September 2016, "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 28, 2007. 11. Cooper Nuclear Station Amendment 245 as approved by the NRC on February 22, 2013. (MLl 3032A526)

12. Cooper Nuclear Station Calculation NEDC 16-024, Revision 0, September 2016, "Review of SIA Calculation 11004 73.301 Cooper Vacuum Assessment", Revision 0 dated December 2015. 13. Cooper Nuclear Station Amendment 120 as approved by the NRC on April 26, 1988. (ML021360424)

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 Release 11.0 (w/ Service Pack 1 ), ANSYS, Inc., August 2007. 17. ANSYS Mechanical APDL and PrepPost, Release 12.1 x64, ANSYS, Inc., November 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-Attachment 9.2 EC 16-046, Rev 1 Page 14 of 76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 14 of31 303," specifically Structural Integrity Associates Calculation No. NPPD-13Q-302, Revision 1, "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," January 31, 2008. 25. Letter NLS2002 l 04 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. Nuclear 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)

Attachment 9.2 EC 16-046, Rev 1 Page 15 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 15 of31 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. Cooper Nuclear Station Amendment 256, Cooper Nuclear Station as approved by the NRC on July 25,2016 (ML16158A022)

29. 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.

Attachment 9.2 EC 16-046 , Rev 1 Page 16 of76 C o op e r N uclear Station PTLR ER 2 016-042 , Re v 1 Page 16of31 Figure 1: CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 54 EFPY Curve A -Pressure Test, Composite Curves --Bel t line ---Bottom Head --Non-Belt l ine -ove r all 1300 : I f , I I 1200 -*--j *------, , I 1100 1000 900 00 800 *;;; Qi "' "' 700 GI > C5 .... .., Ill GI 600 a:: .: .... .E :::; 500 GI :; "' "' GI , I

• I I ' I I I I----------:--,___, *-I------------I , __ I I I . -------1 I --....__ ____ -----I I I 70°F , 814 psi g j / J I ------=-t *-...-------

I I I 1---------'--------1-------I I I -I ' ---* ------------I I I I I I I Safe ------.-I ---* --I Operating I 7 0°F , 42 6 ps i g I ' I Region 400 I --*-300 f--. --...__ ____ -* *-I 70°F , 3 13 ps i g I I 11 0°F , 3 13 p s i g I 200 100 -*--Minim u m RPV P r es s ure = -1 4. 7 ps i g I 0 Minimum Bolt-Up Temperature>

70°F -100 I 0 50 100 150 200 250 300 Min i mum Reactor Vessel Meta l Temperature

(°F)

Attachment 9.2 EC 1 6-046 , Rev 1 Page 17 of76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 17 of31 Figure 2: CNS P-T Curve B (N ormal Operation -Core Not Critical) for 54 EFPY Curve B -Core Not Critical, Composite Curves --Beltline ---Bottom Head -Non-Beltline

-overall 1300 ,. I I ' 1200 ------_...L_ ---------' ' ' 1100 ' I I I 1000 ---_,__ __ ------I I I I I 900 ------, ____ ------I I I I ;;o 800 ------------*;;; ..!:!: I Qi I "' I __ _) "' 700 QI ------------> I I 0 I .. u I ... QI 600 ---__ ,_ ------er:: I .!: I .. 70°F , 499 ps i g I .E I Safe ::::; 500 -.--1 24°F , 1 40°F , 50 3 p si g QI I Operating I "' Region "' I QI lr. 400 I 70°F , 313 psig I 140°F , 313 p s ig I -300 ------------200 70°F , 1 84 psig 100 ---M i nimum RPV Pressure=

-14.7 psig 0 Minimum Bolt-Up Temperature>

70°F -100 0 50 100 150 200 250 300 Minimum Reactor Vessel Metal Temperature

(°F)

Attachment 9.2 EC 16-046 , Rev 1 Page 18 of 76 Cooper Nuclear Station PTLR ER 2016-042 , Rev 1 Page18of31 Figure 3: CNS P-T Curve C (Normal Operation

-Core Critical) for 54 EFPY 1300 1200 1100 1000 900 'iiO 800 *;;:; ..!:!: Qi "' "' 700 (IJ > 0 ... u "' (IJ 600 a: c: ... *e 500 :::; (IJ :i "' "' (IJ r;. 400 300 200 100 0 -100 0 Curve C -Core Critical, Composite Curves --Beltline ---Bottom Head --Non-Beltline

-overall ----------*-**--*---

---

I I I ___ , I I ---+--! I I -t---1 ' ' ' -r-' ' I ' ' ' ' I ' --,---+-

I gg*F , t__ 164°F , I I Safe Operating Region l80°F, 313 psig ----1 50 100 150 200 Minimum RPV Pressure=

-14.7 psig Minimum Core Critical Temperature

> 80°F 250 Minimum Reactor Vessel Metal Temperature

(°F) 300 Attachment 9.2 EC 16-046 , Rev 1 Page 19 of76 ELEME1\lTS MAT NUM Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 19 of31 Figure 4: Cooper Feedwater Nozzle Finite Element Model [19] J\N APR 20 2011 15: 1 8: 48 Plill NO. 1 Attachment 9.2 EC 16-046 , Rev 1 Page 20 of 76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 20of31 Figure 5: Cooper Core Differential Pressure Nozzle Finite Element Model [20) J\N Attachment 9.2 EC 16-046 , Rev 1 Page 21 of 76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page21 of31 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.S 178.0 1169.7 181.2 1218.9 184.3 1268.0 187.2 1317.2 Attachment 9.2 EC 16-046 , Rev 1 Page 22 of 76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 22of31 Table 1: CNS Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Non-Beltline Region Curve A -Pressure Test P-TCurve P-T Curve Temperature Pressure *F psi 70.0 0.0 70.0 312.6 110.0 312.6 110.0 1563.0 Attachment 9.2 EC 16-046 , Rev 1 Page 23 of 76 Cooper Nuclear Station PTLR ER 2016-042 , Re v 1 Page 23 of31 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-TCurve 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 Attachment 9.2 EC 16-046 , Rev 1 Page 24 of 76 Cooper Nuclear Station PTLR ER 2016-042 , Rev 1 P age 24of31 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-TCurve Temperature Pressure "F 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.S 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 Attachment 9.2 EC 16-046 , Rev 1 Page 25 of 76 Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page 25 of31 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Non-Beltline Region Curve B -Core Not Critical P-TCurve P-TCurve Temperature Pressure OF psi 70.0 0.0 70.0 312.6 140.0 312.6 140.0 1563.0 Attachment 9.2 EC 16-046 , Rev 1 Page 26 of76 Cooper Nuclear Station PTLR ER 2016-042, Re v 1 Page 26 of31 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-TCurve 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.S 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 Attachment 9.2 EC 16-046 , Rev 1 Page 27 of 76 Cooper Nuclear Station PTLR ER 2016-042, Re v 1 Page 27of31 Table 3: CNS Core Critical (Curve C) P-T Curves for 54 EFPY Curve C -Core Critical P-T Curve P-TCurve 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 Attachment 9.2 EC 16-046 , Rev 1 Page 28 of 76 Cooper Nuclear Station PTLR ER 2016-042 , Rev 1 Page 28of31 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-TCurve Temperature Pressure "F 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 Attachment 9.2 EC 16-046 , Rev 1 Page 29 of 76 Cooper Nuclear Station PTLR ER 2016-042 , Re v 1 Page 29 of31 Table 3: CNS Core Critical (Curve C) P-T Curves for 54 EFPY (continued)

Bottom Head Region Curve C -Core Critical P-TCurve P-T Curve Temperature Pressure *F 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.S 168.4 1316.8 Attachment 9.2 EC 16-046 , Rev 1 Page 30 of 76 Beltline ID Lowe r S h e ll Plate Lowe r S h e ll Pl ate Plat es Lowe r S h e ll Plate Lowe r Int. S h e ll P l a t e Lowe r Int. S h e ll Pl a t e Lowe r Int. S h e ll Pl a t e Lowe r S h e ll Axia l Welds Lowe r S h e ll Axia l W e ld s Lowe r S h e ll Axial Welds Welds Lower Int. S h e ll Axial Welds Lower Int. S h e ll Axial Weld s Lower Int. S h e ll Axial Welds Lower/Lower I nt. S h e ll Circ W e ld Nozz l es Nozzle N-I 6A Nozz l e N-168 Be l tllne ID Lowe r S h e ll P l a t e Lower S h e ll Plat e Plat es Lowe r S h e ll Pl ate Lower I nt. S h e ll Plat e Lowe r Int. S h e ll Plate Lower In t. She ll Pl a t e Lower She ll Axial Welds Lower hell Axial Welds Lowe r S h e ll Axia l Welds Welds Lower Int. S h e ll Axial Welds Lower Int. S h e ll Axial Welds Lower Int. S h e ll Axial We l ds L owe r/Lower Int. S h e ll C ir c W e ld ozz l es N ozz l e N-1 6A Nozz l eN-1 68 Code No. G-2803-1 G-2803-2 G-2803-3 G-2802-1 G-2802-2 G-2801-7 2-233A 2-2338 2-23 3C l-233A 1-2338 1-233C 1-240 G-2822 G-2822 Code No. G-2803-1 G-2803-2 G-2803-3 G-2802-1 G-2802-2 G-2801-7 2-233A 2-2338 2-233C 1-233A 1-2338 l-233C 1-240 G-2822 G-2822 Table 4: CNS ART Calculations for 54 EFPY Heat No. Flux Type Initial Cu NI RTND T ff) (wt%) (wt%) C2274-1 -14.0 0.20 0.68 C2307-I 0.0 0.21 0.73 C2274-2 --8.0 0.20 0.68 C233 l-2 1 0.0 0.16 0.62 -C2307-2 -20.0 0.2 1 0.76 C2 407-I -1 0.0 0.13 0.65 1 2420 LINDE 1 092 -50.0 0.270 1.035 1 2420 LINDE 1 092 -50.0 0.270 1.035 1 2420 LIND E 1 092 -50.0 0.270 1.035 27204/1 2008 LINDE 1092 -50.0 0.219 0.996 2 7204112008 LIND E 1 092 -50.0 0.219 0.996 27204/12008 LINDE 1 092 -50.0 0.219 0.996 2 1 935 LIND E 1 092 -50.0 0.1 83 0.7 04 E V-26067 -10.0 0.1 3 0.65 E V-26067 10.0 0.1 6 0.62 Fluence Data Heat No. Wall Thickness (In.) Flue nee Attenuation at ID Fu ll l/4t (n/cm 2) e-0.2.i. C2274-I 6.3 7 5 1.59 l.75E+l 8 0.68 C230 7-I 6.375 1.59 1.75E+1 8 0.68 C2274-2 6.375 1.59 1.75E+I 8 0.68 C2331-2 5.375 1.34 2.23E+l 8 0.72 C230 7-2 5.3 7 5 1.34 2.23E+l 8 0.72 C2407-1 5.3 7 5 1.34 2.23E+ 1 8 0.72 12420 6.3 75 1.59 1.72E+18 0.68 1 2420 6.375 1.59 1.72 E+I8 0.68 1 2420 6.3 7 5 1.59 1.72E+I 8 0.68 27204112008 5.375 1.34 1.26E+ 1 8 0.72 27204/1 2008 5.375 1.34 1.26E+ 1 8 0.72 2 7204 1 1 2008 5.3 7 5 1.34 I.26E+I 8 0.72 2 1 935 5.375 1.34 l.75E+l8 0.72 E V-26067 5.375 1.34 5.44E+ 1 7 0.72 E V-26067 5.375 1.34 5.44E+ 1 7 0.72 EPRI Proprietary Information CF ff) 153.0 162.8 1 53.0 (lllJ J l llll l 92.3 254.4 254.4 254.4 231.1 231.1 23 1.1 172.2 92.3 11 8.5 Fluence at l/4t (n/cm 2) 1.1 9E+l 8 1.1 9E+I 8 1.1 9E+I 8 I.62E+l 8 1.62E+l 8 1.62E+1 8 1.17E+18 1.17E+I8 1.17E+1 8 9.1 3E+1 7 9.1 3E+17 9.1 3E+I 7 1.27E+ 1 8 3.94E+1 7 3.94E+ 17 Cooper Nuclear Station PTLR ER 2016-042 , R ev 1 Pa ge 30 o f3 1 ARTNDT Margin Terms Total ART Ma rain ff) aA (0 f) a 1 ("Fl ("F) l"FI 69.3 17.0 0.0 34.0 117.3 73.8 17.0 00 34.0 1 07.8 69.3 17.0 0.0 34.0 95.3 77.7 8.5 0.0 17.0 1 04.7 1 34.2 8.5 00 1 7.0 1 3 1.2 47.9 1 7.0 0.0 34.0 71.9 I 1 4.4 28.0 0.0 56.0 1 20.4 114.4 28.0 0.0 56.0 120.4 114.4 28.0 0.0 56.0 1 20.4 92.2 28.0 0.0 56.0 98.2 92.2 28.0 0.0 56.0 98.2 92.2 28.0 0.0 56.0 98.2 80.2 28.0 0.0 56.0 86.2 23.7 8.3 0.0 16.5 37.4 30.4 10.6 0.0 21.2 70.8 Fluence Factor, FF 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 (suc h information is mark ed with doubl e bra ces " { { xxx " and a bar in t h e ri g ht-h a nd mar gi n) Attachment 9.2 EC 16-046, Rev 1 Page 31 of 76 Appendix A Cooper Nuclear Station PTLR ER 2016-042, Rev 1 Page31 of31 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 1991at11.2 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 BWRVIP 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]. Under the ISP, a capsule was scheduled for removal in 2003 but removal has been deferred to approximately 2017 at 32 EFPY [27]. CNS recently transitioned to 24 month refueling cycles during "even" years so the next capsule removal will occur in 2018 to align with a plant refueling outage as allowed by the ISP [27]. 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-B, 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].}}