NLS2015073, Er 15-019, Rev. 0, Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary), Enclosure 2

Er 15-019, Rev. 0, Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary), Enclosure 2

ML15229A033
Person / Time
Site:	Cooper
Issue date:	08/06/2015
From:	Mcclure T Entergy Corp
To:	Office of Nuclear Reactor Regulation
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ML15229A031	List: NLS2015073, Er 15-019, Rev. 0, Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary), Enclosure 2
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NLS2015073 ER 15-019, Rev. 0
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NLS201 5073 Page 1 of 28 Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY)

(Non-Proprietary)

Cooper Nuclear Station Docket No. 50-298, DPR-4

ATTACHMENT 9.1 ENGINEERING REPORT COVER SHEET & INSTRUCTIONS SHEET 1 OF 2 Engineering Report No.

15 -019 Rev 0

Page 1

of 27 Engineering Report Cover Sheet Engineering Report

Title:

Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY)

(Non-Proprietary)

Engineering Report Type: (3)

New []

Revision LI Cancelled LI Superseded Superseded by:

Rev 1: Clarified methodology discussion and capsule removal schedule.

El ECR No. N/.__A EC No. N/_.AA (4) Report Origin:

[] CNS LI Vendor Vendor Document No.:__________

(5) Quality-Related:

[] Yes

[] No Prepared by:

Tim McClurei 6h -.--

Responsible Engineer (Print Na ne/Sign)

S tan Domikaitisi!..x*

Design Verified:

Reviewed by:

Approved by:

Design Verifier ('f reqbired¶" (Print Name/Sign)

Date:

/

Date:

N/A Reviewer (Print Name/Sign)

Supervisor / Maae"*~m miign)

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 2 of 27 Table of Contents Section 1.0 2.0 3.0 4.0 5.0 6.0 Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Purpose Applicability Methodology Operating Limits Discussion References 3a~

3 4

5 6

11 CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 32 EFPY CNS P-T Curve B (Normal Operation - Core Not Critical) for 32 EFPY CNS P-T Curve C (Normal Operation - Core Critical) for 32 EFPY Cooper Feedwater Nozzle Finite Element Model [ 14]

Cooper Core Differential Pressure Nozzle Finite Element Model [16]

CNS Pressure Test (Curve A) P-T Curves for 32 EFPY CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY CNS Core Critical (Curve C) P-T Curves for 32 EFPY CNS ART Calculations for 32 EFPY Cooper Reactor Vessel Materials Surveillance Program 14 15 16 17 18 Table I Table 2 Table 3 Table 4 Appendix A 19 22 25 26 27

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 3 of 27 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-1 1-022-A, Revision 1 [ 1] and 0900876.401, Revision 0-A contained within BWROG-TP-1 1-023-A, Revision 0 [2].

It should also be noted that the P-T curves referenced in this PTLR have previously been approved by the NRC in Amendment 245 [24]. No changes are being made under this PTLR to the current P-T curves that were approved by the NRC [24] and currently in effect at CNS.

2.0 Applieabili ty This report is applicable to the CNS RPV for up to 32 Effective Full-Power Years (EFPY).

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

TS RCS Pressure and Temperature (P-T) Limits TS Surveillance Requirements

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 4 of 27 3.0 Methodology The limits in this report were derived as follows:

1. The methodology used is consistent with Reference [ 1] and Reference [21, which have been approved by the NRC in References [25] and [26], respectively.

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [3], using the RAMA computer code, as documented in Reference [4].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 [5], as documented in Reference [6].

4. The pressure and temperature limits were calculated consistent with Reference [I1],

"Pressure - Temperature Limits Report Methodology for Boiling Water Reactors,", as documented in NPPD calculation NEDC 07-048, Reference [7].

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

Initial issue of PTLR.

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 [17], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or paramcters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumption 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.

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 5 of 27 4.0 Operatin2 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 beitline 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 32 EFPY for Cooper Nuclear Station, as documented in Reference [7] and approved by the NRC in CNS Amendment 245 [24]. The CNS P-T curves for 32 EFPY are provided in Figures 1 through 3, and a tabulation of the curves is included in Tables I through 3. The adjusted reference temperature (ART) tables for the CNS vessel beltline materials are shown in Table 4 for 32 EFPY (Reference [6]). 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 at or near isothermal conditions (Figure 1: Curve A): < 25°F/hour1 [7].

• Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): _< 100F/hour2 [7].

• RPV bottom head coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 145°F.

Recirculation loop coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 500F.

• RPV flange and adjacent shell temperature limit > 70°F [7].

tInterpreted as the temperature change in any 1-hour period is less than or equal to 250F.

2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100°F.

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 6 of 27 To address the NRC condition regarding lowest service temperature in Reference [ 1] the minimum temperature is set to 70°F, which is equal to the RTNDT, max + 60°F, for all curves. [24]

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 [5] 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 [6]; 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 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 1.41 x 10's n/cm2 at 32 EFPY used in the P-T curve evaluation were obtained from Reference [4] and are calculated in accordance with RG 1.190

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

C23 07-2). 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 32 EFPY for the limiting lower intermediate shell plate is 1.02 x 1018 n/cm2 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 [7]. The nozzle material is not ferritic, however the effect of the penetration on the adjacent shell is considered

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 7 of 27 according to the methodology in Reference [2]. The RPV ID fluence value of 2.94 x 1017 n/cm2 at 32 EFPY used in the P-T curve evaluation of the WLI nozzle was obtained from Reference [4]

and is calculated in accordance with RG 1.190 [3]. This fluence value applies to the limiting WLI nozzle (Heat No. EV-26067). The fluence value for the WLI nozzle is based upon an attenuation factor of 0.72 for a postulated 1/4T flaw. As a result, the l/4T fluence for 32 EFPY for the limiting WLI nozzle is 2.13 x 1017 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 Ii/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< 0lOF/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

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 8 of 27 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.

The initial RTNDT~, 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 > 1017 n/cm 2 for E > 1 MeV) are shown in Table 4 for 32 EFPY [6].

Per Reference [6] and in accordance with Appendix A of Reference [I1], 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) [ 19]. 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 (170°F), the margin term (oa = 1 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 RGl1.99 were used in the determination of the ART values for all CNS beitline 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 [8] for the feedwater (FW) nozzle (non-beltline) pressure and thermal down shock stresses.

Mechanical and PrepPost, Release 11.0 (Service Pack 1) [9] for the development of the generic WLI nozzle stress intensity factors in [2].

• Mechanical APDL and PrepPost, Release 12.1 [ 10] 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

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 9 of 27 stress intensity factors for these nozzles [2, 13, 14, 16]. 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 [ 111] Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [12] 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 [ 13, 14].

Detailed information regarding the analysis can be found in References [ 13] and [ 14]. 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 [13], and a thermal ramp were analyzed [ 14]. 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 1 00° 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 [13, 14]. Therefore, the heat

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 10 of 27 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 [13].

Material properties were taken at 3500F, which is approximately the average temperature for the shutdown transient, from the 1989 ASME Code [ 15]. 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 [ 16]. Detailed information regarding the analysis can be found in Reference [ 16]. 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 in the bottom head.

• A two-dimensional finite element model of the core DP nozzle was constructed (Figure 5). Material properties were taken at 3250 F from the vessel stress report [16]. The use of temperature independent material properties is consistent with original design basis documents.

Intial RTNDT values were reported in the ART calculation in amendment 120 [22].

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 1 l of 27 6.0 References

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, August 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. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 2001.

4. 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 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, SI File No. 1100445.201.

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

6. Cooper Nuclear Station Calculation NEDC07-045, Revision 2, "Review of SIA Calculation COOP-27Q-301, ARTNDT and ART Evaluation", April 2013 that incorporated Structural Integrity Associates Calculation No. 1100445.301, Revision 1, "ARTNDt-and ART Evaluation", July 2010.

7. Cooper Nuclear Station Calculation, NEDC07-048, Revision 6, "Revised Pressure Temperature Curves", April 2013 that incorporated Structural Integrity Associates Calculation No. 1100445.303, Revision 0, "Revised P-T Curve Calculation", August 2011I.

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

9. ANSYS Mechanical and PrepPost, Release 1l1.0 (w/ Service Pack 1), ANSYS, Inc., August 2007.

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 12 of 27

10. ANSYS Mechanical APDL and PrepPost, Release 12.1 x64, ANSYS, Inc., November 2009.

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

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

13. 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-I13-Q-303, specifically Structural Integrity Associates Calculation No. NPPD-13Q-302, Revision 1, "Feedwater Nozzle Stress Analysis," June 1999.

14. 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.

15. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1, Appendices, 1989 Edition.

16. Cooper Nuclear Station Calculation, NEDC07-048, Revision 6, "Revised Pressure Temperature Curves", April 2013 that incorporated Structural Integrity Associates Calculation No. 1100445.304, Revision 0, "Core Differential Pressure Nozzle Finite Element Model and Stress Analysis," August 2011.

17. U. S. Code of Federal Regulations, Title 10, Part 50, Section 59, "Changes, tests and experiments," Aug. 28. 2007.

18. U. S. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Jan. 31, 2008.

19. BWRVIP-1 35, 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.

20. Letter NLS2002 104 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.

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 13 of 27

21. BWRVIP-86, Revision I-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI PROPRIETARY INFORMATION.

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

(ML021360424)

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

(ML033090607)

24. Cooper Nuclear Station Amendment 245 as approved by the NRC on February 22, 2013.

(ML13032A526).

25. U.S. NRC Letter to BWROG, "Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-1 1-022, Revision 1, November 2011, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC NO. ME7649, ML13277A557).

26. U.S. NRC Letter to BWROG, "Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-l 1-023, Revision 0, November 2011, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations" (TAG NO. ME7650, ML13183A017)

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 14 of 27 Figure 1: CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 32 EFPY [71 1,300 1,200 1,100 0.

-J Ln en z

LU U)uJ 0,

1,000 900 800 TOO 600 500 400 300 200 100 0

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 15 of 27 Figure 2: CNS P-T Curve B (Normal Operation - Core Not Critical) for 32 EFPY 171 1,300

'L 1,100

-I 1,9000 o 800 I-.

e,,

700 I-ii m

500 300 J

i-UpperVessel Bolt-upI II 200 Temp:1 IIIII 70 F I I I I I 100 0

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 16 of 27 Figure 3: CNS P-T Curve C (Normal Operation - Core Critical) for 32 EFPY [7]

1,300 1,200 1,100 1,000

~j900 uJ 00 I-j600

-J w

400 300 100 I

I!

I I-i f i i i i i i i i i i i i i i i i i I I I I I I I I 1 I I 1 I I I I I 1 I I I I I I I I I I I I I I I 1 I I I I ! I t I I I I I I I I I I I I I I [

I I I f I I I I I I I I I I I I I I I I

L I

I Minimum Core Critical Temperature *8*

I 0

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 17 of 27 Figure 4: Cooper Feedwater Nozzle Finite Element Model [141 MAT N APR 20 2Q11 15:18:48 PLOT NO.

1

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 18 of 27 Figure 5: Cooper Core Differential Pressure Nozzle Finite Element Model 1161 opeCor-e DP Nozzle, Pressure Stress

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 19 of27 Table 1: CNS Pressure Test (Curve A) P-T Curves for 32 EFPY 171 Beitline Region P-T Curve Temperature 70.00 70.00 70.00 70.00 70.00 70.00 70.00 70.00 70.00 70.00 70.00 70.00 70.00 77.37 85.28 92.11 100.07 109.08 116.71 123.34 129.19 134.42 139.16 143.49 147.47 151.16 154.60 157.82 160.83 P-T Curve Pressure 0

50 100 150 200 250 3O00 312 313 350 400 450 500 550 600 650 700 750 8O00 850 900 950 1000 1050 1100 1150 1200 1250 1300

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 20 of 27 Table 11-able-4: CNS Pressure Test (Curve A) P-T Curves for 32 EFPY (continued)

Plant =

Component =

Bottom Head thickness, t =

Bottom Head Radius, R=

ART =

Safety Factor =

Stress Concentration Factor=

Mm =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment=

Bifrn I*.

(penetrations portion)

=====>

All EFPY i!(no thermal eflects) i ill(bottom head penetrations)

,ii i*.!F (applied after bolt-up, instrument uncertainty) 0i* 5

• inches

~psig (instrument uncertainty)

Gauge Fluid Temperature

(°F) 65.0 65.0 67.0 69.0 71.0 73.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 97.0 99.0 (ksi*inchl(2) 76.66 76.66 78.43 80.28 82.20 84.20 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.43 106.30 109.28 112.38 115.62 118.98 Kim (ksi*inchl( 2) 51.10 51.10 52.29 53.52 54.80 56.13 57.52 58.96 60.47 62.03 63.66 65.35 67.12 68.95 70.86 72.85 74.92 77.08 79.32 Temperature for P-T Curve

(°F) 70 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 Adjusted Pressure for P-T Curve (p~sig) 0 814 834 855 877 900 923 948 973 1,000 1,028 1,056 1,086 1,118 1,150 1,184 1,219 1,256 1,294

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 21 of 27 Table lTobIe-4: CNS Pressure Test (Curve A) P-T Curves for 32 EFPY (continued)

Plant=

Component=

ART=

Vessel Radius, R =

Nozzle corner thickness, t' =

Kt=

Kip-applied=

Crack Depth, a =

Safety Factor =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment=

Reference Pressure =

Unit Pressure =

Flange RTNoT=

CMS Upp.Veu i!iiii*

oF====

All EFPY inches

• inches, approximate i!(no thermal effects) i ksi*inchll2 i:*inches

¶;i"i° (applied after bolt-up, instrument uncertainty) i*inches psig (hydrostatic pressure head for a full *ssel at 70°F) i psig (instrument uncertainty)

• i!,psig (pressure at which the FEA stress coefficients are valid)

• ipsig (hydrostatic pressure)

F===:=>

AllEP Gauge Fluid Temperature (0F) 65.0 65.0 67.0 69.0 71.0 73.0 75.0 77.0 (ksi*inchl(2) 84.20 84.20 86.28 88.44 90.70 93.05 95.49 98.03 (ksi*inch 1*2) 56.13 56.13 57.52 58.96 60.47 62.03 63.66 65.35 P-T Curve Tem pe rature (OF) 70 70 110 110 110 110 110 110 P-T Curve 10CFR50 Adjustments (psig) 0 313 313 1461 1499 1539 1581 1625

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 22 of 27 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY [7]

Beitline Reg~ion P-T Curve Temperature 70.00 70.00 70.00 70.00 70.00 70.22 81.92 84.36 84.55 91.39 99.35 106.22 114.04 123.11 130.80 137.46 143.34 148.60 153.35 157.70 161.70 165.40 168.84 172.07 175.09 177.96 180.65 183.21 185,65 P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 23 of 27 Table 2Tgable-4: CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY (continued)

Plant =

Component =

Bottom Head thickness, t =

Bottom Head Radius, R =

ART =

Safety Factor =

Stress Concentration Factor =

Mm =

Temperature Adjustment=

Height of Water fbr a Full Vessel =

Pressure Adjustment=

Pressure Adjustment=

Heat Up and Cool Down Rate =

Gauge Fluid Toempe rature

(°F) 65.0 65.0 67.0 69.0 71.0 73.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 97.0 99.0 101.0 103.0 105.0 107.0 109.0 111.0 113.0 115.0 117.0 Bt I~s4 (penetrations portion)

• iiinches

• i ihes

• ksC~inch112 ii!

(bottom head penetrat i,&ii, 0F (applied after bolt-u inches psig (hydrostatic pres,,

!*iiiipsig (instrument uncer

°~ F/Hr EFPY ions) p, instrument uncertainty) sure head for a full vessel at 70°F) rtainty)

(ksi*inchlI 2) 76.66 76.66 78.43 80.28 82.20 84.20 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.43 106.30 109.28 112.38 115.62 118.98 122.48 126. 12 129.92 133.86 137.97 142.25 146.70 151.33 156.15 (k~i*inchlI 2 )

37.46 37.46 38.35 39.27 40.23 41.23 42.27 43.36 44.49 45.66 46.88 48.15 49.47 50.85 52.28 53.78 55.33 56.94 58.63 60.38 62.20 64.09 66.07 68.12 70.26 72.48 74.80 77.21 Temperature for P-T Curve

(°F) 70 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 122 Adjusted Pressure for P-T Curve (pig) 0 582 597 613 629 646 664 682 701 721 742 764 786 810 834 859 886 913 942 972 1,003 1,035 1,068 1,103 1,140 1.178 1,217 1.258

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 24 of 27 Table 2T-eble-: CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY (continued)

Plant =

CNS Component Upper Vesee ART =

20.0 Vessel Radius. R =

110.375 Nozzle corner thickness, t =

5:.753 Crack Depth, a=

Safety Factor =

Temperature Adjustment=

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment Reference Pressure =

Unit Pressure =

Flange RTNoT =

Gauge Fluid Temperature (iF) 65.0 65.0 67.0 69.0 71.0 73.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 97.0 99.0 101.0 103.0 105.0 107.0 109.0 111.0 113.0 115.0 38.90 1.438 2.00 5.0 30.0 25,0 1,000 1,563 20.0 0F======>

All EFPY inches inches, approximate ksi-inchl t 2 ksilinch 11 2 inches

°F (applied after bolt-up, instrument uncertainty) inches psig (hydrostatic pressure head for a full vessel at 70°F) psig (instrument uncertainty) psig (pressure at which the FEA stress coefficients are valid) psig (hydrostatic pressure)

°F======>

All EFPY (ksi*inchlI 2 )

84.20 84.20 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.43 106.30 109.28 112.38 115.62 118.98 122.48 126.12 129.92 133.86 137.97 142.25 146.70 151.33 156.15 161.17 166.39 171.83 K1,,

10.37 19.46 20.51 19.26 20.06 20.91 21.80 22.73 23.71 24.74 25.82 26.95 28.15 29.40 30.71 32.09 33.54 35.04 36.63 38.30 40.04 41.87 43.79 45.80 47.90 50.10 52.39 P.?

Curve Temperature 70 )

70 740 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 P-T Curve Pressure (psig) 0 313 313 440 461 482 505 529 554 581 609 638 666 701 734 770 807 846 887 929 974 1021 1071 1122 1176 1233 1292

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 25 of 27 Table 3: CNS Core Critical (Curve C) P-T Curves for 32 EFPY [71 Curve A Leak uvTest TemperatureAPesr

==

• , pg Unt rssr =

~

i psig (hydrostatic pressure)

Flange RTNDT =

° P-T Curve Temperature 80.00 80.00 80.00 80.00 94.91 110.21 121.92 124.36 180.00 180.00 180.00 180.00 180.00 180.00 180.00 180.00 183.34 188.60 193.35 197.70 201.70 205.40 208.84 212.07 215.09 217.96 220.65 223.21 225.65 P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1 300

Cooper Nuclear Station PTLR ER 15-019 Revision 0 Page 26 of 27 Table 4: CNS ART Calculations for 32 EFPY 161 Lower Shell Plate G-2803-I C2274-1 14.0 0.20 0.68 153.0 55.l 17.0 0.0 34.0 103.1 Lower Shell Plate G-2803-2 C2307-l1 0.0 0.21 0.73 1629.8 58.6 17.0 0.0 34.0 92.6 Lower Shell Plate G-2803-3 C2274-2 8.0 0.20 0.68 153.0 55.1l 17.0 0.0 34.0 81.1 Lower Int. Shell Plate 0-2802-2 C2307-2 20.0

[11]

[IU))

((I]

108.8 8.5 0.0 17.0 105.8 Lower Int. Shell Plate G-2801-7 C2407-1

-10.0 0.13 0.65 92.3 38.8 17.0 0.0 34.0 62.8 Lower Shell Axial Welds 2-233A 12420 LINDE 1092

-50.0 0.270 1.035 254.4 90.8 28.0 0.0 56.0 96.8 Lower Shell Axial Welds 2-233B 12420 LINDE 1092

-50.0 0.270 1.035 254.4 90.8 28.0 0.0 56.0 96.8 Lower Shell Axial Welds 2-233C 12420 LINDE 1092

-50.0 0.270 1.035 254.4 90.8 28.0 0.0 56.0 96.8 Welds Lower Int. Shell Axial Welds 1-233A 27204/12008 LINDE 1092

-50.0 0.219 0.996 231.1 73.7 28.0 0.0 56.0 79.7 Lower Int. Shell Axial Welds 1-233B 27204/12008 LINDE 1092

-50.0 0.219 0.996 231.1 73.7 28.0 0.0 56.0 79.7 Lower Int. Shell Axial Welds 1-233C 27204/12008 LINDE 1092

-50.0 0.219 0.996 231.I 73.7 28.0 0.0 56.0 79.7 Lower/Lower lInt. Shell Circ Weld 1-240 21935 LINDE 1092

-50.0 0.183 0.704 172.2 63.9 28.0 0.0 56.0 69.9 Nozzle N-16A 0-2822 EV-26067

-10.0 0.13 0.65 92.3 16.5 8.3 0.0 16.5 23.0 NzlsNozzle N-16B 0-2822 EV-26067 10.0 0.16 0.62 118.5 21.2 10.6 0.0 21.2 52.4

....... oeNo etNo

~

lblles~n)Aeaaln

~

lac Factor, FF Lower Shell Plate G-2803-I C2274-l 6.375 1.59 l.09E+18 0.68 7.44E+17 0.360 Lower Shell Plate 0-2803-2 C2307-1 6.375 1.59 1,09E+18 0.68 7.44E+17 0.360 Lower Shell Plate 0-2803-3 C2274-2 6.375 1.59 l.09E+l18 0.68 7.44E+l17 0.360 Plates Loekt.SelPae 020-C2325371.4 l4E8 0.2 l0E8042 Lower Int. Shell Plate 0-2802-2 C2301-2 5.375 1.34 1.41E+18 0.72 1.02E+18 0,421 Lower Int. Shell Plate 0-2802-7 C2407-2 5.375 1.34 l.41 E+ 18 0.72 1.02 E+ 18 0.421I Lower hell Axial Wlte G-2301-C24207-6.375 1.59 lI07E+I8 0.68 1.30E+18 0.357 Lower Shell Axial Welds 2-233B 12420 6.375 1.59 1.07E+18 0.68 7.30E+17 0.357 Lower Shell Axial Welds 2-233B 12420 6,375 1.59 1.07E+18 0.68 7.30E+17 0,357 Wls Lower ki.Shell Axial Welds 2-233A 1274/200 5.375 1.34 81.07E+18 0.72 5.30E+17 0.357 Wds Lower kInt. Shell Axial Welds 1-233A 27204/12008 5.375 1.34 8.1 IE+I7 0.72 5.87E+17 0.319 Lower kint. Shell Axial Welds 1-233B 27204/l2008 5.375 1.34 8.11IE+I7 0.72 5.87E+17 0.319 Lower/Lower Int. Shell Circ Weld 1-240 21935 5.375 1.34 l.09E+18 0.72 7.90E+17 0.371 Nozzle N-16A G-2822 EV-26067 5.375 1.34 2.94E+17 0.72 2.13E+17 0.179 NzlsNozzle N-16B 0-2822 EV-26067 5.375 1.34 2.94E+17 0.72 2.13E+17 0.179

Cooper Nuclear Station PTLR ER 15-015 Revision 1 Page 27 of 27 Appendix A COOPER REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [1 8], two surveillance capsules were removed from the CNS reactor vessel in 1985 at 6.8 EFPY and 1991 at 11.2 EFPY [20, 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 the 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

[23]. Under the ISP, a capsule was scheduled for removal in 2003 but removal has been deferred to approximately 2017 at 32 EFPY [21]. CNS recently transitioned to 24 month refueling cycles during "event' years so the next capsule removal will occur in 2018 to align with a plant refueling outage as allowed by the ISP [21 ]. 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 [21] 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 [21].

NLS20 15073 Page 1 of 4 Affidavit for Proprietary Information Contained in the Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY)

Cooper Nuclear Station Docket No. 50-298, DPR-4

~UI~RESEARCHI tN~TITJT?

NEIL WILMSHURST Vice President and Chief Nuclear Officer Ref. EPRI Project Number 669 July 1, 2015 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Withholding of the following Proprietary Information Included in:

Entergy, Nuclear Management Manual. Engineering Report

Title:

Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY), Quality Related 3-EN-DC-147, REV. 501 To Whom It May Concern:

This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ('NR") withhold from public disclosure the report identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI) identified in the attached report.

Proprietary and non-proprietary versions of the Report and the Affidavit in support of this request are enclosed.

EPRI desires to disclose the Proprietary Information in confidence to assist the NRC review of the enclosed submittal to the NRC by NPPD. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.

If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (704) 595-2732. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440.

Sincerer Attachment(s) c: Sheldon Stuchell, NRC (sheldon~stuchell@nrc.gov)

Together.

,Shaping the Future of Electricity 1300 West WT.T Harris Boulevard, Charlotte, NC 28262-8550 USA

• 704.595.2732

• Mobile 704.490.2653

• nwilmshurst@epri.com

.ii~ IIRESEARCI-

• NSTITUIE AFFIDAVIT RE:

Request for Withholding of the Following Proprietary Information Included In:

Entergy, Nuclear Management Manual. Engineering Report

Title:

Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY), Quality Related 3-EN-DC-147, REV. 5C1 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 1300 W WT Hamrs Blvd, Charlotte, NC. ("EPRr~) and I have been specifically delegated responsibility for the above-listed report that 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 Coinmission ("NR") for the withholding of the Proprietary Information on behalf of EPRI.

EPRI Proprietary Information is identified in the above referenced report by double brackets. An example of such identification is as follows:

((This sentence is an example.{EI))

Tables containing EPRI Proprietary Information are identified with double brackets before and after the object.

In each case, the superscript notation {E refers to this affidavit as the basis 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.q., 10 C.F.R. § 2.390(a)(4):

a.

The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof,

b.

EPRI considers the Proprietary Information contained therein to constitute trade secrets of EPRI. As such, EPRI holds the Information in confidence and disclosure thereof is strictly limited to individuals and entities who have agreed, in writing, to maintain the confidentiality of the Information.

c.

The information sought to be withheld is considered to be proprietary for the following reasons. EPRI made a substantial economic investment to develop the Proprietary Information and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Proprietary Information. If the Proprietary Information were publicly available to consultants and/or other businesses providing services in the electric and/or nuclear power industry, they would be able to use the Proprietary Information for their own commercial benefit and profit and without expending the substantial economic resources required of EPRI to develop the Proprietary Information.

d.

EPRI's classification of the Proprietary Information as trade secrets is justified by the Uniform Trade Secrets Act which California adopted in 1984 and a version of which has been adopted by over

forty states. The California Uniform Trade Secrets Act., California Civil Code §§3426 - 3426.11, defines a "trade secret" as follows:

"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:

(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."

e.

The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Proprietary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.

f.

A public disclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRI's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally. The Proprietary Information can only be acquired and/or duplicated by others using an equivalent investment of time and effort.

I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and under the laws of the State of North Carolina.

Executed at 1300 W WT Harris Blvd being the premises and place of business of Electric Power Research Institute, Inc.

Neil Wilmshurs{

(State of North Carolina)

(County of Mecklenburg)

Subscribed a n* swoe to, (or affirmed) before me on this.__iday of 20.2_ by

'7'*__ /J~*,*

proved to me on the basis of C-atiV~ctory evidence to be the person~s)-who-appeared lefore me.

Signature QtA *

.. /. ;*2.z (Seal)

My Commission Expires o*--4y of _

/.j 20 '_*