NLS2018012, Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RI5-03

Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RI5-03

ML18082A563
Person / Time
Site:	Cooper
Issue date:	03/14/2018
From:	Dent J Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2018012
Download: ML18082A563 (71)

Text

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Nebraska Public Power District NLS2018012 March 14, 2018 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Always there when you need us 50.55a

Subject:

Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RI5_-03 Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

References:

1.

Email from Thomas Wengert, U.S. Nuclear Regulatory Commission, to Jim Shaw, Nebraska Public Power District, dated February 14, 2018, "Cooper Nuclear Station - Final RAI RE: Alternative Request RI5-03 (EPID L-2017-LLR-064)"

2.

Letter from John Dent, Jr., Nebraska Public Power District, to the U.S.

Nuclear Regulatory Commission, dated August 17, 2017, "10 CFR 50.55a Relief Requests for Fifth Ten-Year Inservice Inspection Interval" (MLl 7241A048)

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District to respond to the Nuclear Regulatory Commission's Request for Additional Information (RAI) (Reference 1) related to the Cooper Nuclear Station Relief Request RI5-03 (Reference 2).

The response to the specific RAI question is provided in Attachment 1 to this letter. Attachment 2 provides a revised Relief Request RI5-03. The requested engineering report is included as an Enclosure.

This letter does not contain any new regulatory commitments.

If you have any questions concerning this matter, please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

NLS2018012 Page 2 of2 Dent, Jr.

Vice President - Nuclear and Chief Nuclear Officer

/dv Attachments: 1) Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RI5-03

2) Revised Relief Request RI5-03

Enclosure:

Engineering Report 2017-027 cc:

Regional Administrator w/ attachments and enclosure USNRC - Region IV Cooper Project Manager w/ attachments and enclosure USNRC - NRR Plant Licensing Branch IV Senior Resident Inspector w/ attachments and enclosure USNRC-CNS NPG Distribution w/o attachments and enclosure CNS Records w/ attachments and enclosure

NLS2018012 Page 1 of2 Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RIS-03 Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) regarding Relief Request RI5-03 is shown in italics. The Nebraska Public Power District (NPPD) response to the request is shown in normal font.

RAJ Rl5-03-l In the August 17, 2017 submittal (Attachment 1, page 51 of68), the licensee stated, in part:

CNS performed a plant specific probabilistic fracture mechanics to supplement the criteria of Code Case N-702 and BWRVIP-241 in order to demonstrate that the PoF

[probability of vessel failure] remains acceptable over the period of extended operation

[PEO]. Conservatively assuming zero inspection for the initial 40 years of operation and examination of 25% [percent] for PEO, the evaluation concluded the average PoF for a LTOP [low-temperature overpressure] event is 2.92 x 10-11 per year (Reference ])for the nozzle inner radius, and <8.33 x 10-13 per year (Reference 1) for the nozzle-to-shell weld, both of which are less than the NRC safety goal of 5 x 1 o-6 per year.

In the above quote, the title for BWRVIP-241 is "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (ADAMS Accession No. MLJJ 119A041).

However, "Reference l" or ER 2017-027, "Review of Structural Integrity Calculations 1400334.301 & 1400334.302 for Code Case N-702 Relief Request, Revision 0," (Proprietary) dated June 7, 2017, was not provided with the submittal. In addition, the PoF for the limiting condition (normal operation) was not mentioned in the submittal.

Because the "Reference l" calculations were not provided with the submittal, the NRC staff does not have enough information to complete its safety evaluation. Therefore, the NRC requests that the licensee provide the "Reference l" calculations in order for the NRC to review the probabilistic fracture mechanics analysis of the nozzle inner radius and the nozzle-to-shell welds of this proposed alternative. In the response, please include the PoF for the limiting condition (normal operation), in accordance with the safety evaluation for BWRVIP-108, "Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii" (ADAMS Accession No. ML073600374).

NLS2018012 Page 2 of2 NPPD Response:

NPPD is providing Engineering Report (ER) 2017-027, "Review of Structural Integrity Calculations 1400334.301 & 1400334.302 for Code Case N-702 ReliefRequest," Revision 1, as an Enclosure. ER 2017-027 and the associated calculations were recently revised due to an error identified in the Structural Integrity Associates probabilistic fracture mechanics software. The error was corrected to yield a minor increase in the PoF per year value due to the LTOP event.

The revised evaluation concludes that the average PoF due to LTOP is 1.675 x 10-10 per year for the N2 nozzle blend (inner) radius and< 8.33 x 10-13 per year for the nozzle-to-shell weld. In addition, the PoF per year due to normal operation is< 8.33 x 10-10* The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to LTOP events, as well as normal operation, are all less than the 5 x 1 o-6 per year criteria established in NUREG 1806, Volume 1.

Therefore, Attachment 2 of this response includes revised paragraph 4 on page 6 (paragraph 4 on page 51 in original submittal) and the associated reference on the Attachment to the August 17, 2017 NPPD submittal.

NLS2018012 Page 1 of 12 Revised Relief Request RIS-03 Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

NLS2018012 Page 2 of 12 10 CFR 50.55a Request No. RIS-03 Implementation of Code Case N-702 Proposed Alternative in Accordance with 10 CFR 50.SSa(z)(l)

Acceptable Level of Quality and Safety ASME Code Component{s) Affected Code Class:

ASME Section XI Code Class 1 Component Numbers:

Various (see Table 1 for detailed list of components)

Code

References:

ASME Section XI, 2007 Edition with 2008 Addenda Code Case N-702 Examination Category:

B-D Item Number(s):

B3.90 and B3.100

Applicable Code Edition and Addenda

ASME Section XI, 2007 Edition through the 2008 Addenda Applicable ASME Code Requirements Table IWB-2500-1, Examination Category B-D, "Full Penetration Welded Nozzles in Vessels" requires a volumetric examination of all nozzles with full penetration welds to the vessel shell

( or head) and integrally cast nozzles each 10-year interval. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented, as required and conditioned by 10 Code of Federal Regulations 50.55a(b )(2)(xv).

The RPV nozzle-to-vessel welds and inner radii subject to this request are listed below in Table 1:

TABLE 1 Identification Description Total Number Minimum Number Number to be examined Nl Recirculation Outlet 2

1 N2 Recirculation Inlet 10 3

N3 Main Steam Outlet 4

1 NS Core Spray 2

1 N6 Head Instrument 2

1 N7 Head Vent 1

1 N8 Jet Pump Instrumentation 2

1

NLS2018012 Page 3 of 12

Reason for Request

10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 NRC Regulatory Guide 1.147 Rev. 17 conditionally accepts the use of Code Case N-702. This code case provides an alternative to performing examination of 100% of the Nozzle-to-Vessel Welds and Inner Radii for Examination Category B-D nozzles with the exception of the Feedwater and CRDRL Nozzles. The alternative is to perform examination of a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size, excluding the Feedwater and CRDRL Nozzles.

Proposed Alternative and Basis for Use Proposed Alternative Pursuant to 10 CFR 50.55a(z)(l), NPPD requests approval to implement the alternative of Code Case N-702 in lieu of the code required 100% examination of all nozzles identified in Table 1.

As an alternative, for the nozzle-to-shell welds and inner radii identified in Table 1, NPPD proposes to examine a minimum of25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702.

Basis for Use BWRVIP has issued two topical reports:

BWRVIP-108, "Technical Basis for the Reduction oflnspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, dated October 2002 (ML023330203).

BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1021005, dated October 2010 (ML11119A041).

The BWRVIP-108 report contains the technical basis supporting ASME Code Case N-702 "Alternative Requirements for BWR Nozzle Inner Radius and Nozzle-to-Shell Welds" for reducing the inspection ofRPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval.

BWRVIP-241 provides supplemental analyses for BWR RPV recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii. BWRVIP-241 was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation for the BWRVIP-108 report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii."

NLS2018012 Page 4 of 12 10 CFR 50.55a Request No. RI5-03 (continued)

Implementation of Code Case N-702 Based on the two evaluations (BWRVIP-241 and BWRVIP-108), the failure probabilities due to a LTOP event at the nozzle blend radius region and the nozzle-to-vessel shell welds for CNS recirculation inlet and outlet nozzles are very low and meet the NRC safety goal.

Regulatory Guide 1.147, Revision 17 conditionally accepts the use of Code Case N-702 with the following condition "The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 ofNRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 ofNRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ML13071A240) are met."

Section 5.0 of the NRC Safety Evaluation for BWRVIP-241 states:

"Licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWR VIP-241 report to their units in the relief request by demonstrating all of the following:

(1) The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour CNS Technical Specification 3.4.9.1, Reactor Coolant System heatup and cooldown rates are limited to a maximum of 100°F when averaged over any one hour period and thus meets the requirement of Condition 1.

Note: Inputs used in 2 through 5 representing the CNS configuration are in bold text.

Recirculation inlet nozzles (N2)

(2) (pr/t)/CRPv ~ 1.15 p = RPV normal operating pressure (psi) (1020 psig per CNS Technical Specifications 3.4.10 for Reactor Steam Dome Pressure) r = RPV inner radius (inch) (110.375) t = RPV wall thickness (inch) (6.875)

CRPv = 19332 (based on the BWRVIP-108 recirculation inlet nozzle/RPV finite element method (FEM) model);

CNS specific calculations for Condition 2 above:

(1020 X 110.375)/6.875)/19332 = 0.85 ~ 1.15 The CNS result is 0.85 and thus meets the requirement of condition 2 to be,::: 1.15.

NLS2018012 Page 5 of 12 10 CFR 50.55a Request No. RI5-03 (continued)

Implementation of Code Case N-702 (3) [p(r/ + r/) I (r/ - r/)]/CNOZZLE :s__ 1.47 P = RPV normal operating pressure (psi) (1020) r0 = nozzle outer radius (inch) (10.219) ri = nozzle inner radius (inch) (6.188)

CNozzLE = 1637 based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model);

CNS specific calculations for Condition 3 above:

[1020(10.2192 + 6.1882)/(10.2192 - 6.1882))/1637 = 1.34 ~ 1.47 The CNS result is 1.34 and thus meets the requirement of condition 3 to be ~1.47 Recirculation outlet nozzles (Nl)

(4) (pr/t)/CRPv.::: 1.15 p = RPV normal operating pressure (psi) (1020) r = RPV inner radius (inch) (110.375) t = RPV wall thickness (inch) (6.875)

CRPv = 16171 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model);

CNS specific calculations for Condition 4 above:

(1020 X 110.375)/6.875)/16171 = 1.013 ~ 1.15 The CNS result is 1.013 and thus meets the requirements of condition 4 to be~ 1.15 (5) [p(r0 2 + r/) I (r/ - r/)]/CNoZZLE :s__ 1.59 P = RPV normal operating pressure (psi) (1020) r0 = nozzle outer radius (inch) (21.656) ri = nozzle inner radius (inch) (12.875)

CNozzLE = 1977 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model).

CNS specific calculation for Condition 5 above:

[1020(21.6562 + 12.8752)/(21.6562 - 12.8752)]/1977 = 1.08 ~ 1.59 The CNS result is 1.08 and thus meets the requirements of condition 5 to be ~ 1.59

NLS2018012 Page 6 of 12 10 CFR 50.55a Request No. RI5-03 (continued)

Implementation of Code Case N-702 The analyses for the N2 nozzles in BWRVIP-108 and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation and do not address the extended operating period. Based on analysis performed in support oflicense renewal for CNS, the beltline was re-evaluated for 60 years based on the axial flux profile and the active fuel and nozzle elevations. In 4.2.1 "Reactor Vessel Fluence" of the License Renewal Application, it is documented that fluence at the recirculation inlet nozzles (the closest ferritic nozzles to the beltline) will not exceed IE+ 17 n/cm2 during the period of extended operation. Since the LRA, CNS has updated the fluence values using the NRC approved RAMA fluence method in support of the current Pressure Temperature Limits Report. As part of that evaluation, the predicted fluence at 54 EFPY was also determined for the N2 nozzles which support the conclusion of the LRA that the fluence at the recirculation inlet nozzles will not exceed lE+ 17 n/cm2* The plates and welds in the beltline remain the limiting materials for the period of extended operation. Therefore, the fluence assumptions used in BWRVIP-108 and BWRVIP-241 remain valid and are applicable to CNS.

The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth. CNS is projected to exceed the total number of thermal cycles used in the BWRVIP analysis during the extended operating period. However, the usage factor for the N2 nozzles is expected to remain below 1.0. Previous BWRVIP documents have demonstrated that SCC crack growth represents the majority of the crack growth and that crack growth due to additional mechanical/thermal fatigue cycles introduced by the extended operation time is insignificant compared to hypothetical SCC growth. Thus, the amount of thermal cycle driven fatigue crack growth due to the extended operation to 60 years is not a controlling factor in the probability of failure of the BWR reactor vessel nozzles.

CNS performed a plant specific probabilistic fracture mechanics to supplement the criteria of Code Case N-702 and BWRVIP-241 in order to demonstrate that the PoF remains acceptable over the period of extended operation. Conservatively assuming zero inspection for the initial 40 years of operation and examination of 25% for PEO, the evaluation concluded the average PoF for a LTOP event is 1.65 x 10-10 per year (Reference 1) for the nozzle inner (blend) radius,

< 8.33 x 10-13 per year (Reference 1) for the nozzle-to-shell weld, and< 8.33 x 10-10 per year (Reference 1) due to normal operation, all of which are less than the NRC safety goal of 5 x 10-6 per year.

The examination history for the subject nozzles is included in Table 2.

Duration of Proposed Alternative The duration of this request is for the extended license period ending January 18, 2034.

NLS2018012 Page 7 of 12 Precedents 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 "Pilgrim Nuclear Power Station - Relief Request PRR-50, Use of Alternatives, Implementation of Code Case N-702," dated January 5, 2016 (ADAMS Accession Number ML15338A309).

"Cooper Nuclear Station - Request for Relief No. RI-04 for the Fourth 10-Year Inservice Inspection Interval Regarding Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds,"

dated October 8, 2010 (ADAMS Accession Number ML102220449).

"Cooper Nuclear Station - Relief Request No. RI-08, Revision O Applicable to Fourth 10-Year Inservice Inspection Interval," dated May 20, 2015 (ADAMS Accession NumberMLl 5134A242).

Reference

1. ER 2017-027, "Review of Structural Integrity Calculations 1400334.301 & 1400334.302 for Code Case N-702 Relief Request," Revision 1, dated March 8, 2018.

NLS2018012 Page 8 of 12 Nozzle ID Nozzle-to-Vessel (NV)

Inner Radius (IR)

NIA (NV)

NIA (IR)

NIB (NV)

NIB (IR)

N2A(NV)

N2A (IR)

N2B (NV)

N2B (IR)

N2C (NV)

N2C (IR)

N2D (NV) 10 CFR 50.55a Request No. RI5-03 (continued)

Implementation of Code Case N-702 Table 2 - Nozzle-to-Shell Welds and Nozzle Blend Radii Component Nominal Pipe Last Category Number Item Number System Size (Inches)

Examination B-D B3.90 Recirculation 28 10/2016 (Outlet)

B-D B3.100 Recirculation 28 10/2016 (Outlet)

B-D B3.90 Recirculation 28 01/2005 (Outlet)

B-D B3.100 Recirculation 28 01/2005 (Outlet)

B-D B3.90 Recirculation 12 01/2005 (Inlet)

B-D B3.100 Recirculation 12 04/2011 (Inlet)

B-D B3.90 Recirculation 12 01/2005 (Inlet)

B-D B3.100 Recirculation 12 04/2011 (Inlet)

B-D B3.90 Recirculation 12 04/2011 (Inlet)

B-D B3.100 Recirculation 12 04/2011 (Inlet)

B-D B3.90 Recirculation 12 01/2005 (Inlet)

Appendix Results VIII Exam NRI Yes NRI VT-1 NRI Yes NRI MVT-1 NRI Yes NRI Yes NRI Yes NRI Yes NRI Yes NRI Yes NRI Yes

NLS2018012 Page 9 of 12 Nozzle ID Nozzle-to-Vessel (NV)

Inner Radius (IR)

N2D (IR)

N2E (NV)

N2E (IR)

N2F (NV)

N2F (IR)

N2G(NV)

N2G (IR)

N2H (NV)

N2H (IR)

N2J (NV)

N2J (IR)

N2K(NV)

Category Number B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nominal Pipe Item Number Systein Size (Inches}

B3.100 Recirculation 12 (Inlet)

B3.90 Recirculation 12 (Inlet)

B3.100

• Recirculation 12 (Inlet)

B3.90 Recirculation 12 (Inlet)

B3.100 Recirculation 12 (Inlet)

B3.90 Recirculation 12 (Inlet)

B3.100 Recirculation 12 (Inlet)

B3.90 Recirculation 12 (Inlet)

B3.100 Recirculation 12 (Inlet)

B3.90 Recirculation 12 (Inlet)

B3.100 Recirculation 12 (Inlet)

B3.90 Recirculation 12 (Inlet)

Last Appendix Results Examination VIII Exam 04/2011 NRI Yes 10/2016 NRI Yes 10/2016 NRI Yes 04/2011 NRI Yes 04/2011 NRI Yes 04/2011 NRI Yes 04/2011 NRI Yes 10/2016 NRI Yes 10/2016 NRI Yes 04/2011 NRI Yes 04/2011 NRI Yes 10/2016 NRI Yes

NLS2018012 Page 10 of 12 Nozzle ID Nozzle-to-Vessel (NV)

Inner Radius (IR)

N2K (IR)

N3A (NV)

N3A (IR)

N38 (NV)

N38 (IR)

N3C (NV)

N3C (IR)

N3D (NV)

N3D (IR)

N4A(NVi1)

N4A (IRi1)

N4B (NV)°)

N4B (IRi'>

N4C (NV)°>

N4C (IR)°>

Category Number 8-D 8-D 8-D B-D 8-D 8-D 8-D 8-D 8-D B-D B-D B-D B-D B-D B-D 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nominal Pipe Item Number System Size (Inches) 83.100 Recirculation 12 (Inlet) 83.90 Main Steam 24 83.100 Main Steam 24 B3.90 Main Steam 24 B3.100 Main Steam 24 83.90 Main Steam 24 83.100 Main Steam 24 83.90 Main Steam 24 83.100 Main Steam 24 B3.90 Feedwater 12 B3.100 Feedwater 12 B3.90 Feedwater 12 B3.100 Feedwater 12 B3.90 Feedwater 12 B3.100 Feedwater 12 Last Appendix Results Examination VIII Exam 10/2016 NRI Yes 10/2016 NRI Yes 09/2016 NRI VT-1 03/2000 NRI NoC3) 03/2000 NRI NoC3) 01/2005 NRI Yes 02/2005 NRI MVT-1 03/2000 NRI NoC3) 03/2000 NRI NoC3) 10/2014 NRI Yes 10/2014 NRI Yes 10/2014 NRI Yes 10/2014 NRI Yes 10/2014 NRI Yes 10/2014 NRI Yes

NLS2018012 Page 11 of 12 Nozzle ID Nozzle-to-Vessel (NV)

Inner Radius (IR)

N4D (NV)'1>

N4D (IRi1>

N5A (NV)

N5A (IR)

N5B (NV)

N5B (IR)

N6A (NV)

N6A (IR)

N6B (NV)

N6B (IR)

N7 (NV)

N7 (IR)

N8A (NV)

Category Number B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D B-D 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nominal PiQe Item Number System Size (Inches)

B3.90 Feedwater 12 B3.100 Feedwater 12 B3.90 Core Spray 10 B3.100 Core Spray 10 B3.90 Core Spray 10 B3.100 Core Spray 10 B3.90 Head 6

Instrumentation B3.100 Head Instrumentation 6

B3.90 Head Instrumentation 6

B3.100 Head Instrumentation 6

B3.90 Head Vent 4

B3.100 Head Vent 4

B3.90 Jet Pump 4

Instrumentation Last AQQendix Results Examination VIII Exam 10/2014 NRI Yes 10/2014 NRI Yes 10/2016 NRI Yes 10/2016 NRI Yes 01 /2005 NRI Yes 03/2011 NRI Yes 02/2005 NRI Yes 02/2005 NRI MVT-1 03/2011 NRI Yes 03/2011 NRI MVT-1 03/2011 NRI Yes 03/2011 NRI MVT-1 03/2000 NRI No<3)

NLS2018012 Page 12 of 12 Nozzle ID Nozzle-to-Vessel (NV)

Inner Radius (IR)

N8A (IR)

N8B (NV)

N8B (IR)

N9 (NVi2>

N9 (IRi2>

Notes:

Category Number B-D B-D B-D B-D 8-D

1) Code Case N-702 excludes these nozzles 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nominal Pige Item Number System Size (Inches)

B3.100 Jet Pump 4

Instrumentation B3.90 Jet Pump 4

Instrumentation B3.100 Jet Pump 4

Instrumentation B3.90 CRD Return Line 5

83.100 CRD Return Line 5

Last Aggendix Results Examination VIII Exam 03/2000 NRI NoC3l 10/2014 NRI Yes 10/2014 NRI Yes 03/2011 NRI Yes 03/2011 NRI Yes

2) CNS CRD Line is capped and as such is no longer the return for CRD. Because N9 is a unique nozzle, it will be examined and an alternative is not being sought for this nozzle.

3) Exams performed prior to adoption of Appendix VIII.

NLS2018012 Enclosure Page 1 of 55 Enclosure Engineering Report 2017-027 Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

I

.!'a NUCLEAR QUALITY RELATED 3-EN-DC-147 REV. 5C1

~Entergy MANAGEMENT MANUAL INFORMATIONAL USE Page 1 of 10 Engineering Reports ATTACHMENT 9.1 ENGINEERING REPORT COVER SHEET & INSTRUCTIONS 5HEET1 OF 2 Engineering Report No.

ER2017-027 Rev l Page _I __ of 10 Engineering Report Cover Sheet Engineering Report

Title:

Review of Structural Integrity Calculations 1400334.301 & 1400334.302 for Code Case N-702 Relief Request Engineering Report Type:

New D Revision IZI Cancelled D

Superseded D

Superseded by:

ECR No. NIA EC No. NIA Report Origin:

D CNS lg} Vendor Vendor Document No.:.~S=ee~T=itl=e'---------

Quality-Related:

~ Yes D No Prepared by:

Tim McClure!

Date: 3-7-J 8 Responsible Engineer (Print Name/Sign)

NIA Date:

Reviewed by:

ER 2017-027 Rev1

ENGINEERING REPORT HNPPD ER2017-027 Rev 1 rtl COOPER NUCLEAR STATION Page 2 of 10 TABLE OF CONTENTS Section Title Page ATTACHMENT 9.1 ENGINEERING REPORT COVER SHEET & INSTRUCTIONS 1

1.0 REVISION

SUMMARY

.................................................................... 3

2.0 INTRODUCTION

............................................................................. 4 3.0 PURPOSE....................................................................................... 3 4.0 TECHNICAL APPROACH AND METHODOLOGY.........................4 5.0 DESIGN INPUTS............................................................................. 5 6.0 ASSUMPTIONS.............................................................................. 5 7.0 EVALUATION.................................................................................. 6 8.0 RESULTS........................................................................................ 6

9.0 REFERENCES

.............................................................................. 10 10.0 ATTACHMENTS............................................................................ 10 10.1 SIA Calculation 1400334.301, Rev 1 (24 pages) 10.2 SIA Calculation 1400334.302, Rev 1 (19 pages) 10.3 Technical Comments and Review Form (1 page)

ER 2017-027 Rev1

ENGINEERING REPORT NNPPD ER2017-027 Rev 1 COOPER NUCLEAR STATION Page 3 of 10 1.0 REVISION

SUMMARY

1.1 Revision O: Initial Revision.

1.2 Revision 1: Corrects an error identified by Structural Integrity Associates (SIA) regarding the calculated PoF per year value due to a L TOP event and adds a new PoF value per year due to normal plant operation [Reference CNS Condition Report CR-CNS-2018-01214]. This revision accepts and incorporates Revision 1 of SIA Calculations 1400334.301 [Attachment 10.1] and 1400334.302

[Attachment 10.2] in their entirety. Based on the revisions to the affected calculations, this report no longer includes proprietary information as part of Revision 1.

2.0 INTRODUCTION

This evaluation provides an Owner acceptance review of Rev 1 of Structural Integrity Calculations 1400334.301, Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N2 Nozzle and 1400334.302, Code Case N-702 Evaluation for Cooper Recirculation Inlet (N2) Nozzle. These calculations provide a technical basis for extending the applicability of ASME Section XI Code Case N-702 through the end of the period of extended operation (PEO). The Code case allows reduction of in-service inspection from 100% to 25% of all nozzle blend radii and nozzle-to-shell welds every 10 years, including one nozzle from each system and pipe size, except for feedwater and control rod drive return nozzles.

These calculations determine the Probability of Failure (PoF) per year due to the limiting Low Temperature Overpressurization Probability (L TOP) event with 25%

inspection for the extended operating term (with zero inspection coverage for the initial 40 years of operation). In addition, the PoF for the limiting condition (normal operation) is also determined in Revision 1. The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to L TOP events are compared to the Sx10-6 per year NRC safety goal established in NUREG 1806, Volume 1, [Reference 9.2.19]

Technical documents BWRVIP-108 [References 9.2.2, 9.2.3] and BWRVIP-241

[Reference 9.2.4] provide the basis for the code case, but only consider 40 year plant operation. In order to extend the applicability of Code Case N-702, a probabilistic fracture mechanics (PFM) evaluation; consistent with the methods of BWRVIP-108 and BWRVIP-241, is performed to ensure that the probability of failure remains acceptable.

The N2 (Recirculation Inlet) nozzles are identified as the bounding nozzles when fluence is not considered [Reference 9.2.4].

The evaluation consists of two parts: Finite Element Model (FEM) Stress Analysis and Probabilistic Fracture Mechanics (PFM) Analysis. The FEM stress analysis is performed in 1400334.301 and the PFM analysis is performed in 1400334.302.

ER 2017-027 Rev1

ENGINEERING REPORT NNPPD ER2017-027 Rev 1 COOPER NUCLEAR STATION Page 4 of 10 3.0 PURPOSE This calculation provides an owner review and acceptance of two (2) vendor prepared calculations that provide a technical basis for the application of ASME Section XI Code Case (CC) N-702 as an alternative to ASME Section XI, Category B-D examination requirements. Code Case N-702 is currently not approved for use by the NRC unless approved via a Relief Request. The inservice inspection requirements of the reactor pressure vessel (RPV) nozzle to shell and nozzle inner radius examinations will be reduced from 100% to 25% with the exception of the four (4) feed water (N4's) and control rod drive return nozzle (N9) if approved by the NRC.

The objectives of calculation 1400334.301 are to:

1. Develop a Finite Element Model (FEM) for the N2 nozzle and to,

2. Determine the stresses caused by applicable Service Level A and B thermal transients and internal pressure.

The stress distributions obtained as output of this analysis will be used as input for the subsequent probabilistic fracture mechanics (PFM) evaluation performed in 1400334.302.

The objective of calculation 1400334.302 is to perform a plant specific probabilistic fracture mechanics evaluation analysis of the bounding Cooper N2 nozzle to extend applicability of the existing relief request to 60 years of operation, or 54 effective full power years (EFPY).

4.0 TECHNICAL APPROACH AND METHODOLOGY SIA Calculation 1400334.301, Rev1 [Attachment 10.1]:

Section 5.0, Methodology describes the method of analysis for 1400334.301, Rev1. This analysis developed a Finite Element Model (FEM) for the N2 nozzle (Reactor Recirculation Inlets) and determined the stresses caused by applicable Service __ Level A and B thermal transients and internal pressure. The stress distributions obtained as output of this analysis will be used as input for the probabilistic fracture mechanics (PFM) performed in 1400334.302, Rev1 [Attachment 10.2]. The following process is used:

1. Select limiting Service Level A/B transients from RPV and nozzle thermal cycle diagram (TCD),

2. Build nozzle FEM and analyze bounding thermal transient and pressure load case,

3. Perform mesh sensitivity check and model validation checks, ER 2017-027 Rev1

ENGINEERING REPORT NNPPD ER2017-027 Rev 1 COOPER NUCLEAR STATION Page 5 of 10

4. Extract stresses for subsequent calculation.

Review of the methodology was considered reasonable and therefore acceptable.

SIA Calculation 1400334.302. Rev1 [Attachment 10.2]:

This evaluation considers the nozzle-to-shell weld and nozzle blend radius on the N2 nozzle per References [9.2.31 and [9.2.4] and confirms that the nozzle still meets the acceptable failure probability considering the bounding fluence at the end of the Period of Extended Operation (PEO). Reference [9.2.6, PDF pg. 171] shows the highest fluence at 3.34x1016 n/cm2.

The acceptance criterion limits the difference in probability of failure per year due to the low temperature over pressure (L TOP) event to be no more than 5x10-5 when changing from full (100%) inc.service inspection to 25% inspection for the PEO. In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. If the resulting probability of failure per year due to an L TOP event (including 1 x 10-3 probability of L TOP event occurrence per year [Reference 9.2.3, pg.

5-131) is less than 5x10-5, then no comparison to the full inspection case is required.

The probabilistic evaluation is performed for the case of 25% inspection for the extended operating period (with zero inspection coverage conservatively assumed for the initial 40 years of operation).

The approach used for this evaluation is consistent with the methodology presented in BWRVIP-05 [Reference 9.2.8].

Review of the methodology was considered reasonable and therefore acceptable.

5.0 DESIGN INPUTS Section 4.0 of each analysis provides the various design inputs/criteria such as nozzle geometry, material properties, and transient definitions associated with stress distributions and fatigue cycles. The inputs and design criteria specific to each analysis were reviewed against CNS design basis requirements and are found to be acceptable.

6.0 ASSUMPTIONS The assumptions listed in Section 3.0 (1400334.301) and Section 6.0 (1400334.302) are considered reasonable and therefore acceptable. A key assumption is that zero examination coverage is assumed for the first 40 years of operation.

ER 2017-027 Rev1

ENGINEERING REPORT HNPPD ER2017-027 Rev 1 COOPER NUCLEAR STATION Page 6 of 10 7.0 EVALUATION SIA Calculation 1400334.301, Section 6.0, Analysis documents results of the analyses used in Section 5.0 of the SIA calculation. Review of the analysis approach is considered reasonable and therefore is acceptable. SIA Calculation 1400334.301 is an input to 1400334.302.

SIA Calculation 1400334.302, Revision 1, Table 4 in Attachment 10.2, summarized the probabilities of failure (PoF) per year due to the limiting L TOP event with 25% inspection for the extended operating term (with zero inspection coverage assumed for the first 40 years of operation). Revision 1 corrected the PoF per year due the L TOP event of the nozzle-blend-radius (Path 1) to 1.675x10-10 with Paths 2-4 being unchanged. A new column was added to the table providing the PoF per year due to normal operation for both the nozzle blend radius and nozzle-to-shell welds (Paths 1 - 4):

Table 4 from 1400334.302: PoF for Period of Extended Operation Zero inspection for initial 40 yl'ars, 25°/o for PEO Location PoF per year due PoF per year due to Al!owable PoF per to L TOP En*nt

• Normal Opc1*ation

,*em* (191 Path 1 1.675 ;.,: 10-10

< S.33 x 10-10 Pntb2

< 8.3Jx 10-B

< 8.33 X } o-lO 5.0 :-: )0"6 Path 3

< 8.33 X JO" JJ

< 8.33 X 10*JO Patb4

< 8_33 X w*J.l

< 8.33 X JO.JO

• Note: Values include 1x10*

3 probability of L TOP event occurrence per year (Reference 9.2.3, pg 5-13).

Paths 1/3 represents the nozzle blend radius PoF and Paths 2/4 represents the nozzle-to-shell weld PoF.

8.0 RESULTS The probabilities of failure (PoF) per year due to the limiting L TOP event with 25%

inspection for the extended operating term (with zero inspection coverage for the initial 40 years of operation) are summarized in Table 4 from 1400~34.302 [Attachment 10.2].

The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to L TOP events as well as normal operation are all less than the sx10*5 per year NRC safety goal established in NUREG 1806, Volume 1 [Reference 9.2.19].

This analysis shows that the N2 nozzles meet the acceptable failure probability even when considering elevated fluence level, thus qualifying all Cooper RPV nozzles with full penetration welds (except feedwater and control rod drive return nozzles) for reduced inspection using ASME Code Case N-702 to the end of the period of extended operation.

ER 2017-027 Rev1

-1

H NPPD COOPER NUCLEAR STATION

9.0 REFERENCES

ENGINEERING REPORT ER2017-027 Rev 1 Page 7 of 10 9.1 References from SIA Calculation 1400334.301:

1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR)

Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.

2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007, SI File No. BWRVIP.108P.

3. BWRVIP-241: BWR Vessel Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA. 1021005. EPRI PROPRIETARY INFORMATION.

4. ASME Boiler and Pressure Vessel Code, Section Ill, 1965 Edition through Winter 1966 Addenda

5. ASME Boiler and Pressure Vessel Code, Section Ill, 1971 Edition.

6. Combustion Engineering Drawing No. 232-233, Rev. 10, "LowerVessel Shell Ass'y, Machining & Welding," SI File No. 1400334.203.

7. Combustion Engineering Drawing No. 232-241, Rev. 5, "Nozzle Details," SI File No. 1400334.204.

8. Combustion Engineering Material Verification Report:

a. Document No. RVG-0000009619, Rev. 0, SI File No. 1400334.212.

b. Document No. RVG-0000009621, Rev. 0, SI File No. 1400334.213.

c. Document No. RVG-0000009478, Rev. 0, SI File No. 1400334.210.

d. Document No. RVG-0000009480, Rev. 0, SI File No. 1400334.209.

e. Document No. RVG-0000009485, Rev. 0, SI File No. 1400334.208.

f. Document No. RVG-0000011697, Rev. 0, SI File No. 1400334.211.

g. Document No. RVG-0000000252, Rev. 0, SI File No. 1400334.215.

h. Document No. RVG-0000000255, Rev. 0, SI File No. 1400334.216.

i. Document No. RVG-0000000286, Rev. 0, SI File No. 1400334.217.

ER 2017-027 Rev1

ENGINEERING REPORT HNPPD ER2017 -027 Rev 1 COOPER NUCLEAR STATION Page 8 of 10

j. Document No. RVG-0000000304, Rev. 0, SI File No. 1400334.218.

9. Structural Integrity Document, "Design Input Request," Rev. 1, SI File No.

1400334.207.

10. Thermal Cycle Diagrams:

a. General Electric Drawing No. 13589990, Sheet 2, Rev. 1, "Nozzle Thermal Cycles (Recirculation Inlet)," SI File No. 1400334.222.

b. General Electric Drawing No. 729E762, Sheet 1, Rev. 1, "Reactor Thermal Cycles," SI File No. 1400334.206.

11. ANSYS Mechanical APDL, Release 14.5 (w/ Service Pack 1 UP20120918),

ANSYS, Inc., September 2012.

12. BWRVIP-108NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2007. 1016123, 9.2 References from SIA Calculation 1400334.302:

1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR)

Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.

2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007, SI File No. BWRVIP.108P.

3. BWRVIP-108NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2007. 1016123.

4. BWRVIP-241: BWR Vessel Internal Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA. 1021005. EPRI PROPRIETARY INFORMATION.

5. SI Calculation 1400334.301, "'Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N2 Nozzle," Revision 0, August 2016.

ER 2017-027 Rev1

ENGINEERING REPORT NNPPD ER2017-027 Rev 1 COOPER NUCLEAR STATION Page 9 of 10

6. 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, CalculationNEDC 07-032, Revision 3, SI File No.

1400334.201.

7. SI Calculation 1400735.303, "Verification of Software VIPERNOZ Version 2.0,"

Revision 0, October 2015.

8. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.

9. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Structural Integrity Associates.

10. Combustion Engineering Drawing No. 232-241, Rev. 5, "Nozzle Details," SI

. File No. 1400334.204..

11. General Electric Drawing No. 13589990, Sheet 2, Rev. 1, "Nozzle Thermal Cycles (Recirculation Inlet)," SI File No. 1400334.222.

12. General Electric Drawing No. 729E762, Sheet 1, Rev. 1, "Reactor Thermal Cycles,"' SI File No. 1400334.206.

13. Combustion Engineering Drawing No. 232-233, Rev. 10, "LowerVessel Shell Ass'y, Machining & Welding," SI File No. 1400334.203.

14. SI Calculation W-EPRl-180-302, "Evaluation of effect of inspection on the probability of failure for BWR Nozzle-to-Shell-Welds and Nozzle Blend Radii Region,"' Revision 0.

15. BWRVIP-60-A: BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment, EPRI, Palo Alto, CA: 2003. 1008871. EPRI PROPRIETARY INFORMATION.

16. Bamford, W. H., "'Application of corrosion fatigue crack growth rate data to integrity analyses of nuclear reactor vessels,"' Journal of Engineering Materials and Technology, Vol. 101, 1979, SI File No. 1300341.213.

ER 2017-027 Rev1

ENGINEERING REPORT HNPPD ER2017-027 Rev 1 COOPER NUCLEAR STATION Page 10 of 10

17. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, 2013 Edition.

18. RCPB/Torus Fatigue Event Descriptions, CUF/EAF Component Record for Year 2014, "RVP Cycle Count Post RE28.PDF," SI File No. 1400334.219.

19. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), NUREG-1806, Vol. 1, August 2007.

20. USN RC Report, "Final Safety Evaluation of the BWR Vessel Internals Project BWRVIP-05 Report," TAC No. M93925, Division of Engineering Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, July 28, 1998.

21. Combustion Engineering Material Verification Report:

a. Document No. RVG-0000009619, Rev. 0, SI File No. 1400334.212.

b. Document No. RVG-0000009621, Rev. 0, SI File No. 1400334.213.

c. Document No. RVG-0000009478, Rev. 0, SI File No. 1400334.210.

d. Document No. RVG-0000009480, Rev. 0, SI File No. 1400334.209.

e. Document No. RVG-0000009485, Rev. 0, SI File No. 1400334.208.

f. Document No. RVG-0000011697, Rev. 0, SI File No. 1400334.211.

g. Document No. RVG-0000000252, Rev. 0, SI File No. 1400334.215.

h. Document No. RVG-0000000255, Rev. 0, SI File No. 1400334.216.

i. Document No. RVG-0000000286, Rev. 0, SI File No. 1400334.217.

j. Document No. RVG-0000000304, Rev. 0, SI File No. 1400334.218.

22. EPRI Letter 2012-138, "BWRVIP Support of ASME Code Case N-702 lnservice Inspection Relief," August 31, 2012, SI File No. 1300341.213.

10.0 ATTACHMENTS 10.1 SI Calculation 1400334.301, Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N2 Nozzle, Rev 1.

10.2 SI Calculation 1400334.302, Code Case N-702 Evaluation for Cooper Recirculation Inlet (N2) Nozzle, Rev 1.

10.3 Technical Review Comments and Resolutions Form ER 2017-027 Rev1

(> Structural Integrity Associates, Inc.

CALCULATION PACKAGE PROJECT NAME:

Coop N-702 Relief Request for 60 Years CONTRACT NO.:

4200002709, Rev 0 CLIENT:

PLANT:

File No.: 1400334.301 Project No.: 1400334 ER2017-027 Page 1 of 24 Quality Program: ~ Nuclear D Commercial Nebraska Public Power District Cooper Nuclear Station CALCULATION TITLE:

Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N2 Nozzle Document Affected Project Manager Preparer( s) &

Revision Pages Revision Description Approval Checker(s)

Si2nature & Date Si2natures & Date 0

1 - 22 Initial Issue Responsible Engineer:

A A-2 Richard Bax 8/9/16 Minji Fong 8/9/16 Responsible Verifier:

I Wilson Wong I

8/9/16 Res~onsible Engineer:

1 4,6 Removed proprietary wJM.-vJ6z5 WJM.-vJ markings

~

Wilson Wong Wilson Wong 3/2/18 3/2/18 Res~onsible Verifier:

~.~--A

"' II" -

Kevin Wong 3/2/18 I i I

Table of Contents ER2017-027 Page 2 of 24

1.0 INTRODUCTION

......................................................................................................... 4 2.0 OBJECTIVES................................................................................................................ 4 3.0 ASSUMPTIONS............................................................................................................ 4 4.0 DESIGN INPUTS.......................................................................................................... 5 4.1 Nozzle Geometry............................................................................................... 5 4.2 Material Properties............................................................................................. 5 4.3 Transient Definitions......................................................................................... 6 5.0 METHODOLOGY........................................................................................................ 6 5.1 Thermal Transient Selection.............................................................................. 6 5.2 Nozzle FEM and Load Case Evaluation............................................................ 6 5.3 Model Validation............................................................................................... 7 5.4 Post-Processing.................................................................................................. 7 6.0 ANALYSIS.................................................................................................................... 7 6.1 Bounding Transient Selection............................................................................ 8 6.2 Unit Internal Pressure Analysis......................................................................... 8 6.3 Thermal Transient Analyses.............................................................................. 9 6.3.1 Thermal Analyses............................................................................................... 9 6.3.2 Thermal Stress Analyses.................................................................................... 9 6.4 Model Validation............................................................................................... 9

7.0 CONCLUSION

............................................................................................................ 10

8.0 REFERENCES

............................................................................................................ 10 APPENDIX A FILENAMES............................................................................................... A-1

List of Tables ER2017-027 Page 3 of24 Table 1: Bounding Transients for Analysis [10]..................................................................... 12 Table 2: Material Properties for SA-533 Grade B Class 1 [ 4, 5]............................................ 13 Table 3: Material Properties for A-508 Class 2 [4, 5]............................................................. 13 Table 4: Material Properties for SA-240 Type 304 [ 4, 5]....................................................... 14 Table 5: Material Properties forE-8018 [4, 5]........................................................................ 14 List of Figures Figure 1: Dimensions Used in the Finite Element Model...................................................... 15 Figure 2: Components Included in the Finite Element Model............................................... 15 Figure 3: 3-D Finite Element Model Mesh for Analyses, Baseline Mesh.............................. 16 Figure 4: Path Locations for Through-Wall Stress Extractions.............................................. 16 Figure 5: Applied Boundary Conditions and Unit Internal Pressure...................................... 17 Figure 6: Total Stress Intensity Plot for Unit Internal Pressure..................... :........................ 17 Figure 7: Applied Thermal Boundary Conditions for Thermal Transient Analyses.............. 18 Figure 8: Applied Mechanical Boundary Conditions for Thermal Stress Analyses.............. 19 Figure 9: Temperature Contour for Loss ofFeedwater Pump Transient at Time=190 sec.... 19 Figure 10: Stress Intensity Plot for Loss ofFeedwater Pump Transient at Time=190 sec.... 20 Figure 11: Total Stress Intensity Contours for Mesh Sensitivity Study-Unit Pressure....... 21 Figure 12: Linearized Membrane-Plus-Bending Stress Intensity History for Path 1, Loss ofFeedwater Pump and Sudden Start of Cold Recirculation Loop Transients..... 22

1.0 INTRODUCTION

ER2017-027 Page 4 of 24 Cooper Nuclear Station intends to extend the applicability of Code Case N-702 [ 1] through the end of the period of extended operation (PEO, 60 years of operation). The Code Case allows for the reduction of in-service inspection from 100% to 25% of all Reactor Pressure Vessel (RPV) nozzle inner radii and nozzle-to-shell welds that must be performed every 10 years, including one nozzle from each system and pipe size, except for the feedwater and control rod drive return nozzles.

Technical documents BWRVIP-108 [2] and BWRVIP-241 [3] provide the technical basis for the code case, but only consider 40 years of plant operation. In order to extend the applicability of Code Case N-702 [1 ], a probabilistic fracture mechanics (PPM) evaluation, consistent with the methods of BWRVIP-108 [2] and BWRVIP-241 [3], must be performed to ensure the probability of failure remains acceptable. The N2 (Recirculation Inlet) nozzles of Cooper Nuclear Station are identified as the bounding RPV nozzles when fluence is not considered [3].

2.0 OBJECTIVES The objectives of this calculation package are to:

1. Develop a Finite Element Model (FEM) for the N2 nozzle and to,

2. Determine the stresses caused by applicable Service Level A and B thermal transients and internal pressure.

The stress distributions obtained as output of this analysis will be used as input for a subsequent probabilistic fracture mechanics (PPM) evaluation to be performed in a separate calculation package.

3.0 ASSUMPTIONS The following assumptions are made in this evaluation:

The N2 nozzle-to-safe end weld is not specifically modeled. Instead the material instantaneously transitions from the nozzle material to the safe end material. Since the location of stress extraction (see Figure 4) is not near the transition, the effect of this assumption on the analysis output is minimal.

Density and Poisson's ratio for all materials are considered temperature independent. In addition, typical values are assumed. This is consistent with the manner in which these properties are presented in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code.

The inside surfaces are subjected to an essentially infinite convective heat transfer coefficient (HTC) of 10,000 Btu1hr-ft2-°F to maximize through-wall thermal gradients and thus thermal stresses.

A heat transfer coefficient of 0.2 Btu1hr-ft2-°F with an ambient temperature of l00°F is applied on the outside surfaces of the RPV and nozzle. These represent typical values for the RPV and

ER2017-027 Page 5 of 24 nozzle's external insulated surfaces and the containment temperature. These values will be used at all times for any evaluated transient.

The thermal sleeve of the N2 nozzle is not modeled. Since the thermal sleeve is not a structural component and the heat transfer coefficient on the inside surfaces are assumed to maximize through-wall thermal gradients, the thermal sleeve is not required and is conservatively excluded.

4.0 DESIGN INPUTS This section introduces the design inputs used for the analyses documented in this calculation package.

The following items are discussed separately below:

Nozzle Geometry, Material Properties, Transient Definitions.

4.1 Nozzle Geometry The geometry of the recirculation inlet nozzle is obtained from component drawings [6, 7]. The RPV inside radius (IR) is 110.375 inches (to base metal, not including the 0.3125 inch cladding [6]), with a vessel wall thickness of6.875 inches [7], resulting in an outside radius (OR) of 117.25 inches. The N2 nozzle has an IR of6.1875 inches (to base metal, not including the 0.21875 inch cladding [7]),

with an outside radius (OR) of 10.21875 inches on the vessel side and 7.1875 inches on the safe end side of the nozzle [7]. The nozzle to vessel blend radius is 4.0 inches on the inside and outside [7].

Figure 1 shows the nozzle geometry with key dimensions identified.

4.2 Material Properties The material component identifications are as listed:

RPV shell:

Nozzle forging:

Cladding Nozzle to vessel weld:

SA-533 Grade B Class 1 [8a and 8b ],

ASTM A-508 Class 2 [8c, 8d, 8e, and 8t],

SA-240 Type 304 [9],

E-8018 electrode [8g, 8h, 8i, and 8j]

(carbon-moly steel properties used for the analysis)

Figure 2 identifies the applicable material type on an image of the nozzle model.

The material properties are taken from the Design Code of Construction [9], the 1965 Edition through the Winter 1966 Addenda of the ASME B&PV Code,Section III [4]. Since this Edition of the Code does not list thermal conductivity or diffusivity, these values are obtained from the 1971 Edition of the Code [5]. All material properties used in the calculations are presented in Table 2 through Table 5.

4.3 Transient Definitions ER2017-027 Page 6 of24 The thermal transient definitions are obtained from Reference [10]. Only normal and upset (Service Level A and B) transients for the RPV and N2 nozzle specific transients are considered [10a, 10b].

5.0 METHODOLOGY This section describes the methodology used to perform the analyses documented in this calculation package, which follows the same procedure as BWRVIP-108 & 241 [2, 3]. The N2 (Recirculation Inlet) nozzles are identified as the bounding RPV nozzles when fluence is not considered (geometry only) [3].

The general analytical methodology is introduced followed by specific discussion of the modeling method.

The following process is used:

1. Select limiting Service Level A/B transients from RPV and nozzle thermal cycle diagram (TCD),

2. Build nozzle FEM and analyze bounding thermal transient and pressure load case,

3. Perform mesh sensitivity check and model validation checks,

4. Extract stresses for subsequent calculation.

5.1 Thermal Transient Selection Two bounding transients are selected for evaluation rather than analyzing all Service Level NB transients defined on the RPV and recirculation inlet nozzle TCDs. Separate load cases are defined for the internal pressure and thermal portions of the transients. All analyses are performed using linear elastic methods for material models and loading; therefore, a single "unit" pressure load case is evaluated from which stress distributions at any pressure-can be scaled from the results of the "unit" load case.

The bounding thermal transients are selected based on maximum temperature fluctuation and rate of change.

5.2 Nozzle FEM and Load Case Evaluation All structural analyses are performed using the finite element method. A three-dimensional (3-D) finite element model is constructed using the ANSYS finite element program [11]. The model is used for linear elastic pressure and thermal transient stress analyses. It is developed as a symmetric quarter model using the dimensions given in Reference [6, 7], and includes a local portion of the RPV shell, the N2 nozzle-to-vessel weld, the N2 nozzle, and a portion of the attached safe end, as shown in Figure 2. The N2 nozzle-to-safe end weld is not modeled because it is sufficiently far from the region

ER2017-027 Page 7 of24 of interest to introduce any significant influence. The model is meshed with the SOLID45 element type, of which the thermal equivalent is SOLID70. The mesh used in this calculation is depicted in Figure 3.

A 1000 psig "unit" internal pressure load case is evaluated from which nozzle stress distributions for any desired pressure can be obtained by linearly scaling the results of the "unit" load case. Nozzle piping loads are not considered since stresses in the blend radius and nozzle-to-vessel weld locations caused by these loads are insignificant according to Reference [12, Section 5.5].

5.3 Model Validation The FEM built for this analysis is validated using three separate checks:

1. Mesh size check To ensure the mesh provides stress results in the region of interest which are insensitive to the local mesh refinement, a second model is created utilizing twice the mesh density in the nozzle-to-vessel weld and blend radius regions. The acceptance criterion for adequate mesh density is a change in peak stress intensity between solutions ofless than 1 %.

2. Time step check Excessively large time steps during thermal transient stress analysis may cause stress peaks to be missed. This can be checked by ensuring there are no sharp changes when plotting linearized through-wall stress intensity time history results, and that there are multiple time points located in the peak and valley regions.

3. Far field stress check If the model boundaries are sufficiently far from the region of interest and if the boundary conditions are correct, hoop stresses in the RPV shell should be within a few percent of a hand calculation approximation. The hoop and axial stress due to internal and end cap pressure can be approximated using the formula for thin walled cylinders: Pr/t for hoop stress and Pr/2t for axial stress, where P is the internal pressure, r is the mean radius, and t is the wall thickness.

5.4 Post-Processing In support of future PFM analysis, four through-wall stress paths, two each at 0° and 90°, are defined within the region of the N2 nozzle blend radius and nozzle-to-vessel weld, as shown in Figure 4. Since the model is symmetric, these paths also represent the stress at 180° and 270°, respectively. Stresses from the thermal transient and pressure load cases are extracted and saved in *.csv file format which can be imported to an Excel workbook for further processing (see Appendix A for file listings).

6.0 ANALYSIS This section documents the results of the analyses described in Section 5.0 above. The following items are discussed in separate sections below:

1. Bounding Transient Selection,

2. Unit Internal Pressure Analysis,

3. Thermal Transient Analyses,

4. Model Validation.

6.1 Bounding Transient Selection ER2017-027 Page 8 of24 The nozzle specific transient "Sudden Start of Cold Recirculation Loop (SSPC)" located between Zone 18 and 19 [10a] has an instantaneous temperature down shock to 130°F from the normal operating temperature, before instantaneously returning to the normal operating temperature after only 54 seconds, bounding all Service Level A/B recirculation inlet nozzle specific transients. Due to the severity of the transient, the vessel "Loss ofFeedwater Pump, Isolation Valve Closed (FWP)" transient located in Zone 11 to 12 of Region B [10b] is selected to bound all other vessel transients in Reference [10b].

6.2 Unit Internal Pressure Analysis A unit internal pressure, P = 1,000 psi, is applied to the interior surfaces of the model. For the pressure run, cladding was removed since cladding cannot be credited as structural material per the ASME Code Section NB-3122 [5]. An end-cap load is applied to the free end of the nozzle piping in the form of tensile axial pressure, as calculated below.

where, Pec1 p

IR1 OR1 p

p. JR/

1000

• 6.1875 2

_ 2 862 ecJ- (OR1 2 -/R/)-(7.1875 2 -6.1875 2 )-

psi End cap pressure on attached piping free end (psi)

= Internal unit pressure (psi)

= Inside radius of modeled nozzle piping (in)= 6.1875 inches (w/o cladding) [7]

= Outside radius of modeled nozzle piping (in) =7.1875 inches [7]

The internal pressure also induces an end-cap load on the axial free end of the modeled vessel shell, as calculated below.

where, Pec2 =

p

=

IR2 =

OR2 =

p

=

P*IR/

1000-110.375 2

_ 7 7*85 ecz (OR2 2 -/Rz2)-(117.25 2 -110.375 2 )-

psi End cap pressure on vessel shell axial free end (psi)

Internal unit pressure (psi)

Inside radius of modeled vessel shell (in)= 110.375 inches (w/o cladding) [6]

Outside radius of modeled vessel shell (in)= 117.25 inches [6, 7]

Symmetric boundary conditions are applied at the vessel's circumferential free end and the overall model's two planes of symmetry. The free end of the nozzle piping and axial free end of the vessel shell

ER2017-027 Page 9 of 24 are coupled in their respective axial directions to simulate the remaining portions of the geometry not included in the model. The applied pressure loads and boundary conditions for this case are shown in Figure 5. A representative stress intensity contour plot for the unit pressure analysis is shown in Figure 6.

6.3 Thermal Transient Analyses The transients to be analyzed are tabulated in Table 1. In order to achieve a final steady state condition, an arbitrary time of 3,600 seconds is added after the last load step of each transient, followed by an imposed steady state load step (at an arbitrary 400 seconds after the 3,600 seconds of additional time).

6.3.1 Thermal Analyses Bulk fluid temperatures and heat transfer coefficients are applied to the inside and outside surface nodes of the model. The inside surfaces are subjected to a conservative high convective heat transfer coefficient (HTC) of 10,000 Btu/hr-ft2-°F. The heat transfer coefficient of 0.2 Btu/hr-ft2°F, along with an ambient temperature of 100°F. The symmetry planes, top/bottom surfaces of the RPV, and "cut" plane in the piping are treated as adiabatic. Figure 7 depicts a representative plot of the thermal boundary conditions applied for the transient analysis.

6.3.2 Thermal Stress Analyses Symmetric boundary conditions are applied at the vessel's circumferential free end and the overall model's two planes of symmetry. The free end of the nozzle piping and axial free end of the vessel shell are coupled in their respective axial directions to simulate the remaining portions of the geometry not included in the model. Figure 8 shows a representative plot of the mechanical boundary conditions applied for the thermal transient stress analysis. A representative temperature contour and total stress intensity contour plot at an arbitrary time step for the Loss ofFeedwater Pump transient are shown in Figure 9 and Figure 10, respectively.

6.4 Model Validation

. A unit internal pressure load with end cap loads and boundary conditions, as discussed in Section 6.2, was applied to both the base and double mesh density models. Figure 11 depicts both meshed models and shows that there is less than 0.01 % difference in the maximum stress intensities at the nozzle blend radius. Therefore, the mesh density originally chosen for this calculation does not need further refinement and is adequate for this calculation.

To validate that the time step sizes chosen for the thermal transient solutions described in Section 6.3 were sufficient, the linearized through-wall stress intensity time history results from the thermal transient load cases along the four paths shown in Figure 4 are reviewed to ensure a smooth stress history response is obtained. Representative plots of the Path I linearized membrane-plus-bending stress intensity history are shown in Figure 12 and clearly show the time points and resulting smooth

ER2017-027 Page 10 of 24 stress response. Peaks and valleys are clearly represented by multiple time points ensuring no peaks were missed and time stepping is adequate.

Using the formulae for thin walled cylinders (precise for R/t>> 10), the hoop stress for the vessel should be:

p

= P*r _1000-113.8125 _ 16555 hoop (t) -

(6.875) psi

where, Phoop = Thin wall cylinder predicted hoop stress (psi)

P

= Internal unit pressure (psi) r

= Mean radius of modeled vessel shell (in)= 113.8125 inches (w/o cladding) [6, 7]

t

= Wall thickness of vessel (in)= 6.875 inches (w/o cladding) [6, 7]

Axial stresses should be half the hoop stresses. Nodal query of hoop stress indicates a variance of less than 5% from the thin wall formulae at the vessel mid-wall, which is acceptable since the r/t ratio for the vessel is approximately 16.6.

7.0 CONCLUSION

Unit pressure and thermal transient stress analyses have been performed. Stress results were extracted from all analyses for through-wall paths at locations of interest along the N2 nozzle blend radius and nozzle-to-vessel weld in support of future PFM calculations. All of the stress results are stored in computer files for later use (see Appendix A for file listings).

8~0 REFERENCES

1.

Code Case N-702, "Alternative Requirements for Boiling Water Reactor (B\\YR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.

2.. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007, SI File No. BWRVIP.108P.

3.

BWRVIP-241: BWR Vessel Internals Project, Probabilistic Fracture Mechanics Evaluationfor the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA.

1021005. EPRI PROPRIETARY INFORMATION.

4.

ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition through Winter 1966 Addenda

5.

ASME Boiler and Pressure Vessel Code,Section III, 1971 Edition.

6.

Combustion Engineering Drawing No. 232-233, Rev. 10, "Lower Vessel Shell Ass'y, Machining &

Welding," SI File No. 1400334.203.

7.

Combustion Engineering Drawing No. 232-241, Rev. 5, "Nozzle Details," SI File No. 1400334.204.

8.

Combustion Engineering Material Verification Report:

a. Document No. RVG-0000009619, Rev. 0, SI File No. 1400334.212.

b. Document No. RVG-0000009621, Rev. 0, SI File No. 1400334.213.

c. Document No. RVG-0000009478, Rev. 0, SI File No. 1400334.210.

d. Document No. RVG-0000009480, Rev. 0, SI File No. 1400334.209.

e. Document No. RVG-0000009485, Rev. 0, SI File No. 1400334.208.

f.

Document No. RVG-0000011697, Rev. 0, SI File No. 1400334.211.

g. Document No. RVG-0000000252, Rev. 0, SI File No. 1400334.215.

h. Document No. RVG-0000000255, Rev. 0, SI File No. 1400334.216.

1.

Document No. RVG-0000000286, Rev. 0, SI File No. 1400334.217.

J.

Document No. RVG-0000000304, Rev. 0, SI File No. 1400334.218.

ER2017-027 Page 11 of 24

9.

Structural Integrity Document, "Design Input Request," Rev. 1, SI File No. 1400334.207.

10. Thermal Cycle Diagrams:

a. General Electric Drawing No. 135B9990, Sheet 2, Rev. 1, "Nozzle Thermal Cycles (Recirculation Inlet)," SI File No. 1400334.222.

b. General Electric Drawing No. 729E762, Sheet 1, Rev. 1, "Reactor Thermal Cycles," SI File No. 1400334.206.

11. ANSYS Mechanical APDL, Release 14.5 (w/ Service Pack 1 UP20120918), ANSYS, Inc.,

September 2012.

12. BWRVIP-108NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii.

EPRI, Palo Alto, CA: 2007. 1016123, SI File No. BWRVIP.108NP.

Notes:

1.

2.

3.

a e :

T bl 1 B d"

oun m2 T Description

Time, sec Loss ofFeedwater Pump, 0

Isolation Valve Closed 190 (FWP)<3l 2290 0

Sudden Start of Cold 0.1 Recirculation Loop (SSPC) 54.l 54.2 rans1ents or nalVSlS

£ A

I. [10]

Inside T nozzle, °F Tvessel, °F Surface h, Btu/hr-n2_0F 522 522 300 300 500 500 522 522 10000 130 522 130 522 522 522 ER2017-027 Page 12 of 24 Outside Surface h, Btu/hr-ft2-oF 0.2 A total of 3,600 seconds is added to the end of each transient followed by an imposed steady state condition 400 seconds later (see Section 6.3 of this calculation).

Flow rates are not considered since the inside surface heat transfer film coefficients are essentially infinite.

The "Loss ofFeedwater Pump, Isolation Valve Closed (FWP)" transient has three similar internal cycles of temperature. Only the first cycle of the transient is evaluated since the first cycle is the most severe. The initial down shock ramp time of the transient is not clear in Reference 10b.

Therefore, a value of 3 minutes plus the indicated 10 seconds [ 1 Ob] is conservatively used for the ramp time.

Table 2: Material Properties for SA-533 Grade B Class 1 [4, 5]

Temperature, Young's Mean Thermal

Modulus, Expansion, OF x106 psi xl0-6 in/in/°F 70 29.9 6.10 200 29.5 6.38 300 29.0 6.60 400 28.6 6.82 500 28.0 7.02 600 27.4 7.23 Density, p = 0.283 lb/in3, assumed temperature independent.

Poisson's Ratio, u = 0.3, assumed temperature independent.

Thermal Conductivity, xl0-4 Btn/sec-in-°F 8.33 8.07 7.69 7.31 6.90 6.55 ER2017-027 Page 13 of 24 Specific

Heat, Btu/lb-°F 0.113 0.117 0.121 0.124 0.128 0.132 Note: Specific Heat values are derived from the equation shown in General Note (5) of Table I-4.0 [5], Specific Heat= TC I (TD x density).

Table 3: Material Properties for A-508 Class 2 [4, 5]

Temperature, Young's Mean Thermal

Modulus, Expansion, OF x106 psi xl o-6 in/in/°F 70 29.9 6.10 200 29.5 6.38 300 29.0 6.60 400 28.6 6.82 500 28.0 7.02 600 27.4 7.23 Density, p = 0.283 lb/in3, assumed temperature independent.

Poisson's Ratio, u = 0.3, assumed temperature independent.

Thermal Conductivity, x10-4 Btu/sec-in-°F 8.33 8.07 7.69 7.31 6.90 6.55 Specific

Heat, Btu/lb-°F 0.113 0.117 0.121 0.124 0.128 0.132 Note: Specific Heat values are derived from the equation shown in General Note (5) of Table I-4.0 [5], Specific Heat= TC I (TD x density).

Table 4: Material Properties for SA-240 Type 304 [4, SJ Temperature, Young's Mean Thermal

Modulus, Expansion, OF x106 psi x10*6 in/in/°F 70 27.4 9.20 200 27.1 9.34 300 26.8 9.47 400 26.4 9.59 500 26.0 9.70 600 25.4 9.82 Density, p = 0.29 lb/in3, assumed temperature independent.

Poisson's Ratio, u = 0.31, assumed temperature independent.

Thermal Conductivity, x10*4 Btu/sec-in-°F 1.93 2.06 2.16 2.27 2.37 2.48 ER2017-027 Page 14 of 24 Specific

Heat, Btu/lb-°F 0.111 0.115 0.117 0.120 0.123 0.125 Note: Specific Heat values are derived from the equation shown in General Note (5) of Table 1-4.0 [5], Specific Heat= TC / (TD x density).

Table 5: Material Properties for E-8018 [4, SJ Temperature, Young's Mean Thermal

Modulus, Expansion, OF xl06 psi xto*6 in/in/°F 70 29.9 6.10 200 29.5 6.38 300 29.0 6.60 400 28.6 6.82 500 28.0 7.02 600 27.4 7.23 Density, p = 0.283 lb/in3, assumed temperature independent.

Poisson's Ratio, u = 0.3, assumed temperature independent.

Thermal Conductivity, x10*4 Btu/sec-in-°F 8.33 8.07 7.69 7.31 6.90 6.55 Specific

Heat, Btu/lb-°F 0.113 0.117 0.121 0.124 0.128 0.132 Notes: 1) Specific Heat values are derived from the equation shown in General Note (5) of Table 1-4.0 [5],

Specific Heat= TC / (TD x density).

2) E-8018 is treated as carbon moly steel for this analysis.

Figure 1: Dimensions Used in the Finite Element Model (Units for dimensions in inches)

COOP IJ2 - V:pemoi M:lde: - 3.'.l N...S.16\\'e,;sd\\\\'eN F.-AAIII

\\ \\

Figure 2: Components Included in the Finite Element Model ER2017-027 Page 15 of 24

ER2017-027 Page 16 of 24 Figure 3: 3-D Finite Element Model Mesh for Analyses, Baseline Mesh Figure 4: Path Locations for Through-Wall Stress Extractions

-/184.82 -6 CS. 73 -!>832.64 4F.56.SS -J~~J. 46 -?904.37 CXU' _'J? - \\/lrx,...-.cz 'b:iel - ::r, "l.l -957. 133 ' 3 * ',!OB!>

1 OOJ Figure 5: Applied Boundary Conditions and Unit Internal Pressure (Units for Pressure in psi) r-ror.L..:u:r,:o,

,'JU'-]

!lJB *1 TDS-l Slltr (A'/j

~c,.,.:

CMO: *~O::"'.'~t;i

l-1J -1:,,:q,'7

M-&--732.:49 1'{ -4c;*.l\\2.r, a-t~1~9l:lj~::i L"l.09.'Z 6131. 81

o!>J. \\I 15976 CXU' _ 'J? - VI pP. '"nc":7. "hie 1

C 35664. < 4CaB6.4 455J3. 5 Figure 6: Total Stress Intensity Plot for Unit Internal Pressure (Units for stress intensity in psi)

ER2017-027 Page 17 of24

.38oE--06. :l:?! 44

.00428 I

.CCf.43

.o* 50:l'

. Cl 114 I

.019?9 a:;a> _ 'J;> - VI pc,.,-,r,7 "b:le 1 3C a) Heat transfer coefficient (HTC) o..E:IC\\":'t T'i.\\;; N.t{

~ mi 100 1'6. 889 VI pc,.,-,r,,: "b:le 1

93

• ll~

?4C.. 667 3C 2B I.,,o 334.444 b) Bulk temperature (TBULK) 478.777 4 lo.i:: 577 ER2017-027 Page 18 of24 Figure 7: Applied Thermal Boundary Conditions for Thermal Transient Analyses (End of Loss ofFeedwater Pump, Isolation Valve Closed (FWP) Transient shown)

(Units for HTC in Btu/sec-in2-°F, TBULK in °F)

ER2017-027 Page 19 of24 Figure 8: Applied Mechanical Boundary Conditions for Thermal Stress Analyses JQ2. I 375. 848 J!>0. 9%

375. 144

=x> V - VJ

..,....,7 "t":d<?l -

31":

Figure 9: Temperature Contour for Loss of Feedwater Pump Transient at Time=190 sec.

(Units for temperature in °F)

200 *263 3615. Cl 1029* 1~

1C444.5 13~~~-"

17714 o::J:l> 'l? - VI

. me? "bcie 1 3D 74* 03. 5

, 1~18 ' 2 30933 ER2017-027 Page 20 of 24 Figure 10: Stress Intensity Plot for Loss of Feedwater Pump Transient at Time=190 sec.

(Units for stress intensity in psi)

ER2017-027 Page 21 of 24

---

~ Fine Mesh Figure 11: Total Stress Intensity Contours for Mesh Sensitivity Study-Unit Pressure (Units for stress intensity in psi)

Vl

a.

25000 20000

?:: 15000 V,

C

.3 C

-:; 10000

~

~

\\/1 5000 0

0 500 1000 1500 Time (sec)

SINT(I)

-M-SINT(O) 2000 2500 Loss ofFeedwater Pump, Isolation Valve Closed (FWP) 35000 30000

~ 25000

?:: 20000 vi C

.8 C 15000 V,

V,

~ 10000

~

\\/1 5000 0

0 so 100 150 Time (sec)

SINT(I)

-M-SINT(O) 200 250 Sudden Start of Cold Recirculation Loop 3000 300 ER2017-027 Page 22 of 24 Figure 12: Linearized Membrane-Plus-Bending Stress Intensity History for Path 1, Loss of Feedwater Pump and Sudden Start of Cold Recirculation Loop Transients Note: SINT(I) and SINT(O) refer to the membrane-plus-bending stress intensity on the inside and outside surfaces of the model, respectively.

APPENDIX A FILENAMES ER2017-027 Page 23 of 24

File Name COOP N2.INP COOP N2 X2.INP MProp _Linear_ COOP.INP COMPONENTS.INP TRANS FWP.INP TRANS SSPC.INP COOP PRESS.INP COOP PRESS X2.INP THM _ FWP _ mntr.inp THM _ SSPC _ mntr.inp GenStress.mac GETPATH.TXT STR

• COE P?.CSV Description Input file to construct the 3-D model for linear-elastic analysis Input file to construct the 3-D model with Refined Mesh ER2017-027 Page 24 of 24 Input file of temperature dependent linear elastic material properties Input file to define thermal analysis components Analysis input file for Loss ofFeedwater Pump, Isolation Valve Closed (FWP) Transient Analysis input file for Sudden Start of Cold Recirculation Loop Transient Analysis input file for Unit Internal Pressure Analysis input file for Unit Internal Pressure using Refined Mesh Load step definition file from thermal analysis -Loss ofFeedwater Pump, Isolation Valve Closed (FWP)

Load step definition file from thermal analysis - Sudden Start of Cold Recirculation Loop Path stress extraction macro file to extract.CSV files Through-wall path definition file Curve fit coefficients outputs of stresses in tabulated forms

• =PRESS, FWP, and SSPC

? = path number 1-4

I I

I)

Structural Integrity Associates, Inc.

CALCULATION PACKAGE PROJECT NAME:

Cooper N702 Relief Request for 60 Years CONTRACT NO.:

4200002709Rev.O CLIENT:

PLANT:

File No.: 1400334.302 Project No.: 1400334 Quality Program: ~ Nuclear ER2017-027 Page 1 of 19 D Commercial Nebraska Public Power District Cooper Nuclear Station CALCULATION TITLE:

Code Case N-702 Evaluation for Cooper Recirculation Inlet (N2) Nozzle Document Affected Project Manager Preparer( s) &

Revision Pages Revision Description Approval Checker(s)

Signature & Date Signatures & Date Responsible Engineer:

0 1 - 17 Initial Issue A-1 -A-2 Wilson Wong Wilson Wong 8/9/16 8/9/16 Responsible Verifier:

Minji Fong 8/9/16 I

I i

I I

Responsible Engineer:

I 1

12, 16 Updated calculation w~vJ~

\\Al~vJ~

A-2 results in Table 4 based I

on CAR-18-003, and Wilson Wong Wilson Wong I

I changed Reference 15 to 3/2/18 3/2/18 remove proprietary markings Responsible Verifier:

A tl*

""Ill'

!;I"."

Kevin Wong 3/2/18 I

Table of Contents ER2017-027 Page 2 of 19

1.0 INTRODUCTION

......................................................................................................... 3 2.0 OBJECTIVE.................................................................................................................. 3 3.0 METHODOLOGY........................................................................................................ 3 3.1 Fatigue Cycles................................................................................................... 3 3.2 Probabilistic Fracture Mechanics Evaluation....................................................4 4.0 DESIGN INPUT............................................................................................................ 5 4.1 Deterministic Parameters................................................................................... 5 4.1.1 ISl....................................................................................................................... 5 4.1.2 Stresses.............................................................................................................. 5 4.1.3 Fatigue Cycles................................................................................................... 5 4.2 Random Variables............................................................................................. 5

4. 2.1 SCC Initiation.................................................................................................... 6 4.2.2 SCC Growth....................................................................................................... 6 4.2.3 Fatigue Crack Growth....................................................................................... 7 5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS....................................... 8 6.0 ASSUMPTIONS............................................................................................................ 9 7.0 RESULTS OF ANALYSES........................................................................................ 10

8.0 CONCLUSION

S......................................................................................................... 10

9.0 REFERENCES

............................................................................................................ 11 Appendix A LIST OF SUPPORTING FILES...................................................................... A-1 List of Tables Table 1: Deterministic Parameter Summary............................................................................ 13 Table 2: Probability of Detection Distribution [14]................................................................ 13 Table 3: Random Variables Parameter Summary for N2 Nozzle............................................ 14 Table 4: PoF for Period of Extended Operation..................................................................... 15 List of Figures Figure 1: Stress Extraction Path Orientations in the N2 Nozzle............................................ 15 Figure 2: Pressure Stress Distributions for the N-702 Evaluation.......................................... 16 Figure 3: Full Power Thermal Expansion Stress Distributions for the N-702 Evaluation..... 16 Figure 4: Bounding Transient Stress Distributions for the N-702 Evaluation....................... 17 Figure 5: Weld Residual Stress Distributions for Paths 2 and 4............................................. 17

1.0 INTRODUCTION

ER2017-027 Page 3 of 19 Nebraska Public Power District intends to extend the applicability of Code Case N-702 [1] for Cooper Nuclear Station through the end of the period of extended operation (PEO). The Code Case allows reduction of in-service inspection from 100% to 25% of all nozzle blend radii and nozzle-to-shell welds every 10 years, including one nozzle from each system and pipe size, except for feedwater and control rod drive return nozzles.

Technical documents BWRVIP-108 [2, 3] and BWRVIP-241 [4] provide the basis for the code case, but only consider 40 year plant operation. In order to extend the applicability of Code Case N-702, a probabilistic fracture mechanics (PFM) evaluation, consistent with the methods ofBWRVIP-108 and BWRVIP-241, is performed to ensure that the probability of failure remains acceptable. The N2 (Recirculation Inlet) nozzles are identified as the bounding nozzles when fluence is not considered [ 4].

The evaluation consists of two parts: Finite Element Model (FEM) Stress Analysis and Probabilistic Fracture Mechanics (PFM) Analysis. The FEM stress analysis is performed in a separate calculation [5]

while this calculation package documents the PFM analysis.

2.0 OBJECTIVE The objective of the evaluations documented in this calculation package is to perform a plant specific analysis of the bounding Cooper N2 nozzle to extend applicability of the existing relief request to 60 years of operation, or 54 effective full power years (EFPY).

3.0 METHODOLOGY This evaluation considers the nozzle-to-shell weld and nozzle blend radius on the N2 nozzle per Reference [3] and [ 4] and confirms that the nozzle still meets the acceptable failure probability considering the bounding fluence at the end of the PEO. Reference [6, PDF pg. 171] shows the highest fluence at 3.34xl016 n/cm2.

The acceptance criterion limits the difference in probability of failure per year due to the low temperature over pressure (LTOP) event to be no more than 5x10-6 when changing from full (100%) in-service inspection to 25% inspection for the PEO. In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. If the resulting probability of failure per year due to an LTOP event (including 1 x 10-3 probability of LTOP event occurrence per year

[3, pg 5-13]) is less than 5xl0-6, then no comparison to the full inspection case is required.

3.1 Fatigue Cycles In the FEM calculation [5], two bounding transients were defined to conservatively include fatigue crack growth contributions from all the normal and upset (Service Levels A and B) thermal transients defined for the N2 nozzles. The nozzle specific transient "Sudden Start of Cold Recirculation Loop (SSPC)"

located between Zone 18 and 19 of the nozzle thermal cycles [ 11] has an instantaneous temperature down shock to 130°F from the normal operating temperature, before instantaneously returning to the

ER2017-027 Page 4 of 19 normal operating temperature after only 54 seconds, bounding all Service Level NB recirculation inlet nozzle specific transients. Due to the severity of the transient, the vessel "Loss of Feedwater Pump, Isolation Valve Closed (FWP)" transient located in Zone 11 to 12 of Region B [12] is selected to bound all other vessel transients in Reference [12]. The number of cycles used in the N-702 evaluation is defined in Section 5.0.

3.2 Probabilistic Fracture Mechanics Evaluation The probabilistic evaluation is performed for the case of 25% inspection for the extended operating period (with zero inspection coverage conservatively assumed for the initial 40 years of operation).

For the nozzle blend radius region, a nozzle blend radius crack model from ASME Code, Appendix G

[17] is used in the probabilistic fracture mechanics evaluation. For this location and crack model, the applicable stress is the stress perpendicular to a path defined 90 degrees from the tangent drawn at the blend radius.

For the nozzle-to-shell weld, either a circumferential or an axial crack, depending on weld orientation, can initiate due to either component fabrication (i.e. considering only welding process) or stress corrosion cracking. From BWRVIP-05 [8], it is shown that the probability of failure for a circumferential crack is less than an axial crack; due to the difference in the stress (hoop versus axial) and the influence on the crack model. However, this probabilistic fracture mechanics evaluation for the nozzl~ and vessel shell weld considers both circumferential and axial cracks ( depending on weld orientation).

An axial elliptical crack model with a crack aspect ratio of a/1 = 0.5 is used in the evaluation for the nozzle-to-shell weld. The inspection probability of detection (PoD) curve from BWRVIP-05 [8] (Table

2) is utilized with a ten year inspection interval. The calculation of stress intensity factor is at the deepest point of the crack.

The approach used for this evaluation is consistent with the methodoiogy presented in BWRVIP-05 [8].

A Monte Carlo simulation is performed using a variant of the VIPER program [9]. The Monte Carlo method can be used to solve probabilistic problems using deterministic computation. A mean value, standard deviation, and distribution curve as defined in the random variables summary (Table 3) defines a set of possible inputs and their probabilities of occurring. Using this domain of possible inputs, a set of inputs are generated for use in determining whether the nozzle will fail using conventional deterministic fracture mechanics methodology. This is repeated 20 million times. The number of simulations in which the nozzle is determined to fail divided by the number ofsimulations run gives the probability of failure.

The VIPER program was developed as part of the BWRVIP-05 effort for Boiling Water Reactor (BWR) reactor pressure vessel (RPV) shell weld inspection recommendations. The software was modified into a separate version, identified as VIPERNOZ, for use in this evaluation. The detailed description of the methodology incorporated in the VIPER/VIPERNOZ program is documented in References [8] and [3].

--i

ER2017-027 Page 5 of 19 The modified software for this project is identified as VIPERNOZ to distinguish from the original VIPER software, and is verified on a project specific basis [7] to ensure the modifications made to the VIPER software are fully quality assured.

4.0 DESIGN INPUT The plant specific input is described below. Section 4.1 presents all inputs modeled deterministically as constants while Section 4.2 describes the probabilistic treatment of inputs considered to be random variables (RV) in the VIPERNOZ code.

4.1 Deterministic Parameters Table 1 summarizes the dimensional and operational inputs used in the N-702 evaluation [10,11,12,13].

Subsections 4.1.1 through 4.1.3 describe the more detailed input parameters used for in service inspection (ISI) interval, stress distributions and fatigue cycles, respectively.

4.1.1 ISi In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. The probability of detection (POD) distribution function associated with inspection is shown in Table 2 [14].

4.1.2 Stresses Stresses due to vessel pressure and bounding thermal transients are determined in the Finite Element Model Development and Thermal Mechanical Stress Analyses for the N2 nozzle [5]. In that calculation package, through wall stress distributions are presented at four locations in the region of the N2 nozzle for use in the N-702 evaluation. Figure 1 shows the locations and orientations of these four through-wall stress paths.

For vessel pressure, an internal pressure of 1,000 psig is applied to the inside surfaces of the RPV and N2 nozzle FE model. A bounding transient is also analyzed and the maximum cyclic stress ranges, based on a linearized through wall stress distribution, are identified. Figures 2 through 4 show the distributions of the stress component acting normal to the crack plane ( e.g. hoop or axial depending on the Path location) for the unit pressure, full power thermal expansion (steady state first load step of transient analysis) and the bounding transient load case step, respectively. Details of the analysis can be found in [5].

4.1.3 Fatigu,e Cycles The thermal transients from Reference [5] were obtained from the thermal cycle diagrams in References

[11] and [12]. The 40 year design basis number of cycles from References [11] and [12] are conservatively considered for fatigue instead of the cycle count to date listed in Reference [18]. The number of cycles considered for each bounding transient is explained in Section 5.0.

4.2 Random Variables Random variables (RV) used in the N-702 evaluation are summarized in Table 3. Subsections 4.2.1 through 4.2.3 describe the more detailed input parameters used for SCC Initiation, SCC Growth and

ER2017-027 Page 6 of 19 fatigue crack growth respectively. Table 3 identifies the specific references for each RV used in this N-702 evaluation.

4.2.1 SCC Initiation The cladding stress corrosion crack (SCC) initiation model in the VIPERNOZ program is a power law relationship. Since there is no cladding specific SCC initiation data, the cast stainless steel SCC data in a BWR environment is used as specified in Reference 8, Section 8.2.2.2, and used in References 3 and 4.

This model has the form; T = 84.2

• I 018 (J"-io.s (1) where:

T = time, hours a = applied stress, ksi The residual plot shows that a lognormal distribution produces the best fit for the data. The lognormal residual plot with the linear fit of the data is shown below:

where:

4.2.2 SCC Growth cl) = 0.9248x - 0.003

<D.= (x - a) I µ a = data mean

µ=data standard deviation X = In (T actuai/T predicted)

(2)

The SCC growth model in VIPERNOZ program is also a power law relationship [15]. The relationship used is; where:

da = 6.82

• 10-12 K4 dt da/dt = stress corrosion crack growth rate, in/hr K

= sustained crack tip stress intensity factor, ksiv'in (3)

ER2017-027 Page 7 of 19 The residual plot shows that a Weibull distribution produces the best fit for the data. The Weibull residual plot with the linear fit of the data is shown below:

where:

y Y = 0.9085x - 0.3389

= In (In (1/ (1-F) ))

F

= cumulative distribution from O to 1 X

= In ((da/dt) actual/ (da/dt) predicted) 4.2.3 Fatigue Crack Growth (4)

The fatigue crack growth data for SA-533 Grade B Class 1 and SA-508 Class 2 (carbon moly steels) in a reactor water environments are reported in Reference [16] for weld metal testing at an R-ratio (algebraic ratio ofKmin/Kmax, "R") of 0.2 and 0.7. To produce a fatigue crack growth law and distribution for the VIPERNOZ software, the data for R= 0.7 was fitted into the form of Paris Law. The R= 0.7 fatigue crack growth law was chosen for conservatism. The curve fit results of the mean fatigue crack growth law is presented with the Paris law shown as follows:

where da = 3. 817

• 10-9 ( Af<)2.921 dn a = crack depth, in n = cycles

~K = Kmax - Kmin, ksi-in°.5 (5)

A comparison to the ASME Section XI fatigue crack growth law in a reactor water environment is documented in Reference [14] and it shows a reasonable comparison where the Section XI law is more conservative on growth rate at high ~K.

Using the rank ordered residual plot, it is shown that a Weibull distribution is representative for the data. The Weibull residual plot with the linear curve fit of the data is shown below:

y = -0.3712 + 4.15x where y = ln(ln(l/(1-F))

X = ln((da/dn)actuai/(da/dn)mean)

F = cumulative probability distribution (6)

5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS ER2017-027 Page B of 19 The stress analyses for the nozzle-to-shell weld and the nozzle blend radius for the N2 nozzle are presented in Reference [5]. The stress analyses are performed for the load cases of unit pressure, and the bounding normal and upset (Service Levels A and B) thermal transient. The azimuthal locations evaluated are 0° and 90°, which also represents the symmetric un-modeled 180° and 270° locations of the nozzle. Two through-wall sections are selected at each azimuthal location. One is at the location of the weld between the RPV and nozzle and the other is at the blend radius location of the nozzle.

The load cases analyzed for the N2 nozzle include:

1. Unit pressure (1000 psi)

2. SSPC Transient (Zone 18 to 19) [11]

3.

• FWP Transient (Zone 11 to 12) [12]

For the thermal transients, only the maximum or minimum through-wall linearized membrane plus bending stress profiles that produce the largest stress ranges for thermal fatigue crack growth are used in the evaluation. These through wall stress profiles are shown in Figures 3 and 4.

The nozzle specific SSPC transient is the bounding Service Level A/B transient and occurs five times over the 40 year design basis cycles. The Improper Start of Cold Recirc Loop transient (Zone 17 to 18) also occurs five times over the 40 year design basis cycles, and is bounded by the SSPC transient. These two transients amount to three cycles for each block of 10 years of operation over 60-years of operation Due to the severity of the SSPC transient, the vessel FWP transient (Zone 11 to 12, 12) is selected to bound all other vessel transients. Since the FWP transient has three similar internal cycles of temperature, only the first cycle of the transient is evaluated in Reference [5] since the first cycle is the most severe. The FWP transient occurs 10 times over the 40 year design basis cycles, but 30 are considered due to the internal cycling.

The number of thermal cycles for the FWP transient is considered to be the total number of cycles for Service Level A/B conditions and normal startup/shutdown cycles [11, 12]. Therefore, the transients bounded are:

Plant Startup (120 cycles),

Plant Cooldown (118 Cycles) including, o Reduction of power, o Hot standby, o

Shutdown Vessel Flooding Loss ofFeedwater Heater (80 Cycles) including, o Turbine Trip at 25% Power o Feedwater Heater Bypass SCRAM (220 Cycles) including,

o Isolation Valve Closed o Turbine Trip, Feedwater on, Isolation Open o Reactor Overpressure o Safety Valve Blowdown o All other SCRAMs This totals:

538 cycles for the design 40 years of operation and ER2017-027 Page 9 of 19 807 cycles when scaled for 60 years of operation, or 135 cycles for each block of 10 years of operation over 60-years of operation.

Weld residual stresses (WRS) are assumed present in the nozzle-to-shell welds. The WRS distribution at the nozzle/shell weld is assumed to be a cosine distribution through the wall thickness with 8 ksi mean amplitude and 5 ksi standard deviation. Figure 5 shows the assumed cosine distribution and the 3rd order polynomial fit used in the evaluation for Paths 2 and 4. No WRS is present in the nozzle blend radius region.

6.0 ASSUMPTIONS The following assumptions.used in the evaluation are.based on previous BWRVIP development projects.

Details of each assumption are provided.

1. Flaws are assumed to be aligned parallel with the weld direction as justified in BWRVIP-05 [8].

2. One stress corrosion crack initiation and 0.1 fabrication flaws is assumed per nozzle blend radius as justified in BWRVIP-108NP [3] and BWRVIP-108 SER [2].

3. One stress corrosion crack initiation and 1.0 fabrication flaw is assumed per nozzle/shell weld as justified in BWRVIP-108NP [3].

4. The NRC Pressure Vessel Research Users' Facility (PVRUF) flaw size distribution is assumed to apply as justified in the W-EPRI-180-302 [14] report.

5. The weld residual stress distribution at the nozzle/shell weld is assumed to be a cosine distribution through the wall thickness with 8 ksi mean amplitude and 5 ksi standard deviation as justified in BWRVIP-108NP [3].

6. Upper shelf fracture toughness is set to 200 ksiv'in with a standard deviation of O ksiv'in for un-irradiated material [2].

7. Standard deviation of the mean Kie is set to 15 percent of the mean value of the Kie as justified in BWRVIP-108 SER [2].

8. No copper content values are available in the nozzle weld or forging CMTRs [21]. Therefore, generic industry values are used from Reference [3] and [ 4].

9. Zero inspection coverage conservatively assumed for the initial 40 years of operation.

10. Since no occurrences of the bounding transients have been recorded in the RPV cycle count to date [18], the 40 year design basis number of cycles [11, 12] are conservatively used for fatigue.

7.0 RESULTS OF ANALYSES ER2017-027 Page 10 of 19 The reliability evaluation is presented using plant specific inspection coverage. The probabilities of failure (PoF) per year due to the limiting LTOP event with 25% inspection for the extended operating term (with zero inspection coverage for the initial 40 years of operation) are summarized in Table 4.

The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to LTOP events are both less than the 5x10-6 per year NRC safety goal from Reference [19].

8.0 CONCLUSION

S The probability of failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii in the Cooper N2 nozzle is below the acceptance criterion of Sx 1 o-6 per year. This analysis shows that the N2 nozzles meet the acceptable failure probability even when considering elevated fluence level, thus qualifying all Cooper RPV nozzles with full penetration welds ( except feedwater and control rod drive return nozzles) for reduced inspection using ASME Code Case N-702 to the end of the period of extended operation.

9.0 REFERENCES

ER2017-027 Page 11 of 19

1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.

2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007, SI File No. BWRVIP.108P.

3. BWRVIP-108NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii.

EPRI, Palo Alto, CA: 2007. 1016123.

4. BWRVIP-241: BWR Vessel Internal Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA.

1021005. EPRI PROPRIETARY INFORMATION.

5. SI Calculation 1400334.301, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N2 Nozzle," Revision 1, March 2018.

6. 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, Calculation NEDC 07-032, Revision 3, SI File No. 1400334.201.

7. SI Calculation 1400334.303, "Verification of Software VIPERNOZ Version 1.1," Revision 1, March 2018.

8. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.

9. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Structural Integrity Associates.

10. Combustion Engineering Drawing No. 232-241, Rev. 5, "Nozzle Details," SI File No. 1400334.204.

11. General Electric Drawing No. 135B9990, Sheet 2, Rev. 1, "Nozzle Thermal Cycles (Recirculation Inlet)," SI File No. 1400334.222.

12. General Electric Drawing No. 729E762, Sheet 1, Rev. 1, "Reactor Thermal Cycles," SI File No.

1400334.206

13. Combustion Engineering Drawing No. 232-233, Rev. 10, "Lower Vessel Shell Ass'y, Machining &

Welding," SI File No. 1400334.203.

14. SI Calculation W-EPRI-180-302, "Evaluation of effect of inspection on the probability of failure for BWR Nozzle-to-Shell-Welds and Nozzle Blend Radii Region," Revision 0.

15. NUREG/CR-6923, Appendix B.8, "Expert Panel Report on Proactive Materials Degradation Assessment," Published February 2007.

ER2017-027 Page 12 of 19

16. Bamford, W. H., "Application of corrosion fatigue crack growth rate data to integrity analyses of nuclear reactor vessels," Journal of Engineering Materials and Technology, Vol. 101, 1979, SI File No. 1300341.213.

17. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, 2013 Edition.

18. RCPB/Torus Fatigue Event Descriptions, CUF/EAF Component Record for Year 2014, "RVP Cycle Count Post RE28.PDF," SI File No. 1400334.219.

19. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), NUREG-1806, Vol. 1, August 2007.

20. USNRC Report, "Final Safety Evaluation of the BWR Vessel Internals Project BWRVIP-05 Report," TAC No. M93925, Division of Engineering Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, July 28, 1998.

21. Combustion Engineering Material Verification Report:

a. Document No. RVG-0000009619, Rev. 0, SI File No. 1400334.212.

b. Document No. RVG-0000009621, Rev. 0, SI File No. 1400334.213.

c. Document No. RVG-0000009478, Rev. 0, SI File No. 1400334.210.

d. Document No. RVG-0000009480, Rev. 0, SI File No. 1400334.209.

e. Document No. RVG-0000009485, Rev. 0, SI File No. 1400334.208.

f.

Document No. RVG-0000011697, Rev. 0, SI File No. 1400334.211.

g. Document No. RVG-0000000252, Rev. 0, SI File No. 1400334.215.

h. Document No. RVG-0000000255, Rev. 0, SI File No. 1400334.216.

1.

Document No. RVG-0000000286, Rev. 0, SI File No. 1400334.217.

J.

Document No. RVG-0000000304, Rev. 0, SI File No. 1400334.218.

22. EPRI Letter 2012-138, "BWRVIP Support of ASME Code Case N-702 Inservice Inspection Relief,"

August 31, 2012, SI File No. 1300341.213.

Table 1: Deterministic Parameter Summary VIPERNOZ Variable Value RPV Thickness 6.875 inches (excluding clad)

RPV Radius 110.375 inches (to vessel surface)

Clad Thickness 0.3125 inches Operating Temperature 522 °F(Region B)

LTOP Event Temperature 100 °F Operating Pressure 1005 psig LTOP Event Pressure 1200 psig Table 2: Probability of Detection Distribution [14]

Flaw Size, in.

Cumulative POD 0.00 0.20 0.05 0.32 0.10 0.46 0.15 0.61 0.20 0.75 0.25 0.85 0.30 0.91 0.35 0.95 0.40 0.96 0.45 0.97 0.50 0.98 0.55 0.99 0.60 1.00 ER2017-027 Page 13 of 19 Reference

[10,13]

[10,13]

[10,13]

[11,12]

[20]

[11,12]

[20]

ER2017-027 Page 14 of 19 Table 3: Random Variables Parameter Summary for N2 Nozzle Random Parameter Mean Std Dev Distribution Ref.

Flaw density, nozzle/shell 1 per weld

'1Mean Poisson

[3,3,3]

weld (fabrication)

Flaw density, nozzle/shell 1 per weld

'1Mean Poisson

[3,3,3]

weld (SCC initiation)

Flaw density, nozzle blend 0.1 per weld

'1Mean Poisson

[2,2,3]

radius (fabrication)

Flaw size (fabrication) n/a n/a PVRUF

[3]

Flaw size (stress corrosion)

Clad thickness n/a Constant

[3,3]

Weld residual stress, 8

inside surface 5

Normal

[3,3,3]

through-wall (ksi) cosine distribution Clad residual stress (ksi)*

32 5

Normal

[3,3,3]

N2 Nozzle

%Cu 0.26 0.045 Normal

[3,3,3]

%Ni 1.46 0.0165 Normal r21,3,3J to shell Initial RT ndt weld (OF)

+10 13 Normal

[21,3,3]

%Cu 0.09189 0.04407 Normal r2,2,31 N2 Nozzle

%Ni 0.72 0.068 Normal

[21,2,3]

forging Initial RTndt

+10 26.48 Normal

[21,2,3]

(OF)

Lower Shell fast neutron 3.34e16 0.2 (20%)

n/a

[6,3]

fluence (n/cm2)

Kie upper shelf (ksi\\iin) 200 0

Normal

[2,22,3]

Residual sec initiation time (hr)

"C = 84.2x 1 ots( o}l0.5 y=0.9248x-Lognormal

[2,3,3]

0.0003 K dependent Residual da/dt = 6.82xI0-12(K)4 y=0.9085x-Weibull

[15,3,3]

SCCG (in/hr)

K >50 ksi'1in 0.3389 K independent da/dt = 2.8xl0-6, na na

[15]

K <50 ksi\\iin sec threshold (ksi'1in) 10 2

Normal

[2,3,3]

Fatigue crack growth (FCG) da/dn=3.82 Residual (in/cycle)

X 10-9( dK)2.927 y=4.155x-Weibull

[3,3,3]

0.3712 FCG threshold (ksi'1in) 0 0

Normal

[3,3,3]

• Note: The mean clad stress used already includes the effects of post-weld heat treatment.

Location Path 1 Path 2 Path 3 Path 4 ER2017-027 Page 15 of 19 Table 4: PoF for Period of Extended Operation Zero inspection for initial 40 years, 25% for PEO PoF per year due PoF per year due to Allowable PoF per to LTOP Event*

Normal Operation year [191 1.675 X 10-IO

< 8.33 X 10-IO

< 8.33 X 10-13

< 8.33 X 10-IO 5.0 x 10-6

< 8.33 X 10-13

< 8.33 X 1 o-lO

< 8.33 X 10-13

< 8.33 X 10-lO

• Note: Values include Ix 10-3 probability ofLTOP event occurrence per year [3, pg 5-13].

Figure 1: Stress Extraction Path Orientations in the N2 Nozzle

55 50

-- Path 1 45 Path 2 en

~ 40

...... Path 3 en

- o-Path 4 en 35 Cl).... - 30

0)

C 25 20 C

~

CJ 15 ca....

0 10 5

0000000 0

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Path length (in)

Figure 2: Pressure Stress Distributions for the N-702 Evaluation 15 10 5

~

0 G)

-5 u,

~ -10

"'i: -15 C

~ -20 0 -25

-30

-35 +----,.----,----.------.---...------.----------'

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Path length (in)

ER2017-027 Page 16 of 19 Figure 3: Full Power Thermal Expansion Stress Distributions for the N-702 Evaluation

70 60

~ 50

= 40 cu...

en 30 O>

C 20 0 10

~

~

0 O -10

-20

-- Path 1 Path 2

-+- Path 3 Path 4

-30 +--~--~-~--~--~----~-------<

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Path length (in)

ER2017-027 Page 17 of 19 Figure 4: Bounding Transient Stress Distributions for the N-702 Evaluation 12 10

-e-Cosine Stress Distr 8

Poly. (Cosine Stress Distr)

~ 6 y = 1E-14xl + 1.0626x2-8.6432x + 11.631 4

G)... -

UJ 2

ftl

~ 0

~

-2 G) 0:::

-4

-6

-8

-10 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Weld Thickness (in)

Figure 5: Weld Residual Stress Distributions for Paths 2 and 4

Appendix A LIST OF SUPPORTING FILES ER2017-027 Page 18 of 19

File Name Pathl.INP Path3.INP Path2.INP Path4.INP Pathl.OUT Path3.0UT Path2.0UT Path4.0UT VIPERNOZlpl.EXE ISPCTPOD.EXE FLWDSTRB.EXE Description ER2017-027 Page 19 of 19 VIPERNOZ input file for Path 1 at nozzle blend radii.

VIPERNOZ input file for Path 3 at nozzle blend radii.

VIPERNOZ input file for Path 2 at nozzle-to-shell-weld.

VIPERNOZ input file for Path 4 at nozzle-to-shell-weld.

VIPERNOZ output file for Path 1 at nozzle blend radii.

VIPERNOZ output file for Path 3 at nozzle blend radii.

VIPERNOZ output file for Path 2 at nozzle-to-shell-weld.

VIPERNOZ output file for Path 4 at nozzle-to-shell-weld.

VIPERNOZ executable program VIPERNOZ probability of detection curve input file VIPERNOZ flaw size distribution curve input file

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

~Entergy MANUAL INFORMATIONAL USE Page 1 of 1 ER2017a027, Attachment 10.3 ATTACHMENT 9.3 TECHNICAL REVIEW COMMENTS AND RESOLUTION FORM SHEET1 OF 1 NNPPD Engineering Report Cooper Nuclear Station Technical Review Comments and Resolutions Form Engineering ER 2017-027, R1 Rev. I Title Review of Structural Integrity Calculations 1400334.301 & 1400334.302 for Reoort Number Code Case N-702 Relief Request Quallty R~lated: ~Yes LJ No Special Notes or Instructions: N/A Comment Section/ Page No.

Review Comment Response/Resolution Reviewer's Number Accept Initials NONE N/A Review limited to Revision 1 of this Engineering Report.

N/A Verified/Reviewed By:

I Phil Lelnin~

J

--.... I Date 13-*7-(8 Resolved By:

IN/A r -

\\.

Slte/Dtmartment:

I EP&C

"-t-PM.x5310

--:i Date:

I ER 2017 a027 Rev1