Text

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)
	v • d • e

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

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 Attachment 1 Page 1 of2 Attachment 1 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 ofoperation and examination of25% [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 ofwhich 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 ofStructural Integrity Calculations 1400334.301 & 1400334.302 for Code Case N-702 ReliefRequest, 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 staffdoes 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 ofInspection 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 1o-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 Attachment 2 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 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702

Reason for Request

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 BWRVIP-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) r 0 = 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(10.219 2 + 6.1882)/(10.219 2 - 6.188 2))/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 2

(5) [p(r0 + r/) I (r/ - r/)]/CNoZZLE :s__ 1.59 P = RPV normal operating pressure (psi) (1020) r 0 = 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(21.656 2 + 12.875 2)/(21.6562 - 12.875 2)]/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 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Precedents "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 OApplicable 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 Attachment 2 Page 8 of 12 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 Nozzle ID Nominal Pipe Last Appendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches) Examination Inner Radius (IR)

Recirculation Yes NIA (NV) B-D B3.90 28 10/2016 NRI (Outlet)

Recirculation VT-1 NIA (IR) B-D B3.100 28 10/2016 NRI (Outlet)

Recirculation Yes NIB (NV) B-D B3.90 28 01/2005 NRI (Outlet)

Recirculation MVT-1 NIB (IR) B-D B3.100 28 01/2005 NRI (Outlet)

Recirculation Yes N2A(NV) B-D B3.90 12 01/2005 NRI (Inlet)

Recirculation Yes N2A (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2B (NV) B-D B3.90 12 01/2005 NRI (Inlet)

Recirculation Yes N2B (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2C (NV) B-D B3.90 12 04/2011 NRI (Inlet)

Recirculation Yes N2C (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2D (NV) B-D B3.90 12 01/2005 NRI (Inlet)

NLS2018012 Attachment 2 Page 9 of 12 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nozzle ID Nominal Pipe Last Appendix Nozzle-to-Vessel (NV) Category Number Item Number Systein Results VIII Exam Size (Inches} Examination Inner Radius (IR)

Recirculation Yes N2D (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2E (NV) B-D B3.90 12 10/2016 NRI (Inlet)

Recirculation Yes N2E (IR) B-D B3.100 12 10/2016 NRI (Inlet)

Recirculation Yes N2F (NV) B-D B3.90 12 04/2011 NRI (Inlet)

Recirculation Yes N2F (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2G(NV) B-D B3.90 12 04/2011 NRI (Inlet)

Recirculation Yes N2G (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2H (NV) B-D B3.90 12 10/2016 NRI (Inlet)

Recirculation Yes N2H (IR) B-D B3.100 12 10/2016 NRI (Inlet)

Recirculation Yes N2J (NV) B-D B3.90 12 04/2011 NRI (Inlet)

Recirculation Yes N2J (IR) B-D B3.100 12 04/2011 NRI (Inlet)

Recirculation Yes N2K(NV) B-D B3.90 12 10/2016 NRI (Inlet)

NLS2018012 Attachment 2 Page 10 of 12 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nozzle ID Nominal Pipe Last Appendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches) Examination Inner Radius (IR)

Recirculation Yes N2K (IR) 8-D 83 .100 12 10/2016 NRI (Inlet) 8-D 83.90 Main Steam 24 10/2016 NRI Yes N3A (NV) 8-D 83.100 Main Steam 24 09/2016 VT-1 N3A (IR) NRI Main Steam 24 03/2000 NoC3)

N38 (NV) B-D B3 .90 NRI Main Steam 03/2000 NoC3)

N38 (IR) 8-D B3.100 24 NRI 8-D 83 .90 Main Steam 24 01 /2005 Yes N3C (NV) NRI 8-D 83.100 Main Steam 24 02/2005 MVT-1 N3C (IR) NRI Main Steam NoC3)

N3D (NV) 8-D 83.90 24 03/2000 NRI Main Steam NoC3)

N3D (IR) 8-D 83.100 24 03/2000 NRI N4A(NVi 1) B-D B3.90 Feedwater 12 10/2014 NRI Yes N4A (IRi 1) B-D B3.100 Feedwater 12 NRI Yes 10/2014 N4B (NV)°) B-D B3.90 Feedwater 12 10/2014 NRI Yes B-D B3.100 Feedwater 12 NRI Yes N4B (IRi'> 10/2014 N4C (NV)°> B-D B3.90 Feedwater 12 NRI Yes 10/2014 N4C (IR)°> B-D B3.100 Feedwater 12 NRI Yes 10/2014

NLS2018012 Attachment 2 Page 11 of 12 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nozzle ID Nominal PiQe Last AQQendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches) Examination Inner Radius (IR)

N4D (NV)'1> B-D B3.90 Feedwater 12 10/2014 NRI Yes N4D (IRi 1> B-D B3.100 Feedwater 12 10/2014 NRI Yes B-D B3.90 Core Spray 10 10/2016 NRI Yes N5A (NV)

B-D B3.100 Core Spray 10 10/2016 Yes N5A (IR) NRI B-D B3.90 Core Spray 10 01 /2005 Yes N5B (NV) NRI B-D B3.100 Core Spray 10 NRI Yes N5B (IR) 03/2011 Head Yes N6A (NV) B-D B3 .90 6 02/2005 NRI Instrumentation B3.100 Head MVT-1 N6A (IR) B-D 6 02/2005 NRI Instrumentation B3.90 Head Yes N6B (NV) B-D 6 03/2011 NRI Instrumentation B3.100 Head MVT-1 N6B (IR) B-D 6 03/2011 NRI Instrumentation B-D B3.90 Head Vent 4 03/2011 NRI Yes N7 (NV)

B-D B3.100 Head Vent 4 03/2011 NRI MVT-1 N7 (IR)

Jet Pump No<3)

N8A (NV) B-D B3.90 4 03/2000 NRI Instrumentation

NLS2018012 Attachment 2 Page 12 of 12 10 CFR 50.55a Request No. RIS-03 (continued)

Implementation of Code Case N-702 Nozzle ID Nominal Pige Last Aggendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches) Examination Inner Radius (IR)

Jet Pump NoC 3l N8A (IR) B-D B3.100 4 03/2000 NRI Instrumentation Jet Pump Yes N8B (NV) B-D B3.90 4 10/2014 NRI Instrumentation Jet Pump Yes N8B (IR) B-D B3.100 4 10/2014 NRI Instrumentation N9 (NVi2> B-D B3.90 CRD Return Line 5 03/2011 NRI Yes NRI Yes N9 (IRi2> 8-D 83.100 CRD Return Line 5 03/2011 Notes:

1) Code Case N-702 excludes these nozzles

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

.!'a NUCLEAR QUALITY RELATED 3-EN-DC-147 I 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 J8 Responsible Engineer (Print Name/Sign)

NIA Date:

Reviewed by:

ER 2017-027 Rev1

ENGINEERING REPORT ER2017-027 Rev 1 rtl HNPPD 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 ER2017-027 Rev 1 NNPPD Page 3 of 10 COOPER NUCLEAR STATION 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 LTOP 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 LTOP 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 ER2017-027 Rev 1 NNPPD Page 4 of 10 COOPER NUCLEAR STATION 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 ER2017-027 Rev 1 NNPPD Page 5 of 10 COOPER NUCLEAR STATION

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 3.34x10 16 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 5x10-5 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 LTOP event (including 1 x 10-3 probability of LTOP 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

-1 ENGINEERING REPORT ER2017-027 Rev 1 HNPPD Page 6 of 10 COOPER NUCLEAR STATION 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 LTOP 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 LTOP 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 3

Note: Values include 1x10* probability of LTOP 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 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 from 1400~34.302 [Attachment 10.2].

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 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

ENGINEERING REPORT H

ER2017-027 Rev 1 NPPD COOPER NUCLEAR STATION Page 7 of 10

9.0 REFERENCES

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

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

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

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: