Text

Inservice Inspection Interval Relief Request 15R-07
	ML16354A749
Person / Time
Site:	Quad Cities
Issue date:	12/19/2016
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	15R-07., RS-16-256
	Download: ML16354A749 (55)
	v • d • e

~- 4300 Winfield Road Warrenville, IL 60555 ExeLon G 630 657 2000 Office RS-16-256 10 CFR 50.55a(z)(1)

December 19, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Quad Cities Nuclear Power Station Inservice Inspection Interval Relief Request 15R-07 In accordance with 10 CFR 50.55x, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC (EGC) requests NRC approval of the attached relief request for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. QCNPS, Units 1 and 2 are currently within their fifth inservice inspection ( ISI) interval , which complies with the 2007 Edition with 2008 Addenda of the American Society of Mechanical Engineers (ASME) Code,Section XI.

The fifth interval began on April 2, 2013, and is currently scheduled to end on April 1, 2023; however, EGC is requesting approval of this relief request for the remaining term of the QCNPS, Units 1 and 2, renewed operating licenses, which currently expire on December 14, 2032.

The subject relief request is regarding the implementation of ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor ( BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," and Boiling Water Reactor Vessel Inspection Program (BWRVIP)-241, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," as documented in the attachments to this letter. The NRC provided a safety evaluation approving the generic technical bases and acceptability criteria for application of Code Case N-702 and BWRVIP-241, which EGC has followed as detailed in the attachments. EGC requests approval of the proposed alternative on or before December 21, 2017, to accommodate its application during an upcoming refueling outage. EGC plans to implement this alternative for the remaining term of the QCNPS, Units 1 and 2 renewed operating licenses.

December 19, 2016 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.

Sincerely, Patrick R. Simpson Manager Licensing Exelon Generation Company, LLC Attachments:

Quad Cities Nuclear Power Station 10 CFR 50.55a Request No. 15R-07

2. Structural Integrity Associates, Inc. File No. 1400735.301, Revision 1, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N1 Nozzle," dated June 23, 2016

3. Structural Integrity Associates, Inc. File No. 1400735.302, Revision 1, "Code Case N-702 Evaluation for Dresden and Quad Cities Recirculation Outlet (N1) Nozzle,"

dated June 23, 2016

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected:

Code Class:

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.90 and B3.100

Description:

Alternative to Nozzle to Vessel Weld and Inner Radius Examinations Component Numbers: N1, N2, N3, N5, N6, N7, and N8 Nozzles (See Tables 5-3 and 5-4 below for complete list of nozzle component identification numbers) 9 Applicable Code Edition and Addenda:

The current interval of the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2007 Edition with the 2008 Addenda is implemented, as required and modified by 10CFR50.55a(b)(2)(xv). The fifth 10-year interval is effective from April 2, 2013, through April 1, 2023.

3. Applicable Code Requirement:

Table IWB-2500-1 "Examination Category B-D, Full Penetration Welded Nozzles in Vessels."

Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number B3.90 "Nozzle-to-Vessel Welds," and B3.100 "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles are examined each interval.

All of the nozzle assemblies identified in Tables 5-3 and 5-4 are full penetration welds.

Page 1 of 9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

4. Reason for Request:

Pursuant to 10CFR50.55a(z)(1), relief is requested from performing the required examinations on 100% of the RPV nozzle assemblies at QCNPS, Units 1 and 2 that are identified in Tables 5-1 and 5-2, and further delineated in Tables 5-3 and 5-4 below.

Tables 5-3 and 5-4 provide a complete listing of the applicable RPV nozzles.

The proposed alternative provides an acceptable level of quality and safety, and the reduction in scope could provide a dose savings of as much as 16 person-roentgen equivalent man (rem) for Unit 1 and 19 person-rem for Unit 2, over the remaining term of the renewed operating licenses.

5. Proposed Alternative and Basis for Use:

As an alternative for the welds and inner radii identified in Tables 5-1 and 5-2, Exelon Generation Company, LLC (EGC) proposes to examine a minimum of 25% of the QCNPS, Units 1 and 2 nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, during the current fifth and upcoming sixth 120-month ISI Program intervals interval in accordance with ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1." Specifically, 25% of the required nozzle types identified in this relief request will be examined in accordance with the conditions for the implementation of Code Case N-702, as defined in NRC Regulatory Guide 1.147, Rev. 17, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," during each of the remaining QCNPS, Units 1 and 2 120-month ISI Program intervals for the remaining QCNPS, Units 1 and 2 period of extended operation. EGC does not plan to deviate from the previously evaluated Code Case N-702 examination frequencies. For the applicable nozzle assemblies identified in Tables 5-1 and 5-2, this would mean at least one from each of the groups identified below:

Table 5-1: QCNPS, Unit 1 Summary Minimum Number Group Total Number to be Examined Recirculation Outlet N1 2 1 Recirculation Inlet N2 10 3 Main Steam (N3) 4 1 Core Spray N5 2 1 Nozzles on Vessel Top Head (N6, N7) 3 1 Jet Pump Instrumentation N8 2 1 Page 2 of 9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

Table 5-2: QCNPS, Unit 2 Summary Minimum Number Group Total Number to be Examined Recirculation Outlet N1 2 1 Recirculation Inlet N2 10 3 Main Steam N3 4 1 Core Spray N5 2 1 Nozzles on Vessel Top Head N6, N7 3 1 Jet Pump Instrumentation N8 2 1 Table 5-3: QCNPS, Unit 1 Applicable Nozzles Component Exam Item Nominal System ID Category Number Pipe Size N1 Nozzle B-D B3.90 Recirc Outlet 28" N1A IRS B-D B3.100 Recirc Outlet 28" N1 B Nozzle B-D B3.90 Recirc Outlet 28" N1 B IRS B-D B3.100 Recirc Outlet 28" N2A Nozzle B-D B3.90 Recirc Inlet 12" N2A IRS B-D B3.100 Recirc Inlet 12" N2B Nozzle B-D B3.90 Recirc Inlet 12" N2B IRS B-D B3.100 Recirc Inlet 12" N2C Nozzle B-D B3.90 Recirc Inlet 12" N2C IRS B-D B3.100 Recirc Inlet 12" N2D Nozzle B-D B3.90 Recirc Inlet 12" N2D IRS B-D B3.100 Recirc Inlet 12" N2E Nozzle B-D B3.90 Recirc Inlet 12" N2E IRS B-D B3.100 Recirc Inlet 12" N2F Nozzle B-D B3.90 Recirc Inlet 12" N2F IRS B-D B3.100 Recirc Inlet 12" N2G Nozzle B-D B3.90 Recirc Inlet 12" N2G IRS B-D B3.100 Recirc Inlet 12" N2H Nozzle B-D B3.90 Recirc Inlet 12" N2H IRS B-D B3.100 Recirc Inlet 12" N2J Nozzle B-D B3.90 Recirc Inlet 12" N2J IRS B-D B3.100 Recirc Inlet 12" N2K Nozzle B-D B3.90 Recirc Inlet 12" N2K IRS B-D B3.100 Recirc Inlet 12" N3A Nozzle B-D B3.90 Main Steam 20" N3A IRS B-D B3.100 Main Steam 20" N313 Nozzle B-D B3.90 Main Steam 20" N313 IRS B-D B3.100 Main Steam 20" N3C Nozzle B-D B3.90 Main Steam 20" N3C IRS B-D B3.100 Main Steam 20" N3D Nozzle B-D B3.90 I Main Steam 20" N3D IRS B-D B3.100 1 Main Steam 20" Page 3 of 9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative, In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

Table 5-3: QCNPS, Unit 1 Applicable Nozzles Component Exam Item Nominal System ID Category Number Pipe Size N5A Nozzle B-D B3.90 Core Spray 10" N5A IRS B-D B3.100 Core Spray 10" N5B Nozzle B-D B3.90 Core Spray _ 10" N5B IRS B-D B3.100 Core Spray 10" N6A Nozzle B-D B3.90 Head Spray 6" N6A IRS B-D B3.100 Head Spray 6" N6B Nozzle B-D B3.90 Head Spray Spare 6" N6B IRS B-D B3.100 Head Spray Spare 6" N7 Nozzle B-D B3.90 Head Vent 4" N7 IRS B-D B3.100 Head Vent 4" N8A Nozzle B-D B3.90 Jet Pump Instrumentation 4" N8A IRS B-D B3.100 Jet Pump Instrumentation 4" N8B Nozzle B-D B3.90 Jet Pump Instrumentation 4" N8B IRS B-D B3.100 Jet Pump Instrumentation 4" Table 5-4: QCNPS, Unit 2 Applicable Nozzles Component Exam Item Nominal System ID Cate or Number Pipe Size N1 Nozzle B-D B3.90 Recirc Outlet 28" N1A IRS B-D B3.100 Recirc Outlet 28" N1 B Nozzle B-D B3.90 Recirc Outlet 28" N1 B IRS B-D B3.100 Recirc Outlet 28" N2A Nozzle B-D B3.90 Recirc Inlet 12" N2A IRS B-D B3.100 Recirc Inlet 12" N213 Nozzle B-D B3.90 Recirc Inlet 12" N2B IRS B-D B3.100 Recirc Inlet 12" N2C Nozzle B-D B3.90 Recirc Inlet 12" N2C IRS B-D B3.100 Recirc Inlet 12" N2D Nozzle B-D B3.90 Recirc Inlet 12" N2D IRS B-D B3.100 Recirc Inlet 12" N2E Nozzle B-D B3.90 Recirc Inlet 12" N2E IRS B-D B3.100 Recirc Inlet 12" N2F Nozzle B-D B3.90 Recirc Inlet 12" N21F IRS B-D B3.100 Recirc Inlet 12" N2G Nozzle B-D B3.90 Recirc Inlet 12" N2G IRS B-D B3.100 Recirc Inlet 12" N2H Nozzle B-D B3.90 Recirc Inlet 12" N2H IRS B-D B3.100 Recirc Inlet 12" N2J Nozzle B-D B3.90 Recirc Inlet 12" N2J IRS B-D B3.100 Recirc Inlet 12" N2K Nozzle B-D B3.90 Recirc Inlet 12" N2K IRS B-D B3.100 Recirc Inlet 12" Page 4 of 9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

Table 5-4: QCNPS, Unit 2 Applicable Nozzles Component Exam Item Nominal System ID Cate o Number Pi a Size N3A Nozzle B-D B3.90 Main Steam 20" N3A IRS B-D B3.100 Main Steam 20" N3B Nozzle B-D B3.90 Main Steam 2_0" N3B IRS B-D B3.100 Main Steam 20" N3C Nozzle_ B-D B3.90 Main Steam 20" N3C IRS B-D B3.100 Main Steam 20" N3D Nozzle B-D B3.90 Main Steam 20" N3D IRS B-D B3.100 Main Steam 20" N5A Nozzle B-D B3.90 Core Spray-------- _ 10" N5A IRS B-D B3.100 Core Spray 10" N5B Nozzle B-D B3.90 Core Spray 10" N5B IRS B-D B3.100 Core Spray 10" N6A Nozzle B-D B3.90 Head Spray 6" N6A IRS B-D B3.100 Head Spray 6" N6B Nozzle B-D B3.90 Head Spray Spare 6" N6B IRS B-D B3.100 Head Spray Spare 6" N7 Nozzle B-D B3.90 Head Vent 4" N7 IRS B-D B3.100 Head Vent 4" N8A Nozzle B-D B3.90 Jet Pump Instrumentation 4" N8A IRS B-D B3.100 Jet Pump Instrumentation 4" N8B Nozzle B-D B3.90 Jet Pump Instrumentation 4" N8B IRS B-D B3.100 Jet Pump Instrumentation 4" Electric Power Research Institute (EPRI) Technical Report (TR)-1003557, "BWRVIP-108: 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,"' provides the basis for the use of ASME Code Case N-702.

The evaluation found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a low temperature overpressure event are very low (i.e., less than (<)1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25% of each nozzle type is technically justified.

This EPRI report was approved by the NRC in a safety evaluation (SE) dated December 19, 2007 (i.e., ADAMS Accession No. ML073600374). Section 5.0, "Plant-Specific Applicability," of the SE indicates that each licensee who plans 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-108 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability criteria from the BWRVIP-108 report to its units in the relief request by showing that all the general and nozzle-specific criteria addressed below are satisfied as described in Section 8 below.

Page 5 of 9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

(1) The maximum RPV heatup/cooldown rate is limited to <115 °F per hour.

QCNPS Technical Specification (TS) 3.4.9, "RCS Pressure and Temperature (P/T) Limits," provides a limiting condition for operation (LCO) and a corresponding surveillance requirement (SR) that ensure the reactor coolant system heatup and cooldown rates are less than or equal to (<_)100°F/hr. The SR (i.e., monitoring of reactor vessel heatup and cooldown rates) is referenced in the QCNPS Updated Final Safety Analysis Report (UFSAR) Section 5.3.2,"Pressure-Temperature Limits," and UFSAR Table 5. 1 -1,"Reactor Coolant System Data."

(2) For the Recirculation Inlet Nozzles, the following criteria must be met:

a. (pr/t)/CRPv<1.15; The calculation for the QCNPS, Units 1 and 2 N2 Nozzle results in 1.065 which is less than 1.15.

b. [p(ro2 +r;2)/(ro2-r;2 )]/CNOZZLE <1.15; The calculation for the QCNPS, Units 1 and 2 N2 Nozzle results in 0.972, which is less than 1.15.

(3) For the Recirculation Outlet Nozzles, the following criteria must be met:

a. (pr/t)/CRPv< 1.15; The calculation for the QCNPS, Units 1 and 2 N1 Nozzle results in 1.273 which is higher than 1.15.

b. [p(ro2+r;2)/(ro2-ri2 )]/CNOZZLE <1.15; The calculation for the QCNPS, Units 1 and 2 N1 Nozzle results in 0.840 which is less than 1.15.

Based upon the above information, all applicable QCNPS, Units 1 and 2 RPV nozzle-to-vessel shell welds and nozzle inner radii sections, with the exception of the recirculation outlet nozzles, meet the general and nozzle-specific criteria in BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and therefore ASME Code Case N-702 is applicable. The recirculation outlet nozzles do not meet all of the criteria in BWRVIP-241. BWRVIP-241, Section 6.0 notes that for plants having recirculation outlet nozzles with Condition 4 greater than 1.15 such as for QCNPS, Units 1 and 2, a plant specific analysis following the approach described in this report may be able to justify values greater than 1.15. Additional discussion is provided in Section 8 below.

Attachment 3 in this relief request considers the nozzle-to-shell weld and nozzle blend radius on the N1 nozzle in accordance with Attachment 3, References 3 and 4, and confirms that the nozzle still meets the acceptable failure probability considering the bounding fluence at the end of the QCNPS period of extended operation. Attachment 3, Reference 6 shows the highest fluence at 5.59 x 10" neutrons per square centimeter.

Because the QCNPS N1 nozzles did not meet the BWRVIP-241 criteria, a bounding analysis was performed to qualify all the nozzles. This bounding analysis is contained in Structural Integrity Associates, Inc. (SIA) File Nos. 1400735.301, Revision 1 and 1400735.302, Revision 1. As required by BWRVIP-241, these analyses have been Page 6of9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

included as Attachments 2 and 3, respectively. The methods approved in BWRVIP-108 and BWRVIP-241 were followed. The site specific analysis concluded that the failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii for the QCNPS, Units 1 and 2 N1 nozzles is below the acceptance criterion of 5 x 10-6 per year. This analysis shows that the N1 nozzles meet the acceptable failure probability even when considering elevated fluence level, thus qualifying all QCNPS, Units 1 and 2 RPV nozzles with full penetration welds, with the exception of the feedwater and control rod drive return nozzles, for reduced inspection using ASME Code Case N-702 to the end of the QCNPS, Units 1 and 2 periods of extended operation.

Therefore, use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) for all applicable full penetration RPV nozzle-to-vessel shell welds and nozzle inner radii sections for the remaining term of the renewed operating licenses for QCNPS, Units 1 and 2.

6. Duration of Proposed Alternative:

Relief is requested for the remaining term of the QCNPS, Units 1 and 2, renewed operating licenses, which currently expire on December 14, 2032.

7. Precedents:

Columbia Generating Station - Relief Request for Alternative 41S1-04 Applicable to the Fourth 10-year Inservice Inspection Program Interval (CAC NO. MF7331) was authorized by NRC SE dated October 5, 2016 (i.e., NRC Accession No. ML16263A233).

LaSalle County Station, Units 1 and 2, Relief from the Requirements of the ASME Code Re: RR 13R-14, Proposed Alternative To The Examination Requirements For Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1) (TAC NOS. MF5654 AND MF5655) was authorized by NRC SE dated October 31, 2015 (i.e., NRC Accession No. ML15226A412).

Pilgrim Nuclear Power Plant Relief Request PRR-24 Regarding Nozzle-to-Vessel Welds and Nozzle Inner Radii Examination (TAC NO MF4187) was authorized by NRC SE dated April 21, 2015 (i.e., NRC Accession No. ML15103A069).

DNPS, Units 2 and 3 Fourth Inspection Interval Relief Request 14R-16 was authorized by and NRC SE dated November 3, 2009 (i.e., NRC Accession No. ML092940436).

Quad Cities Nuclear Power Station, Units 1 and 2 Fourth Inspection Interval Relief Request 14R-17 was authorized by an NRC SE dated February 2, 2010 (i.e., NRC Accession No. ML092860259).

Page 7 of 9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

8. Plant Specific Applicability EGC performed the following analysis to demonstrate the plant specific applicability of criteria from the BWRVIP-108 report to QCNPS, Units 1 and 2 by showing that all the general and nozzle-specific criteria addressed below are satisfied:

(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to

< 115 °F/hour.

Response: QCNPS Technical Specification 3.4.9, "RCS Pressure and Temperature (P/T) Limits," provides a limiting condition for operation (LCO) and a corresponding surveillance requirement (SR) that ensure the reactor coolant system heatup and cooldown rates are s 100°F/hr. The SR (i.e., monitoring of reactor vessel heatup and cooldown rates) is referenced in the QCNPS Updated Final Safety Analysis Report (UFSAR) Section 5.3.2, "Pressure-Temperature Limits," and UFSAR Table 5.1-1, "Reactor Coolant System Data."

(2) For Recirculation Inlet Nozzles

a. (pr/t)/CRPv <l .15 p=RPV Normal Operating Pressure 1005 r=RPV inner radius 125.5 t=RPV Wall thickness 6.125 CRPv= 19332 (pr/t)/CRPV =1.065 <1.15

b. [p(ro2 +rig)/(ro2-ri2)I/CNOZZLE< 1.15 p=RPV Normal Operating Pressure 1005 ro=nozzle outer radius 12.5 ri=nozzle inner radius 5.941 CNOZZLE __ __ 1637

[p(rorig)/(r02-ri2WCNOZZLE =0.972 <1.15 Page 8of9

ATTACHMENT 1 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.55a Request No. 15R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

(3) For Recirculation Outlet Nozzles

a. (pr/t)/CRPV <1.15 p=RPV Normal Operating Pressure 1005 r=RPV inner radius 125.5 t=RPV Wall thickness 6.125 CRPV= 16171 (pr/t)/CRPV = 1.273 >1.15

b. [p(ro2 +ri2)/(ra2-ri2)l/CNOZZLE< 1.15 p=RPV Normal Operating Pressure 1005 ro=nozzle outer radius 26.5 ri=nozzle inner radius 13.1375 CNOZZLE 1977

[p(ro rig)/(ro2 -riz)]/CNOZZLE = 0.840 <1.15 Page 9 of 9

ATTACHMENT 2 Structural Integrity Associates, Inc. File No. 1400735.301, Revision 1, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N1 Nozzle,"

dated June 23, 2016

File No.: 140073.5.301 Structural Integrity Associates, Inc ° Project No.: 1400735 and 1501375 CALCULATION PACKAGE Quality Program: Nuclear Commercial PROJECT NAME:

N-702 I ,valuation Im Dresden and Quad Cities CONTRACT NO.:

00517760, 1Zev 62 and 63 CLIENT: PLANT:

Fxelon Corl)oration Dresden and Quad Cities Generating Stations CALCULATION TITLE:

F inite Element Model Develol)ment and Ther mal/Mechanical Stress Analyses for the N 1 Nozzle Project Manager Preparer(s) &

Document Affected Revision Description Approval Checker(s)

Revision Pages Si nature & Date Si natm-es & Date 0 1 -21 Initial Issue Responsible Engineer:

A A-2 Richard Bax 1/22/16 Wilson Wong 1/22/16 Responsible Verifier:

Minji Fong 1/22/16 1 1, 6, 11 Removed Proprietary u1 o ~ Responsible Engineer:

Markings v' 6~~

~U~~ 6'~

Wilson Wong 6/23/16 Wilson Wong 6/23/16 Responsible Verifier:

r r

4 1 7 -

Minj i Fong 6/23/16 Page 1 of 21 F0306-01 R2

C structural Integrity Associates, Inc.

Table of Contents 1.0 INTIZODUCTION .........................................................................................................4 2.0 0B.1ECTIVF'S ................................................................................................................4 3.0 ASSUMI)TIONS ............................................................................................................4 4.0 DESIGN INPUTS ..........................................................................................................5 4.1 Nozzle Geomctiy ...............................................................................................5 4.2 Material Properties .............................................................................................5 4.3 "I,ransient Defnitions .........................................................................................5 5.0 METI IODOL,OGY ........................................................................................................6 5.1 Thei-mal Ti-ansicnt Selection ..............................................................................6 5.2 Nozzle hIM and Load Case LvalLlatlon ............................................................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

............................................................................................................11 APPENDIX A FILENAMES ............................................................................................... A-1 File No.: 1400735.301 Page 2 of 21 Revision: 1 F0306-01 R2

C structural Integrity Associates, Inc.

List of Tables

able 1: Bounding 'Transients for Analysis 19, 121 .................................................................12

'fable 2: Material Properties for Carhon Moly Steel [4, 51 ......................................................12

'1'ablc 3: Material Properties for AllstCnitiC Stainless Steel 14, 51 ...........................................13 List of Figures Fi Ure 1: Co111po11e11tS 111CIUded in the Finite Element Model ...............................................14 Figure 2: 3-1) Finite Element Model Mesh for Analyses, Baseline Mesh ..............................14 Figure 3: Path Locations for Through-Wall Stress Extractions ..............................................15 Figure 4: Applied Boundary Conditions and Unit Internal Pressure ......................................15 Figure 5: "total Stress Intensity Plot for Unit Internal Pressure ..............................................16 Figure 6: Applied Thermal Boundary Conditions for Thermal Transient Analyses ..............17 Figure 7: Applied Mechanical Boundary Conditions for Thermal Stress Analyses ..............18 Figure 8: Temperature Contour for SCRAM Transient at Time=6000 sec ............................ 18 Figure 9: Stress Intensity Plot 101' SCRAM Transient at 'rime=6000 sec ..............................19 Figure 10: Total Stress Intensity Contours for Mesh Sensitivity Study Unit Pressure .......20 Figure 11: Linearized Membrane-Plus-Bending Stress Intensity History for Path 1, SCRAM and Interruption of Feed Flow ................................................................21 File No.: 1400735.301 Page 3 of 21 Revision: 1 F0306-01 R2

CStructural Integrity Associates, Inc."

1.0 INTRODUCTION

Exelon intends to extend the applicability of Code Case N-702 [1] at both Dresden Units 2&3 and Quad Cities Units 1 &2 through the end of their respective periods of extended operation (PEO). 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 (PFM) 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 N 1 (Recirculation Outlet) and N2 (Recirculation Inlet) nozzles are identified as the bounding RPV nozzles when fluence is not considered [3].

Since the N1 nozzle does not meet the NRC's additional requirements, outlined in Reference [2] and has a larger r/t ratio (and thus higher pressure stress) than the N2 nozzle, a bounding approach is used in this calculation to qualify all applicable nozzles using the N1 nozzle geometry.

2.0 OBJECTIVES The objectives of this calculation package are to:

1. Develop a Finite Element Model (FEM) for the N1 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 (PFM) evaluation to be performed in a separate calculation package.

3.0 ASSUMPTIONS The following assumptions are made in this evaluation:

The N1 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 3) 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 nozzle-to-vessel weld is assumed to have thermal properties equal to the vessel material, since all carbon moly steels have the same thermal properties in the ASME Code [4, 5]. The modeled location and shape of the nozzle-to-vessel weld are approximate and for visualization purposes only.

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The inside surfaces are subjected to an essentially infinite convectivc heat transfer coefficient (HTC) of 1.0,000 Btu/hr-lt`-°F to maximize through wall thermal gradients and thus thermal stresses.

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 recirculation outlet nozzles from Dresden, Units 2 and 3 are identical to the ones from Quad Cities, Units 1 and 2 when the nozzle drawings are compared [6, 7, pg. A-17]. The RPV inside radius (IR) is 125.6875 inches (to base metal, not including the 0.1875 inch cladding [6, 7, pg. A-1 and A-8]), with a vessel wall thickness of 6.125 inches [6, 7, pg. A-1 and A-8], resulting in an outside radius (OR) of 131.8125 inches. The N1 nozzle has an IR of 12.95 inches with cladding (13.1375 inches without cladding), with an outside radius (OR) of 26.5 inches on the vessel side and 14.6875 inches on the safe end side of the nozzle. The nozzle to vessel blend radius is 3.0 inches on the inside and 5.75 inches on the outside [6, 7, pg. A-17]. Figure 1 shows the nozzle geometry with key dimensions identified.

4.2 Material Properties The material component identifications are as listed:

RPV shell: SA-302 Grade B in accordance with Code Case 1339 (evaluated as carbon moly steel) [8, Section 5.3.3.1.1],

Nozzle forging: Mn-Mo steel [6, 7, pg. A-16],

Cladding ASTM 371 ER308L (evaluated as austenitic stainless) [6, 7, pg. A-16]

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

The material properties are taken from the Design Code of Construction [8 Section 5.3.1 ], the 1965 Edition through the Summer 1965 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 and Table 3.

4.3 Transient Definitions The thermal transient definitions are obtained from Reference [9]. Only normal and upset (Service Level A and B) transients for the RPV and N1 nozzle specific transients are considered [9a, 9b].

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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 N1 (Recirculation Outlet) and N2 (Recirculation Inlet) nozzles are identified as the bounding RPV nozzles when fluence is not considered (geometry only) [3]. Since the N nozzle does not meet the NRC additional requirements outlined in Reference [2] and has a larger r/t ratio (and thus higher pressure stress) than the N2 nozzle, a bounding approach is used in this calculation to qualify all nozzles except for the feedwater and control rod drive return nozzles using the N 1 nozzle geometry.

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

The following process is used:

Select limiting Service Level A/B transient 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 A bounding transient is selected for evaluation rather than analyzing all Service Level A/B transients defined on the RPV and recirculation outlet nozzle TCD. Separate load cases are defined for the internal pressure and thermal portions of the transient. 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 transient is 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 [10]. 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 N 1 nozzle-to-vessel weld, the N 1 nozzle, and a portion of the attached safe end, as shown in Figure 1. The N1 nozzle-to-safe end weld is not modeled because it is sufficiently far from the region 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 2.

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 File No.: 1400735.301 Page 6 of 21 Revision: 1 F0306-01R2

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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 [11, 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 Tor adequate mesh density is a change in peak stress intensity between solutions of less than I%.

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 N1 nozzle blend radius and nozzle-to-vessel weld, as shown in Figure 3. 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.

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6.1 Bounding Transient Selection The nozzle specific transient "Interruption of Feed Flow" located between Zone 9 and 10 [9a] has a instantaneous temperature down shock to 70°F from the normal operating temperature, before instantaneously returning to the normal operating temperature after only 60 seconds, bounding all Service Level A/l3 recirculation outlet nozzle specific transients. This transient only applies to Dresden 2 and 3, but will be considered for Quad Cities 1 and 2 for bounding purposes. Due to the severity of the transient, the vessel "SCRAM" transient located in Zone 10 to 11 of Region B [9b] is selected to bound all vessel transients in Reference [9b]. These transients are tabulated in Table 1, and includes the effects of extended power Uprate (EPU) [12].

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.

P

IR, 2 1000.13.1375' Pec1 = , 2 = 2 = 4,002 psi (OR] IR, 14.6875 - 13.1375 where, Peel = End cap pressure on attached piping free end (psi)

P = Internal unit pressure (psi)

IR1 = Inside radius of modeled nozzle piping (in) = 13.1375 inches (w/o cladding) [6, 7, pg. A-17]

OR, = Outside radius of modeled nozzle piping (in) =14.6875 inches [6, 7, pg. A-17]

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

P

IR2 2 1000.125.6875 Peet= 2 _ = 10,016 psi (OR22 IRz (131.8125 125.6875 where, Pec2 = End cap pressure on vessel shell axial free end (psi)

P = Internal unit pressure (psi)

IR2 = Inside radius of modeled vessel shell (in) = 125.6875 inches (w/o cladding) [6, 7, pg. A-1 ]

OR2 =Outside radius of modeled vessel shell (in) = 13 1.812 5 inches [6, 7, pg. A-1 ]

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. The applied load and boundary conditions for this case are shown in Figure 4. A representative stress intensity contour plot for the unit pressure analysis is shown in Figure 5.

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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 60 seconds after the 3,600 seconds of additional time).

6.3.1 Therinal. 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-f1'-°F. The heat transfer coefficient for the RPV and nozzle external surfaces of 0.2 Btu/hr-ft20F, along with an ambient temperature of 100°F, are typical values [13, Section 4.6.9.3] for all times during the transient. The symmetry planes, top/bottom surfaces of the RPV, and "cut" plane in the piping are treated as adiabatic. Figure 6 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 7 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 SCRAM transient are shown in Figure 8 and Figure 9, 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 10 depicts both meshed models and shows that there is less than 0.1 % 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 3 are reviewed to ensure a smooth stress history response is obtained. Representative plots of the Path 1 linearized membrane-plus-bending stress intensity history are shown in Figure 11 and clearly show the time points and resulting smooth 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:

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1). r 1000.128.7525 Phoop = = 21,020 hSl (t) (6.125) where, Pimp = Thin wall cylinder predicted hoop stress (psi)

P = Internal unit pressure (psi) r = Mean radius of modeled vessel shell (in) = 128.75 inches (w/o cladding) [6, 7]

t = Wall thickness of vessel (in) = 6.125 inches (w/o cladding) [6, 7]

Axial stresses should be half the hoop stresses. Nodal query of hoop stress indicate a variance of less than 1% from the thin wall formulae at the vessel mid-wall, which is acceptable since the r/t ratio for the vessel is approximately 21.

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 N1 nozzle blend radius and nozzle-to-vessel weld in support of fixture PFM calculations. All of the stress results are stored in computer files for later use (see Appendix A for file listings).

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8.0 REFERENCES

1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section X1, 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.I08P.

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.

1021005N P.

4. ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Summer 1965 Addenda

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

6. Babcock & Wilcox Certified Design Document for Quad Cities I&II, Contract No. 610-0122-51/52, SI File No. CECO-36Q-210.

7. Babcock & Wilcox RPV Certified Stress Reports:

a. Dresden 2, Contract No. 610-0098-51, Rev. 1, SI File No. 1400735.205

b. Dresden 3, Contract No. 610-0111-51, Rev. 2, SI File No. 1400735.206.

8. Excerpts from Quad Cities UFSAR, Revision 6, October 2001, Section 5.3, SI File No. 1400735.208.

9. Thermal Cycle Diagrams:

a. General Electric Drawing No. 158137279, Sheet 1, Revision 1, "Nozzle Thermal Cycles (Recirculation Outlet)," SI File No. 1400735.203.

b. General Electric Drawing No. 921 D265, Sheet 1, Revision 1, "Reactor Thermal Cycles,"

SI File No. 1400735.209.

10. ANSYS Mechanical APDL and PrepPost, Release 14.5 (w/ Service Pack 1 UP20120918), ANSYS, Inc., September 2012.

11. 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.

12. General Electric Design Specifications

a. Report No. 26A5588, Revision 0, "Reactor Vessel -Power Uprate, Quad Cities 1& 2,"

SI File No. 140073 5.201.

b. Report No. 26A5587, Revision 0, "Reactor Vessel - Power Uprate, Dresden 2 & 3,"

SI File No. 1400735.201

13. GE Purchase Specification 21A9477, Revision 7, "Reactor Vessel," SI File No. 0801345.203.

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Table 1: Bounding Transients for Analysis 19, 121 Inside Outside Surface Surface Time, Tnnzzle, Tvessei, Pressure, Description h, h, sec °F °F prig Btu/hr- Btu/hr-ftz ° I,' i'tz-O F 0 530 530 SCRAM 4320 400 400 8640 530 530 0 530 530 1000 10000 0.2 Intel*rupNol, of reed 0.1 70 530 Flow 60.1 70 530 60.2 530 530 Notes:

1. A total of 3,660 seconds are added to the end of each transient followed by an imposed steady state condition 60 seconds later (see Section 6.3 of this calculation).

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

Table 2: Material Properties for Carbon Moly Steel 14, 51 Young's Mean Thermal Specific Temperature, Thermal Conductivity, Modulus, Expansion, Heat, OF x10, psi x10-' in/in /°F x104 Btu/sec-in-°F Btu/lb-°F 70 29.9 6.10 8.33 0.113 200 29.5 6.38 8.07 0.117 300 29.0 6.60 7.69 0.121 400 28.6 6.82 7.31 0.124 500 28.0 7.02 6.90 0.128 600 27.4 7.23 6.55 0.132 Density. p = 0.283 Win3, assumed temperature independent.

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

Note: Specific Heat values are derived from the equation shown in General Note (5) of Table I-4.0 [5], Specific Heat = TC / (TD x density).

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Table 3: Material Properties for Austenitic Stainless Steel 14, 51 Young's Mean Thermal Specific Temperature, Thermal Conductivit Modulus, Expansion, y' Heat, O F x 10' psi x 10- ' in/in/°R x 10-a Btu/sec-in-'F Btu/1b-°F 70 27.4 9.20 1.93 (M I I 200 27.1 9.34 2.06 0.115 300 26.8 9.47 2.16 0.117 400 26.4 9.59 2.27 0.120 500 26.0 9.70 2.37 0.123 600 25.4 9.82 2.48 0.125 Density, p = 0.29 Win', asSUmed temperature independent.

Poisson's Ratio, u 0.3 1, assumed temperature independent.

Note: Specific I-leat values are derived from the equation shown in General Note (5) of Table 1-4.0 [5], Specific Heat - TC / (TD x density).

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-: D-Wr K%7

.6875 in. OR 3.1375 in. IR i in. OR 13 OR Blend Radius end Radius 12 adding SS)

Figure 1: Components Included in the Finite Element Model (Inside radius (I R) dimensions do not include the 0.1875 inch thick cladding.)

Ores Qq N1 - Vi znoz Model - 30 Figure 2: 3-D Finite Element Model Mesh for Analyses, Baseline Mesh File No.: 1400735.301 Page 14 of 21 Revision: 1 F0306-01R2

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Figure 3: Path Locations for Through-Wall Stress Extractions r'Tr. rxra

- 1' Figure 4: Applied Boundary Conditions and Unit Internal Pressure (Units for Pressure in psi)

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[-U u Im i STF.; -

rnw sir MI -o Ct4t ~.11'~Q STT) :Z69

S7C ~i?4G1 Figure 5: Total Stress Intensity Plot for Unit Internal Pressure (Units for stress intensity in psi)

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.01429 a) Pleat transfer coefficient (HTC) b) Bulk temperature (TBULK)

Figure 6: Applied Thermal Boundary Conditions for Thermal Transient Analyses (End of Single Relief or Safety Valve Blowdown SCRAM Transient shown)

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

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31E?-J c~ .7

'l'lA$;-6.lUU sxa =a2Q.:

SXa -ABU.!

45C.445 DrPs Figure 8: Temperature Contour for SCRAM Transient at Time=6000 sec.

(Units for temperature in °F)

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^a 5~_ro

t IM-6WO

utrr x is-o 12~Y ~9,1G

09-Wbo/

0024.7 m,

Figure 9: Stress Intensity Plot for SCRAM Transient at Time=6000 sec.

(Units for stress intensity in psi)

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Original Fine Me-,h Figure 10: Total Stress Intensity Contours for Mesh Sensitivity Study Unit Pressure (Units for stress intensity in psi)

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10000 9000 8000 Q

7000 6000 V)

Q) 5000 4

C:

tn 4000 V) 3000 2000 j

--Sint(I) 1000

~ Sint(0) 0 0 2000 4000 6000 8000 10000 12000 Time (s)

SCRAM nterruption of Feed Flow Figure 11: Linearized Membrane-Plus-Bending Stress Intensity History for Path 1, SCRAM and Interruption of Feed Flow Note: SINT(1) and SINT(0) refer to the membrane-plus-bending stress intensity on the inside and outside surfaces of the model, respectively.

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APPENDIX A FILENAMES File No.: 1400735.301. Page A-1 of A-2 Revision: 1 F0306-01R'22

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File Name Description STACK.INP Controller input file to run thermal and mechanical analyses Dres_QC_N1.INP Input file to construct the 3-D model for linear-elastic analysis Dres_QC_N1_X2.INP Input file to construct the 3-D model with Refined Mesh Input file of temperature dependent linear elastic material MProp_Linear_Dres_QC.INP properties THM_Dres_QC_1.INP Analysis input file for SCRAM Transient THM_Dres_QC_2.INP Analysis input file for Int. Feed Flow Transient Dres_QC_PRESS.INP Analysis input file for Unit Internal Pressure Dres_QC_PRESS_X2.INP Analysis input file for Unit Internal Pressure using Refined Mesh THM_Dres_QC_1_mntr.inp Load step definition file from thermal analysis - SCRAM THM_Dres_QC_2_mntr.inp Load step definition file from thermal analysis - Int. Feed Flow Thermal temperature time history extraction macro file to create CMNTR.MAC THM

mntr.inp files GenStress.mac Path stress extraction macro file to extract .CSV files GETPATH.TXT Through-wall path definition file Curve fit coefficients outputs of stresses in tabulated forms STR_*_COE_P?.CSV * = PRESS, Dres_QC_1, Dres_QC_2

? = path number 1-4 File No.: 1400735.301 Page A-2 of A-2 Revision: 1 F0306-01 R2

ATTACHMENT 3 Structural Integrity Associates, Inc. File No. 1400735.302, Revision 1, "Code Case N-702 Evaluation for Dresden and Quad Cities Recirculation Outlet (N1) Nozzle,"

dated June 23, 2016

rile No.: 1400735.302 Structural Integrity Associates, Inc .0 Project No.: 1400735 and 1501375 CALCULATION PACKAGE Quality Program: Nuclear Commercial PROJECT NAME:

N-702 Evaluation for Dresden and Quad Cities CONTRACT NO.:

00517760 1lev 62 and 63 CLIENT: PLANT:

I,xelon Corporation DI-esdell a nd Quad Cities generating Sta tions CALCULATION TITLE:

Code Case N-702 Evaluation for Dresden and Quad Cities Recirculation Outlet (N 1) Nozzle Project Manager Preparer(s) &

Document Affected Revision Description Approval Checker(s)

Revision Pages Signature & Date Signatures & Date Responsible Engineer:

0 1 -17 Initial Issue A-I - A-2 Daniel Sommerville Wilson Wong 1/22/16 1/22/16 Responsible Verifier:

Jim Wu 1/22/16 Responsible Engineer:

1 3, 11, 12, 14 Removed Proprietary W,~~ W,~~

Markings and Changed 11~

References 6 and 15 Wilson Wong Wilson Wong 6/23/16 6/23/16 Responsible Verifier:

Jim Wu 6/23/16 Page 1 of 17 F0306-01 R2

Structural Integrity Associates, Inc .9 Table of Contents

1.0 INTRODUCTION

.........................................................................................................3 2.0 013.11;CTIVI :..................................................................................................................3 3.0 MF'I'l K)DO LOG Y ........................................................................................................3 3.1 Fatigue Cycles ...................................................................................................4 3.2 Probabilistic Fracture Mechanics Evaluation ....................................................4 4.0 DFSIGN INPU'l .............................................................................................................5 4.1 Deterministic Parameters ...................................................................................5 151 ....................................................................................................................... 5 4.1.2 SIresses ..............................................................................................................5

-1.1.3 Fatigue C)~cles ................................................................................................... S 4.2 Random Variables .............................................................................................6

-1.2.1 SCC Initiation ....................................................................................................6 4.2.2 SCC Growth ....................................................................................................... 6 4.2.3 1 at igue 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 N1 Nozzle ............................................14 Table 4: PoF for Period of Extended Operation .....................................................................15 List of Figures Figure 1: Stress Extraction Path Orientations in the N 1 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 File No.: 1.400735.302 Page 2of17 Revision: 1 F0306-01R2

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1.0 INTROI)UCTION Exelon intends to extend the applicability of Code Case N-702 1 1 1 at both Dresden Units 2&3 and Quad Cities Units 1 &2 through the end of their respective periods 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 of BWRVIP-108 and BWRVIP-241, is performed to ensure that the probability of failure remains acceptable. The N1 (Recirculation Outlet) and N2 (Recirculation Inlet) nozzles are identified as the bounding nozzles when fluence is not considered[4].

Since all 4 units under consideration have identical N 1 and N2 geometries and the N 1 nozzle has a larger r/t ratio (and thus higher pressure stress) than the N2 nozzle, a bounding approach is used in this calculation to qualify all applicable nozzles in which the N1 nozzle geometry is considered in conjunction with the bounding fluence from the applicable nozzles.

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 Dresden/Quad Cities N1 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 NI 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] shows the highest fluence at 5.5 9x 101 ~ 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 5x10-6, then no comparison to the full inspection case is required.

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3.1 Fatigue Cycles In the Fl--'.M 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 N 1 nozzles. The nozzle specific transient "Interruption of Feed Flow" located between Zone 9 and 10 of the nozzle thermal cycles [I I] has a 450°F temperature shock and returns to 520°F after only 60 seconds, bounding all Service Level A/B recirculation outlet nozzle specific transients. This transient only applies to Dresden 2 and 3, but will be considered for Quad Cities 1 and 2 for bounding purposes.

Due to the severity of the transient, the vessel "SCRAM" transient located in Zone 10 to 11 of Region B

[ 12] is selected to bound all 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 [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 nozzle 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 methodology 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, std devation, 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 of simulations run gives the probability of failure..

File No.: 1400735.302 Page 4 of 17 Revision: 1 F0306-01R2

Structural Integrity Associates, lmc The VIPER program was developed as part of the BWRVIP-05 effort l-or Boiling Water Reactor (BWR) reactor pressure vessel (RI'V) 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].

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, and 12]. Subsections 4.1.1through 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 N nozzle [5]. In that calculation package, through wall stress distributions are presented at four locations in the region of the N1 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 NI 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 Fatigue Cycles The thermal transients from Reference [5] were obtained from the thermal cycle diagrams in References 11 and 12, and the number of cycles to be considered for fatigue are listed in Reference 18. The number of cycles considered for each bounding transient is explained in Section 5.0.

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4.2 Random Variables Random variables (RV) used lil the N-702 evaluation are summarized iii Table able 3. Subsections 4.2.1 through 4.2.3 describe the more detailed input para111eterS Used for SCC Initiation, SCC Growth and fatigue crack growth respectively. Table 3 identifies the specific rcicrences for each RV used in this N-702 evaluation.

-1.2.1 SCC Initiation The cladding stress corrosion crack (SCC) initiation model 111 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 13WR 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 *10'6 10-' (1) where: T= time, hours 6 = applied stress, ksi The residual plot shows that a lognormal distribution produces the best tit for the data. The lognormal residual plot with the linear tit of the data is shown below:

(1) = 0.9248x - 0.003 (2) where: ( _ (x - (T) / µ 6 = data mean

µ = data standard deviation x = In (Tactual/Tpredicted) 4.2.2 SCC Growth The SCC growth model in VIPERNOZ program is also a power law relationship [15]. The relationship used is; da

=6.82*10-1'K' (3) dt where: da/dt = stress corrosion crack growth rate, in/hr K = sustained crack tip stress intensity factor, ksi4in File No.: 1400735.302 Page 6 of 17 Revision.: 1 F0306-01R2

C structural Integrity Associates, Inc 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:

Y = 0.9085x - 0.3389 (4) where: Y = In (In (I/ (1-F) ))

F = cumulative distribution from 0 to 1 x = In ((da/dt) actual / (da/dt) predicted) 4.2.3 Fatigue Crack Growth 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 of Kmin/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:

2.927 da = 3.817

10 " (AK) (5) do where a = crack depth, in n = cycles OK = Kmax Kmin, ksi-in0.5 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 OK.

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 (6) where y = ln(ln(1 /(1-F))

x = ln((da/dn)acwa]/(da/dn),,,.)

F = cumulative probability distribution File No.: 1400735.302 Page 7 of 17 Revision: 1 x0306-01 R2

C Structural Integrity Associates, Inc 5.0 STRESS RESULTS AND FATIGUE CYCLU" LOADINGS The stress analyses for the nozzle-to-shell weld and the nozzle blend radius for the N 1 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 ol'the nozzle.

The load cases analyzed for the N1 nozzle include:

1. Unit pressure (1000 psi)

2. SCRAM Transient (Zone 10 to 11) [ 12]

3. Interruption of Feed Flow Nozzle Transient (Zone 9 to 10) [ l 1 ]

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 transient Interruption of Feed Flow (Zone 9 to 10, 80 cycles, 11, 13, 18) is the bounding Service Level A/B transient because it has a 450°F temperature shock and returns to 530°F after only 60 seconds. Reference 11 states that this transient is only applicable to Dresden, but will be considered in this analysis for both plants to bound the analysis. The transient occurs in 20 out of the 80 cycles of the Interruption of Feed Flow transient. The second part of the transient is bounded by the first shock, but occurs 60 times during the 80 cycles of the Interruption of Feed Flow transient. The Improper Start of Cold Recirc Loop (Zone 13 to 14, 11) nozzle transient (10 cycles, 12) is also bounded by the first Interruption of Feed Flow shock.

Thus, 90 cycles of the third load case analyzed in Reference [5] will be considered for the 40 year revised design basis cycles. This amount to 23 cycles for each block of 10 years of operation over 60-years of operation Due to the severity of the transient, the Vessel SCRAM transient (Zone 10 to 11, 12) is selected to bound all other transients.

The number of thermal cycles for the SCRAM transient is considered to be the total number of cycles for Service Level A/B conditions and normal startup/shutdown cycles [11, 12, 18]. Therefore the transients bounded were:

Plant startup (298 cycles),

o Includes Turbine roll with Feedwater Injection

Plant cooldown (286 Cycles),

o Includes Reduction of power File No.: 1400735.302 Page 8of17 Revision: 1 F0306-01R2

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Loss oi~ Feedwater Flow (80 Cycles),

SCRAM (294 Cycles),

This totals:

958 cycles for the design 40 years of operation and

1437 cycles when scaled for 60 years of operation, or 240 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 tit 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- 180-3 02 [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 ksNin with a standard deviation of 0 ksNin for 1111-irradiated material [2].

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

8. No CMTR is available for the nozzle-to-vessel weld in the RPV fabrication records [21 ], thus the copper/nickel content and initial RT dt values are taken from Reference [3].

9. No CMTR is available for the Dresden nozzles. Typically when material chemistries are not available, generic industry values are used from Reference [3] and [4]. However, since the values available for Quad Cities bound the generic values, the Quad Cities values are considered bounding for this calculation.

10.No copper content is available in the nozzle forging CMTR [21 ], thus the nozzle forging copper content of 0.09189% from Reference [4] is used.

11. Zero inspection coverage conservatively assumed for the initial 40 years of operation File No.: 1400735.302 Page 9of17 Revision: 1 F0306-01R2

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7.0 RESULTS OF ANALYSES The reliability evaluation is presented using plant specific inspection coverage. The probabilities of failure (Po F) 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 5 x 10-' per year acceptance criterion 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 Dresden/Quad Cities N 1 nozzle is below the acceptance criterion of 5 x 10-6 per year. This analysis shows that the N 1 nozzles meet the acceptable failure probability even when considering elevated f7uence level, thus qualdf ing all Dresden 2/3 and Quad Cities 1/2 RPV nozzles with frill 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.

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9.0 REFERENCES

1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section X1, 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-1 08)," December 19, 2007, SI File No. BWRVIP.108P.

3. B WR VIP- 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.

1021005NP.

5. SI Calculation 1400735.301, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N I Nozzle," Revision 1, June 2016.

6. General Electric Fluence Evaluation GE-NE-0000-0011-0531-R3-NP, "Dresden and Quad Cities Neutron Flux Evaluation," Rev. 3, SI File No. 1400735.213.

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. Babcock & Wilcox RPV Certified Stress Report for Dresden 3, Contract No. 610-0111-51, Rev. 2, SI File No. 1400735.206.

11. General Electric Drawing No. 158137279, Sheet 1, Revision 1, "Nozzle Thermal Cycles (Recirculation Outlet)," SI File No. 1400735.203.

12. General Electric Drawing No. 921D265, Sheet 1, Revision 1, "Reactor Thermal Cycles," SI File No.

140073 5.209.

13. General Electric Report No. 26A5588, Revision 0, "Reactor Vessel - Power Uprate," SI File No.

1400735.201.

14. SI Calculation W-EPRI- 180-3 02, "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.

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16. Bamford, W. 11., "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 X1, Appendix G, 2013 Edition.

18. Exelon Transmittal of Design Input (TODI) No: QDC-13-049, Rev. 0, SI File No. 1301300.205.

19. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS RuIC (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. RPV Fabrication Records

a. Quad Cities 1, SI File No. 1400735.212.

b. Quad Cities II, SI File No. 1400735.211.

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

August 31, 2012, SI File No. 1300341.213.

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Table 1: Deterministic Parameter Summary VIP[?RNOZ Variable Value Reference RPV Thickness 6.125 inches (excluding clad) [ 101 RPV Radius 125.6875 inches (to vessel surface) [ 10]

Clad Thickness 0.1875 inches 101*

Operating Temperature 530 °F(Region I3) [ 1 1,12, 13]

LT013 Event Temperature 100 01" [20]

Operating Pressure 1005 psig [ 1 1,12, 13]

LT013 Event Pressure 1200 psig [20]

"Note: It was determined in Reference [5] that the N 1 nozzles o(~'all four units under consideration are identical.

Table 2: Probability of Detection Distribution 1 141 Flaw Size, in. Cumulative POD 0.00 0.00 0.05 0.10 0.10 0.46 0.15 0.80 0.20 0.92 0.25 0.95 0.30 0.98 0.35 0.99 0.40 0.99 0.45 1.00 0.50 1.00 0.55 1.00 0.60 1.00 File No.: 1400735.302 Page 13 of 17 Revision: 1 F0306-01 R2

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Table 3: Random Variables Parameter Summary for N 1 Nozzle Random Parameter Mean Std Dev Distribution Ref.

l law density, nozzle/shell ,

1 per weld ~MCall Poisson [3,3,3]

weld (fabrication)

Flaw density, nozzle/shell I per weld Mean Poisson (3,3,3]

weld (SCC initiation)

F law density, Nozzle blend I>

1 per weld 4mean Poisson [2,2,3]

radius ((-abricatioll)

I-,law size (fabr lcatloil) n/a n/a PVRUF [3]

Flaw size (stress corrosion) Clad thickness n/a Constant [3,3]

8 Weld residual stress, inside surface 5 Normal [3,3,3]

through-wail (ksi) cosine distribution Clad residual stress (ksi)a- 32 5 Normal [3,3,3]

Cu 0.26 0.045 Normal [3,3,3]

N 1 Nozzle

-% Ni 1.2 0.0165 Normal [3,3,3]

to shell Initial R"1'~it weld initi -20 13 Normal [3,3,3]

% Cu 0.09189 0.04407 Normal [2,2,3]

N1 Nozzle %Ni 0.70 0.068 Normal [21,2,3]

forging Initial RTa`

40 26.48 Normal [21,2,3]

(° F)

Lower Shell fast neutron 5.59e17 0.2(20%) n/a [6,3]

K1c upper shelf (ksi4in) 200 0 Normal [2,22,3]

Residual SCC initiation time (hr) r = 84.2x1018 (6)-10.' y=0.9248x- Lognormal [2,3,3]

0.0003 K dependent Residual da/dt = 6.82x10-12(K)`1 y=0.9085x- Weibull [15,3,3]

K >50 ksi4in 0.3389 SCCG (in/hr)

K independent da/dt = 2.8x 10-6, Na na [15]

K <50 ksi Vin SCC threshold (ksi4in) 10 2 Normal [2,3,3]

Residual Fatigue crack growth (FCG) da/dn=3.82 y=4.155x- Weibull [3,353]

(in/cycle) x10-9(dK)2.927 0.3712 FCG threshold (ksi4in) 0 0 Normal [3,3,3]

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

File No.: 1400735.302 Page 14 of 17 Revision: 1 F0306-01 R2

C structural Integrity Associates, Inc .9 Table 4: PoF for Period of Extended Operation PoF per year due to LTOP Event Location Zero inspection Allowable PoF per for initial 40 years, Y ear 19 25% for PE0*

-s Path 1 7.8 X 10 Path 2 3.33 X 10-"

5.0 X 10-6 Path 3 < 8.33 X 10-13 Path 4 < 8.33 x 10-13

Note: Values include I X 10-3 probability of LTOP event occurrence per year [3. pg 5-13].

e.LU~

MAT tSki PATH

- Vipernoz Nhdel - 3D Figure 1: Stress Extraction Path Orientations in the N1 Nozzle File No.: 1400735.302 Page 15 of 17 Revision: 1 F0306-01 R2

Structural Integrity Associates, Inc" 55 50 + Path 1 45 Path 2 N n Path 3 40 U) 35 o Path 4 L

30 c 25 p 20 Y

15 L

U 10 5

01 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 Path length (in)

Figure 2: Pressure Stress Distributions Tor the N-702 Evaluation 10 4t o

a lilt p o ~i Q ppp °°° I V p° s Path 1 Path 2 L Path 3 o Path 4

-20 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 Path length (in)

Figure 3: Full Power Thermal Expansion Stress Distributions for the N-702 Evaluation File No.: 1400735.302 Page 16 of 17 Revision: I F0306-01 R2

C Structural Integrity Associates, Inc 70 60 -*+- Path 1 -- Path 2

~ 50 Y c. Path 3 o Path 4 N 40 Cn L 30 M

20 0 10 0

CU L

U -10

-20

-30 ' i I Ij _

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 Path length (in)

Figure 4: Bounding 'Transient Stress Distributions for the N-702 Evaluation 12

-*-Cosine Stress Distr 10 Poly. (Cosine Stress Distr) 8 y = -4E-14x3 + 1.543x2 -10.415x + 11.631

' Y 6 N

4 2

.o 0

-2

-6

-8

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

Figure 5: Weld Residual Stress Distributions for Paths 2 and 4 File No.: 1400735.302 Page 17of17 Revision: 1.

F0306-0 I R2

Structural Integrity Associates, Inca Appendix A LIST OF SUPPORTING FILES File No.: 1400735.302 Page A-1 of A-2 Revision: 1 F0306-01 R2

Structural Integrity Associates, Inc .9 File Name Description Path 1.IN13 VIPFIRNOZ input file for Path I at nozzle blend radii.

Patl13.INP VIPIRNOZ input file for Path 3 at nozzle blend radii.

Patll2.INP VIPIRNOZ input file. for Path 2 at nozzle-to-shell-weld.

Patl14.INP VIPERNOZ_, input file for Path 4 at nozzle-to-shell-weld.

Path LOUT Vll3E.RNOZ, output file for Path I at nozzle blend radii.

Path3.0UT VIPERNOZ output file for Path 3 at nozzle blend radii.

Path2.OUT VIPERNOZ, output file for Path 2 at nozzle-to-shell-weld.

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

VIPERNOZ v2.EXE VIPERNOZ executable program ISPCTPOD.EXE VIPERNOZ probability of detection curve input file l~LWDS' RB.EXE VIPERNOZ flaw size distribution curve input file File No.: 1400735.302 Page A-2 of A-2 Revision: 1 F0306-01 R2

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