NMP1L3137, Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)

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Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)
ML17067A056
Person / Time
Site: Nine Mile Point  Constellation icon.png
Issue date: 03/07/2017
From: Jim Barstow
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP-RR-001, NMP1L3137
Download: ML17067A056 (47)
v • d • e


Text

200 Exelon Way Exelon Generation . Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.55a NMP1L3137 March 7, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRG Docket Nos. 50-220 and 50-410

Subject:

Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)

In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC (Exelon), is requesting relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components." Relief Request NMP-RR-001 proposes an alternative to the requirements contained in Table IWB-2500-1 concerning nozzle-to-vessel weld and nozzle inner radii examination requirements.

We request your review and approval by March 7, 2018.

No regulatory commitments are contained in this letter. Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Relief Request to Utilize ASME Code Case N-702

2) Plant Specific Applicability
3) Structural Integrity Associates, Inc. Report No. 1501576.401.R1, "Evaluation of Nine Mile Point Units 1 and 2 Using Code Case N-702"
4) Structural Integrity Associates, Inc. Report No. 1501576.301, "Effects of NMP 2 SRV Slowdown Transient on Probability of Failure" cc: Regional Administrator - NRG Region I NRG Senior Resident Inspector - NMP NRG Project Manager, NRA* NMP A. L. Peterson, NYSERDA

Attachment 1 Relief Request to Utilize ASME Code Case N-702

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 1 of 12)

1. ASME CODE COMPONENTS AFFECTED Code Class: 1 Component Numbers: Reactor Vessel Nozzles - See Enclosure 1 for complete list of nozzle identifications Examination Category: 8-D (Inspection Program B)

Item Number: 83.90 and 83.100

Description:

Alternative to ASME Section XI, Table IWB-2500-1

2. APPLICABLE CODE EDITION AND ADDENDA The fourth 10-year lnservice Inspection (ISi) program at Nine Mile Point Nuclear Station, Unit 1 (NMP1) is based on the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (B&PV) Code,Section XI, 2004 Edition. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2001 Edition is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv).

The third 10-year ISi program at Nine Mile Point Nuclear Station, Unit 2 (NMP2) is based on the ASME B&PV Code,Section XI, 2004 Edition. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2001 Edition is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv).

3. APPLICABLE CODE REQUIREMENTS The applicable requirements are contained in Table IWB-2500-1, "Examination Category B-0, Full Penetration Welded Nozzles in Vessels - Inspection Program B." Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number 83.90, "Nozzle-to-Vessel Welds," and 83.100, "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 Enclosure 1 are full penetration welds. The plant, interval, ASME Section XI Code Edition and Addenda, interval start date and interval end date are provided in the table below.

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 2 of 12)

PLANT INTERVAL EDITION START END Nine Mile Point Nuclear Fourth 2004 Edition August 23, 2009 August 22, 2019 Power Station, Unit 1 Nine Mile Point Nuclear Third 2004 Edition April 5, 2008 June 15, 2018 Power Station, Unit 2 This relief request will remain in effect for the remainder of the 10-year intervals and the remaining term of the NMP1 and NMP2 renewed operating licenses, which currently expire on August 22, 2029 and October 31, 2046, respectively.

4. REASON FOR REQUEST The Federal Register Notice (FRN) published November 5, 2014, contains the rulemaking that amends 10 CFR 50.55a to incorporate by reference Regulatory Guide (AG) 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1." As stated in the FAN, licensees may use the Code Cases listed in AG 1.147 as alternatives to engineering standards for the construction, inservice inspection, and inservice testing of nuclear power plant components. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," is listed in AG 1.147, Table 2, "Conditionally Acceptable Section XI Code Cases." The Condition associated with Code Case N-702 is as follows:

"The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013(ML13071A240) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case."

In the section of the FRN associated with NRG Responses to Public Comments on Draft Regulatory Guides, the NRC responses to comments specific to Code Case N-702 start on page 9 of 40 (79 FR 65783). An excerpt from the FRN is included as follows:

"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 the use of ASME Code Case N-702 as an alternative. However, licensees should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by addressing the conditions and limitations specified in Section 5.0 of the NRC Safety Evaluation for BWRVIP-241."

The proposed alternative provides an acceptable level of quality and safety based on the technical content of BWRVIP-108 and BWRVIP-241, as endorsed by the NRC Safety Evaluation Reports (SERs).

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 3 of 12)

5. PROPOSED ALTERNATIVE AND BASIS FOR USE In accordance with 10 CFR 50.55a(z)(1 ), relief is requested from performing the required examinations on 100 percent of the nozzle assemblies identified in Tables 5-1 and 5-2 below (see Enclosure 1 for the list of RPV nozzles applicable to this relief request). As an alternative, for all welds and inner radii identified in Tables 5-1 and 5-2, Exelon Generation Company, LLC (Exelon) proposes to examine a minimum of 25 percent of the nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702 ("Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1")for the remaining term of the NMP1 and NMP2 renewed operating licenses. For the nozzle assemblies identified in Enclosure 1, this would mean 25 percent from each of the groups identified below:

Table 5-1 Nine Mile Point Nuclear Power Station, Unit 1 Summary Group* Total Minimum Number Comments Number to be Examined Recirculation Outlet 5 2 Two (2) nozzles have (N1) been inspected this interval. No rejectable indications have been identified.

Recirculation Inlet 5 2 No nozzles have been (N2) inspected this interval.

Main Steam 2 1 No nozzles have been (N3) inspected this interval.

Emergency 2 1 One ( 1) nozzle has been Condenser Supply inspected this Interval.

(NS) No rejectable indications were identified.

Core Spray (N6) 2 1 One (1) nozzle has been inspected this Interval.

No rejectable indications were identified.

Safety Valve 18 5 Ten (10) nozzles have Nozzles been inspected this (N7) interval. No rejectable indications were identified.

  • The nozzle-to-vessel weld and inner radius examinations are performed together.

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 4 of 12)

Table 5-2 Nine Mile Point Nuclear Power Station, Unit 2 Summary Group* Total Minimum Number Comments Number to be Examined Recirculation Outlet 2 1 No nozzles have been (N1) inspected this interval.

Recirculation Inlet 10 3 Six (6) nozzles have (N2) been inspected this interval. No rejectable indications.

Main Steam (N3) 4 1 Three (3) nozzles have been inspected this interval. No rejectable indications.

Residual Heat 3 1 Three (3) nozzles have Removal (N6) been inspected this interval. No rejectable indications.

Reactor Vessel 2 1 One (1) nozzle has been Instrumentation (N9) inspected this interval.

No rejectable indications.

  • The nozzle-to-vessel weld and inner radius examinations are performed together.

The exams in Tables 5-1 and Table 5-2 will be scheduled in accordance with Table IWB-2412-1, Inspection Program B.

Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item No. B3.100, "Nozzle Inside Radius Section"). This VT-1 examination is outlined in Code Case N-648-1 ("Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1"). Exelon will perform either volumetric examination or VT-1 examination of the inner radius as required by Code Case N-702.

Electric Power Research Institute (EPRI) Technical Report 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 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., <1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified.

EPRI Report BWRVIP-241 received a final NRC Safety Evaluation Report on April 19, 2013(ML13071A240). In the NRC Safety Evaluation Report, Section 5.0, "Conditions and Limitations" indicates that each licensee who plans to request relief from the ASME

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 5 of 12)

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 the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVI P-241 report to their units in the relief request as demonstrated in Attachment 2.

As noted in Attachment 2, the NMP1 Recirculation Inlet Nozzle N2 does not meet the general nozzle-specific criterion 3 in BWRVIP-241; however, the NMP1 N2 location is specifically addressed by BWRVIP-241 and NMP1 meets all limitation and conditions of the final NRC Safety Evaluation Report. BWRVIP-108 and BWRVIP-241 addressed the original 40 year operating license. The additional justification for applicability through the 60 year term is provided in the attached Structural Integrity Associates, Inc. (SIA) Report No. 1501576.401.R1 (Attachment 3).

As also noted in Attachment 2, the NMP2 Recirculation Outlet Nozzle N1 location does not meet the general nozzle-specific criterion 4 in BWRVIP-241. Consistent with the note in BWRVIP-241 Section 6, a plant specific analysis of the NMP2 N1 nozzle, , has been performed which demonstrates that the probability of failure (PoF) meets the BWRVIP-108 acceptance criteria through the 60 year term. The NMP2 plant specific analysis is based on the LaSalle N 1 nozzle finite element analysis and probability of failure evaluation. The LaSalle evaluation is applicable to NMP2; however, the NMP2 licensing basis for the safety valve blowdown transient considers this event an upset event whereas the LaSalle licensing basis considered this event an emergency event and therefore this event was not considered in the LaSalle Unit 2 PoF analyses.

The Attachment 4 SIA Report No. 1501576.301 evaluated the effects of the NMP2 Safety Relief Valve (SRV) blowdown transient on the probability of failure and concludes the effects of the SAV blowdown transient result in a PoF that remains acceptable through the 60 year operating term.

6. DURATION OF PROPOSED ALTERNATIVE This relief request will remain in effect for the remainder of the 10-year intervals defined in Section 3 of this relief request and the remaining term of the NMP1 and NMP2 renewed operating licenses, which currently expire on August 22, 2029 and October 31, 2046, respectively.
7. PRECEDENTS
1. "Columbia Generating Station - Relief Request for Alternative 41Sl-04 Applicable to the Fourth 10-Year lnservice Inspection Program Interval (CAC No. MF7331 )"

October 5, 2016 (ML16263A233)

2. "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),"

October 30, 2015 (ML15226A412)

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 6 of 12)

3. "Pilgrim Nuclear Power Plant - Relief Request PRR-24 Regarding Nozzle-to-Vessel Welds and Nozzle Inner Radii Examinations (TAC No. MF4187)," April 21, 2015 (ML15103A069)

Relief Request NMP*RR-001 to Utilize ASME Code Case N-702 in Accordance with 1o CFR 50.55a(z)(1)

(Page 7 of 12)

Enclosure 1 Applicable Nine Mile Point, Unit 1 Nozzles Component Category Item System Nominal Last Last Last Appendix ID Number Number Pipe Exam Exam Exam VIII Exam Size Interval Period Result*

N1A 8-D 83.90 Recirc Suction 28" 3 3 NRI Yes N1A-IR 8-D 83.100 Recirc Suction 28" 3 3 NRI VT-1 **

N18 8-D 83.90 Recirc Suction 28" 4 2 NRI Yes N18-IR 8-D 83.100 Recirc Suction 28" 4 2 NRI VT-1 **

N1C 8-D 83.90 Recirc Suction 28" 3 2 NRI Yes N1C-IR 8-D 83.100 Recirc Suction 28" 3 2 NRI VT-1 **

N1D 8-D 83.90 Recirc Suction 28" 4 2 NRI Yes N1D-IR 8-D 83.100 Recirc Suction 28" 4 2 NRI VT-1**

N1E 8-D 83.90 Recirc Suction 28" 3 2 NRI Yes N1 E-IR 8-D 83.100 Recirc Suction 28" 3 2 NRI VT-1 **

N2A 8-D 83.90 Recirc 28" 3 3 NRI Yes Discharge N2A-IR 8-D 83.100 Recirc 28" 3 3 NRI Yes Dischan::ie N28 8-D 83.90 Recirc 28" 3 3 NRI Yes Discharoe N28-IR 8-D 83.100 Recirc 28" 3 3 NRI Yes Discharge N2C 8-D 83.90 Recirc 28" 3 3 NRI Yes Discharqe N2C-IR 8-D 83.100 Recirc 28" 3 3 NRI Yes Discharge N2D 8-D 83.90 Recirc 28" 3 3 NRI Yes Discharqe N2D-IR 8-D 83.100 Recirc 28" 3 3 NRI Yes Discharoe N2E 8-D 83.90 Recirc 28" 3 3 NRI Yes Discharge N2E-IR 8-D 83.100 Recirc 28" 3 3 NRI Yes Discharne N3A 8-D 83.90 Main Steam 32" 3 3 NRI Yes N3A-IR 8-D 83.100 Main Steam 32" 3 3 NRI VT-1 **

N38 8-D 83.90 Main Steam 32" 3 3 NRI Yes N38-IR 8-D 83.100 Main Steam 32" 3 3 NRI VT-1**

N5A 8-D 83.90 Emergency 18" 3 3 NRI Yes Condenser Supply N5A-IR 8-D 83.100 Emergency 18" 3 3 NRI VT-1 **

Condenser Supply N58 8-D 83.90 Emergency 18" 4 1 NRI Yes Condenser Supply N58-IR 8-D 83.100 Emergency 18" 4 1 NRI VT-1 **

Condenser Suoolv

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 8 of 12)

Applicable Nine Mile Point, Unit 1 Nozzles Component Category Item System Nominal Last Last Last Appendix ID Number Number Pipe Exam Exam Exam VIII Exam Size Interval Period Result N6A 8-D 83.90 Core Spray 6" 3 2 NRI Yes N6A-IR 8-D 83.100 Core Spray 6" 3 2 NRI Yes N68 8-D 83.90 Core Spray 6" 4 1 NRI Yes N68-IR 8-D 83.100 Core Spray 6" 4 1 NRI Yes N7A 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7A-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1**

N78 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N78-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7C 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7C-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7D 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7D-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7E 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7E-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7F 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7F-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7G 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7G-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7H 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7H-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7J 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7J-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1**

N7K 8-D 83.90 Safety Valve 6" 3 2 NRI Yes N7K-IR 8-D 83.100 Safetv Valve 6" 3 2 NRI VT-1 **

N7L 8-D 83.90 Safety Valve 6" 4 1 NRI Yes N7L-IR 8-D 83.100 Safetv Valve 6" 4 1 NRI VT-1 **

N7M 8-D 83.90 Safety Valve 6" 3 2 NRI Yes N7M-IR 8-D 83.100 Safetv Valve 6" 3 2 NRI VT-1 **

N7N 8-D 83.90 Safety Valve 6" 3 1 NRI Yes N7N-IR 8-D 83.100 Safetv Valve 6" 4 1 NRI VT-1 **

N7P 8-D 83.90 Safety Valve 6" 3 1 NRI Yes N7P-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7R 8-D 83.90 Safety Valve 6" 3 1 NRI Yes N7R-IR 8-D 83.100 Safety Valve 6" 4 1 NRI VT-1 **

N7S 8-D 83.90 Safety Valve 6" 3 2 NRI Yes N7S-IR 8-D 83.100 Safety Valve 6" 4 2 NRI VT-1 **

N7T 8-D 83.90 Safety Valve 6" 3 2 NRI YES N7T-IR 8-D 83.100 Safety Valve 6" 4 2 NRI VT-1 **

N7U 8-D 83.90 Safety Valve 6" 3 2 NRI Yes N7U-IR 8-D 83.100 Safety Valve 6" 4 2 NRI VT-1 **

  • NRI - No Recordable Indications
    • This VT-1 examination is outlined in Code Case N-648-1 ("Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1")

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 9 of 12)

Applicable Nine Mile Point, Unit 2 Nozzles Component Category Item System Nominal Last Last Last Appendix ID Number Number Pipe Exam Exam Exam VIII Exam Size Interval Period Result*

N1A 8-D 83.90 Reactor 24" 2 3 NRI Yes Coolant Outlet N1A-IR 8-D 83.100 Reactor 24" 2 3 NRI VT-1**

Coolant Outlet N18 8-D 83.90 Reactor 24" 2 3 NRI Yes Coolant Outlet N18-IR 8-D 83.100 Reactor 24" 2 3 NRI VT-1 **

Coolant Outlet N2A 8-D 83.90 Reactor 12" 3 1 NRI Yes Coolant Inlet N2A-IR 8-D 83.100 Reactor 12" 3 1 NRI Yes Coolant Inlet N28 8-D 83.90 Reactor 12" 2 3 NRI Yes Coolant Inlet N28-IR 8-D 83.100 Reactor 12" 2 3 NRI Yes Coolant Inlet N2C 8-D 83.90 Reactor 12" 3 1 NRI Yes Coolant Inlet N2C-IR 8-D 83.100 Reactor 12" 3 1 NRI Yes Coolant Inlet N2D 8-D 83.90 Reactor 12" 2 1 NRI Yes Coolant Inlet N2D-IR 8-D 83.100 Reactor 12" 2 1 NRI Yes Coolant Inlet N2E 8-D 83.90 Reactor 12" 3 1 NRI Yes Coolant Inlet N2E-IR 8-D 83.100 Reactor 12" 3 1 NRI Yes Coolant Inlet N2F 8-D 83.90 Reactor 12" 3 1 NRI Yes Coolant Inlet N2F-IR 8-D 83.100 Reactor 12" 3 1 NRI Yes Coolant Inlet N2G B-D 83.90 Reactor 12" 2 3 NRI Yes Coolant Inlet N2G-IR 8-D 83.100 Reactor 12" 2 3 NRI Yes Coolant Inlet N2H B-D 83.90 Reactor 12" 3 1 NRI Yes Coolant Inlet N2H-IR 8-D 83.100 Reactor 12" 3 1 NRI Yes Coolant Inlet N2J 8-D 83.90 Reactor 12" 3 1 NRI Yes Coolant Inlet N2J-IR 8-D 83.100 Reactor 12" 3 1 NRI Yes Coolant Inlet N2K 8-D 83.90 Reactor 12" 2 3 NRI Yes Coolant Inlet N2K-IR 8-D 83.100 Reactor 12" 2 3 NRI Yes Coolant Inlet

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 10 of 12)

Applicable Nine Mile Point, Unit 2 Nozzles Component Category Item System Nominal Last Last Last Appendix ID Number Number Pipe Exam Exam Exam VIII Exam Size Interval Period Result*

N3A 8-D 83.90 Main Steam 26" 3 1 NRI Yes N3A-IR 8-D 83.100 Main Steam 26" 3 1 NRI VT-1**

N38 8-D 83.90 Main Steam 26" 3 2 NRI Yes N38-IR 8-D 83.100 Main Steam 26" 3 1 NRI VT-1 **

N3C 8-D 83.90 Main Steam 26" 3 2 NRI Yes N3C-IR 8-D 83.100 Main Steam 26" 3 1 NRI VT-1**

N3D 8-D 83.90 Main Steam 26" 2 3 NRI Yes N3D-IR 8-D 83.100 Main Steam 26" 3 1 NRI VT-1 **

N6A 8-D 83.90 Residual Heat 12" 3 2 NRI Yes Removal N6A-IR 8-D 83.100 Residual Heat 12" 3 2 NRI Yes Removal N68 8-D 83.90 Residual Heat 12" 3 1 NRI Yes Removal N68-IR 8-D 83.100 Residual Heat 12" 3 1 NRI Yes Removal N6C 8-D 83.90 Residual Heat 12" 3 1 NRI Yes Removal N6C-IR 8-D 83.100 Residual Heat 12" 3 1 NRI Yes Removal N9A 8-D 83.90 Reactor Vessel 4" 2 3 NRI Yes Instrumentation N9A-IR 8-D 83.100 Reactor Vessel 4" ~ 3 NRI Yes Instrumentation N98 8-D 83.90 Reactor Vessel 4" 3 1 NRI Yes Instrumentation N98-IR 8-D 83.100 Reactor Vessel 4" 3 1 NRI Yes Instrumentation

  • NRI - No Recordable Indications
    • This VT-1 examination is outlined in Code Case N-648-1 ("Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1")

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 11 of 12)

Attachment 2 Plant Specific Applicability BWRVIP-241 received a final NRG Safety Evaluation Report (SER) on April 19, 2013 (ML13071A240). As discussed in Section 5.0, "Conditions and Limitations," 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-241 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 of the BWRVI P-241 report to their units in the relief request by demonstrating all of the following:

(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115°F/hour.

The maximum RPV heatup/cooldown rate is limited to less than 100°F/hour as defined in the Pressure Temperature Limits Reports (PTLR's) referenced by Technical Specification 3.2.1 for NMP1 and by Technical Specification 3.4.11 for NMP2.

Recirculation Inlet Nozzles (N2l (2) (pr/t)/CRPV S 1.15 2 2 2 2 (3) [p(ro +ri )/(ro -ri )]/CNoZZLE S 1.47 Recirculation Outlet Nozzles (N 1)

(4) (pr/t)/CRPV S 1.15 The terms to be used in the NRG SER, Section 5, applicability evaluations Criteria 2-5 are:

CRPv = recirculation inlet nozzles N2 (from BWRVIP-241 model) = 19332 CNozzLE =recirculation inlet nozzles N2 (from BWRVIP-241 model)= 1637 CRPV =recirculation outlet nozzles N1 (from BWRVIP-241 model)= 16171 CNozzLE = recirculation outlet nozzles N1 (from BWRVIP-241 model) = 1977 p = RPV normal operating pressure (psi) r = RPV inner radius (inch) t = RPV wall thickness (inch) ri = Nozzle inner radius (inch) r0 =Nozzle outer radius (inch)

Relief Request NMP-RR-001 to Utilize ASME Code Case N-702 in Accordance with 10 CFR 50.55a(z)(1)

(Page 12 of 12)

Nine Mile Point Nuclear Station. Unit 1 The results show that NMP1 meets the Criteria 4 and 5 for the N1 nozzle and the Criteria 2 for the N2 nozzle. NMP1 does not meet the screening Criteria 3 for the N2 nozzle.

This location was evaluated in BWRVIP-241 and determined to be acceptable in the BWRVIP-241 SER.

Outlet Nozzles (N1)

Criteria (4) Criteria (5)

CNOZZLE CRPV p r t(min) rj ro S1 .15 S1.59 1977 16171 1030 106.71875 7.125 13.0625 24.59375 0.95 0.93 Inlet Nozzles (N2)

Criteria (2) Criteria (3)

CNOZZLE CRPV p r t(min) rj ro S1.15 S1 .47 1637 19332 1030 113.3125 8.75 13.0625 14.53125 0.69 5.927*

  • Does not meet Criteria.

Nine Mile Point Nuclear Station. Unit 2 The results show that NMP2 meets the Criteria 2 and Criteria 3 for the N2 nozzle and the Criteria 5 for the N1 nozzle. The N1 nozzle plant specific evaluation justifies the acceptability.

Outlet Nozzles (N1)

Criteria (4) Criteria (5)

CNOZZLE CRPV p r t(min) rj ro S1 .15 S1.59 1977 16171 1020 126.6875 6.5626 11.03125 20.0625 1.22* 0.96 Inlet Nozzles (N2)

Criteria (2) Criteria (3)

CNOZZLE CRPV p r t(min) rj ro S1.15 S1.47 1637 19332 1020 126.6875 6.5625 5.8125 13.125 1.03 0.94

  • Does not meet Criteria.

Attachment 3 Structural Integrity Associates, Inc. Report No. 1501576.401.R1, "Evaluation of Nine Mile Point Units 1and2 Using Code Case N-702"

~Structural Integrity Associates, Inc.

5215 Hellyer Ave.

Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 WWY1.strucbnt.com wwong@structinlcom February 15, 2017 Report No. 1501576.401.Rl Quality Program: ['.gl Nuclear D Commercial

Subject:

Evaluation of Nine Mile Point Units 1 and 2 using Code Case N-702 INTRODUCTION Structural Integrity Associates (SIA) is contracted by Exelon to perform a plant specific analysis for inspection relief request per Code Case N-702 over the extended period of operation (60 years) for Nine Mile Point reactor pressure vessel (RPV) nozzles. The technical basis of the inspection relief request is based on the results in Reference [2]. It is stated in Section 2 of Reference [2] that "It should be noted that only the recirculation inlet and outlet nozzles need to be checked because the P(FIE)s, conditional probability of failure (CPoF) of even F due to the occurrence of event E, for other nozzles are an order of magnitude lower."

The Reference [2] evaluation is only applicable for the original plant design life of 40 years. In order to apply the evaluation to an extended operational license period of 60 years, a plant specific evaluation considering fracture toughness material properties, thermal transients, cycle counts, and increased fluence in the RPV at the end of evaluation period (EoEP) must be performed.

TECHNICAL APPROACH For NMP 2, the recirculation outlet (Nl) nozzle is considered the limiting nozzle based on the criteria set forth in the safety evaluation [4] of BWRVIP-108. NMP 2 is a BWR-5 design much like LaSalle Units 1 and 2. Thus, the RPV configurations, material specification, and operating conditions are very similar. Since an evaluation for inspection relief has been performed for the NI nozzle at LaSalle Unit 2 [5] based on the methodology in References [3, 4], the approach in this evaluation will be to assess all the relevant parameters significant to the evaluation of the Toll-free 877-474* 7693 Akron, OH Austin, TX Charlotte, NC Chattanooga, TN 330-899-9753 512-533-9191 704.597.5554 423-553-1180 Chicago, IL Denver, co san Diego, CA san Jose, CA state College, PA Toronto, Canada sn-474-7693 303-792-0077 858-455-6350 408-978-8200 814*954*7776 905-829-9817

February 15,2017 Report No. 1501576.401.Rl Page 2of17 probability of failure (PoF) for inspection relief between these two plants. These relevant parameters, based on the approach in Reference [2] are:

1. Nozzle configurations and dimensions,
2. Base metal and weld chemistry,
3. Fluence,
4. Thermal transients If these parameters are found to be similar as addressed in subsequent sections, then it can be concluded that the results from the LaSalle Unit 2 evaluation [5] can be applied to the inspection relief request for the NMP 2 N 1 nozzles.

For NMPI, the recirculation inlet (N2) nozzle was evaluated in Reference [2]. The selection of this nozzle was based on the criteria set forth in the safety evaluation [4] of BWRVIP-108, and is considered the limiting nozzle for NMP 1. The additional fluence and thermal cycle counts from 40 years to 60 years of operation should be addressed for continued applicability of inspection relief per Code Case N-702.

NOZZLE CONFIGURATIONS AND DIMENSIONS The Nl nozzle geometry and dimensions for NMP 2 are obtained from References [6] and [7]

and shown in Figure 1 and Figure 2. The Nl nozzle geometry and dimensions for LaSalle Unit 2 are obtained from Reference [8] and shown in Figure 3.

The comparison of dimensions for these two nozzles are summarized in Table 1. The difference between the RPV radius, vessel wall thickness, Nl nozzle bore diameter, and Nl nozzle inside radius is Jess than Yi% between LaSalle and NMP 2.

The only significant difference is the length of the nozzle and nozzle blend radius. The overall nozzle length up to the safe-end-to-nozzle weld differs by 3.75" (about 13%). Since the relevant locations are the nozzle blend region and nozzle-to-shell weld and this difference is at the far end of the nozzle where the wall thickness is more than 3.5 times thinner than the vessel wall, any stress due to piping/nozzle interface loads would be limited to safe end side of the nozzle.

The nozzle blend radius on the NMP 2 Nl nozzle is larger than the blend radius at LaSalle, which will result in lower peak surface stresses for the same applied load. The blend radius only affects surface stress, and has little effect on through wall stress.

Since the dimensions of the NMP 2 Nl nozzles that affect the stresses at the locations of interest are within Yi% of the dimensions of the LaSalle Nl nozzle, the stresses in the relevant locations will be similar to those in the LaSalle Nl nozzle under the same loading conditions MATERIAL CHEMISTRY The NMP 2-Nl nozzle forging material chemistry is obtained from Reference [9] and presented in Table 2. For the nozzle forging, the weight% Ni is 0.76 but the data on weight% Cu is not available. Drop weight testing was performed at -10 °F to qualify for the RTndt of -20 °F.

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February 15, 2017 Report No. 1501576.401.Rl Page 3of17 The weld chemistry for the nozzle-to shell weld is not available for NMP 2.

The material chemistry for LaSalle Units 1 and 2 are also shown in Table 2. Plant specific material chemistry information is not available, so generic data from References [3] and [ 1O]

were used. Reference [ 1O] is the study on the evaluation of chemistry data for BWR fleet vessel nozzle forging material. The weight% Cu used in LaSalle evaluation [5] has the largest observed number in the fleet data. Based on the LaSalle weight% Cu, it is reasonable to conclude that the Nine Mile Point Unit 2 nozzle forging and nozzle to shell weld chemistry are bounded by or similar to the material chemistry used in the LaSalle evaluation.

FLUENCE The fluence for NMP 2 is obtained from References [11, 12, and 13] and summarized in Table 3.

The maximum fluence at the NMP 2-Nl nozzle is l.8lxl0 16 n/cm 2 at 54 effective full power years (EFPY). The fluence for NMP 2 nozzle N6 and lower shell plates are also included in the table for reference. The maximum fluence among all the nozzles and shell plates l.58xl0 18 n/cm 2 , at 54 EPFY.

For LaSalle, the fluence is obtained from References [14] and [15]. The maximum fluence at the shell plates is l.22xl0 18 n/cm 2 at 54 EPFY. Since the fluence at the LaSalle Nl nozzle is not available, the nozzle with the maximum fluence, the LaSalle-N6 nozzle at 5.36xl0 17 n/cm 2 for 54 EFPY, was used for the Nl nozzle. The maximum fluence at any NMP 2 nozzle is 5.34xl0 17 n/cm 2 for 54 EPFY, which is lower than the LaSalle N6 nozzle fluence and well above the NMP 2-Nl nozzle fluence.

The NMP1-N2 nozzle is located on the bottom head (Figure 4) away from the beltline, therefore the fluence level at the NMP1-N2 nozzle remains below lxl0 17 n/cm 2 even at 46 EFPY [21]. In Reference [20], it is stated that ifthe fluence does not exceed the threshold of lxl0 17 n/cm 2 , it does not require the material within the beltline region of the RPV to be considereq for material surveillance program requirements [20].

THERMAL TRANSIENTS The design thermal transients for 40 years for NMP 2 are obtained from References [17] and [23]

and shown in Table 4. The actual number of occurrences from of the start of the plant to December 2002 are obtained from Reference [24], and projected to 60 years as shown in Table 4.

The total projected number of thermal cycles for NMP 2 is 1368 for 60 years.

The total number of thermal transient cycles for LaSalle Unit 1 and 2 are obtained from Reference [5] and also summarized in Table 4. The total number of thermal transient cycles evaluated is 774 for 40 years. In Reference [5], these design occurrences are projected to 60 years for the period of extended operation, which totals 1161 cycles.

The temperature and pressure during the NMP 2 bounding transients of Safety Valve Blowdown, Loss of Feedwater Pumps, and SCRAM/Turbine Generator Trip are presented in Figure 5, Structural Integrity Associates, Inc...

February 15,2017 Report No. 1501576.401.Rl Page 4of17 Figure 6 and Figure 7, respectively [17]. The Loss ofFeedwater Pump transient has a maximum temperature of 567 °F and a minimum temperature of 400 °F. The SCRAM/Turbine Generator Trip has a maximum temperature of 528 °F and a minimum temperature of 400 °F In the LaSalle evaluation [5], the bounding transients are SCRAM/Turbine Generator Trip and Loss ofFeedwater Pumps Isolation Valve Close. From Reference [18], the LaSalle SCRAM/Turbine Generator Trip has a maximum temperature of 573 °F with a minimum temperature of 400 °F. The Scram has a maximum temperature of 528 °F and a minimum temperature of 400°F. The Safety Valve Blowdown is classified as an emergency condition, therefore, it was not included in the LaSalle evaluation. All other transients were analyzed as SCRAM/Turbine Generator Trip except the Loss of Feedwater Pump, as shown in Table 4.

The NMP 2 Loss of Feedwater Pumps Isolation Valves Close transient has a lower maximum temperature, and the SCRAM/Turbine Generator Trip has the same maximum temperature. The minimum temperatures are the same for these two transients between NMP 2 and LaSalle.

The LaSalle Loss of Feedwater Pumps transient is bounded by the NMP Safety Valve Blowdown transient which was not analyzed in LaSalle. As shown in Figure 6, the NMP 2 Safety Valve Blowdown has a larger temperature difference (528°F to 100°F [17] versus LaSalles' 573° to 400°F [I8]) and faster initial rate of change.

Reference [25] determines the effects ofNMP 2's SRV Slowdown transient, as well as the increased total number of thermal transient cycles (Table 4) and concludes that the resulting PoF increases, but remains below the applicable screening criteria.

CONCLUSIONS NMP 2 NJ Nozzle:

A comparative evaluation was performed for the NMP 2 NI nozzle using the results from LaSalle Units I and 2. These two plants are of the same BWR generation, therefore if a comparative evaluation confirms all relevant parameters are essentially the same, then the results and conclusions from the LaSalle evaluation can be used to justify the inspection relief for the RPV nozzles in NMP 2 for the extended license period.

The relevant parameters examined are RPV and nozzle configuration, material chemistry, tluence, and thermal transients. The configurations and dimensions between NMP 2 and LaSalle are essentially the same, except the total length of the nozzle and nozzle blend radius. This difference does not cause significant differences in stress at the nozzle blend radius and nozzle-to-shell weld. The maximum tluence used in the LaSalle evaluation is higher than the tluence levels in the NMP 2 NI nozzles. Based on the available but limited material chemistry data, it's reasonable to conclude the NMP 2 NI nozzle chemistry is similar to or bounded by the LaSalle nozzle chemistry.

The total number of projected thermal transients in NMP 2 are higher than the total number of analyzed design cycles in LaSalle for 60 years, and NMP 2 has a more severe bounding transient.

Structural Integrity Associates, Inc.~

February 15, 2017 Report No. 1501576.401.Rl Page 5of17 These differences are addressed in Reference [25] which shows that even with LaSalle's bounding tluence, the addition ofNMP 2's bounding transient, and increased total number of thermal transients, the PoF (6.85x10-7 for Nl Nozzle blend radius and <3.3x10- 12 for nozzle to shell) remains less than the required screening criteria of 5x 10-6 per Reference [22].

Therefore, NMP 2 is justified to extend their existing inspection relief request over the extended period of operation (60 years).

NMP 1 N2 Nozzle:

For NMPl, the N2 nozzle was evaluated in Reference [2]. The selection of this nozzle was based on the criteria set forth in the safety evaluation of Reference [4] which is based on the vessel and nozzle dimensions, without consideration of other parameters such as tluence and location, and is considered the limiting nozzle for NMPl in Reference [2].

However, for extended operation to 60 years, the additional fluence and thermal cycle counts should be addressed for continued applicability of inspection relief request per Code Case N-702.

Thermal fatigue crack growth due to extended operation to 60 years is not a controlling factor according to Reference [20] and fluence is below lxl0 17 n/cm 2 threshold at 46 EFPY due to the nozzle location away from the beltline. Also, since the bottom head is a semispherical head, the nozzle is subjected to lower stress (i.e. the hoop stress) than a nozzle in the cylinder.

Based on the results in Reference [2], the NMPl N2 nozzles had no failures in the simulations for 40 years of operation (<Ix l 0-9 probability of failure), therefore, it is expected that the PoF for the NMP 1 N2 nozzle remains well below the PoF criterion of (<5x 10-6 for 60 years of operation).

Structural Integrity Associates, Inc. ~

February 15, 2017 Report No. 1501576.401.Rl Page 6of17 Prepared by: Verified by:

vV~~~

Wilson Wong Engineer 2/15/17 Date ~4?r-Associate 2/15/17 Date Reviewed by:

_.--I~ 2/15/17 Terry J.~ Date Senior Associate Approved by:

2/15/17 Wilson Wong Date Engineer Structural Integrity Associates, Inc.'"*

February 15, 2017 Report No. 1501576.401.Rl Page 7of17 REFERENCES

l. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle to Shell Welds,Section XI, Division l," February 20, 2004.
2. 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. 1201005, October, 2010, EPRI PROPRIETARY INFORMATION.
3. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.
4. NRC 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.

5. SI Calculation No. 1400187.302, 'Probability of Failure for LaSalle Unit 2 Nl Nozzle-to Shell Welds and Nozzle Blend Radii Regions," Rev. 2, May 2015.
6. CBI Drawing, VPF # 3516-242, "N l Nozzle Assembly (Recirculation Outlet)", SI File No. 1501576.202.
7. CBI Drawing, VPF # 3516-240, "Recirculation Outlet Nozzle Forging (Nl)," SI File No.

1501576.203.

8. CB&I Drawing No. 72-2046, Revision 5, "Recirculation Outlet Nozzle Nl," SI File No.

1400187.214.

9. Bonney Forge & Foundry, Inc. Material Test Report, S.O. No. 3572-3, January 8, 1974, SI File No. 1501576.215.
10. BWRVIP-173-A, "Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials,' EPRI, Palo Alto, CA 2011, 1022835, SI File Number BWRVIP-173-A. EPRI PROPIETARY INFORMATION.
11. CENG Report PTLR-1, "Pressure and Temperature Limits Report (PTLR)." Revision 02.00, September 2012, SI File No.1501576.210.
12. CENG Report PLTR-2, "Pressure and Temperature Limits Report (PTLR)," Revision 0, June 2014, SI File No. 1501576.209.
13. Email, George Inch (GenCo-Nu)/Wilson Wong (SIA), "NMP2 N6 Nozzle fluence best estimate," March 07, 2016, SI File No. 1501576.2011.
14. Transware Report, "LaSalle County Generating Station Unit l Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 with Projection to 32 and 54 EFPY,' Revision 0, January 2014, SI File No. 1400187.208.
15. Transware Report, "LaSalle County Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 with Projection to 32 and 54 EFPY,' Revision 0, January 2014, SI File No. 1400187.209.

Structural Integrity Associates, Inc."'

February 15, 2017 Report No. 1501576.401.Rl Page 8of17

16. SI Calculation No. FP-NMP2-307 (formerlyNMP-09Q-307) 'Cycle Based Fatigue Development for RPV Locations for NMP-2,' Project No. 1200346 Rev. 4, June 2014.
17. General Electric Drawing No. 762E673, "Reactor Vessel Thermal Cycle," Sheets 1 and 2, Rev. 2, SI File No. NMP-09Q.226.
18. Thermal Cycle Diagrams
a. General Electric Drawing Number 158B8136, Sheet 1, Revision 6, "Reactor Vessel Nozzle Thermal Cycles," SI File No. 1400187.207
b. General Electric Drawing Number 731 E776, Sheets 1 and 2, Revision 3, "Reactor Vessel Thermal Cycles," LaSalle Unit 1, SI File No. 1400187.205
c. General Electric Drawing Number 761E581, Sheets 1 and 2, Revision 1, "Reactor Vessel Thermal Cycles," LaSalle Unit 2, SI File No. 1400187.206.
19. CBIN VPF 3516-50-2, "Thermal Analysis of Recirculation Inlet Nozzle, SI File No.

NMP-09Q-215.

20. EPRI Letter, 2012-138, "BWRVIP Support of ASME Code Case N-702 Inservice Inspection Relief," August 31, 2012, SI File No. 1300341.213.
21. E-mail, Inch, G. B. (GenCo-Nuc)/Wong, W. (SIA), "RE: Fatigue cycles NMPl," March 10,2016, SIFileNo.1501576.103.
22. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limits in the PTS Rule (10CFR 50.61), NUREG-1806, Vol. 1, August, 2007.
23. Nine Mile Point Unit 2 UFSAR, Section 3.9B, SI File No. NMP-09Q.232.
24. Constellation Energy, "Application for Renewed Operating Licenses, Technical Information, Nine Mile Point Nuclear Station Units 1, and 2," ADAMS Accession No. ML041490223.
25. SI Calculation No. 1501576.301, 'Effects ofNMP2 SRV Blowdown Transient on Probability of Failure," Rev. 0, November 2016.

Structural Integrity Associates, Inc.""

February 15, 2017 Report No. 1501576.40 l.Rl Page 9of17 Table 1 Recirculation Outlet Nozzle (Nl) Comparison Nine Mile Point LaSalle Unit2 Unit 1and2 Vessel Radius 126 11116"[7] 126.5" f81 Vessel Thickness 6 9/16" [7] 6 9/16" [81 Nozzle Inside Radius 11 1116" f71 11 1/16" [81 Nozzle Bore Thickness 9.03" [7] 9.00" [8]

Nozzle Blend Radius 4 21 /32" f71 4 7/32" [81 Nozzle Outside Radius 4 5/8" [7] 4 W' [8]

Nozzle Length 26.91" [6] 30 21/32" [8]

Safe end Outside Radius 12.56" [6] 12.375" [8]

Table 2 Weld Chemistry Nine Mile Point Unit 2 LaSalle Units 1 and 2 Forging Weld Forging Weld

%Cu -- -- 0.09189 0.26

%Ni 0.76 -- 0.78 1.2 Initial RTndt °F -20 -- 24.13 -20 Table 3 Fluence at end of 60 years Fluence (E> 1 Me V) n/cmz NMP LaSalle Unit 1 Unit2 Unit 1 Unit2 EPFY 36 46 32 54 32 54 32 54 Nl -- <lx10 17 9.97x10 15 l.81x10 16 N6 -- -- 2.42xl0 17 4.33x10 17 2.60xl0 17 5.36xl0 17 3.05xl0 17 5.36x10 17 Lower Shell Plates 1.42xl0 18 -- 9.37x10 17 l.58x10 18 2.34xl0 17 l.06x10 18 6.59xl0 17 l.06x10 18 NzMax -- -- 3.16x10 17 5.34x10 17 2.60xl0 17 3.75x10 17 3.05x10 17 5.36xl0 17 Note : N6 is RHR-LPCI nozzle Structural Integrity Associates, Inc.~

February 15, 2017 Report No. 1501576.401.Rl Page 10of17 Table 4 Thermal Transient Cycle Counts NMP2 LaSalle Design Cycles to Projected Analysis Analysis Design (40 years) Dec 2002 to 60 (60 years) (60 years)

(40 years)

[17, 23] [24] years [25] [5]

Design Hydrostatic Test 130 20 80 130 StartUp 117 120 99 396 Natural Circulation Start Up 3 Loss of Feed water Heater 1 -- -- -- 80 Partial Feedwater heater bypass 70 4 16 1288 1116+15 Scram 140 82 328+36 180 Scram/Turbine Generator Trip 40 Shutdown Vessel Flooding 111 98 392 111 Unbolt 123 10 40 123 Loss of Feedwater Pump2 10 9 36+36 72 10 15+15 Single SRV Blowdown 8 2 8 8 -- --

Total Number of Cycles 752 1368 1368 754 1161 I )Defined as testing condition for NMP 2 in Reference (17), and is therefore not considered in this evaluation.

2)Per Reference [5 and 17), The Loss ofFeedwater Pump transient contains 3 internal cycles, one of which is bounded by the SCRAM/Turbine Generator Trip transient.

Structural Integrity Associates, Inc... .

February 15, 2017 Report No. 1501576.40 I.RI Page 11 of 17

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February 15, 2017 Report No. 1501576.401.Rl Page 14of17 Instrument Nozzie Vent Nozzle 4-5/16 11 64 - b-l/4 11 Stud'S Eleva. tion ~ 1 -ll 3/lj. 11 Instrument Nozzle Steam Outlet Nozzie Instrument

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February 15, 2017 Report No. 1501576.401.Rl Page 16of17

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Attachment 4 Structural Integrity Associates, Inc. Report No. 1501576.301, "Effects of NMP 2 SRV Slowdown Transient on Probability of Failure"

~Structural Integrity Associates, Inc. File No.: 1501576.301 Project No.: 1501576 CALCULATION PACKAGE Quality Program: ~Nuclear 0 Commercial PROJECT NAME:

Nine Mile Point N702 Evaluation CONTRACT NO.:

00517760 Release 00085 and 00107 CLIENT: PLANT:

Exelon Generation Company LLC Nine Mile Point Unit 2 CALCULATION TITLE:

Effects ofNMP 2 SRV Blowdown Transient on Probability of Failure Project Manager Preparer(s) &

Document Affected Revision Description Approval Checker(s)

Revision Pages Si2nature & Date Si2natures & Date 0 l - 14 Initial Issue Preparer:

---1~~ ivV~~~

Terry J. Herrmann Wilson Wong 1119/16 11/9/16 Checker:

~~ Minji Fong 11/9/ 16 Page l of 14 F0306-0I R2

Structural Integrity Associates, lnc.l!J Table of Contents 1.0 OBJECTIVE .................................................................................................................. 4 2.0 METHODOLOGY ........................................................................................................ 4 3.0 DESIGN INPUTS .......................................................................................................... 4 4.0 FINITE ELEMENT MODEL ........................................................................................ 5 4.1 Thermal Analysis ............................................................................................... 5 4.2 Thermal Stress Analysis .................................................................................... 5 5.0 RESULTS ...................................................................................................................... 5 5.1 Thermal Stress Analysis .................................................................................... 5 5.2 Probability ofFailure ......................................................................................... 5

6.0 CONCLUSION

.............................................................................................................. 6

7.0 REFERENCES

.............................................................................................................. 7 File No.: 1501576.301 Page 2of14 Revision: 0 F0306-0IR2

Structural Integrity Associates, Inc:~

List of Tables Table 1: Thermal Transient Cycle Counts ................................................................................. 8 Table 2: NMP 2 SRV Blowdown Transient. ............................................................................. 8 List of Figures Figure 1: Components Included in the Finite Element Model ................................................. 9 Figure 2: 3-D Finite Element Model Mesh for Analyses ....................................................... 10 Figure 3: Applied Mechanical Boundary Conditions for Thermal Stress Analyses .............. 11 Figure 4: Temperature Contour for SRV Blowdown at Time=600 sec .................................. 12 Figure 5: Stress Intensity Plot for SRV Blowdown at Time=600 sec .................................... 13 Figure 6: Path Locations for Through-Wall Stress Extractions .............................................. 14 File No.: 1501576.301 Page 3of14 Revision: 0 F0306-0IR2

Structural Integrity Associates, Inc:'!>

1.0 OBJECTIVE Nine Mile Point Unit 2 (NMP 2) intends to apply the methods of Code Case N-702 [1] using guidance from BWRVlP-108 [2] and BWRVlP-241 [3] to extend an existing inspection relief request. NMP 2 is a BWR-5 design much like LaSalle Units l and 2. Thus, the Reactor Pressure Vessel (RPV) configurations, material specification, and operating conditions are very similar.

However, the NMP 2 Safety Relief Valve (SRV) Blowdown transient is considered an Upset condition instead of an Emergency condition as defined in the design basis thennal cycle diagram for LaSalle.

The objective of this calculation is to detennine the stresses due the SRV Blowdown transient and resulting change in probability of failure due to the addition of the SRV Blowdown transient. Reference

[11] provides a through-wall cracking frequency criterion per reactor year. The probability of failure expressed in this calculation is equivalent to the through-wall cracking frequency. Therefore, the reported probability of failure can be taken as the through-wall cracking frequency.

2.0 METHODOLOGY The NMP 2 SRV Slowdown transient is applied to the finite element model of the Nl nozzle from the LaSalle Unit 2 N702 evaluation [4] with no changes to the model or material properties. A comparison of the LaSalle Nl nozzle with that ofNMP 2 was perfonned in Reference [6], which established that the LaSalle model is appropriate for use for the NMP 2 evaluation.

The stresses from the finite element run are extracted along paths representing the locations of a nozzle blend radius and nozzle-to-vessel weld flaw. These stresses are then input into the VIPERNOZ input files as an additional transient, with existing transient cycle counts increased per Table 1, with no other changes to the file. Note that the VIPERNOZ computer program is the same program used in the SWRVIP-108NP report that was accepted by the NRC in their SER [2].

3.0 DESIGN INPUTS The finite element model geometry and material properties are unchanged from Reference [4].

The design thermal transients for 40 years for NMP 2 are obtained from References [7] and [8] and shown in Table l. The actual number of occurrences from of the start of the plant in 1987 through December 2002 are obtained from Reference [9], and projected to 60 years as shown in Table 1. The total projected number of thennal cycles for NMP 2 is 1368 for 60 years. The thennal transient definition for the NMP 2 SRV Slowdown obtained from Reference [7] is tabulated in Table 2.

The total number of thennal transient cycles for LaSalle Unit 1 and 2 are obtained from Reference [5]

and also summarized in Table 1. The total number of thennal transient cycles evaluated is 774 for 40 years. In Reference [5], these design occurrences are projected to 60 years for the period of extended operation, which totals 1161 cycles.

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The VIPERNOZ input file with all variables as described in Reference [5] is modified to include the SRV Blowdown transient and increased number of cycles for existing transients as shown in Table 1.

4.0 FINITE ELEMENT MODEL A three-dimensional (3-0) finite element model was constructed in Reference [4] using the ANSYS finite element program [10]. It was developed as a symmetric quarter model and includes a local portion of the reactor pressure vessel, the Nl nozzle-to-vessel weld, the Nl nozzle, and a portion of the attached safe end, as shown in Figure 1. The NI nozzle-to-safe end weld was not modeled because it is not near the region of interest. The model is meshed with the SOLID45 element type from the ANSYS library of elements, for which the thermal equivalent is SOLID70. The mesh is depicted in Figure 2.

4.1 Thermal Analysis Bulk fluid temperatures and heat transfer coefficients are applied to the inside surface nodes of the model. The nozzle and inside surface of the vessel are both subjected to a conservatively high convective heat transfer coefficient (HTC) of I0,000 Btu/hr-ft2-°F, while the entirety of the outside surface is assumed to be perfectly insulated with no heat transfer.

4.2 Thermal Stress Analysis 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 3 shows a representative plot of the mechanical boundary conditions applied for the thermal transient stress analyses.

5.0 RESULTS 5.1 Thermal Stress Analysis A representative temperature contour and total stress intensity contour plot for the SRV Blowdown transient is shown in Figure 4 and Figure 5, respectively.

Four through-wall stress paths, two each at 0° and 90°, are defined within the region of the NI nozzle blend radius and nozzle-to-vessel weld, as shown in Figure 6. Since the model is symmetric, these two paths also represent the stress at I80° and 270°, respectively.

5.2 Probability of Failure The thermal stresses calculated for the NMP 2 SRV Blowdown transient on LaSalle's Nl nozzle model from Reference [4] are inserted into the VIPERNOZ input files as an additional transient. Cycle counts for existing transients in the input file from Reference [5] are also slightly increased to match the predicted 60 year numbers for NMP 2 as shown in Table 1. The resulting probability of failure due to low temperature over-pressure (L TOP) event is 6.85x 10-7 and <3.33x 10- 12 per reactor year for the NMP2 Nl nozzle blend radii and nozzle-to-shell-weld, respectively.

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6.0 CONCLUSION

The stresses due the SRV Blowdown transient and resulting change in probability of failure due to the addition of the NMP2 SRV Blowdown transient is determined. The probability of failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii remain below the criteria of 5x1 o-6 per year [ 11]

even with consideration of the severe NMP2 SRV Blowdown transient and LaSalle's higher fluence.

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7.0 REFERENCES

l. 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. NRC 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 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.
4. SI Calculation No. 1400187.301, 'Finite Element Model Development and Thermal Mechanical Stress Analyses for the Unit 2 Nl Nozzle," Rev. l, Feb 2015.
5. SI Calculation No. 1400187.302, 'Probability of Failure for LaSalle Unit 2 Nl Nozzle-to Shell Welds and Nozzle Blend Radius Regions," Rev. 2, May 2015.
6. SI Report No. 1501576.401, 'Evaluation of Nine Mile Point Units land 2 using Code Case N-702,'

Rev. 0, November 2016.

7. General Electric Drawing No. 762E673, "Reactor Vessel Thermal Cycles," Sheets l and 2, Rev. 2, SI File No. NMP-09Q.226.
8. Nine Mile Point Unit 2 UFSAR, Section 3.9B, SI File No. NMP-09Q.232.
9. Constellation Energy, "Application for Renewed Operating Licenses, Technical Information, Nine Mile Point Nuclear Station Units l, and 2," ADAMS Accession No. ML04 l 490223.
10. ANSYS Mechanical APDL, Release 14.5 (w/ Service Pack l UP20120918), ANSYS, Inc.,

September 2012.

11. 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.

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Table 1: Thermal Transient Cycle Counts NMP2 LaSalle Design Cycles to Projected Current Design Analysis Transient (40 years) Dec 2002 to 60 Analysis (40 years) (60 years)

[7, 8] [9] years (60 years) [5] [5]

Design Hydrostatic Test 130 20 80 130 StartUp 117 120 99 396 Natural Circulation Start Up 3 Loss of Feedwater Heater 1 -- -- -- 80 Partial F eedwater heater bypass 70 4 16 1288 1116+ 15 Scram 140 82 328+36 180 Scram/Turbine Generator Trip 40 Shutdown Vessel Flooding 111 98 392 111 Unbolt 123 10 40 123 Loss of Feedwater Pump2 10 9 36+36 72 10 15+15 Single SRV Blowdown 8 2 8 8 -- --

Total Number of Cycles 752 1368 1368 754 1161 I )Defined as testing condition for NMP2 in Reference (7], and is therefore not considered in this evaluation.

2)Per Reference (4 and 7], The Loss ofFeedwater Pump transient contains 3 internal cycles, one of which is bounded by the SCRAM/Turbine Generator Trip transient.

Table 2: NMP 2 SRV Blowdown Transient Temp, h, Transient Time, sec OF Btu/hr-ft2-°F 0 528 10000 SRV Blowdown 600 375 10000 10500 100 10000 Note: A total of3,660 seconds (in addition to the time shown) was added to the end of the transients to capture transient stress lag and to achieve a final transient steady state condition.

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Nozzle to Vessel Weld LGS - Vipemoz M::x:lel - 30 Figure 1: Components Included in the Finite Element Model File No.: 1501576.301 Page 9 of 14 Revision: 0 F0306-0l R2

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LGS - Vipernoz M:x:lel - 30 Figure 2: 3-D Finite Element Model Mesh for Analyses File No.: 1501576.301 Page 10ofl4 Revision: 0 F0306-0IR2

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r LGS - Vipernoz M:x:lel - 3D Figure 3: Applied Mechanical Boundary Conditions for Thermal Stress Analyses File No.: 1501576.301 Page 11 of 14 Revision: 0 F0306-0I R2

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])0).0.L SOIDTICN S'.IEP-2 SOB =25 Tn-E-600 TEMP St-N -375.67 SMX -526.149 375.67 409.11 509.429 392 .39 425.83 492.709 526.149 LGS - Vipernoz M:xiel - 30 Figure 4: Temperature Contour for SRV Blowdown at Time=600 sec.

(Units for temperature in terms of °F)

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l\XIll\L SO!mICN STEP-26 SOB =l TrnEl-600 SINT (A\il'.;)

RSYS-0 lliK - *4827 54 51-N -598.081 SMX ~26546. 7 598.081 3481.26 6364.44 17897.1 20780.3 23663.5 26546.7 LGS - Vipernoz M:::del - 30 Figure 5: Stress Intensity Plot for SRV Blowdown at Time=600 sec.

(Units for stress intensity in terms of psi)

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Structural Integrity Associates, lnc.'t, EIEMNTS Ml\T NUM PATH 0° Face 90° Face LGS - Vipernoz M:x:lel - 30 Figure 6: Path Locations for Through-Wall Stress Extractions File No.: 1501576.301 Page 14ofl4 Revision: 0 F0306-0IR2

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