L-MT-23-010, Subsequent License Renewal Application Supplement 1

Subsequent License Renewal Application Supplement 1

ML23094A136
Person / Time
Site:	Monticello
Issue date:	04/03/2023
From:	Domingos C Northern States Power Company, Minnesota, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-MT-23-010
Download: ML23094A136 (1)

Text

fl Xcel Energy 2807 West County Road 75 Monticello, MN 55362 April 3, 2023 L-MT-23-010 10 CFR 54.17 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Subsequent License Renewal Application Supplement 1

References:

1) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, Monticello Nuclear Generating Plant Docket No. 50-263, Renewal License Number DPR-22 Application for Subsequent Renewal Operating License dated January 9, 2023, ML23009A353.

Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy hereafter "NSPM", is submitting a supplement to the Subsequent License Renewal Application, listed in Reference 1. This supplement provides clarifying information regarding updates to Tables 4.2.3-1 and 4.2.3-2 and an updated reference. The enclosure to this letter provides that information.

In the enclosure, changes are described along with the affected section(s) and page number(s) of the docketed SLRA (Reference 1) where the changes are to apply. For clarity, revisions to the SLRA are provided with inserted text by bold red underline.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

Document Control Desk L-MT-23-010 Page 2 I declare under penalty of perjury, that the foregoing is true and correct.

r'l L( . ~,. ---~

Executed on April , 2023.

~ . Domingos -*-*.

Site Vice President, Monticello and Prairie Island Nuclear Generating Plants Northern States Power Company - Minnesota Enclosure (1) cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

ENCLOSURE 1 REVISED SLRA SECTIONS: 4.2.3, TABLES 4.2.3-1 AND 4.2.3-2, AND 4.7 5 Pages follow

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-010 Page 1 of 5 Non-Proprietary Information in ART Tables Update the ART Tables with EPRI Non-Proprietary Information Affected SLRA Sections: 4.2.3, Tables 4.2.3-1 and 4.2.3-2, and 4.7 SLRA Page Numbers: 4.2-16, 4.2-18, 4.2-19, and 4.7-3 Description of Change:

Update the ART tables for the Lower-Intermediate Shell Plates (Course 2) in Table 4.2.3-1, 0T ART Values for MNGP RPV Components at 72 EFPY, and Table 4.2.3-2, 1/4T ART Values for MNGP RPV Components at 72 EFPY with EPRI non-proprietary information. Six values for components I-14 and I-15 (two for % Cu, % Ni, and CF in each table) were originally identified as EPRI proprietary but none of these values are actually proprietary.

In addition, Structural Integrity Associates Calculation No. 2100300.302, Revision 3 is added to TLAA Reference Section 4.7 and referenced in SLRA Section 4.2.3, as the basis for the ART evaluation.

References:

1. Structural Integrity Associates Calculation No. 2100300.302, Revision 3, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts, March 14, 2023.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-010 Page 2 of 5 SLRA Section 4.2.3 on page 4.2-16 is revised as follows:

4.2.3 Adjusted Reference Temperature (ART) for RPV Materials Due to Neutron Embrittlement TLAA Description The ART of the limiting beltline or extended beltline material is used to adjust the beltline P-T limit curves to account for irradiation effects. RG 1.99 provides the methodology for determining the ART of the limiting material. The initial nil ductility reference temperature, RTNDT, is the temperature at which a non-irradiated metal (ferritic steel) changes in fracture characteristics from ductile to brittle behavior.

RG 1.99 requires calculation of ART and Reference Temperature Shift (RTNDT) values. The ART values are then used to determine the local fracture toughness of the RPV wall and P-T limits, according to ASME Code,Section XI, Non-mandatory Appendix G Evaluations. Neutron irradiation increases the RTNDT beyond its initial value.

10 CFR Part 50, Appendix G defines the fracture toughness requirements for the life of the vessel. The shift in the initial RTNDT (RTNDT) is evaluated as the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. This increase (RTNDT) means that higher temperatures are required for the material to continue to act in a ductile manner. The ART is defined as: Initial RTNDT +

(RTNDT) + Margin. Since the RTNDT values are a function of neutron fluence, these ART calculations meet the criteria of 10 CFR 54.3(a) and have been identified as TLAAs requiring evaluation for the 80-year SPEO.

TLAA Evaluation In 2011, a calculation was developed of the ART and RTNDT values developed for all MNGP plates, welds and nozzles exposed to fluence levels greater than 1.0x1017 n/cm2. This calculation was based on the updated fluence calculations provided at that time, including the increase in neutron flux due to EPU. The ART and RTNDT values were calculated at 36, 40, and 54 EFPY. The reported values for 54 EFPY are intended to be applicable through the end of MNGPs current extended operating period (i.e., 60 years).

Tables 4.2.3-1 and 4.2.3-2, below, provide the surface (0T) and 1/4T fluence and fluence factor (FF) values for MNGP at 72 EFPY and the ART calculation results for 72 EFPY (Reference 4.7.38).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-010 Page 3 of 5 SLRA Tables 4.2.3-1 and 4.2.3-2 on pages 4.2-18 and 4 .2-19 are revised as follows:

Table 4.2.3-1 OT ART Values for MNGP RPV Components at 72 EFPY 72EFPY tnttial 72 EFPY Compor1ent

%Ni OT Fluence ARTNDT Heat Lot %Cu CF RTNDT Fluence 0 1 (°F) aA (*F) OT ART No. Fac;:tor f ("F)

(*F) (*F)

(n/cm 2 \

Lower Shell Plates (Course 1) 1-16 A0946-1 N/A 0.14 0.56 98 27 1.06E+18 0.429 42.1 0 17.0 90.1 1-17 C219J..1 NIA 0.17 0.5 119 0 1.06E+18 0.429 50.8 0 17.0 107.8 Lower-Intermediate Shell Plates (Course 2) 1-14 C2220-1 N/A 0.16 0.64 180 27 5.94E+18 0.854 153.8 0 8 .5 197.8 1-1 5 C2220-2 N/A 0.16 0.64 180 27 5.94E+18 0.854 153.8 0 8.5 197.8 Upper/Int Shell Plates (Course JJ 1-12 C2089-1 N/A 0.35 0.5 200 0 3.23E+17 0.229 45.7 0 17.0 79.7 1-13 C2613-1 N/A 0.35 0.49 198 27 3.23E+17 0.229 45.5 0 17.0 106.5 Lower Shell (Course 1) Axial Welds VLAA-1 &

- E8018N 0.1 0.99 135 -65.6 2.35E+18 0.609 82.1 12.7 28.0 78.0 VLAA-2 Lower-Intermediate Shell (Course 2) Axial Welds:

VLBA-1 &

VLBA-2 - E8018N 0.1 0.99 135 -65.6 2.10E+18 0 .581 78.4 12.7 28.0 74.3 Upper/Int Shell (Course 3) Axial Welds:

VLCB-1 &

VLCB-2 - E8018N 0.1 0.99 135 -65.6 2.12E+17 0 .178 24.1 12.7 12.0 -6.6 Circumferential Welds VCBA-2&

VCBA-3 - E8018N 0.1 0.99 135 -65..6 3.79E+18 0.732 98.7 0 28.0 89.1 N2 Nozzle N2 Nozzle E21WJ N/A 0 .18 0.86 142 40 7.08E +17 0 .351 49.9 0 17.0 123.9

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-010 Page 4 of 5 Table 4.2.3-2 1/4T ART Values for MNGP RPV Components at 72 EFPY 72EFPY fnlllal 72 EFPY Component 1/4T Ftuence MTNoT Heat lot %Cu %NI CF RTHOT Fluence o, (*F) c:r4' (4F) 114T ART No. Factorf (*F)

(*F) t*F)

(nlcm 2 )

Lower Shell Plates (Course 1) 1-16 A0946-1 NIA 0.14 0.56 98 27 2.80E+ 18 0.653 64.1 0 17.0 125.1 1-17 C2193-1 NIA 0.17 0.5 119 0 2.80E+ 18 0.653 n .3 0 17.0 111 .3 Lower-lntermedJate Shell Plates (Course 2) 1-14 C2220-1 NIA 0.16 0.64 180 27 4.38E+18 0.770 138.7 0 8.5 182.7 1-15 C2220-2 NIA 0.16 0.64 180 27 4.38E+18 o.no 138.7 0 8.5 182.7 Upper/Int Shell Plates (Course 3) 1-12 C2089-1 NIA 0.35 0.5 200 0 2.38E+17 0.191 38.2 0 17.0 72.2

.. , 3 C2613-1 NIA 0.35 0.49 198 27 2.38E+17 0.191 37.9 0 17.0 98.9 Lower Shell (Course 1) Axlal Welda Vl.M-1 &

vt.AA2 - E8018N 0.1 0.99 135 --65.6 1.73E+18 0.535 72.2 12.7 28.0 68.1 Lower-Intermediate Shell (Course 2) Axlal Welds~

VLBA-1 &

\ll.BA-2

- E8018N 0.1 0.99 135 -65.6 1.55E+18 0.510 68.8 12.7 28.0 64.7 Upper/Int Shell (Course 3) Aldal Welds:

VLCB-1 & . E8018N 0.1 0.99 135 -65.6 1.56E+17 0.147 19.8 12.7 9.9 -13.5 VLCB-2 Circumferential Welds VCBA-2&

VCBA-3 - E8018N 0.1 0.99 135 -65.6 2.80E+ 18 0.653 88.0 0 28.0 78.4 N2Noule N2Nozzle E21\/W NIA 0.18 0.86 142 40 5.23E+17 0 .300 42.6 0 17.0 116.6

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-010 Page 5 of 5 SLRA Section 4.7 on page 4.7-3 is revised as follows:

4.7.30 MRP-235, Revision 1, Materials Reliability Program: Fatigue Management Handbook, EPRI Palo Alto, CA, 2008.

4.7.31 NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components (ADAMS Accession No. ML031480219).

4.7.32 NUREG/CR-6909, Revision 1, Effect of LWR Water Environments on the Fatigue Life of Reactor Materials Final Report, May 2018 (ADAMS Accession No. ML16319A004).

4.7.33 MNGP, Plant Unique Analysis Report, Volumes 2-5, Revision 1.

4.7.34 EOCI-61, Electric Overhead Crane Institute, Specification for Electric Overhead Traveling Cranes, 1961.

4.7.35 CMAA-70, Crane Manufacturers Association of America, Specifications for Electric Overhead Traveling Cranes, Revision 1975.

4.7.36 NSP (D M Musolf) letter to the NRC, Travel Path for Spent Fuel Shipping Cask, dated October 24, 1984.

4.7.37 MNGP, License Amendment Request: Extended Power Uprate (TAC MD9990),

November 2008 (ADAMS Accession No. ML083230111).

4.7.38 Structural Integrity Associates Calculation No. 2100300.302, Revision 3, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts, March 14, 2023.