L-MT-23-047, License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data

License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data

ML23363A174
Person / Time
Site:	Monticello
Issue date:	12/29/2023
From:	Scott S Northern States Power Company, Minnesota, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-MT-23-047
Download: ML23363A174 (1)

Text

fl Xcel Energy 2807 West County Road 75 Monticello, MN 55089 December 2, 2023 L-MT-23-047 10 CFR 50.90 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 License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests a license amendment to replace the current neutron fluence methodology with a newer methodology, and to revise the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). The TS change updates Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to reflect the current version (Revision 1) of the Structural Integrity Associates (SIA) methodology report SIR-05-044-A, Pressure Temperature Limits Report Methodology for Boiling Water Reactors.

The current MNGP PTLR hydrostatic pressure and leak test curve has minimal margin to the 212°F operating restriction for reactor pressure vessel (RPV) testing. To gain additional operational margin NSPM proposes to revise the PTLR. Two substantial changes are proposed for this PTLR revision: first, replacement of the current neutron fluence methodology with the TransWare Enterprises, Inc., Radiation Analysis Modeling Application as the licensing basis methodology, and second, incorporation of new MNGP plant-specific surveillance capsule data. provides a description of the proposed changes and includes the technical evaluation and associated no significant hazards determination and environmental evaluation. to the enclosure provides the existing TS page marked-up to show the proposed change. Attachment 2 to the enclosure provides the retyped TS page. provides a copy of the revised MNGP PTLR report. Enclosure 3 provides a copy of the calculation for the adjusted reference temperatures and reference temperature shifts for the RPV components. Enclosure 4 provides a copy of the calculation for generating the MNGP PTLR curves.

Document Control Desk L-MT-23-047 Page 2 of 2 In accordance with 10 CFR 50.91, Notice for public comment; State consultation paragraph (b), NSPM is notifying the State of Minnesota by providing a copy of this application, with this enclosure and attachments, to the State of Minnesota designated official.

NSPM requests issuance of this proposed license amendment within twelve months following completion of NRC acceptance review.

If there are any questions or if additional information is needed, please contact Mr. Richard Loeffler at (612) 342-8981 or Rick.A.Loeffler@xcelenergy.com.

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

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on December __, 29 2023.

Digitally signed by Sara L.

Scott Sara L. Scott Date: 2023.12.29 13:50:53 for

-06'00' Shawn Hafen Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures / Attachments cc: Administrator, Region III, US NRC Project Manager, Monticello, US NRC Resident Inspector, Monticello, US NRC State of Minnesota

L-MT-23-047 NSPM Page 1 of 2 LICENSE AMENDMENT REQUEST REVISION TO THE MNGP PRESSURE TEMPERATURE LIMITS REPORT TO CHANGE THE NEUTRON FLUENCE METHODOLOGY AND INCORPORATE NEW SURVEILLANCE CAPSULE DATA TABLE OF CONTENTS Enclosure (Encl.) 1.0 1.0

SUMMARY

DESCRIPTION .................................................................................. 1 2.0 DETAILED DESCRIPTION................................................................................... 1 2.1 Background ..................................................................................................... 1 2.2 System Design and Operation......................................................................... 2 2.3 Current Technical Specification Requirements................................................ 2 2.4 Reason for the Proposed Changes ................................................................. 3 2.5 Description of the Proposed Licensing Basis Changes ................................... 3

3.0 TECHNICAL EVALUATION

.................................................................................. 3 3.1 General Discussion ......................................................................................... 3 3.2 Development of the Revised P-T Limit Curves - Generic Letter (GL) 96-03 Considerations................................................................................................. 4 3.3 Neutron Fluence Determination....................................................................... 4 3.4 Reactor Vessel Material Surveillance Program ............................................... 7 3.5 Chemistry and Adjusted Reference Temperature............................................ 8 3.6 MNGP Feedwater and Recirculation Inlet Nozzles Finite Element Analyses .. 8 3.7 Pressure-Temperature Limit Curves Generation for 72 EFPY ........................ 8

4.0 REGULATORY EVALUATION

............................................................................. 9 4.1 Applicable Regulatory Requirements/Criteria .................................................. 9 4.2 Precedent ......................................................................................................12 4.3 No Significant Hazards Consideration Analysis.............................................12 4.4 Conclusions...................................................................................................15

5.0 ENVIRONMENTAL CONSIDERATION

..............................................................15

6.0 REFERENCES

16

L-MT-23-047 Page 2 of 2 ATTACHMENTS Att. 1 Technical Specification Pages (Markup)

Att. 2 Technical Specification Pages (Retyped)

Encl. 2.0 Xcel Energy Corporation, Monticello Nuclear Generating Plant (MNGP),

Pressure and Temperature Limits Report (PTLR) Up to 72 Effective Full-Power Years (EFPY), Revision 2 Encl. 3.0 Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts (NSPM Calculation No.22-035) (SI No. 2100300.302, Revision 4)

Encl. 4.0 Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY (NSPM Calculation No.23-012) (SI No. 2200284.303, Revision 0)

L-MT-23-047 NSPM LICENSE AMENDMENT REQUEST REVISION TO THE MNGP PRESSURE TEMPERATURE LIMITS REPORT TO CHANGE THE NEUTRON FLUENCE METHODOLOGY AND INCORPORATE NEW SURVEILLANCE CAPSULE DATA 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), is submitting a license amendment request (LAR). This LAR replaces the current neutron fluence methodology with a newer methodology, described further below, and revises the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP) to reflect the most recent version of a Structural Integrity Associates (SIA), Inc., licensing topical report (LTR) for developing a Pressure Temperature Limits Report (PTLR). The specific proposed change to Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), updates the methodology listed under item 5.6.5.b.1 from the 2007 (Revision 0) to the current, 2013 version (Revision 1), of the SIA LTR SIR-05-044-A, Pressure Temperature Limits Report Methodology for Boiling Water Reactors (Reference 1).

The current MNGP PTLR hydrostatic pressure and leak test curve has minimal margin to the 212°F operating restriction for reactor pressure vessel (RPV) testing. To gain additional operating margin for the remainder of the current renewed operating license period, NSPM proposes to revise the PTLR. Two substantial changes are included in this proposed PTLR revision. First, the current General Electric neutron fluence methodology is replaced by the TransWare Enterprises, Inc., (hereafter TransWare) Radiation Analysis Modeling Application (RAMA) as the licensing basis methodology to estimate RPV fluence. Second, results from evaluation of the MNGP 120-degree surveillance capsule removed during the spring 2021 MNGP Refueling Outage (RFO) are incorporated into the supporting analyses.

2.0 DETAILED DESCRIPTION

2.1 Background

In February 2013, through Amendment No. 172 (Reference 2) NSPM was approved to revise the MNGP TS in accordance with the guidance of NRC Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, (Reference 3) and Technical Specification Task Force (TSTF) - TSTF-419, Revise PTLR Definition and References in ISTS [Improved Standard Technical Specifications]

5.6.6, RCS PTLR (Reference 4) to revise and relocate the MNGP Pressure Temperature (P-T) limit curves to a PTLR.

Page 1 of 18

L-MT-23-047 NSPM The 2013 amendment revised the P-T limits based on the methodology documented in Revision 0 of the SIA LTR SIR-05-044-A (Reference 5). The fast neutron fluence calculations supporting that amendment were performed in accordance with the established General Electric calculational methodology (Reference 6). The PTLR (currently Revision 1) is applicable for the MNGP through the end of the license renewal period, i.e., up to 54 effective full power years (EFPY).

2.2 System Design and Operation The Limiting Condition for Operation (LCO) of Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits, establishes operating limits that provide margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). 10 CFR 50, Appendix G, Fracture Toughness Requirements, specifies material fracture toughness requirements for ferritic materials of the RCPB of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences (AOOs) and during system hydrostatic tests. These P-T limits are instituted through this specification which directs reference to the PTLR (and the curves and associated tables therein).

Each P-T limit curve defines an acceptable region for plant operation. The usual use of the curves is for operational guidance during heatup or cooldown operations and during AOOs -

with the reactor being in a critical condition when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable operating region. The curves also provide guidance during certain other pressure testing conditions (i.e., inservice leak rate testing and / or hydrostatic testing).

Due to the effects of neutron irradiation embrittlement accumulated by the reactor, the P-T limit curves contained in plant TSs are updated periodically to ensure that the limit curves are always valid beyond the EFPYs that the plant has accumulated.

2.3 Current Technical Specification Requirements Specification 3.4.9 provides the reactor coolant system pressure, temperature, heatup, and cooldown rates, and the recirculation pumps starting temperature requirements be maintained within limits through their specification within the PTLR and provides the Conditions, Required Actions, and Completion Times that must be met in order to maintain the required safety margins. The Conditions, Required Actions, and Completion Times remain unchanged by this proposed LAR. The proposed change revises Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), which lists under item 5.6.5.b the 2007 (Revision 0) version of the SIA LTR SIR-05-044-A as the NRC approved PTLR methodology basis.

Page 2 of 18

L-MT-23-047 NSPM 2.4 Reason for the Proposed Changes The current MNGP PTLR hydrostatic pressure and leak test curve has minimal margin to the 212°F operating restriction for RPV testing. The 120-degree surveillance capsule was removed during the spring 2021 MNGP RFO. The development of the new RAMA neutron fluence projections for the RPV combined with the surveillance capsule results support development of a revised PTLR that exhibits increased operational margin for the hydrostatic pressure and leak test curve, which could be used for the remainder of the current license renewal period.

2.5 Description of the Proposed Licensing Basis Changes The licensing basis changes proposed in this LAR update the neutron fluence projections for the RPV EDVHGRQQHZVXUYHLOODQFHFDSVXOHGDWDDQGQHZQHXWURQIOXHQFHSURMHFWLRQVZLWK

WKH5$0$PHWKRGRORJ\7KH76DGPLQLVWUDWLYHVSHFLILFDWLRQLHSpecification 5.6.5, governing development and approval of the PTLRLVEHLQJUHYLVHGIURPUHIOHFWLQJWKH2007 (Revision 0) to the 2013 (Revision 1) versionRIWKH37/5PHWKRGRORJ\WRSLFDOUHSRUW, which is the current version of the SIR 05-044-A LTR. There are no changes to Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits. to this enclosure provides the existing TS page marked-up to show the proposed change. Attachment 2 to this enclosure provides the retyped TS page.

3.0 TECHNICAL EVALUATION

3.1 General Discussion 10 CFR 50 Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the RCPB of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including AOOs and during system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Due to the effects of neutron irradiation embrittlement accumulated by the reactor vessel, the P-T limit curves contained in plant TSs are updated periodically to ensure that the limit curves are always valid for beyond the EFPY that the plant has accumulated.

P-T Curve Axes Inversion for the MNGP While most plants use P-T curves, MNGP uses the inverse of P-T curves, i.e., T-P curves for operation. The MNGP operators have been trained to operate ABOVE these curves.

Use of the inverse of the P-T curves has no effect on the validity of the curves to protect the RPV from fracture. To avoid operator confusion and prevent error-likely situations, MNGP will continue to use the T-P curve format within the PTLR.

Page 3 of 18

L-MT-23-047 NSPM For the purposes of discussion within this LAR, however, NSPM will use the more common P-T curve nomenclature.

3.2 Development of the Revised P-T Limit Curves - Generic Letter (GL) 96-03 Considerations The revised PTLR was developed in accordance with the template PTLR of the SIR-05-044-A LTR and meets the seven GL 96-03 criteria:

 Section 3.0, "Methodology," of the PTLR refers to the neutron fluence calculational

methodology references and provides the values of neutron fluences used in the

adjusted reference temperature (ART) calculation.

 Appendix A of the PTLR describes the MNGP reactor vessel materials surveillance

program. The BWRVIP LTRs describe the administration of the material surveillance

program including the surveillance capsule reports for the MNGP.

 Low Temperature Overpressure Protection System limits are not applicable to Boiling

Water Reactors (BWRs).

 Section 3.0, "Methodology," of the PTLR describes the method for calculating theART values using RG 1.99, Revision 2.

 Section 5.0, "Discussion," of the PTLR describes the application of fracturemechanics in the construction of P-T limits and provides information regarding theANSYS finite element analyses for the feedwater nozzle (non-beltline) and recirculationinlet nozzle (beltline) performed to generate part of the P-T limits.

 Section 4.0, Operating Limits, of the PTLR discusses the minimum temperature

requirements in 10 CFR 50, Appendix G which are applied to P-T limits for bolt-up

temperature and hydrotest temperature. The SIR-05-044-A LTR provides detailed

information regarding the minimum temperature requirements for bolt-up temperature

and hydrotest temperature.

 Appendix A of the PTLR, which discusses the MNGP reactor vessel materials

surveillance program, includes how multiple surveillance capsules are used in the ART

calculation. The referenced reports and calculations describe how the data from

multiple surveillance capsules are used and the determination of the chemistry factor

from the surveillance data are used in the ART calculations.

3.3 Neutron Fluence Determination 10 CFR 50 Appendices G and H, Fracture Toughness Requirements, and Reactor Vessel Material Surveillance Program Requirements, respectively, present requirements that guide fluence determinations. Appendix G specifies fracture toughness requirements for the carbon Page 4 of 18

L-MT-23-047 NSPM and low-alloy ferritic materials of the pressure-retaining components of the RCPB. Appendix H specifies requirements for a material surveillance program that serves to monitor changes in the fracture toughness properties of the ferritic materials in the reactor beltline region.

Implementing guidance addressing the two appendices is provided in two regulatory guides.

US NRC Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Material, (Reference 7) addresses the requirements of 10 CFR 5, Appendix G for determining the QHXWURQ fluence used in the evaluation of fracture toughness in light water nuclear reactor pressure vessel ferritic materials. The ART values for the limiting beltline materials were calculated (see Enclosure 3) in accordance with this regulatory guide.

RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, (Reference 8) addresses the requirements for determining the fast neutron fluence and uncertainty in the fluence predictions used in fracture toughness evaluations. The previous neutron fluence calculations supporting the P-T limits in the current PTLR (approved by Amendment 172 in 2013 (Reference 2)) were performed in accordance with the General Electric fluence methodology, NEDO-32983P-A, Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations (Reference 6). The present PTLR (currently Revision 1) is applicable for the MNGP through the end of the first license renewal period, i.e., up to 54 EFPY.

The RAMA fluence methodology was developed by TransWare under sponsorship for the Electric Power Research Institute, Inc. (EPRI) and the Boiling Water Reactor Vessel Internals Program (BWRVIP), for the purpose of calculating fast neutron fluence in RPVs and reactor vessel internal components. As prescribed in RG 1.190, RAMA has been benchmarked and qualified against industry standard benchmarks for both BWR and pressurized water reactor (PWR) designs. In addition, RAMA has been compared with several plant-specific dosimetry measurements and reported fluence from several commercial operating reactors. The results of the benchmarks and comparisons to measurements show that RAMA accurately predicts specimen activities, RPV fluence, and vessel component fluence in all light water reactor types. This prior work was extended in the Seabrook Station license renewal analysis (Reference 9) further validating the use of RAMA for all light water reactor designs.

The RAMA methodology has been used for determining fast neutron fluence in both BWR and PWR pressure vessels with no discernable bias in the computed results. Utilization of the RAMA fluence methodology is subject to several conditions, including that a plant geometry-specific validation must be performed, as discussed below.

In May 2005, the NRC issued a safety evaluation (SE), enclosed within EPRI report BWRVIP-114NP-A, BWR Vessel and Internals Project RAMA Fluence Methodology Theory Manual, Final Report, (Reference 10), that evaluated the RAMA fluence methodology.

Subsection 4.1, BWR RPV Neutron Fluence, of the SE states, the staff concludes that the BWRVIP methodology, as described in these reports, provides an acceptable best-estimate plant-specific prediction of the fast (E 1.0 MeV) neutron fluence for BWR RPVs. The conclusion goes on to state:

Page 5 of 18

L-MT-23-047 NSPM With respect to the calculation of BWR RPV neutron fluence, the staff concludes that based on the plant-specific benchmark data presently available, no calculational bias is required for the application of the methodology to plants of similar geometrical design to Susquehanna and Hope Creek, i.e., BWR-IV plants. However, in order to provide continued confidence in the proposed neutron fluence methodology for the BWR RPVs, the acceptance of this methodology is subject to the following conditions for plants which do not have geometries similar to the cited BWR-IV's:

x To apply the RAMA methodology to plant groups which have geometries that are different than the cited BWR-IV's, at least one plant-specific capsule dosimetry analysis must be provided to quantify the potential presence of a bias and assure that the uncertainty is within the RG 1.190 limits.

and x Justification is necessary for a specific application based on geometrical similarity to an analyzed core, core shroud, and RPV geometry. That is, a licensee who wishes to apply the RAMA methodology for the calculation of RPV neutron fluence must reference, or provide, an analysis of at least one surveillance capsule from a RPV with a similar geometry.

On January 9, 2023, NSPM submitted an application for subsequent license renewal (SLR) for the MNGP (Reference 11). Neutron fluence projections were performed for the MNGP RPV and reactor vessel internals components and plant structures applying the RAMA fluence methodology for the projected twenty-year period of subsequent extended operation, i.e.,

through 72 EFPY - conservatively bounding the 54 EFPY fluence assumed through the end of the current Monticello Renewed Facility Operating License. On July 11, 2023, NSPM submitted a third supplement to the SLR application (Reference 12) containing non-proprietary and proprietary versions of a TransWare topical report discussing the MNGP fluence methodology and qualification of that model (Reference 13). This TransWare report supported development of the MNGP PTLR discussed herein.

With respect to the indented section above describing conditions that must be met for application of the RAMA fluence methodology, this report also addresses the above conditions.

Specifically, since the MNGP is a BWR-III plant design:

x Plant-specific capsules dosimetry analysis were provided to quantify the potential presence of a bias and ensure that the uncertainty is within the RG 1.190 limits.

x The MNGP RPV was modeled providing an analyzed core, core shroud, and RPV geometry.

Page 6 of 18

L-MT-23-047 NSPM x Also, Subsection 4.1, first paragraph of the SE states, This acceptance is limited to the axial region defined by the core active fuel height. However, since the 2005 SE, conservatisms in later model applications have allowed extension to include the extended beltline region.

In the 2005 NRC SE the RAMA fluence methodology was indicated as approved for the Susquehanna and Hope Creek BWRs for the applications discussed therein. Subsequently, the use of RAMA has been approved for several other BWR licenses as discussed in the precedents section of this LAR.

3.4 Reactor Vessel Material Surveillance Program 10 CFR 50 Appendix H requires a material surveillance program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from exposure to neutron irradiation and the thermal environment. Under the program, fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel. Section III of Appendix H specifies that a material surveillance program is required for light water nuclear power reactors if the peak fast neutron fluence with energy greater than 1 MeV (E > 1 MeV) at the end of the design life of the vessel is expected to exceed 1017 n/cm2. Section III also allows for an Integrated Surveillance Program (ISP) in which representative materials for the reactor are irradiated in one or more other reactors of sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.

MNGP is a member of the BWRVIP ISP which is administered by the EPRI and the BWR Owners Group. The ISP combines the domestic BWR surveillance programs into a single integrated program. This program uses similar heats of materials in the surveillance programs of various BWRs nuclear plants to represent the limiting materials in other BWR RPVs.

The scope of the program is described in the BWRVIP ISP guidance, and the technical basis of the program is described in BWRVIP-78, BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan (Reference 14). The ISP capsule removal schedule is included in BWRVIP-86, Revision 1-A, BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program Implementation Plan, (Reference 15). On April 22, 2003, NSPM committed to implement the BWRVIP ISP in place of its original surveillance programs for the MNGP in Amendment No. 135 (Reference 16). MNGP is currently operating in and is licensed to use the BWRVIP ISP during the Renewed License period of extended operation. (1)

1. Adoption of BWRVIP-321-A, Boiling Water Reactor Vessel and Internals Project, Plan for Extension of the BWR Integrated Surveillance (ISP) Through the Second License Renewal (SLR), is projected for the proposed SLR period.

Page 7 of 18

L-MT-23-047 NSPM 7he most recently removed surveillance capsule, the 120-degree surveillance capsule, was irradiated from initial startup through 30 cycles of operation before it was removed from the RPV during the spring 2021 refueling outageLQDFFRUGDQFHZLWK&)5$SSHQGL[+

This was the last of the three surveillance capsules installed in the MNGP reactor. Results for this capsule are available in BWRVIP-347, BWR Vessel and Internals Project, Testing and Evaluation of the Monticello 120° ISP(E) Surveillance Capsule (Reference 17) and the test results will be added to the next revision of BWRVIP-135, BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations, (Reference 18) and were used in the preparation of the PTLR revision discussed and provided herein.

3.5 Chemistry and Adjusted Reference Temperature Chemistry in the context of P-T curve calculation is related to the copper and nickel contents in the RPV shell material. The copper and nickel metal contents are needed to calculate the ART. The reactor vessel beltline copper and nickel values were obtained from the evaluation of the MNGP reactor vessel plate, weld, and forging materials in the SIA calculation for evaluation of the ART and reference temperature shifts which included the results of the three surveillance capsules. The copper and nickel values were used with Table 1, Chemistry Factor for Welds, °F, of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1, Adjusted Reference Temperature, of the guide for the welds. The copper and nickel values were used with Table 2, Chemistry Factor for Base Metal, °F, of RG 1.99 to determine a CF per Paragraph 1.1 of the guide for the plates and forgings. provides a copy of the calculation describing the method for calculating the ART using RG 1.99, Revision 2.

3.6 MNGP Feedwater and Recirculation Inlet Nozzles Finite Element Analyses Plant-specific MNGP feedwater nozzle (non-beltline) and recirculation inlet nozzle (beltline) analyses were performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients. The results are design inputs providing the quarter-T nozzle stress intensity factors for the feedwater and recirculation inlet nozzles in Enclosure 4.

3.7 Pressure-Temperature Limit Curves Generation for 72 EFPY provides a copy of the calculation for development of the P-T limit curves forWKH

beltline, bottom head, and non-beltline regions of the MNGP RPV for ()3<VRI operation in accordance with the guidance of Revision 1 of the SIA LTR 6,5$

which satisfies the requirements of 10 CFR 50 Appendix G and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Nonmandatory Appendix G.

Page 8 of 18

L-MT-23-047 NSPM The curves were developed for the following plant conditions: Pressure Test (Curve A),

Normal Operation - Core Not Critical (Curve B), and Normal Operation - Core Critical (Curve C). Separate curves are provided for each of the following three regions of the RPV as well as a composite curve for the entire RPV:

1. The beltline region (includes nozzles where 1/4T fluence > 1 x 1017 n/cm2),

2. The bottom head region,

3. The non-beltline region, including the top head flange,

4. Composite curve (bounding curve for all regions)

For the beltline region, the P-T curves incorporate components with the neutron fluence greater than 1 x 1017 n/cm2 (E >1 MeV). The instrument nozzles are not in the beltline region and are not included in the P-T curve evaluations. The Feedwater nozzles are assumed to be the bounding component for non-beltline components.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The NRC has established requirements in 10 CFR 50, Appendix G, "Fracture Toughness Requirements," in order to protect the integrity of the RCPB in nuclear power plants.

Appendix G requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G to Section XI of the ASME B&PV Code were used to generate the P-T limits. Also, Appendix G requires that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

10 CFR 50.36, "Technical specifications," provides the regulatory requirements for the content required in the TSs which includes limiting conditions for operation (LCO's), surveillance requirements and administrative controls.

MNGP was designed before the publishing of the 70 General Design Criteria for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission (AEC) for public comment in July 1967, and constructed prior to the 1971 publication of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR 50. As such, the MNGP was not licensed to the Appendix A, General Design Criteria (GDC).

Page 9 of 18

L-MT-23-047 NSPM MNGP USAR, Section 1.2, lists the principal design criteria (PDCs) for the design, construction and operation of the plant. USAR Appendix E provides a plant comparative evaluation to the 70 proposed AEC design criteria. It was concluded that the plant conforms to the intent of the 70 proposed AEC GDCs. The applicable PDCs, July 1967 - 70 AEC GDCs, and applicable current 10 CFR 50, Appendix A GDC are discussed below.

x PDC 1.2.11 -- Class I Equipment and Structures Class I structures, systems and components are those whose failure could cause significant release of radioactivity or which are vital to a safe shutdown of the plant under normal or accident conditions and to the removal of decay and sensible heat from the reactor.

x AEC 70 GDC 33 -- Reactor Coolant Pressure Boundary Capability The reactor coolant pressure boundary shall be capable of accommodating without rupture and with only limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.

x GDC 14 - Reactor coolant pressure boundary.

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

x GDC 15 - Reactor coolant system design.

The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

x AEC 70 GDC 35 -- Reactor Coolant Boundary Brittle Fracture Prevention Under conditions where reactor coolant pressure boundary system components constructed of Ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperatures shall be at least 120°F above the nil ductility transition (NDT) temperature of the component material if the resulting energy is expected to be absorbed within the elastic strain energy range.

Page 10 of 18

L-MT-23-047 NSPM x AEC 70 GDC 34 -- Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given (a) to the notch-toughness properties if materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loading, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions.

x GDC 31 - Fracture prevention of the reactor coolant pressure boundary.

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

x AEC 70 GDC 36 -- Reactor Coolant Pressure Boundary Surveillance Criteria 36 - Reactor Coolant Pressure Boundary (Category A) Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leak tight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming with ASTM-E-185-66 shall be provided.

x GDC 32 - Inspection of reactor coolant pressure boundary.

Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

NSPM has evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria. It was concluded that the proposed TS changes will continue to assure that the design requirements and acceptance criteria of MNGP pressure / temperature reload limit analyses are met. Based on this, there is reasonable assurance that the health and safety of the public, following approval of this TS change, is unaffected.

Page 11 of 18

L-MT-23-047 NSPM 4.2 Precedent The proposed LAR is similar to the following NRC approved license amendments for BWRs where the RAMA neutron fluence calculational methodology was utilized.

1. The Peach Bottom Atomic Power Station, Units 2 and 3, received amendments in October 2019 where the RAMA neutron fluence methodology was approved for use by the NRC through the subsequent license renewal period (Reference 19).

2. Units 1 and 2 at the Limerick Generating Station received amendments in September 2021 where the RAMA neutron fluence methodology was approved for use by the NRC through the license renewal period in addition to the P-T curves being relocated to a PTLR (Reference 20).

3. The Hope Creek Generating Station received an amendment in December 2017 where the RAMA neutron fluence methodology was approved for use by the NRC through the license renewal period in addition to the P-T curves being relocated to a PTLR (Reference 21).

4. The Columbia Generating Station received an amendment in November 2022 replacing the existing P-T curves within the TS with a PTLR. The curves are valid based on analyses projected through the license renewal period (Reference 22)

5. Units 1 and 2 at the LaSalle County Station received amendments in November 2023 where the RAMA neutron fluence methodology was approved for use by the NRC through the license renewal period and the P-T curves were relocated to a PTLR (Reference 23)

Therefore, based on the considerations discussed above, NSPM has determined that the proposed change does not require any exemptions or relief from regulatory requirements other than the TS, and does not affect conformance with the intent of any GDC differently than described in the USAR.

4.3 No Significant Hazards Consideration Analysis In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests an amendment to facility Renewed Operating License DPR-22, Technical Specifications (TS) for Monticello Nuclear Generating Plant (MNGP).

It is proposed to revise the MNGP Technical Specifications (TS), specifically Specification 5.6.5, Pressure Temperature Limits Report (PTLR), to revise the PTLR curves to reflect new surveillance capsule results, apply a different fluence methodology, and extend the applicability of the curves WKURXJK  effective full power years (EFPY).

Page 12 of 18

L-MT-23-047 NSPM These new curves have been developed applying the analytical methodology described in 5HYLVLRQRIWKHStructural Integrity Associates (SIA) Report SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," which has received NRC approval. The curves were developed applying a different fluence methodology - TransWare implementation of the Radiation Analysis Modeling Application (RAMA) fluence methodology.

NSPM has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed license amendment requests adoption of the NRC approved RAMA neutron fluence calculational methodology together with an update of the TS to reflect the current NRC approved version of a SIA PTLR development methodology report for preparation of MNGP Pressure-Temperature (P-T) limit curves. The revised MNGP PTLR was developed based on these methodologies and templates provided within these reports.

10 CFR 50 Appendix G establishes requirements to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Implementing the NRC approved RAMA and the SIA methodology for calculating the P-T limit curves provides an equivalent level of assurance that RCPB integrity will be maintained, as required by 10 CFR 50 Appendix G.

Additionally, 10 CFR 50, Appendix H, provides the NRC criteria for design and implementation of reactor pressure vessel (RPV) material surveillance programs for operating lightwater reactors. Implementing these NRC approved methodologies does not reduce the ability to protect the RCPB as specified in Appendix G, nor do these changes increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new RAMA fluence methodology for calculating P-T curves provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page 13 of 18

L-MT-23-047 NSPM

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed license amendment for adoption of the NRC approved RAMA neutron fluence calculational methodology together with an update of the TS to reflect the current NRC approved version of a SIA PTLR development methodology does not alter or involve any design basis accident initiators. RCPB integrity will continue to be maintained in accordance with 10 CFR 50 Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), and the installed equipment is not being operated in a new or different manner.

Accordingly, no new failure modes are introduced which could introduce the possibility of a new or different kind of accident from any previously evaluated.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed license amendment for adoption of the NRC approved RAMA neutron fluence calculational methodology together with an update of the TS to reflect the current NRC approved version of a SIA PTLR development methodology. Calculating the MNGP P-T limits using these NRC approved methodologies, ensures adequate margins of safety relating to RCPB integrity are maintained. The proposed changes do not alter the manner in which the Limiting Conditions for Operation P-T limits for the RCPB are determined. There are no changes to the operability requirements for equipment assumed to operate for accident mitigation.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, NSPM concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c); accordingly, a finding of no significant hazards consideration is justified.

Page 14 of 18

L-MT-23-047 NSPM 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed changes do not change a requirement with respect to installation or use of a facility or component located within the restricted area, as defined in 10 CFR 20, Standards for Protection Against Radiation, nor do they change an inspection or surveillance requirement. The proposed changes do not involve (i) a significant hazards consideration, or (ii) authorize a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) result in a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for a categorical exclusion set forth in 10 CFR 51.22, Criteria for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, specifically paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed change.

Page 15 of 18

L-MT-23-047 NSPM

6.0 REFERENCES

1. Boiling Water Reactor Owner's Group (BWROG) Licensing Topical Report (LTR)

BWROG TP-11-022-A, Revision 1 (Structural Integrity Associates, Inc. Report SIR-05-044, Revision 1-A), Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, dated August 2013 (Agency Documents and Management System (ADAMS) Accession No. ML13277A557)

2. NRC letter to NSPM, Monticello Nuclear Generating Plant - Issuance of Amendment to Revise and Relocate Pressure Temperature Curves to a Pressure Temperature Limits Report (TAC No. ME7930), dated February 27, 2013 (ADAMS Accession Number ML13025A155)

3. NRC Generic Letter 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, dated January 31, 1996 (ADAMS Accession Number ML031110004)

4. Technical Specification Task Force (TSTF) -419, Revise PTLR Definition and References in ISTS [Improved Standard Technical Specification] 5.6.6, RCS [Reactor Coolant System] PTLR, dated September 16, 2001 (ADAMS Accession Number ML012690234)

5. BWROG LTR BWROG TP-11-022-A, Revision 0 (Structural Integrity Associates, Inc.

Report SIR-05-044, Revision 1-A), Pressure Temperature Limits Report Methodology for Boiling Water Reactors, dated August 2007

6. GE Energy, Nuclear Report NEDO-32983P-A, Revision 2, Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations, dated January 2006 (ADAMS Accession Number ML072480121)

7. NRC Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, dated May 1988 (ADAMS Accession Number ML003740284)

8. NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001 (ADAMS Accession Number ML010890301)

9. NRC Safety Evaluation to NextEra Energy Seabrook, LLC, Safety Evaluation Report, With Open Items Related to the License Renewal of Seabrook Station, Docket Number 50-443, dated June 2012

10. NRC letter to Bill Eaton, BWRVIP Chairman Entergy Operations, Inc., Safety Evaluation of Proprietary EPRI Reports, BWR Vessel and Internals Project, RAMA Page 16 of 18

L-MT-23-047 NSPM Fluence Methodology Manual (BWRVIP-114), RAMA Fluence Methodology Benchmark Manual Evaluation of Regulatory Guide 1.190 Benchmark Problems (BWRVIP-115), RAMA Fluence Methodology Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 (BWRVIP-117), and RAMA Fluence Methodology Procedures Manual (BWRVIP-121), and Hope Creek Flux Wire Dosimeter Activation Evaluation for Cycle 1 (TWE-PSE-001-R-001)

(TAC No. MB9765), dated May 13, 2005

11. NSPM letter to NRC, Monticello Nuclear Generating Plant Docket No. 50-263, Renewal License Numbers DPR-22 Application for Subsequent Renewal Operating License, (Letter L-MT-23-001) dated January 9, 2023 (ADAMS Accession No. ML23009A353)

12. NSPM letter to NRC, Subsequent License Renewal Application Supplement 3, (Letter L-MT-23-030) dated July 11, 2023 (ADAMS Accession Number ML23193B026)

13. NSPM Calculation No: 23-008, Revision 1, TransWare Topical Report, Monticello Nuclear Generating Plant Fluence Methodology Report (TransWare Doc. No.

MNT-FLU-001-R-001-LNP, Revision 0, April 2023) and Attachment 1, Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30 (TransWare Doc. No. MNT-FLU-001-R-001-LNP, Attachment 1, Revision 1, June 2023) [This calculation and its attachment were enclosed within Letter L-MT-23-030.]

14. BWRVIP-78NP, BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan, dated April 2000 (ADAMS Accession No. ML003704011)

15. BWRVIP-86, Revision 1-A, BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, (ADAMS Accession Nos.

ML023190487)

16. NRC letter to Nuclear Management Company (NMC), LLC, Monticello Nuclear Generating Plant - Issuance of Amendment Re: Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program (TAC No.

MB6460), dated April 22, 2003 (ADAMS Accession No. ML030830591)

17. BWRVIP-347, BWR Vessel and Internals Project, Testing and Evaluation of the Monticello 120° ISP(E) Surveillance Capsule, Final Report, October 2022

18. BWRVIP-135, Revision 4 (current), BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. (ADAMS Accession Number ML22332A455) (non-proprietary version)

19. NRC letter to Exelon Generation Company, LLC, Safety Evaluation Report Related to the Subsequent License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3, Docket Nos. 50-277 and 50-278, dated February 2020 Page 17 of 18

L-MT-23-047 NSPM

20. NRC letter to Exelon Nuclear, Limerick Generating Station, Units 1 and 2 - Issuance of Amendment Nos. 253 and 215 Re: Technical Specification Changes Related to Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report (EPID L-2020-LLA-0221), dated September 28, 2021, (ADAMS Accession Number ML21181A044)

21. Letter from the U.S. NRC to PSEG Nuclear LLC, Hope Creek Generating Station -

Issuance of Amendment to Revise and Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report (CAC No. MF9502, EPID L-2017-LLA-0204), dated December 14, 2017 (ADAMS Accession Number ML17324A840)

22. NRC letter to Energy Northwest, Columbia Generating Station - Issuance of Amendment No. 268 to Revise Technical Specification 3.4.11 RCS Pressure and Temperature (P/T) Limits (EPID L-2021-LLA-0191), dated November 23, 2022 (ADAMS Accession Number ML22263A445)

23. NRC letter to Constellation Energy Generation, LLC, LaSalle County Station, Units 1 and 2 - Issuance of Amendment Nos. 260 and 245 to Renewed Facility Operating Licenses Re: Relocation of Pressure and Temperature Limit Curves to the Pressure Temperature Report (EPID L-2022-LLA-0173), dated November 8, 2023 (ADAMS Accession Number ML23286A260)

Page 18 of 18

ENCLOSURE 1, ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISION TO THE MNGP PRESSURE TEMPERATURE LIMITS REPORT TO CHANGE THE NEUTRON FLUENCE METHODOLOGY AND INCORPORATE NEW SURVEILLANCE CAPSULE DATA TECHNICAL SPECIFICATION PAGES (MARKUP)

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ENCLOSURE 2 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISION TO THE MNGP PRESSURE TEMPERATURE LIMITS REPORT TO CHANGE THE NEUTRON FLUENCE METHODOLOGY AND INCORPORATE NEW SURVEILLANCE CAPSULE DATA XCEL ENERGY CORPORATION MONTICELLO NUCLEAR GENERATING PLANT (MNGP)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

UP TO 72 EFFECTIVE FULL-POWER YEARS (EFPY)

REVISION 2 (32 Pages Follow)

Monticello Nuclear Generating Plant PTLR Revision 2 Page 1 of 32 Xcel Energy Corporation Monticello Nuclear Generating Plant (MNGP)

Pressure and Temperature Limits Report (PTLR) up to 72 Effective Full-Power Years (EFPY)

Revision 2

Monticello Nuclear Generating Plant PTLR Revision 2 Page 2 of 32 Table of Contents Section Page 1.0 Purpose 4 2.0 Applicability 4 3.0 Methodology 5 4.0 Operating Limits 6 5.0 Discussion 7 6.0 References 12 Appendix A Monticello Reactor Vessel Materials Surveillance Program 32

Monticello Nuclear Generating Plant PTLR Revision 2 Page 3 of 32 List of Figures Figure 1: MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) for 72 EFPY ...................14 Figure 2: MNGP P-T Curve B (Normal Operation - Core Not Critical) for 72 EFPY .................15 Figure 3: MNGP P-T Curve C (Normal Operation - Core Critical) for 72 EFPY ........................16 Figure 4: MNGP Feedwater Nozzle 3-D Finite Element Model [13] ...........................................17 Figure 5: MNGP Feedwater Nozzle Stress Extraction Path [13] ..................................................18 Figure 6: MNGP Recirculation Nozzle Finite Element Model [14] ..............................................19 Figure 7: MNGP Recirculation Nozzle Stress Extraction Path [14]..............................................20 List of Tables Table 1: MNGP Pressure Test (Curve A) P-T Curves for 72 EFPY .............................................21 Table 2: MNGP Core Not Critical (Curve B) P-T Curves for 72 EFPY .......................................24 Table 3: MNGP Core Critical (Curve C) P-T Curves for 72 EFPY ..............................................27 Table 4: MNGP 1/4T ART Table for 72 EFPY .............................................................................30 Table 5: Nozzle Stress Intensity Factors ........................................................................................31

Monticello Nuclear Generating Plant PTLR Revision 2 Page 4 of 32 1.0 PURPOSE The purpose of the Monticello Nuclear Generating Plant (MNGP) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cool-down and Hydrostatic/Class 1 Leak Testing.

2. RCS Heat-up and Cool-down rates.

3. Reactor Pressure Vessel (RPV) to RCS coolant 7 7HPSHUDWXUH UHTXLUHPHQWVGXULQJ

Recirculation Pump startup.

4. 539ERWWRPKHDGFRRODQWWHPSHUDWXUHWR539FRRODQWWHPSHUDWXUH7UHTXLUHPHQWV

during Recirculation Pump startup.

5. RPV boltup temperature limits.

This report has been prepared in accordance ZLWKWKHUHTXLUHPHQWVRI the current and previous revisions of Licensing Topical Reports SIR-05-044 contained within BWROG-TP-11-022-A, Revision 1 [1].

2.0 APPLICABILITY This report is applicable to the MNGP RPV for up to 72 Effective Full-Power Years (EFPY).

The following MNGP Technical Specifications (TS) are affected by the information contained in this report:

TS 3.4.9 RCS Pressure/Temperature (P/T) Limits

Monticello Nuclear Generating Plant PTLR Revision 2 Page 5 of 32 3.0 METHODOLOGY The limits in this report were derived as follows:

1. The methodology used is in accordance with Reference [1], Pressure - Temperature Limits Report Methodology for Boiling Water Reactors, August 2013, incorporating the NRC Safety Evaluation in Reference [2].

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [3] as documented in Reference [5].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99) [4], as documented in Reference [5, 6].

4. The pressure and temperature limits, which were calculated in accordance with Reference

[1], are documented in Reference [6].

5. This revision of the pressure and temperature limits report is to incorporate the following changes:

x Revision 2: to incorporate new irradiation fluence data [5, 10] that go out to 72 EFPY of the RPV and new chemistry factor from the 120 degree surveillance capsule [24].

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Safety Analysis Report (USAR), can be made pursuant to 10 CFR 50.59 [8], provided the above methodologies are utilized. After issuance, the revised PTLR is submitted to the NRC for awareness.

7KHUHTXLUHPHQWIRU&)5$SSHQGL[+VXEPLVVLRQRIFDSVXOHUHSRUWLQIRUPDWLRQis not directly associated with PTLR processing.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 6 of 32 4.0 OPERATING LIMITS The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operaWLQJOLPLWVIRUSUHVVXUHDQGWHPSHUDWXUHDUHUHTXLUHGIRUWKUHHFDWHJRULHVRIRSHUDWLRQ

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 72 EFPY for MNGP, as documented in Reference [6],

and are provided in Figure 1 through Figure 3 for MNGP. A tabulation of the curves is included in Table 1 through Table 3. The adjusted reference temperature (ART) tables for 72 EFPY for the MNGP vessel beltline materials are shown in Table 4 [5].

The resulting P-T curves are based on the geometry, design and materials information for the MNGP vessel. The following conditions apply to operation of the MNGP vessel:

x Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figure 1: Curve A): )KRXU 1 [1].

x Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - core non-critical, and Figure 3: Curve C - core critical): 100)KU 2 [6].

x RPV bottom head coolant temperature to RPV coolant temperature 'T limit during Recirculation Pump startup: d qF [1].

x Recirculation loop coolant temperature to RPV coolant temperature 'T limit during Recirculation Pump startup: d qF [1].

1 Interpreted as the temperature change in any 1-KRXUSHULRGLVOHVVWKDQRUHTXDOWo 25°F.

2 Interpreted as the temperature change in any 1-KRXUSHULRGLVOHVVWKDQRUHTXDOWR)

Monticello Nuclear Generating Plant PTLR Revision 2 Page 7 of 32 x RPV head flange, RPV flange and adjacent shell temperature limit during vessel bolt-up

60) [6].

 DISCUSSION The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [4] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the MNGP vessel plate, weld, and forging materials [5]; this evaluation included the results of three surveillance capsules. The Cu and Ni values were used with Table 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds. The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings.

Per Reference [5] and in accordance with Appendix A of Reference [1], the MNGP representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [11]. The representative plate material for MNGP (C2220) in the ISP is the same as the lower intermediate shell plate material in the vessel beltline region of MNGP. For the plate heat C2220, since the scatter in the fitted results is less than 1-VLJPD ) WKHPDUJLQWHUP = 17°F) is cut in half for the material when calculating the ART. The representative heat of the weld material (5P6756) in the ISP is not the same as the limiting weld material in the vessel beltline region of MNGP. Therefore, the CFs from the tables in RG1.99 were used in the determination of the ART values of all MNGP materials except for plate heat C2220. Reference [5] used a chemistry factor (CF) of 180 from Reference [11]. However, the latest ISP data show that the CF value for plate heat C2220 changes to 174 [24] which is used in the ART calculation in P-T curve evaluations

[6].

Monticello Nuclear Generating Plant PTLR Revision 2 Page 8 of 32 The peak RPV ID fluence value of 5.94 x 1018 n/cm2 at 72 EFPY used in the P-T curve evaluations was obtained from Reference [10]. The fluence value applies to the limiting beltline lower intermediate shell plates (Heat No. C2220-1 and C2220-2). The fluence value for the lower intermediate shell plates is based upon an attenuation of 0.738 for a postulated 1/4 flaw.

&RQVHTXHQWO\, the 1/4T fluence for 72 EFPY for the limiting lower intermediate shell plates is 4.38 x 1018 n/cm2 [5]. Using the CF value of 174 from the latest ISP data [24], the limiting ART value for beltline plates and welds is 178.1°F for MNGP [6].

The RPV ID fluence value of 7.08 x 1017 n/cm2 at 72 EFPY used in the P-T curve evaluation of the recirculation inlet nozzle was obtained from Reference [10]. The fluence value applies to the limiting recirculation inlet nozzle (Heat No. E21VW). The fluence value for the recirculation inlet nozzle is based upon an attenuation of 0.738 for a postulated 1/4 flaw. As a result, the 1/4T fluence for the limiting recirculation inlet nozzle is 5.23 x 1017 n/cm2 at 72 EFPY for MNGP.

There are no additional forged or instrument nozzles in the extended beltline at 72 EFPY. The limiting ART value for the recirculation inlet nozzle is 116.6°F for MNGP at 72 EFPY [5].

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cool-down and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable fracture toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, and for a given pressure, the coolant saturation

Monticello Nuclear Generating Plant PTLR Revision 2 Page 9 of 32 temperature is well above the P-7FXUYHOLPLWLQJWHPSHUDWXUH&RQVHTXHQWO\WKHPDWHULDO

fracture toughness at a given pressure would exceed the allowable fracture toughness.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of d 100°F/hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound RPV thermal transients. For the hydrostatic pressure and leak test curves (Curve A), a coolant heatup and cooldown temperature of d 25°F/hr must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial RTNDT, chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 1017 n/cm2 for E >

1 MeV) is shown in Table 4 for 72 EFPY [6]. Use of initial RTNDT values in the determination of P-T curves for MNGP was approved by the NRC in Reference [9].

The only computer code used in the determination of the MNGP P-T curves was the ANSYS Mechanical, Release 18.1 [12] finite element computer program. ANSYS finite element analyses were used to develop the stress distributions through the feedwater nozzle (non-beltline) and recirculation inlet nozzle (beltline) as well as the vessel shell, and these stress distributions were used in the determination of the stress intensity factors for the feedwater and recirculation inlet nozzles [13, 14] and vessel shell. At the time that each of the analyses above was performed, the ANSYS program was controlled under the vendors 10 CFR 50 Appendix B [15] Quality

$VVXUDQFH3URJUDPIRUQXFOHDUTXDOLW\-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [16] was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 10 of 32 The plant-specific MNGP feedwater nozzle analyses were performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analyses can be found in Reference [13]. The following inputs were used as input to the finite element analysis:

x A one-TXDUWHUV\PPHWULFWKUHH-dimensional finite element model of the feedwater nozzle was constructed and is shown in Figure 4. Temperature dependent material properties, taken from the MNGP Code of Record [17], were used in the evaluation.

x Heat transfer coefficients were calculated at different flow rates. The analysis used the conservative forced convection coefficients and applied it to all wetted surfaces [13].

Therefore, the heat transfer coefficients used in the analysis bound the actual operating conditions in the feedwater nozzle at MNGP.

x With respect to operating conditions, stress distributions were developed for two bounding thermal transients. A thermal shock, which represents the maximum thermal shock for the feedwater nozzle during normal operating conditions, and a thermal ramp were analyzed [13]. The thermal stress distributions, corresponding to the limiting times presented in Reference [13], along a linear path through the nozzle corner is used as shown in Figure 57KHERXQGDU\LQWHJUDOHTXDWLRQLQIOXHQFHIXQFWLRQ %,(,) 

methodology presented in Reference [1] is used to calculate the thermal stress intensity factor, KItGXHWRWKHWKHUPDOVWUHVVHVE\ILWWLQJDWKLUGRUGHUSRO\QRPLDOHTXDWLRQWRWKH

path stress distribution for the thermal load case.

x With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the 3-D model in Reference [13]. The pressure stress distribution was taken along a linear path through the nozzle corner as shown in Figure 5. The BIE/IF methodology presented in Reference [1] was used to calculate the applied pressure stress intensity factor, KIpE\ILWWLQJDWKLUGRUGHUSRO\QRPLDOHTXDWLRQWRWKHSDWKVWUHVV

distribution for the pressure load case. The resulting KIp may be linearly scaled to determine the KIp for various RPV internal pressures.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 11 of 32 The plant-specific MNGP recirculation inlet nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients. Detailed information regarding the analysis can be found in Reference [14].

The following inputs were used as input to the finite element analysis:

x A one-TXDUWHUV\PPHWULFWKUHH-dimensional finite element model of the recirculation inlet nozzle was constructed and is shown in Figure 6. Temperature dependent material properties, taken from the MNGP Code of Record [18], were used in the evaluation.

x Heat transfer coefficients were calculated with the most severe thermal shock for the nozzle blend radius in safety valve blow down (SVBD). The heat transfer coefficients were conservatively calculated based on the full temperature difference of the transient, rather than the RPV to coolant temperature difference [14]. Therefore, the heat transfer coefficients used in the analysis bound the actual operating conditions in the recirculation inlet nozzle at MNGP.

x With respect to operating conditions, the thermal transient that would produce the highest tensile stresses at the 1/4T location is the 100°F/hr SVBD transient [14]. Therefore, the stresses represent the bounding stresses in the recirculation inlet nozzle associated with 100°F/hr heatup/cooldown limits associated with the P-T curves for a nozzle in the beltline region.

x With respect to pressure stress, a unit pressure of 1010 psig was applied to the internal surfaces of the 3-D model in Reference [14]. The pressure stress distribution was taken along a linear path through the nozzle corner as shown in top of Figure 7. The BIE/IF methodology presented in Reference [1] was used to calculate the applied pressure stress intensity factor, KIpE\ILWWLQJDWKLUGRUGHUSRO\QRPLDOHTXDWLRQWRWKHSDWKVWUHVV

distribution for the pressure load case. The resulting KIp may be linearly scaled to determine the KIp for various RPV internal pressures.

Table 5 summarizes the pressure stress intensity factor and maximum thermal stress intensity factor for both feedwater and recirculation inlet nozzle.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 12 of 32

6.0 REFERENCES

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, August 2013. (ADAMS Accession No. ML13277A557)

2. U.S. NRC Letter to BWROG dated May 16, 2013, Final Safety Evaluation for Boiling Water Reactor Owners Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors (TAC NO. ME7649, ADAMS Accession No. ML13277A557).

3. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

4. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

5. SI Calculation No. 2100300.302P, Revision 2, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts. CONTAINS PROPRIETARY INFORMATION.

6. SI Calculation No. 2200284.303, Revision 0, Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY.

7. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix G, Fracture Toughness 5HTXLUHPHQWV, December 12, 2013.

8. U.S. Code of Federal Regulations, Title 10, Part 50, Section 59, Changes, tests and experiments, August 28, 2007.

9. Application for Renewed Operation License, Monticello Nuclear Generating Plant, March 2005.

10. Monticello Nuclear Generating Station Reactor Pressure Vessel Fluence Evaluation -

6XEVHTXHQW/LFHQVH5HQHZDO'7UDQV:DUH(QWHUSULVHs, MNT-FLU-001-R-002, Revision

0. June 2022. SI File NO. 2100300.201.

11. BWRVIP-135, Revision 4: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. EPRI PROPRIETARY INFORMATION.

12. ANSYS Mechanical APDL (UP20170403) and Workbench (March 31, 2017), Release 18.1, SAS IP, Inc.

13. SI Calculation No. 2200284.302P, Revision 0, Finite Element Stress Analysis of Monticello RPV Feedwater Nozzle. CONTAINS PROPRIETARY INFORMATION.

14. SI Calculation No. 2200284.301P, Revision 0, Finite Element Stress and Fracture Mechanics Analysis of Monticello RPV Recirculation Inlet Nozzle. CONTAINS PROPRIETARY INFORMATION.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 13 of 32

15. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants.

16. U.S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, License Qualification for Performing Safety Analyses, June 24, 1999.

17. ASME Boiler and Pressure Vessel Code Section III, Rules for Construction of Nuclear Vessels, 1977 Edition with Addenda through Summer1978.

18. ASME Boiler and Pressure Vessel Code, Section III including Appendices, 1980 Edition with Addenda through Winter 1980.

19. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel 0DWHULDO6XUYHLOODQFH3URJUDP5HTXLUHPHQWV, January 31, 2008.

20. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI PROPRIETARY INFORMATION.

21. BWRVIP-250NP, BWR Vessel and Internals Project, Testing and Evaluation of the Monticello Unit 1 120° Surveillance Capsule. EPRI, Palo Alto, CA: 2011. 1022850.

22. SI File No. 2000115.401, Revision 0, Updated Evaluation of the EPRI BWRVIP ISP Capsule Withdrawal Schedule, May 3, 2021.

23. NRC (L. M. Padovan) letter to NMC (D. L. Wilson), "Monticello Nuclear Generating Plant - Issuance of Amendment re: Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program (TAC No. MB6460)", dated April 22, 2003.

24. BWRVIP Letter 2022-053, September 14, 2022, from Bob Carter to Russell Lidberg,

Subject:

Notification of New BWRVIP Integrated Surveillance Program (ISP) Data Applicable to the Monticello Reactor Pressure Vessel (RPV).

Monticello Nuclear Generating Plant PTLR Revision 2 Page 14 of 32 Figure 1: MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) for 72 EFPY MNGP P-T Curve A - Pressure Test, Composite Curves

- B e l tline - - - Bottom Head - - Non-Beltline - overall 300 COMPLIANCE REQUIRES OPERATION ABOVE THE CURVES 250

...::JCl.I

....ra I

I

+

• I I

I I

I I

I I

8_ 200 E

Cl.I

~

iii t

~

ai 150 Cl.I

....0u ra

~ 100 E

J iI tI ~ tI I~

E I I I I

I I

I I

I

• 2 l T T r

~

50 Bolt-Up Temperature= 60°F I I

~

0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Pressure Limit in Reactor Vessel (psig)

Note: The minimum reactor vessel metal temperature at 0 psig is applicable for RPV operation under a vacuum.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 15 of 32 Figure 2: MNGP P-T Curve B (Normal Operation - Core Not Critical) for 72 EFPY MNGP P-T Curve B - Core Not Critical, Composite Curves

-Beltline - - - Bottom Head - - Non-Beltline - overall 300 COMPLIANCE REQUIRES OPERATION ABOVE THE CURVES G:' 250

~

41 r I

s I (II r rI 41

a. 200 E

41 I- .' . . .' ' ...' . .'

iii 41

+ + '

+ + ' + + '

~

1i 150 41 0

u (II 41 100 0:::

E

s E

• 2

+

I I

r ,

,,-:' -- ' r I

• +

r

~ so Bolt-Up Temperature = 60°F t:

1: '

0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Pressure Limit in Reactor Vessel (psig)

Note: The minimum reactor vessel metal temperature at 0 psig is applicable for RPV operation under a vacuum.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 16 of 32 Figure 3: MNGP P-T Curve C (Normal Operation - Core Critical) for 72 EFPY MNGP P-T Curve C- Core Critical, Composite Curves

- Be lt line - - - Bottom Head - - Non-Beltline - ove rall 300 COMPLIANCE REQUIRES OPERATION ABOVE THE CURVES QI

....:I 8_ 200 E

~

iii QI

~

ai 150 QI

....0u

~"' 100 E .

I E

• 2

~ so Minimum Criticality Temperature = 70°F * + *.: :

,f + ......

0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Pressure Limit in Reactor Vessel (psig)

Note: The minimum reactor vessel metal temperature at 0 psig is applicable for RPV operation under a vacuum.

Monticello Nuclear Generating Plant PTLR Revision 2 Page 17 of 32 Figure 4: MNGP Feedwater Nozzle 3-D Finite Element Model [13]

ELEME'NTS ANSYS Rel ease 18 .1 Build 18 .1 MAT NUM CCT 28 2022 10: 38 :43 Plill NJ. 1 FEFDWATER NJZZIB ELEME'NTS MAT NUM FEFDWATER NJZZIB

Monticello Nuclear Generating Plant PTLR Revision 2 Page 18 of 32 Figure : MNGP Feedwater Nozzle 6WUHVV([WUDFWLRQ3DWK [13]

NCDAL SOLUTICN STEP=l SUB =1 TIMEF=l SY (AVG)

RSYS=l OMX =.065998 SJvN =-5534.02 SMX =50205.8

Monticello Nuclear Generating Plant PTLR Revision 2 Page 19 of 32 Figure 6: MNGP Recirculation Nozzle Finite Element Model [14]

-- ANSYS Rl8. l

~tit;1DU0 RPV Pocirculatloo Inlvt. Noz.z;la 2 ANSYS Rl8.l CCT 25 2022 21 , oq ,23 PLC/I' Jol). 1

-89)8 . 86 -7 25 . 5 -6722 . 44 -5619 . 4 -4516 . 03 -2309 . 62 -103.206

-3412 .82 -12C6 . 41 1000 t'D'lticello PP\,' f

Recir:culdtPoo n1et Nozzle 2, Pressure~

Monticello Nuclear Generating Plant PTLR Revision 2 Page 20 of 32 Figure 7: MNGP Recirculation Nozzle Stress Extraction 3DWK [14]

OCOAL SOI.UrICN STE.P=l SUB =1 TIME=l SY (AVG)

RSYS=0 OMX = . 067428 SMll -4910 . 64 SMX =48833 . 1

-4910 . 64 060 7032 . 42 300 18975. 5 30918 . 5 42861. 6 8 24947 36890 . 1 48833. l Monticello RPV Recircuiltiori fnlet Nozzle 2, Pr<is suJe Loading

Monticello Nuclear Generating Plant PTLR Revision 2 Page 21 of 32 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 72 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 60.0 0.0 60.0 329.7 90.3 379.7 109.1 429.6 122.7 479.5 133.4 529.5 142.2 579.4 149.6 629.3 156.1 679.2 161.9 729.2 171.4 778.2 179.3 827.3 186.2 876.3 192.2 925.4 197.6 974.4 202.4 1023.5 206.9 1072.5 210.9 1121.6 214.7 1170.6 218.2 1219.7 221.5 1268.7 224.5 1317.8 227.4 1366.8 230.2 1415.9 232.8 1464.9 235.2 1514.0 237.6 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 22 of 32 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 72 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 60.0 0.0 60.0 812.8 64.8 859.7 69.2 906.6 73.2 953.4 77.0 1000.3 80.5 1047.2 83.7 1094.1 86.8 1141.0 89.6 1187.9 92.3 1234.8 94.9 1281.7 97.4 1328.6 99.7 1375.4 102.0 1422.3 104.1 1469.2 106.1 1516.1 108.1 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 23 of 32 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 72 EFPY (continued)

Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 60.0 0.0 60.0 312.6 100.0 312.6 100.0 936.3 103.6 984.5 106.9 1032.7 110.0 1080.9 113.0 1129.1 115.8 1177.3 118.4 1225.6 120.9 1273.8 123.3 1322.0 125.6 1370.2 127.7 1418.4 129.8 1466.6 131.8 1514.8 133.7 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 24 of 32 Table 2: MNGP Core Not Critical (Curve B) P-T Curves for 72 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 60.0 0.0 59.6 141.2 99.6 187.6 121.1 234.1 135.8 280.5 147.1 326.9 156.1 373.4 163.8 419.8 170.4 466.2 176.1 512.6 181.3 559.1 189.7 606.9 196.9 654.7 203.2 702.5 208.8 750.3 213.9 798.1 218.4 845.9 222.6 893.7 226.5 941.5 230.1 989.3 233.4 1037.1 236.6 1084.9 239.5 1132.7 242.3 1180.5 245.0 1228.4 247.5 1276.2 249.8 1324.0 252.1 1371.8 254.3 1419.6 256.4 1467.4 258.4 1515.2 260.3 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 25 of 32 Table 2: MNGP Core Not Critical (Curve B) P-T Curves for 72 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 60.0 0.0 60.0 533.1 66.6 582.2 72.4 631.2 77.6 680.3 82.4 729.3 86.7 778.3 90.6 827.4 94.3 876.4 97.8 925.5 101.0 974.5 104.0 1023.6 106.8 1072.6 109.5 1121.6 112.1 1170.7 114.5 1219.7 116.8 1268.8 119.0 1317.8 121.1 1366.8 123.2 1415.9 125.1 1464.9 127.0 1514.0 128.8 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 26 of 32 Table 2: MNGP Core Not Critical (Curve B) P-T Curves for 72 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 60.0 0.0 60.0 312.6 130.0 312.6 130.0 1022.7 132.7 1071.8 135.3 1120.9 137.8 1170.0 140.2 1219.1 142.4 1268.3 144.5 1317.4 146.6 1366.5 148.6 1415.6 150.5 1464.8 152.3 1513.9 154.1 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 27 of 32 Table 3: MNGP Core Critical (Curve C) P-T Curves for 72 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 0.0 70.0 122.2 128.5 170.8 155.2 219.3 172.2 267.8 184.7 316.4 194.6 364.9 202.8 413.4 209.8 462.0 215.9 510.5 221.3 559.1 229.7 606.9 236.9 654.7 243.2 702.5 248.8 750.3 253.9 798.1 258.4 845.9 262.6 893.7 266.5 941.5 270.1 989.3 273.4 1037.1 276.6 1084.9 279.5 1132.7 282.3 1180.5 285.0 1228.4 287.5 1276.2 289.8 1324.0 292.1 1371.8 294.3 1419.6 296.4 1467.4 298.4 1515.2 300.3 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 28 of 32 Table 3: MNGP Core Critical (Curve C) P-T Curves for 72 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature °F Pressure psi 70.0 0.0 70.0 376.2 81.5 425.6 90.9 475.1 98.7 524.5 105.5 574.0 111.5 623.4 116.9 672.9 121.7 722.3 126.1 771.8 130.2 821.2 133.9 870.7 137.4 920.1 140.6 969.6 143.7 1019.0 146.6 1068.5 149.3 1117.9 151.9 1167.4 154.3 1216.8 156.7 1266.3 158.9 1315.7 161.1 1365.2 163.1 1414.6 165.1 1464.1 167.0 1513.5 168.8 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 29 of 32 Table 3: MNGP Core Critical (Curve C) P-T Curves for 72 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 70.0 0.0 70.0 276.4 83.7 312.6 203.0 312.6 203.0 1563.0

Monticello Nuclear Generating Plant PTLR Revision 2 Page 30 of 32 Table 4: 01*37$577DEOHIRU()3<

72 72EFPY Initial Fluence EFPY Component  % 7 57NDT i Heat Lot  % Cu CF RTNDT Factor 7

No. Ni Fluence (°F) (°F) (°F)

(°F) f ART QFP2)

(°F)

/RZHU6KHOO3ODWHV &RXUVH

A0946-I-16 N/A 0.14 0.56 98 27 2.80E+18 0.653 64.1 0 17.0 125.1 1

I-17 C2193-1 N/A 0.17 0.5 119 0 2.80E+18 0.653 77.3 0 17.0 111.3 Lower-,QWHUPHGLDWH6KHOO3ODWHV &RXUVH

I-14 C2220-1 N/A 0.16 0.64 174 27 4.38E+18 0.770 134.1 0 8.5 178.1 I-15 C2220-2 N/A 0.16 0.64 174 27 4.38E+18 0.770 134.1 0 8.5 178.1 8SSHU,QW6KHOO3ODWHV &RXUVH

I-12 C2089-1 N/A 0.35 0.5 200 0 2.38E+17 0.191 38.2 0 17.0 72.2 I-13 C2613-1 N/A 0.35 0.49 198 27 2.38E+17 0.191 37.9 0 17.0 98.9

/RZHU6KHOO &RXUVH $[LDO:HOGV VLAA-1 &

- E8018N 0.1 0.99 135 -65.6 1.73E+18 0.535 72.2 12.7 28.0 68.1 VLAA-2 Lower-,QWHUPHGLDWH6KHOO &RXUVH $[LDO:HOGV

VLBA-1 &

- E8018N 0.1 0.99 135 -65.6 1.55E+18 0.510 68.8 12.7 28.0 64.7 VLBA-2 8SSHU,QW6KHOO &RXUVH $[LDO:HOGV

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

&LUFXPIHUHQWLDO:HOGV VCBA-2 - E8018N 0.1 0.99 135 -65.6 2.80E+18 0.653 88.0 0 28.0 78.4 VCBB-3 - E8018N 0.1 0.99 135 -65.6 2.38E+17 0.191 25.8 0 12.9 -14.0 N2 Nozzle N2 Nozzle E21VW N/A 0.18 0.86 142 40 5.23E+17 0.300 42.6 0 17.0 116.6

Monticello Nuclear Generating Plant PTLR Revision 2 Page 31 of 32 Table : Nozzle Stress Intensity Factors Nozzle Applied Pressure, KIp-app 7KHUPDO.It Feedwater 70.59 for 1,000 psi pressure 10.37 Recirculation Inlet 75.20 for 1,010 psi pressure 25.28 KI in units of ksi-in0.5

Monticello Nuclear Generating Plant PTLR Revision 2 Page 32 of 32 APPENDIX A MONTICELLO REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program 5HTXLUHPHQWV [19], the 300-degree surveillance capsule was removed and tested from the Monticello reactor vessel in 2007. The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. The methods and results of testing are presented in References [20, 21]. The 120-degree capsule was withdrawn in the spring 2021 refueling outage [22, 24]. The latest testing results [24] will be added to the next revision of BWRVIP 135, Reference [11], and have been used in the preparation of this report. This was the final capsule installed in the MNGP reactor.

MNGP is licensed to use the BWRVIP ISP during the Renewed License period of extended operation. The BWRVIP ISP PHHWVWKHUHTXLUHPHQWVRI&)5$SSHQGL[+IRU,QWHJUDWHG Surveillance Programs, and has been approved by NRC. Xcel Energy committed to use the ISP in place of its original surveillance programs in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated April 22, 2003 [23].

ENCLOSURE 3 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISION TO THE MNGP PRESSURE TEMPERATURE LIMITS REPORT TO CHANGE THE NEUTRON FLUENCE METHODOLOGY AND INCORPORATE NEW SURVEILLANCE CAPSULE DATA EVALUATION OF ADJUSTED REFERENCE TEMPERATURES AND REFERENCE TEMPERATURE SHIFTS (NSPM CALCULATION NO.22-035)

(SI NO. 2100300.302, REVISION 4)

(21 Pages Follow)

I'(},

QF0549, Rev. 15 (FP-E-CAL-01) Page 1 of 2 Xcel Energy- I Calculation Signature Sheet Approval: 602000018581 Document Information NSPM Calculation (Doc) No: 22-035 I Revision: 2

Title:

EVALUATION OF ADJUSTED REFERENCE TEMPERATURES AND REFERENCE TEMPERATURES SHIFTS Facility: ~ MT PI I Unit: ~ 1 2 Safety Class: ~ SR Aug Q Non SR Type: Calc Sub-Type: OTH NOTE: Print and sign name in signature blocks, as required.

Major Revisions I N/A EC Number: 601000003566 ~ Vendor Calc:

Vendor Name or Code: Structural Vendor Doc No: 21000300.302 Integrity Associates Description of Revision: Correct OT fluence value for shell course 1 in Table 2 and correctons in Table 2-4 The following calculation and attachments have been reviewed and deemed acceptable as a legible QA record ~

Prepared by: (sign) / (print) Vendor Date:

Reviewed by: (sign) / (print) Matthw Date:

Sears via MOC ITEM 60000112096 Type of Review:

Design Verification Engr Review ~ OAR EOC Method Used (For DV Only):

Review Alternate Calc Test Approved by: (sign) / (print) Paul Date:

Young via MOC ITEM 60000112095 Form retained in accordance with requirements identified in FP-G-RM-01, Quality Assurance Records Control

I'(},

QF0549, Rev. 15 (FP-E-CAL-01) Page 2 of 2 Xcel Energy- I Calculation Signature Sheet Minor Revisions II N/A EC No: ID Vendor Calc:

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Superseded Calculations:

Facility Calc Document Number Title Does the Calculation:

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TABLE OF CONTENTS CALCULATION 22-035 REV 2 Item No. Number of Pages QF0549-Calculation Cover Sheet 2 Table of Contents 1 Calculation 18 Total = 21 pages

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Structural Integrity Associates, Inc.

Project No.: 2100300 Quality Program Type: Nuclear Commercial CALCULATION PACKAGE PROJECT NAME:

MNGP Subsequent License Renewal TLAA CONTRACT NO.:

4000020831 CLIENT: PLANT:

Xcel Energy, Inc. Monticello Nuclear Generating Plant CALCULATION TITLE:

Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts Project Manager Preparer(s) &

Document Affected Revision Description Approval Checker(s)

Revision Pages Signature & Date Signatures & Date 0 1 - 14 Initial Issue Prepared By:

A A-2 Approved By:

Jianxin Wang Daniel B. Denis, PE 07/19/2022 07/19/2022 Checked By:

Sam Ranganath 07/19/2022 1 10-14 Adjusted the 3/4T Approved By: Prepared By:

fluence value results from dpa to attenuation Daniel B. Denis, PE Jianxin Wang method and initial RTNDT 07/27/2022 07/27/2022 values for plates I-16 and I-17 Checked By:

Sam Ranganath 07/27/2022

2 6-8, 11-15 Calculate circumferential Approved By: Prepared By:

A-2 weld VCBB-3 separately.

Incorporated client Daniel B. Denis, PE Jianxin Wang comments on text on 09/16/2022 09/16/2022 Page 6, 7 and 8.

Checked By:

Sam Ranganath 09/16/2022 3 2, 3, 7, 10, 12, Remove proprietary Approved By: Prepared By:

13, 14 marks/notes per References [17] [18] Daniel B. Denis, PE Jianxin Wang 03/14/2023 03/14/2023 Checked By:

Sam Ranganath 03/14/2023 4 7, 8, 12, 13, Correct the 0T fluence Approved By: Prepared By:

14 value for shell course 1 in Table 2. Correct the I s~~ ~--w~

term for VCBB-2 and Stephen Parker Jianxin Wang VCBB-3 in Table 2-4. 05/23/2023 05/23/2023 Checked By:

ZJJ~ ~

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Table of Contents 1.0 INTRODUCTION......................................................................................................... 5

2.0 METHODOLOGY ........................................................................................................ 5

3.0 DESIGN INPUTS ........................................................................................................ 7

4.0 ASSUMPTIONS .......................................................................................................... 8

5.0 CALCULATIONS......................................................................................................... 9

6.0 CONCLUSIONS .......................................................................................................... 9

7.0 REFERENCES.......................................................................................................... 10

APPENDIX A SUPPORTING FILES ....................................................................................A-1

List of Tables Table 1. Maximum E > 1.0 MeV Neutron Fluence for Monticello RPV Beltline Region at 72 EFPY ................................................................................................................... 12

Table 2. Surface ART Values for Monticello RPV Components at 72 EFPY ........................ 13

Table 3. 1/4T ART Values for Monticello RPV Components at 72 EFPY ............................. 14

Table 4. 3/4T ART Values for Monticello RPV Components at 72 EFPY ............................. 15

List of Figures Figure 1 Monticello RPV Beltline Region at 72 EFPY [16] .................................................... 16

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1.0 INTRODUCTION

Radiation embrittlement of reactor pressure vessel (RPV) materials causes a decrease in fracture toughness. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2 (RG1.99R2) [1]

describes general procedures to evaluate the effects of neutron irradiation embrittlement on the alloy steel used in RPVs. In order to perform this evaluation, RG1.99R2 requires calculation of Adjusted Reference Temperature (ART) and Reference Temperature Shift (¨RTNDT) values. The ART values are then used to determine the local fracture toughness of the RPV wall and pressure-temperature limits, according to ASME Code, Section XI, Non-mandatory Appendix G [2] evaluations.

In 2011, SI performed a calculation of the ART and ¨RTNDT values developed for all MNGP plates, welds and nozzles exposed to fluence levels greater than 1.0x1017 n/cm2 [3]. Those calculations were 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 effective full power years (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).

The purpose of this calculation is to develop 1/4T and 3/4T ART and ¨RTNDT values for each MNGP RPV ferritic material exposed to end-of-life for fluence greater than 1.0x1017 n/cm2 at the projected fluence levels for 80 years (72 EFPY) with updated fluence values [16].

2.0 METHODOLOGY When surveillance data are limited or unavailable, RG1.99R2 [1] specifies that ART is calculated with the following equation:

(1)

The Initial RTNDT term refers to the reference temperature of nil ductility transition for the non-irradiated material.

The reference temperature shift, RTNDT, is defined in RG1.99R2 [1] as the shift in the reference temperature resulting from neutron irradiation. RTNDT is calculated from the product of the chemistry factor (CF) and fluence factor (FF) as follows:

'RTNDT CF FF (2)

The CF is a function of the weight percent copper (Cu) and weight percent nickel (Ni) of the weld and base metal (plate or forging) materials. Tables 1 and 2 of RG1.99R2 [1] provide the standard CF values used in this calculation.

The FF is based on the accumulated fast neutron exposure (E > 1 MeV) and is typically corrected by the thickness at the location of interest. The FF can be read directly from Figure 1 of RG1.99R2, or calculated using the following equation [1]:

FF f 0.280.10log( f ) (3)

Due to attenuation effects, the fluence decreases with distance into the RPV wall. Per RG1.99R2 [1],

the calculated or measured fluence from the inside surface of the RPV is attenuated using the following formula:

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f f surf e 0.24 x (4)

Where: f = fast neutron fluence (1019 n/cm2, E > 1 MeV) fsurf = fast neutron fluence at the RPV inside surface (i.e., at base metal / cladding interface, same units as f) x = depth into the RPV wall from the inside surface (inches)

For ASME Code, Section XI, non-mandatory Appendix G [2] evaluations, the x value is taken at one-quarter of the base metal thickness (1/4T) and three-quarter thickness of the base metal (3/4T). The fast neutron fluence can be attenuated through the stainless steel cladding on the inside surface of the RPV. By design, however, the cladding is treated purely as a lining, and not as a load-bearing member. Thus, for the purposes of this evaluation, the inside surface neutron fluence is considered to be at the base metal / cladding interface.

Margin (M), a conservative term defined in RG1.99R2 [1], accounts for uncertainty in the initial reference temperature and for variance in RTNDT. The margin is calculated using the following formula:

Margin 2 V I 2  V ' 2 (5)

Where: I = the standard deviation for the initial RTNDT (°F)

= the standard deviation for RTNDT (°F)

RG1.99R2 [1] states that the standard value of is 28°F for welds and 17°F for base metal (plates or forgings), and need not exceed 0.5 times the mean reference temperature shift (0.5

• RTNDT).

The I term, which is related to the uncertainty in the precision of the Initial RTNDT, is applied for values that are determined by measurement and also when generic or default values are used. For MNGP components where a I value is not explicitly identified, I is assumed to be equal to 0°F (Assumption 1 in Section 4.0) from heat-specific data.

When surveillance data exist (e.g., the ISP Representative Material or other Supplemental Surveillance Program (SSP) material) containing an identical match for the heat number of the vessel beltline material being evaluated, a separate procedure is used to evaluate the ART. This procedure first determines the credibility of the data and, using best estimate chemistry values, calculates a fitted CF.

The fitted CF is then compared to the Table CF (defined above in Equation 2), and the greater of the two is used in subsequent ART calculations. If the surveillance data are credible, the margin () may be cut in half, as specified in Section 2.1 of RG1.99R2 [1]. Detailed procedures to evaluate surveillance data in the manner described above can be found in Section 3 of Reference [4].

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3.0 DESIGN INPUTS The MNGP RPV is constructed of a series of plates, numbered 10 through 17 from top to bottom [7].

Two plates are joined at each elevation via circumferential and vertical welds. As shown in Figure 1, from Reference [16], at 72 EFPY the upper elevation of the RG1.99R2 fluence threshold (1.0x1017 n/cm2) is 190.48 inches above the bottom of active fuel (BAF) and the top of the beltline at 72 EFPY is at an elevation of 383.72 inches. Reference [7] specifies that the weld separating the lower intermediate shell plates (14 and 15) from the upper intermediate shell plates (12 and 13) is located at an elevation of 366.125 inches. Therefore, the upper intermediate plates must also be included in the ART evaluation.

The chemical composition of the MNGP RPV plates is obtained from several sources. The nickel content of the lower plates (A0946-1) and upper intermediate plates (C2089-1) is obtained from Reference [8]. The copper content of the lower plates is obtained from Table 4-1 of Reference [9].

Copper content is not available for the upper-intermediate plates; for conservatism, the bounding value of 0.35% copper specified in Section 1.1 of RG1.99R2 [1] is applied to these components (Assumption 2 in Section 4.0). of Reference [10] specifies updated copper and nickel values for the lower intermediate plates (C2220-1); these values supersede any prior information for these components. Reference [10]

also specifies a fitted chemistry factor of 180.0°F based on the BWRVIP-135, Rev. 4 ISP data, which exceeds the default chemistry factor specified in the tables of Reference [1]. According to the discussion in the Attachment to Reference [10], the surveillance data used to determine the modified chemistry factor is credible. Therefore, the margin term is cut in half for the lower intermediate plates.

The 120° capsule information was unavailable at original authorship of this calculation. The fitted CF utilizing that information has been confirmed to be lower than the 180.0°F CF value, so the current analysis is conservative for Subsequent License Renewal. The value is currently EPRI proprietary and has not been directly referenced.

Initial RTNDT values for the MNGP RPV plates are obtained from Table 5-1 of Reference [11]. In certain cases, multiple values are provided, based on different evaluation methods that are equally relevant. In such cases, it is assumed that selecting the minimum reported value is applicable for the ART calculations Assumption 3 in Section 4.0.

The vertical and circumferential welds that join the RPV plates must also be considered during the ART evaluation. Information on specific welds is not available; rather, Reference [12] provides parameters for a bounding beltline weld. Chemical composition information for the beltline weld is provided in Table 4-1 of Reference [12]. As described in Sections 3.1 and 3.2 of the same document, the Initial RTNDT value for the bounding beltline weld is calculated from 45 tests performed on a sample specimen. The average calculated value is -65.6°F, with a standard deviation of 12.7°F. For the ART evaluation, these values are applied as the Initial RTNDT and I, respectively. These data have been publicly docketed (in submittals and in RVID2) and are considered non-proprietary.

According to the drawing in Reference [7], the centerline N2 recirculation inlet nozzles in the MNGP RPV are located at an elevation of 186 inches above the bottom of the reactor vessel. According to Reference [16], at 72 EFPY the lower elevation of the 1.0x1017 n/cm2 fluence threshold corresponds to an RPV elevation of 190.48 inches. However, the elevation of the uppermost blend radius of the N2 File No.: 2100300.302 Page 7 of 16 Revision: 4 F0306-01R4

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nozzle is within the beltline, as shown in Appendix E of Reference [6]. Therefore, the N2 nozzles must also be included in the ART evaluation for 72 EFPY.

Similar to the upper intermediate shell plates, documentation of the copper content of the N2 nozzles is not available. Section 3.2 of Reference [13] provides a conservative estimate of copper content of 0.18% based on a statistical evaluation of beltline nozzles in other BWR plants (Assumption 4 in Section 4.0). Nickel content for each nozzle is identified in the RPV test reports in Reference [14]. The average of the reported values is 0.86%; this value, the best-estimate nickel content, is used to determine an N2 ART value. The Initial RTNDT value is obtained from Table 5-2 of Reference [11],

where a value of 40°F is common to all of the N2 nozzles.

Based on the boundary of the extended beltline [16] and examination of the RPV drawing [7], the N2 nozzles are the only forged nozzles in the extended beltline at 72 EFPY. There are no instrument nozzles in the extended beltline at 72 EFPY.

The maximum projected fluence levels for 72 EFPY at surface were taken from the latest report [16] but the vessel thickness for the Monticello RPV beltline region materials of 5.0625 [7] at base metal and cladding interface is used for fluence attenuation calculation. Although the fluence levels in the RPV at the 1/4T and 3/4T depths through the vessel thickness calculated per Equation 4 are less than the values in Reference [16], the methodology is consistent with the previous licensing document [11].

4.0 ASSUMPTIONS The assumption made in order to define the evaluation approach and perform the analysis are summarized in the following list. The application of these assumptions is indicated throughout the document using a set of parentheses containing the appropriate assumption number; for example, Assumption #3 would be indicated as (Assumption 3 in Section 4.0).

1. According to RG1.99R2, the I term is equal to the standard deviation of the Initial RTNDT when that quantity is estimated from physical measurements [1]. However, for the MNGP evaluation, a number of components do not have a measured Initial RTNDT; rather, a bounding value is estimated via alternative means. Values calculated by this method include substantial conservatism, rendering it unnecessary to create additional conservatism via the I term.

Consequently, for MNGP ART calculations, I is set equal to zero unless the Initial RTNDT for the component in question is estimated directly from measured data (e.g., in the case of the welds) or another source documents the specific I term to utilize.

2. The copper content of the MNGP upper intermediate RPV shell plates is not documented.

RG1.99R2 states that in cases where chemical composition is unknown, a conservative value of 0.35% copper may be used [1]. This approach is used herein to evaluate the ART values for the upper intermediate plates.

3. The Initial RTNDT values listed in Tables 5-1 and 5-2 of Reference [11] are calculated by one of four different methods, as described in the footnotes accompanying the tables. In many cases, the values reported in Reference [11] have been conservatively increased from the estimated value. Additionally, multiple evaluation methods are often applicable for a particular RPV component. All of the methods are valid, so it is assumed that the minimum initial RTNDT value reported for each component may be used for the ART evaluation. The values obtained by application of this assumption are consistent with those in MNGPs licensing basis documents

[15].

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4. Documentation of the copper content of the MNGP N2 nozzles is unavailable. However, this information is available for beltline nozzles at other BWR plants. Section 3.2 of Reference [13]

offers an estimate of the copper content in nozzle forgings by means of statistical evaluation of available industry forging data. It is assumed that this approach is conservative and therefore applicable for the purposes of MNGP ART calculations.

5.0 CALCULATIONS The methodology in Section 2.0 is used to evaluate the ART and RTNDT values for MNGP, based on the design inputs in Section 3.0 and consistent with the assumptions in Section 4.0. The design inputs, and resultant 0T, 1/4T and 3/4T ART values are given in Table 2, Table 3, and Table 4 for 72 EFPY.

6.0 CONCLUSION

S This document contains ART and RTNDT values calculated in accordance with RG1.99R2 [1] for all MNGP plates, welds, and forgings exposed to fluence greater than 1.0x1017 n/cm2. Design inputs are collected from a variety of sources, as discussed in Section 3.0. The calculated ART and RTNDT values at 0T, 1/4T and 3/4T are provided for 72 EFPY in Table 2, Table 3, and Table 4.

The bounding 0T ART value for the RPV plates and welds is 197.8°F and for the N2 nozzles is 123.9°F at 72 EFPY. The bounding 1/4T ART value for the RPV plates and welds is 182.7°F and for the N2 nozzles is 116.6°F at 72 EFPY. The bounding 3/4T ART value for the RPV plates and welds is 154.3°F and for the N2 nozzles is 100.5°F at 72 EFPY.

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7.0 REFERENCES

1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, Appendix G, Fracture Toughness Criteria for Protection Against Failure, 2004 Edition.

3. SI Calculation 1000847.301, Rev. 0, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts, Performed for Monticello P-T Curves Revision According to the PTLR Methodology, January 2011.

4. BWRVIP-135, Revision 4: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA, 2021. 3002020996.. EPRI PROPRIETARY INFORMATION.

5. Not used.

6. GE Hitachi Nuclear Energy Report No. NEDC-33307P, Pressure-Temperature Curves for Nuclear Management Company LLC Monticello Nuclear Generating Plant, Revision 0, February 2008, GE PROPRIETARY INFORMATION, SI File No. 1000847.203P.

7. Chicago Bridge and Iron Company Drawing No. 1, Revision 8, General Plan, 172 I.D. x 63-2 Ins Heads Nuclear Reactor, NX-8290-13, SI File No. NSP-21Q-210.

8. Chicago Bridge and Iron Company Drawing No. R-7, Revision 0, Skirt Knuckle, Heads & Shell

& Misc Heat Number Summary for 17-2 ID x 63-2 INS. HDS. Nuclear Reactor, NX-8290-133, SI File No. NSP-21Q-213.

9. GE Nuclear Energy Report No. SASR 88-99, Implementation of Regulatory Guide 1.99, Revision 2 for the Monticello Nuclear Generating Plant, Revision 1, January 1989, SI File No.

NSP-21Q-202.

10. Letter from B. Carter (EPRI) to D. Potter (MNGP), Evaluation of the Monticello 300° Surveillance Capsule Data, BWR Vessel and Internals Project (BWRVIP), March 23, 2009, EPRI PROPRIETARY INFORMATION, SI File No. 1000207.202P.

11. Structural Integrity Associates, Inc. Report No. SIR-97-003, Review of the Test Results of Two Surveillance Capsules, and Recommendations for the Materials Properties and Pressure-Temperature Curves to be used for the Monticello Reactor Pressure Vessel, Revision 3, March 1999, SI File No. NSP-21Q-401.

12. GE Nuclear Energy Report No. SASR 87-61, Revision of Pressure-Temperature Curves to Reflect Improved Beltline Weld Toughness Estimate for the Monticello Nuclear Generating Plant, Revision 1, December 1987, SI File No. NSP-21Q-201.

13. BWRVIP-173: BWR Vessel and Internals Project, Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2007. 1014995. EPRI PROPRIETARY INFORMATION.

14. Pressure Vessel Record, Exhibit D, Certified Test Reports, Author, Date, and Revision Not Identified, SI File No. NSP-21Q-233.

15. Monticello Nuclear Generating Plant, Application for Renewed Operating License, Docket No.

50-263, March 2005.

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16. Monticello Nuclear Generating Station Reactor Pressure Vessel Fluence Evaluation -

Subsequent License Renewal, TransWare Enterprises, MNT-FLU-001-R-002, Revision 0. June 2022. SI File NO. 2100300.201.

17. Email from Nathan Palm (EPRI) to Max Smith (XCEL ENERGY), dated 03/09/2023, RE: Epri Confirmation of Classifying, SI File No 2100300.210.

18. Email from Max Smith (XCEL ENERGY) to Dan Denis (SI), dated 03/10/2023, FW: BWRVIP 135 Rev 4 Proprietary Issue, SI File No 2100300.210.

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Table 1. Maximum E > 1.0 MeV Neutron Fluence for Monticello RPV Beltline Region at 72 EFPY Beltline Peak I.D. Fluence 1/4T Fluence 3/4T Fluence Component n/cm2 [16] n/cm2 n/cm2 Lower Shell (Course 1) 3.79E+18 2.80E+18 1.52E+18 Lower/Int Shell (Course 2) 5.94E+18 4.38E+18 2.39E+18 Upper/Int Shell (Course 3) 3.23E+17 2.38E+17 1.30E+17 Lower (Course 1) Axial Welds 2.35E+18 1.73E+18 9.45E+17 (VLAA-1 and VLAA-2)

Lower- Int. (Course 2) Axial Welds 2.10E+18 1.55E+18 8.44E+17 (VLBA-1 and VLBA-2)

Upper/Int Shell (Course 3) 2.12E+17 1.56E+17 8.52E+16 (VLCB-1 and VLCB-2)

Horizontal Weld 3.79E+18 2.80E+18 1.52E+18 (VCBA-2)

Horizontal Weld 3.23E+17 2.38E+17 1.30E+17 (VCBB-3)

N2 Nozzles 7.08E+17 5.23E+17 2.85E+17 Notes:

1. Thickness of 5.0625 from base metal and cladding interface is used for fluence attenuation at 1/4T and 3/4T.

2. The fluence values at 1/4T and 3/4T are calculated using the attenuation method per Equation (4) to be consistent with the previous licensing document [11].

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Table 2. Surface ART Values for Monticello RPV Components at 72 EFPY 72EFPY Initial 72 EFPY Co mpo nent 0T Fluence RT ND T i Heat Lo t  % Cu  % Ni CF RT ND T 0T ART No . Fluence Facto r f (°F) (°F) (°F)

(°F) 2 (°F)

(n/cm )

Lo wer Shell Plates (Co urs e 1)

I-16 A0946-1 N/A 0.14 0.56 98 27 3.79E+18 0.732 71.8 0 17.0 132.8 I-17 C2193-1 N/A 0.17 0.5 119 0 3.79E+18 0.732 86.7 0 17.0 120.7 Lo wer-Intermediate Shell Plates (Co urse 2)

I-14 C2220-1 N/A 0.16 0.64 180 27 5.94E+18 0.854 153.8 0 8.5 197.8 I-15 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 (Co urse 3)

I-12 C2089-1 N/A 0.35 0.5 200 0 3.23E+17 0.229 45.7 0 17.0 79.7 I-13 C2613-1 N/A 0.35 0.49 198 27 3.23E+17 0.229 45.5 0 17.0 106.5 Lo wer Shell (Co urse 1) Axial W elds 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 Lo wer-Intermediate Shell (Co urs e 2) Axial W elds:

VLBA-1 &

- E8018N 0.1 0.99 135 -65.6 2.10E+18 0.581 78.4 12.7 28.0 74.3 VLBA-2 Upper/Int Shell (Co urs e 3) Axial W elds:

VLCB-1 &

- E8018N 0.1 0.99 135 -65.6 2.12E+17 0.178 24.1 12.7 12.0 -6.6 VLCB-2 Circ umferential W elds VCBA-2 - E8018N 0.1 0.99 135 -65.6 3.79E+18 0.732 98.7 12.7 28.0 94.6 VCBB-3 - E8018N 0.1 0.99 135 -65.6 3.23E+17 0.229 30.9 12.7 15.5 5.3 N2 No zzle N2 Nozzle E21VW N/A 0.18 0.86 142 40 7.08E+17 0.351 49.9 0 17.0 123.9 File No.: 2100300.302 Page 13 of 16 Revision: 4 F0306-01R4 13 Structural Integrity Associa/es, Inc.* info@structint.com m 1-877-45!-POWER " structint.com @)

Table 3. 1/4T ART Values for Monticello RPV Components at 72 EFPY 72EFPY Initial 72 EFPY Co m po nent 1/4T Fluenc e RT ND T i Heat Lo t  % Cu  % Ni CF RT ND T 1/4T ART No . Fluenc e Fac to r f (°F) (°F) (°F)

(°F) 2 (°F)

(n/cm )

Lo wer Shell Plates (Co urs e 1)

I-16 A0946-1 N/A 0.14 0.56 98 27 2.80E+18 0.653 64.1 0 17.0 125.1 I-17 C2193-1 N/A 0.17 0.5 119 0 2.80E+18 0.653 77.3 0 17.0 111.3 Lo wer-Intermediate Shell Plates (Co urse 2)

I-14 C2220-1 N/A 0.16 0.64 180 27 4.38E+18 0.770 138.7 0 8.5 182.7 I-15 C2220-2 N/A 0.16 0.64 180 27 4.38E+18 0.770 138.7 0 8.5 182.7 Upper/Int Shell Plates (Co urse 3)

I-12 C2089-1 N/A 0.35 0.5 200 0 2.38E+17 0.191 38.2 0 17.0 72.2 I-13 C2613-1 N/A 0.35 0.49 198 27 2.38E+17 0.191 37.9 0 17.0 98.9 Lo wer Shell (Co urs e 1) Axial W elds VLAA-1 & VLAA-

- E8018N 0.1 0.99 135 -65.6 1.73E+18 0.535 72.2 12.7 28.0 68.1 2

Lo wer-Intermediate Shell (Co urse 2) Axial W elds:

VLBA-1 & VLBA-

- E8018N 0.1 0.99 135 -65.6 1.55E+18 0.510 68.8 12.7 28.0 64.7 2

Upper/Int Shell (Co urs e 3) Axial W elds:

VLCB-1 & VLCB-

- E8018N 0.1 0.99 135 -65.6 1.56E+17 0.147 19.8 12.7 9.9 -13.5 2

Circumferential W elds VCBA-2 - E8018N 0.1 0.99 135 -65.6 2.80E+18 0.653 88.0 12.7 28.0 83.9 VCBB-3 - E8018N 0.1 0.99 135 -65.6 2.38E+17 0.191 25.8 12.7 12.9 -3.6 N2 No z zle N2 Nozzle E21VW N/A 0.18 0.86 142 40 5.23E+17 0.300 42.6 0 17.0 116.6 File No.: 2100300.302 Page 14 of 16 Revision: 4 F0306-01R4 13 Structural Integrity Associa/es, Inc.* info@structint.com m 1-877-45!-POWER " structint.com @)

Table 4. 3/4T ART Values for Monticello RPV Components at 72 EFPY 72EFPY Initial 72 EFPY Co m po nent 3/4T Fluenc e RT ND T i Heat Lo t  % Cu  % Ni CF RT ND T 3/4T ART No . Fluenc e Fac to r f (°F) (°F) (°F)

(°F) 2 (°F)

(n/c m )

Lo wer Shell Plates (Co urs e 1)

I-16 A0946-1 N/A 0.14 0.56 98 27 1.52E+18 0.506 49.7 0 17.0 110.7 I-17 C2193-1 N/A 0.17 0.5 119 0 1.52E+18 0.506 59.9 0 17.0 93.9 Lo wer-Interm ediate Shell Plates (Co urs e 2)

I-14 C2220-1 N/A 0.16 0.64 180 27 2.39E+18 0.613 110.3 0 8.5 154.3 I-15 C2220-2 N/A 0.16 0.64 180 27 2.39E+18 0.613 110.3 0 8.5 154.3 Upper/Int Shell Plates (Co urs e 3)

I-12 C2089-1 N/A 0.35 0.5 200 0 1.30E+17 0.131 26.1 0 13.0 52.1 I-13 C2613-1 N/A 0.35 0.49 198 27 1.30E+17 0.131 25.9 0 13.0 78.8 Lo wer Shell (Co urs e 1) Axial W elds VLAA-1 &

- E8018N 0.1 0.99 135 -65.6 9.45E+17 0.406 54.7 12.7 27.4 49.5 VLAA-2 Lo wer-Interm ediate Shell (Co urs e 2) Axial W elds:

VLBA-1 &

- E8018N 0.1 0.99 135 -65.6 8.44E+17 0.384 51.8 12.7 25.9 43.8 VLBA-2 Upper/Int Shell (Co urs e 3) Axial W elds :

VLCB-1 &

- E8018N 0.1 0.99 135 -65.6 8.52E+16 0.098 13.3 12.7 6.6 -23.7 VLCB-2 Circ um ferential W elds VCBA-2 - I E8018N I 0.1 I 0.99 135 I -65.6 I 1.52E+18 I 0.506 I 68.2 12.7 28.0 I 64.1 VCBB-3 - I E8018N I 0.1 I 0.99 135 I -65.6 I 1.30E+17 I 0.131 I 17.6 12.7 8.8 I -17.1 N2 No z z le N2 Nozzle E21VW I N/A I 0.18 I 0.86 142 I 40 I 2.85E+17 I 0.213 I 30.2 0 15.1 I 100.5 File No.: 2100300.302 Page 15 of 16 Revision: 4 F0306-01R4 13 Structural Integrity Associa/es, Inc.* info@structint.com m 1-877-45!-POWER " structint.com @)

, so* 21 o* 240* 210° 300* 330* o* 30° so* go* 120* 1so* 1ao*

I 11 I I I I I l 11I I I I I I I II I I I I I I I I I ii I I I I ll I I I I I I I I 11I I11I I I II I I I I I 11I I ii I I I I ll N

Shell Course 3 m m N1 2B 0 ~ N~A ~

1===::::::;====:::::t==========;:::====:::t::======I VCBB-3 N

Shell Course 2 d:

(!j

....J

~

....J RPV Boltline Region*

VCBA-2 N2F N2G N8B N2H N2J N2K 0N1A N2A N2B N8A N2C N2D N2E N1B 0 0 Inside View Notes: This drawing 1s not to scale.

RPV belllina region is shown for 72 EFPY Figure 1 Monticello RPV Beltline Region at 72 EFPY [16]

File No.: 2100300.302 Page 16 of 16 Revision: 4 F0306-01R4 ti -1 Structural Integrity Associates, Inc.* info@structint.com m 1-877-4S!-POWER e structint.com ~

APPENDIX A SUPPORTING FILES File No.: 2100300.302 Page A-1 of A-2 Revision: 4 F0306-01R4 e- 1 Structural Integrity Associates. Inc.* info@structint.com m 1-877-4S1-POWER G structint.com (@)

Supporting Files

1. 2100300.302 Rev 4.xlsm Excel file contains the detailed ART calculations File No.: 2100300.302 Page A-2 of A-2 Revision: 4 F0306-01R4 e

Structural Integrity Associates, Inc.~ info@structint.com ~ 1-877-451-POWER '9 structint.com ~

ENCLOSURE 4 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISION TO THE MNGP PRESSURE TEMPERATURE LIMITS REPORT TO CHANGE THE NEUTRON FLUENCE METHODOLOGY AND INCORPORATE NEW SURVEILLANCE CAPSULE DATA MONTICELLO PRESSURE-TEMPERATURE LIMIT CURVES GENERATION FOR 72 EFPY (NSPM CALCULATION NO.23-012)

(SI NO. 2200284.303, REVISION 0)

(78 Pages Follow)

QF0549, Rev. 15 (FP-E-CAL-01) Page 1 of 2 I fl Xcel Energy* I Calculation Signature Sheet Approval: 602000018581 Document Information NSPM Calculation (Doc) No: 23-012 I Revision: 0

Title:

Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY Facility: [g] MT PI I Unit: [g] 1 2

Safety Class: [g] SR Aug Q Non SR Type: Calc Sub-Type:

NOTE: Print and sign name in signature blocks, as required.

Major Revisions N/A I

EC Number: 601000004404 [g] Vendor Calc: 2200284.303 Vendor Name or Code: Structural Vendor Doc No: 2200284.303 Integrity Associates (SIA)

Description of Revision: New calculation issuance The following calculation and attachments have been reviewed and deemed

[g]

acceptable as a legible QA record Prepared by: (sign) Vendor / (print) SIA Date: 12/18/23 Reviewed by: (sign) / (print) Russell Date: See MOC Lidberg MOC 600001117674 600001117674 Type of Review:

Design Verification Engr Review [g] OAR EOC Method Used (For DV Only):

Review Alternate Calc Test Approved by: (sign) / (print) Gus Date: MOC Hernandez MOC 600001117673 600001117673 Form retained in accordance with requirements identified in FP-G-RM-01, Quality Assurance Records Control

QF0549, Rev. 15 (FP-E-CAL-01) Page 2 of 2 I fl Xcel Energy* I Calculation Signature Sheet Minor Revisions II ~ N/A EC No: ID Vendor Calc:

Minor Rev. No:

Description of Change:

Pages Affected:

The following calculation and attachments have been reviewed and deemed acceptable as a legible QA record Prepared by: (sign) / (print) Date:

Reviewed by: (sign) / (print) Date:

Type of Review:

Design Verification Engr Review OAR EOC Method Used (For DV Only):

Review Alternate Calc Test Approved by: (sign) / (print) Date:

Summary of Verification (summary is required for Design Verification):

No Comments See attached QF0528 Superseded Calculations:

Facility Calc Document Number Title MNGP 11-005 Revised P-T Curves Calculation Does the Calculation:

YES No Affect piping or supports? (If YES, Attach MT Form 3544.) MONTI ONLY YES No Require Fire Protection Review? (Using QF2900, Fire Protection Program Impact Screen, determine if a Fire Protection Review is required.) If YES, document the engineering review in the EC. If NO, then attach completed QF2900 to the associated EC.

Form retained in accordance with requirements identified in FP-G-RM-01, Quality Assurance Records Control

QF0528 (FP-E-MOD-02) Rev. 3 l'{l Xcel Energy-1 Design Revie Comment Form Sheet 1 of 1 DOCUMENT NUM ER/ TITLE: File No: 2200284.303P REVISION: 0 DATE: 12/2 /22 ITEM REVIE ER S COMMENTS PREPARER S REVIE ER S RESOLUTION DISPOSITION 1 (RL) Reference 1 lists SIR-05-044 revision ill add a statement Acceptable

1. MNGP Tech-Specs section 5.6.5 at the end of Section lists SIR-05-044-A as the approved 1: The method used in methodology for development of P-T this calculation meets curves. e either need a statement the requirements of both that the methodology used to develop the current revision and this PTLR revision meets the previous revision of SIR-05-044 [1].

requirements of both SIR-05-044 revisions or MNGP will need to do a LAR to update the Tech-Specs.

2 (RL) Section 2 has a repeat sentence In ill delete the Acceptable some cases, a region may contain more than repeated one.

one component which is considered for development of the associated P-T curve.

3 (RL) The 72 EFPY T fluence values in The current 1/4T Acceptable table 2 (pg 13) are different than the fluence values were values listed in the Transware report calculated based on MNT-FLU-001-R-002, table 3-2. attenuation method Please e plain why. per Regulatory Guide 1.99 Revision 2, which is consistent with previous TLAA. It was agreed between SI and Monticello when developing the ART values. The values are close to Transware s report.

Form retained in accordance with record retention schedule identified in FP-G-RM-01.

Page 1 of 2

QF0528 (FP-E-MOD-02) Rev. 3 l'(l Xcel Energy-1 Design Revie Comment Form I

ITEM REVIE ER S COMMENTS PREPARER S REVIE ER S RESOLUTION DISPOSITION Reviewer: ~Lde~ Date: 1/3/23 Preparer: Vendor Date: 1/3/23 Form retained in accordance with record retention schedule identified in FP-G-RM-01.

Page 2 of 2

Calculation 23-010/EC 601000004404: Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY Table of Contents # of Pages QF-0549 Calculation Signature Sheet 2 QF-0528 Design Review Comment Form 2 Table of Contents 1 QF-0547 Suitability Review 2 QF-0545 Design Information Transmittal Form 1 QF2900 Fire Protection Program Impact Screen 12 Calculation 58

QF0547 (FP-E-MOD-11-XE) Rev. 8 Page 1 of 2 External Design Document Suitability 11 /l Xcel Energy* I Review Checklist II External Design Document Being Reviewed:Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY

Title:

Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY Number: 2200284.303 Rev: 0 Date: //23 This design document was received from:

Organization Name: Structural Integrity PO or DIA

Reference:

4000024678 Associates The purpose of the suitability review is to ensure that a calculation, analysis or other design document provided by an External Design Organization complies with the conditions of the purchase order and/or Design Interface Agreement (DIA) and is appropriate for its intended use. The suitability review does not serve as an independent verification. Independent verification of the design document supplied by the External Design Organization should be evident in the document, if required.

The reviewer should use the criteria below as a guide to assess the overall quality, completeness and usefulness of the design document. The reviewer is not required to check calculations in detail.

REVIEW Reviewed N/A 1 Design Inputs used by the External Design Organization are appropriate. ~

2 Assumptions are described and reasonable. ~

3 Applicable codes, standards and regulations are identified and met. ~

4 Applicable construction and operating experience is considered. ~

5 Applicable structure(s), system(s), and component(s) are listed. ~

6 Formulae and equations are documented. Unusual symbols are defined. ~

7 Acceptance criteria are identified, adequate and satisfied. ~

8 Results are reasonable compared to inputs. ~

9 Source documents are referenced. ~

10 The document is appropriate for its intended use. ~

11 The document complies with the terms of the Purchase Order and/or DIA. ~

12 Inputs, assumptions, outputs, etc. which could affect plant operation are ~

enforced by adequate procedural controls.

13 Plant impact has been identified and either implemented or controlled. If not ~

identified in the document itself, identify the plant impacts and their associated tracking A/Rs and descriptions are listed in Table 1.

14 Design and Operational Margin have been considered and documented Comments:

Completed by: Russell Lidberg 600001117674 Date: //23 Form retained in accordance with record retention schedule identified in FP-G-RM-01.

QF0547 (FP-E-MOD-11-XE) Rev. 8 Page 2 of 2 External Design Document Suitability 11 /l Xcel Energy* I Review Checklist II TABLE 1 Initiate an AR to track open items and plant impacts (e.g., procedure revisions, validation of assumptions, database updates, etc.), if any.

Item AR Tracking PLANT IMPACT DESCRIPTION No. Number 1 none 2

3 4

5 6

7 8

9 10 11 12 13 14 15 Form retained in accordance with record retention schedule identified in FP-G-RM-01.

QF0545, Rev 06 (FP-E-MOD-05) Page 1 of 1 I fl Xcel Energy* I Design Information Transmittal DIT Approval:

From: Russell Lidberg, Program Engineering, MNGP cel Energy To : Dan Denis, P.E., Senior Consulaltant, Structural Integrity Associates Mod or Trac ing Number: 600000988024 Date: 12/16/22 DIT No: PTLR-001 Mod

Title:

PTLR Update to 72 EFPY Unit 1 ~ Unit 2 Quality Plant: MNGP Common Classification Safety Related S CT:PT R pdate to FP Check if applicable:

This DIT confirms information previously transmitted orally on - by

~ This information is preliminary. See e planation belo .

S RC F INF RM TI N (Source documents should be uniquely identified)

RVIP Letter 2022-053 D SCRIPTI N F INF RM TI N ( rite the information being transmitted or list each document being transmitted)

Notification of New RVIP Integrated Surveillance Program (ISP) Data Applicable to the Monticello Reactor Pressure Vessel (RPV)

DISTRI TI N (Recipients should receive all attachments unless otherwise indicated. All attachments are uncontrolled unless otherwise indicated)

Dan Denis, P.E., Senior Consulaltant, Structural Integrity Associates PPR D Russell Lidberg Program Engineer ~Lu:/6 ~ 12/16/22 Approver Name Position Signature Si t Date A copy of the DIT (along with any attachments not on file) should be sent to the Modification Folder.

Form retained in accordance with record retention schedule identified in FP-G-RM-01.

QF2900, Rev. 2 (FP-PE-FP-01) Page 1 of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen Approval: 602000013088 Plant MT PI Activity Number: 601000004404 Activity Owner: Russell Lidberg Type of ECR Activity:

rief Description of Activity: Revising the PTLR to incorporate updates to fluence pro ections and surveillance capsule data

References:

PTLR,23-010, 23-011,23-012

1. Fire Protection Impact Considerations (Questions should be assessed for impact to Fire Protection Fundamentals, Calculations, Procedure, Programs, Systems and Components)

Does the Activity:

1. Add, remove, or create an opening (not filled by an approved penetration seal, Yes ~ No door, or damper) in any Fire arrier wall, ceiling , or floor .

2. Add, remove, or modify any fire door, radiant heat shield, thermal shield, or fire Yes ~ No damper with components other than replacement in ind

3. Add, remove, or modify the air flow CFM or the VAC discharge within three Yes ~ No feet of a detector (e.g., air flow can affect the amount of air available for combustion and the timing of fire detection actuation)

4. Permanently add, remove, or modify any fire protective coating or wrap on any Yes ~ No electrical raceway component (e.g., cable, conduit, or cable tray) (Ref the raceway drawings)

5. Add or remove any coatings (i.e., foam insulation, sound dampeners, and floor Yes ~ No coatings)

6. Add, remove, or modify fireproofing or passive fire protection of any structural Yes ~ No steel (Ref the system drawing)

. Add, relocate, or modify any wall, fence, door, building, trailer, or other structure that could affect the access or egress to safe shutdown components Yes ~ No This includes impacts on fire brigade access, emergency lighting, and illumination levels.

8. Potentially affect personnel safety or SSCs within 50 feet of a large power Yes ~ No transformer ( 10 MVA) due to fire or e plosion (including debris, pro ectiles, and blast effect) from a catastrophic failure of the transformer (SOER 10-1)

Yes ~ No 9. Add or remove a component in a FPP credited system

10. Modify the normal operating or failure position of a component in any FPP Yes ~ No credited system (Ref the system P ID)

11. Alter normal or emergency system performance or operational characteristics Yes ~ No associated with a FPP credited system (e.g., flow rates, temperatures, available volumes or capacities, pressures)(Ref the system P ID)

12. Modify any process monitoring (e.g., flow, temperature, level, pressure)

Yes ~ No instrumentation, including tubing or indication for any FPP credited system Yes ~ No 13. Alter normal or available tan inventory water levels in a FPP credited system

14. Add, modify, or remove a component such that any unanaly ed flow bloc age, Yes ~ No flow diversion, or inventory loss path are introduced in any FPP credited system Yes ~ No 15. Alter any line si e or configuration within a FPP credited system

16. Affect a Fire Detection System component (e.g., power supply, cable type or Yes ~ No routing, transmitter, control switch, control module)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page 2 of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen

1. Fire Protection Impact Considerations (Questions should be assessed for impact to Fire Protection Fundamentals, Calculations, Procedure, Programs, Systems and Components)

Does the Activity:

1 . Affect Fire Detection actuation logic, set points, interloc s, or software Yes ~ No applications Yes ~ No 18. Add, delete, or relocate detectors, or change detector type

19. Introduce potential obstructions (including possible overlapping obstructions) to the effective operation of fire suppression sprin lers, halon, or CO2 discharge Yes ~ No no les Obstructions could include addition or relocation of cable trays, VAC ductwor , or panels, which could affect air flow, spray patterns, or bloc access to fire protection equipment.

20. Potentially impact the effectiveness of a gaseous suppression system (Any Yes ~ No change to increase room volume or affect room integrity or nominal room operating temperature.)

21. Affect the location or type of manual fire suppression equipment (e.g., fire Yes ~ No e tinguishers, hose stations, hydrants)

22. Affect access to manual or automatic fire suppression equipment or controls Yes ~ No (e.g., alarms, pull stations, fire e tinguishers, hydrants, hose stations, isolation valves)

Yes ~ No 23. Impact suppression system piping or hangers

24. Affect the performance characteristics of any Fire Suppression system Yes ~ No (E amples include water system flow rate or supply pressure, gaseous system pressure or volume, or system initiation time.)

25. Impact the magnitude of e pected fires by permanently adding or removing Yes ~ No fi ed combustibles or flammable materials Yes ~ No 26. Increase or decrease the quantity of oil in an area 2 . Impact any oil collection system, including the Reactor Coolant Pump Oil Yes ~ No Collection System (if applicable)

28. Add or remove equipment containing oils (i.e., pumps, motors, oil filled Yes ~ No transformers, air compressors)

Yes ~ No 29. Add any cable that is not IEEE-383-19 4 or an approved alternative.

30. Impact the performance or capacity of any fire propagation or water control Yes ~ No features, including curbs, drains, or di es in a FPP credited system

31. Impact Fire Protection programmatic / procedural elements including:

x Control of transient combustibles, including storage of combustibles, control of ha ardous materials, and combustible or flammable gases x Coatings program controls involving coating thic ness increases or combustible ratings x Controls for ignition sources, including the ot or program and temporary Yes ~ No heating devices x FP Impairment logging, trac ing, and compensatory measures x Fire rigade staffing, structure, training/drills, equipment, communications, pre-fire plans, fire-related operating procedures, and off-site firefighting assistance x Fire protection surveillance procedures Yes ~ No 32. Affect fire brigade training related to controlling the release of radioactivity

33. Impact the communication systems credited for use by the Fire rigade or by Yes ~ No Operations during Post-Fire SSD activities

34. Install or reconfigure any ma or plant structures (e.g., walls, floors) that could Yes ~ No impact the credited radio systems effectiveness Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page 3 of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen

1. Fire Protection Impact Considerations (Questions should be assessed for impact to Fire Protection Fundamentals, Calculations, Procedure, Programs, Systems and Components)

Does the Activity:

35. Affect fire brigade capability by impacting access to suppression equipment or accessibility to any fire area Consider any change in building access, egress, Yes ~ No paths of travel, and change in door status from normally unloc ed to normally loc ed. (Assume a large fire fighter wearing breathing apparatus and encumbered with equipment.)

36. Impact to the ISFSI Fire a ards Analysis including:

x The receiving location, physical location or quantity of any combustible gases or flammable or combustible liquids stored in tan s or contained in plant equipment x Combustible loading within the ISFSI protected area fence Yes ~ No x Equipment used to load or transport the cas onto the transit vehicle or to the ori ontal Storage Module ( SM) x The path of travel for loaded cas s x Security procedures for allowing access into the Owner Controlled Area (OCA) for delivery truc s containing flammable or combustible liquids or gases NEIL Impact (reference FP-E-NEIL-01) 3 . Impact a NEIL insured Structure, fire protection system or component x Add a new structure x Change the occupancy classification of any part of a NEIL insured structure x Add a new NEIL required fire protection system x Add, modify, or remove an e isting NEIL required fire detection or fire suppression system x Create an addition to an e isting NEIL insured structure x Replace roof dec ing or covering x Does the change affect an interior finish such that it would not meet NEIL Yes ~ No requirements x Reduce the fire rating of a NEIL required fire rated barrier x Add to, renovate, or alter the fire protection water supply or distribution systems, or use the fire protection water supply and distribution systems for other than emergency use x Add oil filled components over 50 gallons oil capacity, or increase the oil capacity of an e isting component greater than 50 gallons x Add to, renovate, or alter oil collection systems, fire barriers, or fire protection systems for oil filled components Is there a potential impact to Classical FP program requirements as indicated by a Yes No Yes on any of questions 1-36 above (if yes, provide details below)

If a potential impact is identified, does the activity ma e a change to the Fire Protection Yes No Program (FP Program engineer document review below)

Is NEIL Impacted y this Change (as indicated by a Yes on question 3 Yes No above) (If so forward this form to the Fire Marshal to complete Section 8)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page 4 of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen

1. Fire Protection Impact Considerations (Questions should be assessed for impact to Fire Protection Fundamentals, Calculations, Procedure, Programs, Systems and Components)

I Does the Activity:

2. Post-Fire Safe Shutdown Capability Considerations (Questions should be assessed for impact to achieving and maintaining safe shutdown in the event of a fire)

Does the Activity Yes ~ No 1. Add or remove a component in a NSCA/Appendi R credited system

2. Modify the normal operating or failure position of a component in any Yes ~ No NSCA/Appendi R credited system

3. Add, modify, or remove a component such that any unanaly ed flow bloc age, Yes ~ No flow diversion, or inventory loss path are introduced in any NSCA/Appendi R credited system Yes ~ No 4. Alter any line si e or configuration within a NSCA/Appendi R credited system

5. Add a branch line or modify an e isting branch line that may affect the Yes ~ No mechanical boundary of any NSCA/Appendi R credited system

6. Alter normal or emergency system performance or operational characteristics Yes ~ No associated with a NSCA/Appendi R credited system (e.g., flow rates, temperatures, available volumes or capacities, pressures)

. Modify any process monitoring (e.g., flow, temperature, level, pressure)

Yes ~ No instrumentation, including tubing or indication for any NSCA/Appendi R credited system

8. Alter normal or available tan inventory water levels in a NSCA/Appendi R Yes ~ No credited system

9. Impact any NSCA/ Appendi R room heat up calculations or thermal hydraulic Yes ~ No analyses Yes ~ No 10. Modify or replace an MOV actuator in any NSCA/Appendi R credited system

11. Add, delete or modify the cable si e, cable design (including use of spare Yes ~ No terminals or conductors), or cable routing for any NSCA/Appendi R credited system

12. Modify the control, power, indication or annunciation circuit for any Yes ~ No NSCA/Appendi R credited component/system Yes No 13. Alter switchyard brea er alignments or interconnects

14. Alter the communication systems credited for use by the Fire rigade or by Yes ~ No Operations during Post-Fire SSD activities

15. Install or reconfigure any ma or plant structures (e.g., walls, floors) that could Yes ~ No impact the credited radio systems effectiveness

16. Permanently add, remove, or modify any fire protective coating or wrap on any Yes ~ No electrical raceway component (e.g., cable, conduit, or cable tray) 1 . Modify or impact any performance characteristic of any fi ed eight hour battery Yes ~ No emergency lighting unit required by Appendi R (MT only)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page 5 of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen

18. Introduce potential obstructions to emergency lighting units or their associated Yes ~ No postfire shutdown access or egress paths

19. Is any plant equipment being modified such that electrical coordination is not Yes ~ No being maintained

20. Is Engineering udgment being used to achieve electrical coordination (i.e.,

Yes ~ No crediting cable length for time current characteristic curves with overlap)

Is there a potential impact to the NSCA/ Appendi R Safe Shutdown Capability as Yes ~ No indicated by a Yes on any of questions 1-20 above (if yes, provide details below)

If a potential impact is identified, does the activity ma e a change to the Yes ~ No NSCA/ Appendi R Safe Shutdown Capability (FP Program engineer document review below)

3. O . Section 3-6 do not apply to Monticello.

Non-Power Operations Assessment Considerations (Questions should be assessed for impact to NPO fundamentals, Strategies, Procedures, Calculations, Analysis, Systems and Components.)

Does the Activity:

Yes No 1. Add, or remove a component in a NPO credited system

2. Modify the normal operating or failure position of a component in any NPO Yes No credited system

3. Add, modify, or remove a component such that any unanaly ed flow bloc age, Yes No flow diversion, or inventory loss path are introduced in any NPO credited system Yes No 4. Alter any line si e or configuration within a NPO credited system

5. Add a branch line or modify an e isting branch line that may affect the Yes No mechanical boundary of any NPO credited system

6. Alter normal or emergency system performance or operational characteristics Yes No associated with a NPO credited system (e.g., flow rates, temperatures, available volumes or capacities, pressures)

. Modify any process monitoring (e.g., flow, temperature, level, pressure)

Yes No instrumentation, including tubing or indication for any NPO credited system Yes No 9. Modify or replace an MOV actuator in any NPO credited system No 8. Alter normal or available tan inventory water levels in a NPO credited system Yes

10. Add, delete or modify the cable si e, cable design (including use of spare Yes No terminals or conductors), or cable routing for any NPO credited system

11. Modify the control, power, indication or annunciation circuit for any NPO Yes No credited component/system Yes No 12. Alter switchyard brea er alignments or interconnects

13. Alter the communication systems credited for use by the Fire rigade or by Yes No Operations during Post-Fire SSD activities

14. Install or reconfigure any ma or plant structures (e.g., walls, floors) that could Yes No impact the credited radio systems effectiveness

15. Permanently add, remove, or modify any fire protective coating or wrap on any Yes No electrical raceway component (e.g., cable, conduit, or cable tray)

16. Is any plant equipment being modified such that electrical coordination is not Yes No being maintained 1 . Is Engineering udgment being used to achieve electrical coordination (i.e.,

Yes No crediting cable length for time current characteristic curves with overlap)

I Yes I No Is there a potential impact to the Non-Power Operations (NPO) analysis as indicated by a Yes on any of questions 1-1 above (if yes, provide details below)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page 6 of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen Yes I No I Operations (NPO) analysis If a potential impact is identified, does the activity ma e a change to the Non Power (FP Program engineer document review below)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page of 12 Ill Xcel Energy* I Fire rote tion rogr m m t reen

4. O . Section 3-6 do not apply to Monticello.

Radioactive Release Considerations (Questions should be assessed for impact to Rad Release fundamentals, Fire Fighting Strategies, Training, Procedures, Calculations, Analysis, Engineering Controls, Systems and Components.)

Does the Activity:

Yes No 1. Affect VAC flow rates and paths within Radiation Control Areas

2. Affect the ability to control or monitor the release of radioactive materials during Yes No fire suppression activities Yes No 3. Affect fire brigade training related to controlling the release of radioactivity

4. Permanently remove a penetration seal, add a new seal with an unapproved Yes No seal type or material, or replace a penetration seal with an unapproved seal type or material for an RCA

5. Add, remove, or create an opening (not filled by an approved penetration seal, Yes No door, or damper) in any fire barrier wall, ceiling, or floor in an RCA

6. Impact the performance or capacity of any fire propagation or water control Yes No features, including curbs, drains, or di es in a FPP credited system

. Add or remove a potential containment release path. Evaluation should include Yes No potential intersystem LOCA paths.

Is there a potential impact to Radioactive Release considerations as indicated by a Yes No Yes on any of questions 1- above (if yes, provide details below)

If a potential impact is identified, does the activity ma e a change to the Fire Yes No Protection Program credited Radioactive Release Considerations (FP Program engineer document review below)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

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5. O . Section 3-6 do not apply to Monticello.

NFPA 805 Section Methodology Requirements or Previously Approved Alternatives (Questions should be assessed for impact to NSCA fundamentals, Fire Modeling, Procedures, Calculations, Analysis, and Evaluations.)

Does the Activity:

Yes NoNo 1. Impact the methodology of NFPA 805 section 2.4 Yes No a. Fire Modeling (2.4.1)

Yes No b. Nuclear Safety Capability Assessment (2.4.2)

Yes No c. Fire Ris Evaluation (2.4.3)

Yes

d. Plant Change Evaluation (2.4.4)

Is there a potential impact to NFPA 805 Methodology requirements (if yes, provide Yes No details below)

If a potential impact is identified, are NFPA 805 Methodology requirements or Yes No Previously Approved Alternatives impacted by this change (FP Program engineer document review below)

For any YES answers above provide details how the activity impacts the NFPA 805 Methodology:

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

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6. O . Section 3-6 do not apply to Monticello.

Fire PRA (Questions should be assessed for impact to Rad Release fundamentals, Fire Modeling, Procedures, Calculations, Analysis, and Evaluations.)

Does the Activity:

Yes No 1. Add or remove a component in a FPRA credited system

2. Modify the normal operating or failure position of a component in any FPRA Yes No credited system

3. Add, modify, or remove a component such that any unanaly ed flow bloc age, Yes No flow diversion, or inventory loss path are introduced in any FPRA credited system

4. Alter normal or emergency system performance or operational characteristics Yes No associated with a FPRA credited system (e.g., flow rates, temperatures, available volumes or capacities, pressures)

5. Modify any process monitoring (e.g., flow, temperature, level, pressure)

Yes No instrumentation, including tubing or indication for any FPRA credited system Yes No 6. Modify or replace an MOV actuator in any FPRA credited system

. Add, delete or modify the cable si e, cable design (including use of spare Yes No terminals or conductors), or cable routing for any FPRA credited system

8. Add any cable that is not IEEE-383-19 4 or an approved alternative per NFPA Yes No 805 FAQ 06-0022 Rev 3

9. Modify the control, power, indication or annunciation circuit for any FPRA Yes No credited component/system Yes No 10. Alter switchyard brea er alignments or interconnects

11. Modify or impact any performance characteristic of any fi ed eight hour battery Yes No emergency lighting unit

12. Introduce potential obstructions to emergency lighting units or their associated Yes No post-fire shutdown access or egress paths Yes No 13. Add a new ignition source

14. Modify, relocate, remove, or change the equipment name of an e isting ignition Yes No source

15. Alter the physical dimensions or cabinet venting characteristics of an e isting Yes No ignition source

16. Add, modify, or remove ventilation or cable penetration openings in an electrical Yes No cabinet (e.g., electrical panel, unction bo , MCC, switchgear)

Yes No 1 . Introduce permanent intervening combustibles or flammable materials Yes No 18. Increase or decrease the quantity of oil in an area

19. Alter any oil collection system, including the Reactor Coolant Pump Oil Yes No Collection System (if applicable)

20. Add or remove equipment containing oils (i.e., pumps, motors, oil filled Yes No transformers, air compressors)

21. Does the change add, delete, or relocate fire detectors, or change fire detector Yes No type

22. Does the change add, delete, change type of fi ed suppression system, or Yes No change effective one covered by the suppression system

23. Affect the performance characteristics of any Fire Suppression system Yes No (E amples include distance from suppression or system initiation time.)

Yes No 24. Add, remove, or create an opening in any wall, ceiling, or floor

25. Permanently add, remove, or modify any fire protective coating or wrap on any Yes No electrical raceway component (e.g., cable, conduit, or cable tray)

26. Add, remove, or modify fireproofing or passive fire protection of any structural Yes No steel Form retained in accordance with the records retention schedule identified in FP-G-RM-01

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6. O . Section 3-6 do not apply to Monticello.

Fire PRA (Questions should be assessed for impact to Rad Release fundamentals, Fire Modeling, Procedures, Calculations, Analysis, and Evaluations.)

Does the Activity:

2 . Add, remove, or modify any fire door, radiant heat shield, thermal shield, or fire Yes No damper with components other than replacement in ind

28. Add, relocate, or modify any wall fence, door, building, trailer, or other structure Yes No that could affect the access or egress to FPRA components

29. Result in an increase or decrease in the amount of open floor space in a room Yes No or area

30. Impact the performance or capacity of any fire propagation or water control Yes No features, including curbs, drains, or di es in a FPRA credited system

31. Alter any local or Control Room instrumentation or controls, including changes Yes No to layout or function, of a credited FPRA system Yes No 32. E ceed the ma fill capacity of any cable tray

33. Affect plant operating procedures such as altering normal or emergency systems operation or alignments (including offsite power) or Operations Yes No responses to abnormal (including fire and annunciators) or emergency conditions

34. Is any plant equipment being modified such that electrical coordination is not Yes No being maintained

35. Is engineering udgment being used to achieve electrical coordination (i.e.,

Yes No crediting cable length for time characteristic curves with overlap)

Is there a potential impact to the Fire PRA as indicated by a Yes on any of questions Yes No 1-35 above (if yes, provide details below)

If a potential impact is identified, does the activity ma e a change to the Fire Yes No Protection Program credited Fire PRA (FP Program engineer document review below)

Form retained in accordance with the records retention schedule identified in FP-G-RM-01

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. C D R Does the Activity:

Yes ~ No re irements re im te t e nge t en

a. Forward the form to the Fire Marshal to complete Section 8.

otenti im t s not i enti ie in e tions t en

a. Activity Owner / RE shall sign and date below as preparer.

b. No additional FP Program evaluation is required.

c. hen used with an engineering change. this form should be retained with the engineering change pac age.

d. 10 CFR 50.59 Applicability Determination should be mar ed No for affecting FPP.

otenti im t s i enti ie in e tions t en ition revie is re ire

a. Activity Owner / RE shall sign and date below as preparer.

b. Contact the Fire Protection Program Engineer to review if the activity ma es a change to the Fire Protection Program t eF rogr m ngineer revie in e tions etermines t t t e tivit m es nge to t e F

a. FP Program Engineer shall sign and date below as the reviewer

b. Initiate a Fire Protection Change Review (QF2901) or Fire Protection Change Evaluation (QF2902), as applicable.

c. 10 CFR 50.59 Applicability Determination should be mar ed Yes for affecting FPP.

If required to be retained per FP-PE-FP-01, QF2900 Fire Protection Program Impact Screen S ALL be retained for the life of the plant with the change pac age.

R Preparer(s):

Russell Lidberg / ~Ld6Vlfr 600001117674 /23 Print Name / Sign Date Fire Protection Program Engineer (if required):

n/a Print Name / Sign Date Form retained in accordance with the records retention schedule identified in FP-G-RM-01

QF2900, Rev. 2 (FP-PE-FP-01) Page 12 of 12 Ill Xcel Energy* I Fire Protection Program Impact Screen

8. N I Revie This section is to be completed by the Fire Marshal/ Fire Protection Coordinator when the NEIL property loss impact question at the end of Section 1 is marked yes.

If a planned addition, renovation or alteration involves a structure, system or component that is or will be insured by NEIL (See FP-E-NEIL-01 for list of applicable SSCs) and the change is permanent (in place over 180 days) and any of the following questions are answered yes, a NEIL design review is required in accordance with FP-E-NEIL-01 and the NEIL Loss Control Standards. Does the change activity:

x Add a new structure?

x Change the occupancy classification of any part of a NEIL insured structure?

x Add a new NEIL required fire protection system?

x Add, modify, or remove an e isting NEIL required fire detection or fire suppression system?

x Create an addition to an e isting NEIL insured structure?

x Replace roof dec ing or covering?

x Does the change affect an interior finish such that it would not meet NEIL requirements?

x Reduce the fire rating of a NEIL required fire rated barrier?

x Add to, renovate, or alter the fire protection water supply or distribution systems, or use the fire protection water supply and distribution systems for other than emergency use?

x Add oil filled components over 50 gallons oil capacity, or increase the oil capacity of an e isting component greater than 50 gallons?

x Add to, renovate, or alter oil collection systems, fire barriers, or fire protection systems for oil filled components?

If any of the above questions were answered yes , provide details of compliance with NEIL requirements, any NEIL deviations and NEIL acceptance below.

Fire Marshal:

n/a Print Name / Sign Date Fire Protection Program ngineer if required :

Print Name / Sign Date Form retained in accordance with the records retention schedule identified in FP-G-RM-01

File No.: 2200284.303 e

Structural Integrity Associates, Inc.

Project No.: 2200284 Quality Program Type: Nuclear Commercial CALCULATION PACKAGE PROJECT NAME:

Monticello P-T Limit Curves Generation for 72 EFPY (with SLR TLAA Synergy)

CONTRACT NO.:

4000024678 CLIENT: PLANT:

Xcel Energy Monticello Nuclear Generating Plant CALCULATION TITLE:

Monticello Pressure-Temperature Limit Curves Generation for 72 EFPY Project Manager Preparer(s) &

Document Affected Revision Description Approval Checker(s)

Revision Pages Signature & Date Signatures & Date 0 1 - 34 Initial Issue Approved By: Prepared By:

(303P.R0) A A-3 ~-w~

B B-15 C C-4 ~ Jianxin Wang 12/30/2022 Mo Uddin, PhD D D-2 12/30/2022 Checked By:

J:'JJJ ~

Daniel B. Denis, PE 12/30/2022 303.R0 1, 2, 12, 13, EPRI Proprietary Footers, Information has been Approved By:

D-2 determined to be Non-Proprietary based on J:'JJJ ~

discussion with Nathan Daniel B. Denis, PE Palm and BWRVIP 12/18/2023 2023-039

This page has been intentionally left blank.

Table of Contents 1.0 INTRODUCTION ......................................................................................................... 5

2.0 METHODOLOGY ........................................................................................................ 5

3.0 ASSUMPTIONS ........................................................................................................ 11

4.0 DESIGN INPUTS ...................................................................................................... 12

5.0 CALCULATIONS....................................................................................................... 14

5.1 Pressure Test (Curve A) ................................................................................ 15

5.2 Normal Operation - Core Not Critical (Curve B) ............................................ 15

5.3 Normal Operation - Core Critical (Curve C) ................................................... 15

5.4 Overall Composite Curves ............................................................................ 16

6.0 CONCLUSIONS ........................................................................................................ 16

7.0 REFERENCES.......................................................................................................... 17

APPENDIX A P-T CURVE INPUT LISTING ........................................................................A-1

APPENDIX B SUPPORTING CALCULATIONS ..................................................................B-1

APPENDIXC DEVELOPMENT OF SATURATION STEAM CURVE FITS ........................ C-1

APPENDIX D SUPPORTING FILES................................................................................... D-1

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List of Tables Table 1. Summary of Minimum Temperature Requirements for P-T Limit Curves. .............. 10

Table 2. Updated MNGP 72 EFPY 1/4T ART Calculation with Latest ISP CF Value.............. 13

Table 3. MNGP Beltline Region, Curve A, for 72 EFPY ....................................................... 19

Table 4. MNGP Bottom Head Region, Curve A, for 72 EFPY .............................................. 20

Table 5. MNGP Non-Beltline Region, Curve A, for 72 EFPY ............................................... 21

Table 6. MNGP Overall Composite Curve, Curve A, for 72 EFPY ........................................ 22

Table 7. MNGP Beltline Region, Curve B, for 72 EFPY ....................................................... 23

Table 8. MNGP Bottom Head Region, Curve B, for 72 EFPY .............................................. 24

Table 9. MNGP Non-Beltline Region, Curve B, for 72 EFPY ............................................... 25

Table 10. MNGP Overall Composite Curve, Curve B, for 72 EFPY ...................................... 26

Table 11. MNGP Beltline Region, Curve C, for 72 EFPY..................................................... 27

Table 12. MNGP Bottom Head Region, Curve C, for 72 EFPY ........................................... 28

Table 13. MNGP Non-Beltline Region, Curve C, for 72 EFPY ............................................. 29

Table 14. MNGP Overall Composite Curve, Curve C, for 72 EFPY...................................... 30

List of Figures Figure 1. MNGP P-T Curve A (Hydrostatic Pressure and Leak Test), 72 EFPY .................. 31

Figure 2. MNGP P-T Curve B (Normal Operation - Core Not Critical), 72 EFPY ................. 32

Figure 3. MNGP P-T Curve C (Normal Operation - Core Critical), 72 EFPY ....................... 33

Figure 4. MNGP Overall Composite Curves A, B, and C, 72 EFPY ...................................... 34

File No.: 2200284.303 Page 4 of 34 Revision: 0 F0306-01R4

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1.0 INTRODUCTION

Pressure-temperature (P-T) limit curves for the beltline, bottom head, and non-beltline regions of the Monticello Nuclear Generating Plant (MNGP) reactor pressure vessels (RPV) were developed for 54 effective full power years (EFPY) in Reference [13]. This calculation updates the P-T curves for 72 EFPY of operation. The P-T curves are prepared using the method documented in the Boiling Water Reactor Owners Group (BWROG) Licensing Topical Reports (LTRs), Pressure Temperature Limits Report Methodology for Boiling Water Reactors [1] which satisfies the requirements of 10CFR50 Appendix G [3] and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Nonmandatory Appendix G [4]. The method used in this calculation meets the requirements of both the current revision and previous revision of SIR-05-044 [1].

2.0 METHODOLOGY A full set of P-T curves, applicable to the following plant conditions, are prepared:

1. Pressure Test (Curve A),

2. Normal Operation - Core Not Critical (Curve B), and

3. Normal Operation - Core Critical (Curve C).

For each plant condition above, separate curves are provided for each of the following three regions of the RPV as well as a composite curve for the entire RPV:

1. The beltline region (includes nozzles where 1/4T fluence > 1 x 1017 n/cm2),

2. The bottom head region,

3. The non-beltline region, including the top head flange,

4. Composite curve (bounding curve for all regions)

In some cases, a region may contain more than one component which is considered for development of the associated P-T curve. For the beltline region, the P-T curves incorporate components with fluence >

1 x 1017 n/cm2 (E > 1 MeV). The instrument nozzles are not in the beltline region per Reference [19] and will not be included in the P-T curves evaluations. The Feedwater nozzle is assumed to be the bounding component for non-beltline, see Assumption 2 in Section 3.0. For MNGP, the curve for each vessel region identified above is composed from the bounding P-T limits determined from the following:

1. Beltline:

a. Beltline shell

b. Recirculation inlet nozzle, N2

2. Non-beltline

a. Feedwater (FW) nozzle

b. 10CFR50 Appendix G limits [3]

3. Bottom Head:

a. Bottom head penetrations (in-core monitor housings, control rod drive housings)

Consequently, separate P-T curves are prepared for each component considered for each region, then a bounding curve is drawn from the individual P-T curves. Complete sets of P-T curves, as identified above, are provided for 72 EFPY of operation for the limiting Service Level A/B (Normal/Upset) thermal transient.

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The methodology for calculating P-T curves, described below, is taken from Reference [1] unless specified otherwise.

The P-T curves are calculated by means of an iterative procedure, in which the following steps are performed:

Step 1: A fluid temperature, T, is assumed. The P-T curves are calculated considering a postulated flaw with a 6:1 aspect ratio that extends 1/4 of the way through the vessel wall. The temperature at the postulated flaw tip is conservatively assumed equal to the coolant temperature.

Step 2: The static fracture toughness, KIc, is computed using the following equation from [4]:

K Ic 33 .2  20 .734 e 0.02 T  ART (1)

Where: KIc = the lower bound static initiation critical fracture toughness (ksi¥in).

T = the metal temperature at the tip of the postulated 1/4T flaw (°F).

ART = the Adjusted Reference Temperature (ART) for the limiting material in the RPV region under consideration (°F).

Step 3: The allowable stress intensity factor due to pressure, KIp, is calculated as:

K Ic  K It K Ip (2)

SF Where: KIp = the allowable stress intensity factor due to membrane (pressure) stress (ksi¥in).

KIc = the lower bound static fracture toughness calculated in Eq. (1)

(ksi¥in).

KIt = the thermal stress intensity factor (ksi¥in) from through wall thermal gradients.

SF = the ASME Code recommended safety factor, based on the reactor condition. For hydrostatic and leak test conditions (i.e., P-T Curve A), SF = 1.5. For normal operation, both core non-critical and core critical (i.e., P-T Curves B and C), SF = 2.0.

When calculating values for Curve A, the thermal stress intensity factor is neglected (KIt = 0),

since the hydrostatic leak test is performed at or near isothermal conditions.

For Curve B and Curve C calculations, KIt is computed in different ways based on the evaluated region. For the beltline, with the exception of nozzles, and bottom head regions, KIt is determined using the following equation [4] for a postulated inside surface connected flaw:

K It 0.953 u 10 3 CR t 2.5 (3)

Where: CR = the cooldown rate of the vessel (°F/hr).

t = the RPV wall thickness (in).

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For the FW nozzle/upper vessel region and the N2 recirculation nozzle, KIt is obtained from the stress distribution output of plant-specific finite element analyses (FEA). A polynomial curve-fit is determined for the through-wall stress distribution at the bounding time point. The linear elastic fracture mechanics (LEFM) solution for KIt is obtained from Reference [1]:

2a a2 4a 3 º K It _ Poly Sa <<0.706C0t  0.537C1t  0.448C 2t  0.393C3t >> (4)

¬ S 2 3S 1/4 Where: a = 1/4T postulated flaw depth, a = 1/4 t (in).

t = thickness of the cross-section through the nozzle at the limiting path near the inner blend radius (in).

C0t,C1t, = thermal stress polynomial coefficients, obtained from a curve-C2t,C3t fit of the extracted stresses from a transient FEA [11, 12].

The thermal stress polynomial coefficients are based on the assumed polynomial form of V x C0  C1 x  C2 x2  C3 x3 . In this equation, x represents the radial distance in inches from the inside surface to any point on the crack face.

The transient FEA is performed assuming a fixed thermal shock between a high and a low temperature. In reality, the actual thermal shock varies for each evaluation step, as the maximum temperature is bounded by the pressure-temperature saturation curve. Thus, the value of KIt calculated in Equation 4 can be scaled to account for the maximum thermal shock, as shown in the following expression:

§ T  Tlow

• K It K It _ Poly Fscaling K It _ Poly ¨ sat ¸ (5)

¨T T ¸

© high low ¹ Where: KIt = the scaled thermal stress intensity factor, which is subsequently used in Equation 2 ( ksi inch )

KIt_Poly = the thermal stress intensity factor computed from the polynomial expression defined in Equation 4 ( ksi inch )

Fscaling = the scaling factor to apply to the polynomial stress intensity factor Tsat = the saturation temperature of the reactor (°F)

Tlow = the lower limit of the thermal shock applied to the FEA (°F)

Thigh = the upper limit of the thermal shock applied to the FEA (°F)

Tsat is determined from the pressure-temperature saturation curve. A power fit of this curve is developed in Appendix C, resulting in the following equation:

1 0.2198 Tsat 119.3 0.7987 Psat Psat (6)

In the above equation, Psat is the saturation pressure corresponding to Tsat. For the purposes of this evaluation, Psat is conservatively applied at the final P-T curve pressure (PP-T), which is calculated below in Equation 12. This results in an iterative calculation File No.: 2200284.303 Page 7 of 34 Revision: 0 F0306-01R4 13 Structural Integrity Associates, Inc.* info@structint.com ~ 1-877-451-POWER C., structint.com (@)

process for each evaluation step, where a saturation pressure is assumed, a scaling factor is determined, the final pressure is computed, and the assumed saturation pressure is adjusted until the results achieve a suitable level of convergence.

Step 4: The allowable internal pressure of the RPV is calculated differently for each evaluation region.

For the beltline region, with the exception of nozzles, the allowable pressure is determined as follows:

K Ip t Pallow *1000 (7)

M m Ri Where: Pallow = the allowable RPV internal pressure (psig).

KIp = the allowable stress intensity factor due to membrane (pressure) stress, as defined in Eq. (2) (ksi¥in).

t = the RPV wall thickness (in).

Mm = the membrane correction factor for an inside surface axial flaw:

Mm = 1.85 for ¥t < 2 Mm = 0.926 ¥t for 2 ¥t 3.464 Mm = 3.21 for ¥t > 3.464.

Ri = the inner radius of the RPV, per region (in).

For the bottom head region, the allowable pressure is calculated with the following equation:

2 K Ip t Pallow *1000 (8)

SCF M m Ri Where: SCF = conservative stress concentration factor to account for bottom head penetration discontinuities; SCF = 3.0 per Reference [1].

Pallow, KIp, t, Mm and Ri are defined in Eq. (7).

For the FW nozzle/ upper vessel region, and the N2 nozzle, the allowable pressure is determined from a ratio of the allowable and applied stress intensity factors. The applied factor can be determined from an FEA that determines the stresses due to the internal pressure on the nozzle and RPV. The methodology for this approach is as follows:

K Ip Pref Pallow (9)

K Ip  app Where: Pref = RPV internal pressure at which the FEA stress coefficients (Eq.

(10)) are determined (psi).

KIp-app = the applied pressure stress intensity factor (ksi¥in).

Pallow and KIp are defined as in Eq. (7).

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The applied pressure stress intensity factor for the FW nozzle and N2 nozzle is determined using a polynomial curve-fit approximation for the through-wall pressure stress distribution from a plant-specific FEA and the LEFM solution given in Eq. (8) [1]:

2a a2 4a 3 º K Ip  app Sa <<0.706C 0 p  0.537C1 p  0.448C 2 p  0.393C3 p >> (10)

¬ S 2 3S 1/4 Where: a = 1/4 through-wall postulated flaw depth, a = 1/4 t (in).

t = thickness of the cross-section through the limiting nozzle inner blend radius corner (in).

C0p,C1p, = pressure stress polynomial coefficients, obtained from a curve-C2p,C3p fit of the extracted stresses from an FEA.

Step 5: Steps 1 through 4 are repeated in order to generate a series of P-T points; the fluid temperature is incremented with each repetition. Calculations proceed in this iterative manner until 1,300 psig is reached. This value bounds the design pressure given in Section 4.

Step 6: Table 1 below summarizes the minimum temperature requirements contained in 10CFR50, Appendix G [3, Table 1], which are applicable to the material highly stressed by the main closure flange bolt preload (non-beltline curve). Additional minimum temperature requirements for bolt-up are included as shown in Table 1 below.

Note that the minimum bolt-up temperature of 60°F, is used here, consistent with the position given in Reference [1]. Further, some utilities specifically request that the minimum moderator temperature used in the plant shutdown margin evaluation be applied as a minimum bolt-up temperature requirement; it is also included in Table 1 but not required by MNGP. An additional 60°F margin is recommended in 10 CFR50, Appendix G [3, Table 1]. For P-T Curves A and B, this 60°F margin is only a recommendation, but for Curve C, the 60°F margin is required.

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Table 1. Summary of Minimum Temperature Requirements for P-T Limit Curves.

Pressure Curve Minimum Metal Temperature P-T Limits Range Maximum of: ASME Appendix G [4]

x RTNDT,max, requirements P < 20% Ph x 60°F [1],

A x TSDM ASME Appendix G [4]

P > 20% Ph RTNDT,max + 90°F requirements Maximum of: ASME Appendix G [4]

x RTNDT,max, requirements P < 20% Ph x 60°F [1],

B x TSDM ASME Appendix G [4]

P > 20% Ph RTNDT,max + 120°F requirements Maximum of: ASME Appendix G [4]

x RTNDT,max + 60°F, requirements + 40°F P < 20% Ph x 60°F [1],

C x TSDM Maximum of: ASME Appendix G [4]

P > 20% Ph x RTNDT,max + 160°F, requirements + 40°F x TISHT Where: Ph is the pre-service hydrotest pressure, 1563 psig for MNGP [8].

RTNDT,max is the maximum RTNDT of the vessel materials highly stressed by the bolt preload.

TSDM is the temperature used in the shutdown margin evaluation.

TISHT is the minimum temperature at which the maximum in-service hydrotest pressure (1025 psig) [8] is allowed per Curve A.

Step 7: Uncertainty in the RPV pressure and metal temperature measurements is incorporated by adjusting the P-T curve pressure and temperature using the following equations:

T P T T UT (11)

PPT Pallow  PH UP (12)

Where: TP-T = The allowable coolant (metal) temperature (°F).

UT = The coolant temperature instrument uncertainty (°F).

PP-T = The allowable reactor pressure (psig).

PH = The pressure head to account for the water in the RPV (psig).

Can be calculated from the following expression: PH U ' h .

= Water density at ambient temperature (lbm/in3).

h = Elevation of full height water level in RPV (in).

UP = The pressure instrument uncertainty (psig).

Steps 1 through 7, above, are implemented for all components and in all regions.

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Nozzles in the beltline introduce stress concentration effects and have the potential to be more limiting than the generic beltline P-T curves. Nozzles or discontinuities outside the beltline are considered to be bounded by the upper vessel / feedwater nozzle or bottom head region P-T curves [1]. Beltline nozzles may be bounded by the upper vessel / feedwater nozzle curve if all of the following are met: the feedwater nozzle experiences more severe thermal transients, the feedwater nozzle RTNDT is greater than or equal to the beltline nozzle ART, and the beltline and feedwater nozzle have similar transition geometry (blend radius).

The P-T Curves for hydrostatic leak test (Curve A) and normal operation - core not critical (Curve B) may be computed by following Steps 1 through 7. Values for Curve C, the core-critical operating curve, are generated from the requirements of 10CFR50 Appendix G [3] and the Curve A and Curve B limits. Table 1 of Reference [3] requires that core critical P-T limits be 40°F above any Curve A or Curve B limits at all pressures. 10CFR50 Appendix G [3] also stipulates that, above the 20% pressure transition point, the Curve C temperatures must be either the reference temperature (RTNDT) of the closure flange region plus 160°F, or the temperature required for the hydrostatic pressure test, whichever is greater.

For P-T Curves A and B, the initial fluid temperature assumed in Step 1 is typically taken at the bolt-up temperature of the closure flange minus coolant temperature instrument uncertainty. According to Reference [3], the minimum bolt-up temperature is equal to the limiting material RTNDT of the regions affected by bolt-up stresses. Consistent with Reference [1], the minimum bolt-up temperature shall not be lower than 60°F. Thus, the minimum bolt-up temperature shall be 60°F or the material RTNDT, whichever is higher.

For P-T Curve C, when the reactor is critical, the initial fluid temperature is equal to the calculated minimum criticality temperature in this region. Table 1 of Reference [3] indicates that, for a BWR with normal operating water levels, the allowable temperature for initial criticality at the closure flange region is equal to the reference temperature (RTNDT) at the flange region plus 60°F.

3.0 ASSUMPTIONS The 10CFR50 Appendix G [3] and ASME Code [4] requirements and methods are considered to be supported in their respective technical basis documentation. Therefore, the assumptions inherent in the ASME B&PV Code methods utilized for this evaluation are not specifically identified and justified in this calculation. Only those assumptions specific to this calculation are identified and justified here. The following assumptions are used in preparation of the MNGP P-T curves:

1. The full-vessel height is used in the calculation of the static head contributed by the coolant in the RPV.

This assumption is conservative in that the static head at the non-beltline regions is slightly lower than that of the bottom head curve; however, the difference in static head is small.

Therefore, the added complexity in considering different static head values for each region of the vessel is not considered beneficial.

2. The FW nozzle is the bounding non-beltline component of the RPV.

This assumption is made because:

a. The geometric discontinuity caused by the nozzle penetration in the RPV shell causes a stress concentration which results in larger pressure induced stresses than would be calculated in the shell regions of the RPV.

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b. The FW nozzle experiences more severe thermal transients than most of the other nozzles because of the feedwater injection temperature [5], which causes larger thermal stresses than are experienced in the shell region of the RPV.

c. Although some other nozzles can experience thermal transients, which would cause thermal stresses larger than those calculated for the shell regions of the RPV, and some nozzles are larger diameter than the FW nozzle, which could result in a slightly larger KIp, the combined stresses from the applied thermal and pressure loads are considered to bound all other non-beltline discontinuities [1, Section 2.5.3].

3. Application of a SCF = 3.0 to the membrane pressure stress in the bottom head bounds the effect of the bottom head penetrations on the stress field in this region of the vessel.

Bottom head penetrations will create geometric discontinuities in the bottom head hemisphere resulting in high localized stresses. This effect must be considered in calculating the stress intensity factor from internal pressure. Rather than performing a plant-specific analysis, SI applies a conservative SCF for a circular hole in a flat plate subjected to a uniaxial load to the membrane stress in the shell caused by the internal pressure. The assumption of SCF = 3.0 is conservative because:

a. It applies a peak SCF to the membrane stress which essentially intensifies the stress through the entire shell thickness and along the entire crack face of the postulated flaw rather than intensifying the stress local to the penetration and considering the stress attenuation away from the penetration,

b. Review of SCFs for circular holes in plates subjected to an equi-bi-axial stress state as well as SCFs for arrays of circular holes in shells, shows that the SCF is likely closer to 2-2.5 rather than 3.0 [5].

Consequently, the method utilized by SI is expedient, as intended, and conservatively bounds the expected effect of bottom head penetrations because a bounding SCF is used and applied as a membrane stress correction factor.

4.0 DESIGN INPUTS The design inputs, also included in Appendix A, used to develop the MNGP P-T curves are identified below.

1. Limiting RTNDT and ART:

Non-beltline RTNDT: 40°F [6, Table 3]

(Bounding RTNDT for non-beltline region, excluding bottom head.)

Closure Flange RTNDT: 10°F [6, Table 3]

(Bounding RTNDT for materials highly stressed by bolt preload)

Bottom Head RTNDT: 26°F [6, Table 3]

Quarter T Recirculation Inlet (N2) Nozzle ART (72 EFPY): 116.6°F [10, p. 13]

Quarter T Beltline ART (72 EFPY): The limiting 1/4T beltline ART value was calculated to be 182.7°F [10, p. 13] for plates heat C2220 with a chemistry factor (CF) of 180 from Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [15]. However, the latest ISP data shows that the CF value for plate heat C2220 changes to 174 [16].

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The corresponding limiting 1/4T beltline ART value changed to 178.1°F as shown in the updated 1/4T ART calculations in Table 2 below.

(The limiting ART value of all beltline materials including plates and welds, used for P-T limit curve calculations for each EFPY)

Table 2. Updated MNGP 72 EFPY 1/4T ART Calculation with Latest ISP CF Value.

72 72EFPY Initial EFPY Component  % 1/4T Fluence RTNDT i Heat Lot  % Cu CF RTNDT 1/4T No. Ni Fluence Factor f (°F) (°F) (°F)

(°F) ART (n/cm2)

(°F)

Lower Shell Plates (Course 1)

A0946-I-16 N/A 0.14 0.56 98 27 2.80E+18 0.653 64.1 0 17.0 125.1 1

C2193-I-17 N/A 0.17 0.5 119 0 2.80E+18 0.653 77.3 0 17.0 111.3 1

Lower-Intermediate Shell Plates (Course 2)

C2220- 174 I-14 N/A 0.16 0.64 27 4.38E+18 0.770 134.1 0 8.5 178.1 1 [16]

C2220- 174 I-15 N/A 0.16 0.64 27 4.38E+18 0.770 134.1 0 8.5 178.1 2 [16]

Upper/Int Shell Plates (Course 3)

C2089-I-12 N/A 0.35 0.5 200 0 2.38E+17 0.191 38.2 0 17.0 72.2 1

C2613-I-13 N/A 0.35 0.49 198 27 2.38E+17 0.191 37.9 0 17.0 98.9 1

Lower Shell (Course 1) Axial Welds VLAA-1 &

- E8018N 0.1 0.99 135 -65.6 1.73E+18 0.535 72.2 12.7 28.0 68.1 VLAA-2 Lower-Intermediate Shell (Course 2) Axial Welds:

VLBA-1 &

- E8018N 0.1 0.99 135 -65.6 1.55E+18 0.510 68.8 12.7 28.0 64.7 VLBA-2 Upper/Int Shell (Course 3) Axial 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 - E8018N 0.1 0.99 135 -65.6 2.80E+18 0.653 88.0 0 28.0 78.4 VCBB-3 - E8018N 0.1 0.99 135 -65.6 2.38E+17 0.191 25.8 0 12.9 -14.0 N2 Nozzle N2 Nozzle E21VW N/A 0.18 0.86 142 40 5.23E+17 0.300 42.6 0 17.0 116.6 Notes:

1. All from values are the same from Reference [10, Table 3] except the bold highlighted values. The CF values of shell plates I-14 and I-15 are based on Reference [16].

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2. Minimum Bolt-Up Temperature:

Bolt-Up Temperature: 60°F [13, p. 8]

3. RPV Dimensions:

Full vessel height: 758 inches [7]

(Used to calculate maximum water head during pressure test and conservatively applied for normal operation as well.)

RPV inside radius: 103.1875 inches [7] (to base metal)

RPV shell thickness: 5.0625 inches [7] (to base metal)

Bottom head inside radius: 103.1875 inches [7] (to base metal)

Bottom head shell thickness: 5.9375 inches [7] (to base metal)

4. Heat-up / Cool-down Rate: 100°F/hr [9, p. A-7]

5. Quarter T Nozzle Stress Intensity Factors:

FW Nozzle [11, Table 7]:

1000 psi Internal Pressure: 70.59 ksi-in0.5 Limiting thermal transient: 10.37 ksi-in0.5 Recirculation Inlet (N2) Nozzle [12, Table 13]:

1010 psi Internal Pressure: 75.20 ksi-in0.5 Limiting thermal transient: 25.28 ksi-in0.5

6. Operating Pressure Design Pressure: 1250 psig [8]

Normal Operating Pressure 1025 psig [8]

7. Hydro-test pressure:

Pre-Service: 1563 psig (i.e. 1.25*Design pressure)

In-Service: 1025 psig (i.e. 1.0*Normal operating pressure)

8. Applicable ASME XI Code Year [4]: 2004 Edition [13]

5.0 CALCULATIONS The P-T curves in this calculation were developed using an Excel spreadsheet listed in Appendix D, which is independently verified for use on a project-specific basis in accordance with SIs Nuclear QA program [17, 18]. P-T limits are evaluated for 72 EFPY. P-T limits are calculated from 0 to 1300 psig.

Supporting calculations for all P-T curves are included in Appendix B and represent the P-T curves for individual RPV components. The tabulated results in Table 3 through Table 14 present bounding composite P-T curves for the three RPV regions (beltline, non-beltline, and bottom head). As discussed in Section 5.1 for instance, the beltline curve A in Table 3 bounds three underlying component curves (beltline shell, Feedwater nozzles, and N2 Recirculation Inlet nozzles), shown in Table B-1 through Table B-3.

The bottom head methodology for calculating the allowable pressure shown in Eq. (8), using an SCF of 3.0 to account for bottom head penetration discontinuities, is applied for the thinner side plates of the MNGP bottom head, which bounds the thicker portion of the bottom head center plates with respect to the resulting P-T limits.

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5.1 Pressure Test (Curve A)

The minimum bolt-up temperature of 60°F minus instrument uncertainty (0°F) is applied to all regions as the initial temperature in the iterative calculation process. The static fracture toughness (KIc) is calculated for all regions using Equation (1). The resulting value of KIc, along with a safety factor of 1.5 is used in Equation (2) to calculate the pressure stress intensity factor (KIp). The allowable RPV pressure is calculated for the beltline, bottom head and upper vessel regions using Equations (7, 8, and 9), as appropriate. For the feedwater nozzle / upper vessel region, the additional constraints specified in Step 6 of Section 2.0 are applied. Final P-T limits for temperature and pressure are obtained from Equations (11) and (12), respectively.

Since the thermal stress intensity factor is taken as zero for Curve A, the cool-down rate does not affect the results for Curve A.

Values for the beltline region curves for 72 EFPY are listed in Table 3. Data for the bottom head region curve for 72 EFPY is listed in Table 4. Data for the non-beltline (feedwater nozzle / upper vessel) region curve for 72 EFPY is listed in Table 5. The data for each region is plotted, and the resulting composite Curve A for 72 EFPY is provided in Figure 1 and tabulated in Table 6. Additional data and curves for each region are included in Appendix B.

5.2 Normal Operation - Core Not Critical (Curve B)

The minimum bolt-up temperature of 60°F for MNGP minus coolant temperature instrument uncertainty (0°F), is applied to all regions as the initial temperature in the iterative calculation process. The static fracture toughness (KIc) is calculated for all regions using Eq. (1). The thermal stress intensity factor (KIt) is calculated for the FW nozzle and N2 recirculation Inlet nozzle using Eq. (4).

The resulting values of KIc and KIt, along with a safety factor of 2.0, are used in Eq. (2) to calculate the pressure stress intensity factor (KIp). The allowable RPV pressure is calculated for the beltline, bottom head, and non-beltline regions using Eq. (7, 8, and 9), as appropriate. For the non-beltline (FW nozzle /

upper vessel) region, the additional constraints specified in Step 6 of Section 2.0 are applied. Final P-T limits for temperature and pressure are obtained from Eq. (11 and 12), respectively.

The data resulting from each P-T curve calculation is tabulated. Values for the beltline region at 72 EFPY are listed in Table 7. Data for the bottom head region are listed in Table 8. Data for the non-beltline (feedwater nozzle / upper vessel) region are listed in Table 9. The data for each region is plotted, and the resulting data for composite Curve B for 72 EFPY is provided in Figure 2 and tabulated in Table 10. Additional data and curves for each region are included in Appendix B.

5.3 Normal Operation - Core Critical (Curve C)

The pressure and temperature values for Curve C are calculated in a similar manner as Curve B, with several exceptions. The initial evaluation temperature is calculated as the limiting non-beltline RTNDT that is highly stressed by the bolt preload (in this case, that of the closure flange region: 10°F per Section 4.0) plus 60°F, resulting in a minimum criticality temperature of 70°F). When the pressure exceeds 20% of the pre-service system hydrostatic test pressure (20% of 1,563 psig = 313 psig), the P-T limits are specified as 40°F higher than the Curve B values. The minimum temperature above the 20% of the pre-service system hydrostatic test pressure is always greater than the reference temperature (RTNDT) of the closure region plus 160°F or is taken as the minimum temperature required for the hydrostatic pressure test. The final Curve C values are taken as the absolute maximum between the regions of the beltline, the bottom head, and the non-beltline.

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The data resulting from each P-T curve calculation is tabulated. Values for the beltline region at 72 EFPY are listed in Table 11. Data for the bottom head region are listed in Table 12. Data for the non-beltline (FW nozzle / upper vessel) region are listed in Table 13. The data for each region is plotted, and the resulting data for composite Curve C for 72 EFPY is provided in Figure 3 and tabulated in Table 14.

Additional data and curves for each region are included in Appendix B.

5.4 Overall Composite Curves Overall composite curves A, B, and C are plotted in Figure 4.

6.0 CONCLUSION

S P-T curves are developed for MNGP using the methodology, assumptions, and design inputs defined in Sections 2.0, 3.0, and 4.0, respectively. P-T curves are developed for the beltline, bottom head, and non-beltline regions, considering limiting thermal transients at 72 EFPY, for the following plant conditions: Pressure Test (Curve A), Normal Operation - Core Not Critical (Curve B), and Normal Operation - Core Critical (Curve C). Tabulated pressure and temperature values are provided for all regions and EFPY in Table 3 through Table 14. The accompanying P-T curve plots are provided in Figure 1 through Figure 4.

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7.0 REFERENCES

1. Licensing Topical Report (LTR) BWROG-TP-11-022-A (SIR-05-044), Revision 1, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, August 2013, ADAMS Accession No. ML13277A557.

2. Not used.

3. Title 10, Code of Federal Regulations, Part 50, Appendix G, Fracture Toughness Requirements, November 29, 2019.

4. ASME Boiler and Pressure Vessel Code, Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G, Fracture Toughness Criteria for Protection Against Failure, 2004 Edition.

5. Pilkey, W. D., Pilkey, D. F., Petersons Stress Concentration Factors, 3rd Ed., C. 2008, John Wiley & Sons, Inc.

6. Structural Integrity Associates Calculation No. NSP-21Q-303, Revision 1, Determination of the Initial RTNDT and ART Values for the Monticello RPV Materials.

7. CB&I Drawing No. 1, Revision 8, General Plan. 172 ID x 63 2 Ins Heads Nuclear Reactor, NX-8290-13, SI File No. NSP-21Q-210.

8. GE Design Specification No. 23A1581, Revision 3, Reactor Vessel - Recirculation Inlet Safe End, SI File No. 1000720.202.

9. GE Report No. SASR 88-99, Revision 1, Implementation of Regulatory Guide 1.99, Revision 2 for the Monticello Nuclear Generating Plant, January 1989, SI File No. NSP-21Q-202

10. SI Calculation No. 2100300.302P, Revision 2, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts. CONTAINS PROPRIETARY INFORMATION.

11. SI Calculation No. 2200284.302P, Revision 0, Feedwater Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development. CONTAINS PROPRIETARY INFORMATION.

12. SI Calculation No. 2200284.301P, Revision 0, Finite Element Stress and Fracture Mechanics Analysis of Monticello RPV Recirculation Inlet Nozzle. CONTAINS PROPRIETARY INFORMATION.

13. SI Calculation No. 1000847.303, Revision 2, Revised P-T Curves Calculation,

14. GE Design Specification No. 22A6996, Revision 0, Reactor Vessel System Cycling, SI File No.

1000847.201.

15. BWRVIP-135, Revision 4: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014. 3002003144. EPRI PROPRIETARY INFORMATION.

16. BWRVIP Letter 2022-053, September 14, 2022, from Bob Carter to Russell Lidberg,

Subject:

Notification of New BWRVIP Integrated Surveillance Program (ISP) Data Applicable to the Monticello Reactor Pressure Vessel (RPV).

17. SI Nuclear QA Procedure, Design and Analysis Control, Revision 8.1, SI File No. QP03-01.

18. SI Nuclear QA Procedure, Calculation Verification Activities, Revision 9, SI File No. QP03-07.

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19. Monticello Nuclear Generating Station Reactor Pressure Vessel Fluence Evaluation -

Subsequent License Renewal, TransWare Enterprises, MNT-FLU-001-R-002, Revision 0. June 2022. SI File No. 2100300.201 File No.: 2200284.303 Page 18 of 34 Revision: 0 F0306-01R4

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Table 3. MNGP Beltline Region, Curve A, for 72 EFPY CurveAPressureTest

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 329.7

90.3 379.7

109.1 429.6

122.7 479.5

133.4 529.5

142.2 579.4

149.6 629.3

156.1 679.2

161.9 729.2

171.4 778.2

179.3 827.3

186.2 876.3

192.2 925.4

197.6 974.4

202.4 1023.5

206.9 1072.5

210.9 1121.6

214.7 1170.6

218.2 1219.7

221.5 1268.7

224.5 1317.8

227.4 1366.8

230.2 1415.9

232.8 1464.9

235.2 1514.0

237.6 1563.0

 

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Table 4. MNGP Bottom Head Region, Curve A, for 72 EFPY CurveAPressureTest

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 812.8

64.8 859.7

69.2 906.6

73.2 953.4

77.0 1000.3

80.5 1047.2

83.7 1094.1

86.8 1141.0

89.6 1187.9

92.3 1234.8

94.9 1281.7

97.4 1328.6

99.7 1375.4

102.0 1422.3

104.1 1469.2

106.1 1516.1

108.1 1563.0

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Table 5. MNGP Non-Beltline Region, Curve A, for 72 EFPY CurveAPressureTest

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 312.6

100.0 312.6

100.0 936.3

103.6 984.5

106.9 1032.7

110.0 1080.9

113.0 1129.1

115.8 1177.3

118.4 1225.6

120.9 1273.8

123.3 1322.0

125.6 1370.2

127.7 1418.4

129.8 1466.6

131.8 1514.8

133.7 1563.0

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Table 6. MNGP Overall Composite Curve, Curve A, for 72 EFPY CurveAPressureTest

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 312.6

100.0 312.6

100.0 403.1

115.0 449.7

126.5 496.3

135.9 542.8

143.7 589.4

150.6 636.0

156.5 682.6

161.9 729.2

171.4 778.2

179.3 827.3

186.2 876.3

192.2 925.4

197.6 974.4

202.4 1023.5

206.9 1072.5

210.9 1121.6

214.7 1170.6

218.2 1219.7

221.5 1268.7

224.5 1317.8

227.4 1366.8

230.2 1415.9

232.8 1464.9

235.2 1514.0

237.6 1563.0

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Table 7. MNGP Beltline Region, Curve B, for 72 EFPY CurveBCoreNotCritical

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

59.6 141.2

99.6 187.6

121.1 234.1

135.8 280.5

147.1 326.9

156.1 373.4

163.8 419.8

170.4 466.2

176.1 512.6

181.3 559.1

189.7 606.9

196.9 654.7

203.2 702.5

208.8 750.3

213.9 798.1

218.4 845.9

222.6 893.7

226.5 941.5

230.1 989.3

233.4 1037.1

236.6 1084.9

239.5 1132.7

242.3 1180.5

245.0 1228.4

247.5 1276.2

249.8 1324.0

252.1 1371.8

254.3 1419.6

256.4 1467.4

258.4 1515.2

260.3 1563.0

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Table 8. MNGP Bottom Head Region, Curve B, for 72 EFPY CurveBCoreNotCritical

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 533.1

66.6 582.2

72.4 631.2

77.6 680.3

82.4 729.3

86.7 778.3

90.6 827.4

94.3 876.4

97.8 925.5

101.0 974.5

104.0 1023.6

106.8 1072.6

109.5 1121.6

112.1 1170.7

114.5 1219.7

116.8 1268.8

119.0 1317.8

121.1 1366.8

123.2 1415.9

125.1 1464.9

127.0 1514.0

128.8 1563.0

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Table 9. MNGP Non-Beltline Region, Curve B, for 72 EFPY CurveBCoreNotCritical

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 312.6

130.0 312.6

130.0 1022.7

132.7 1071.8

135.3 1120.9

137.8 1170.0

140.2 1219.1

142.4 1268.3

144.5 1317.4

146.6 1366.5

148.6 1415.6

150.5 1464.8

152.3 1513.9

154.1 1563.0

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Table 10. MNGP Overall Composite Curve, Curve B, for 72 EFPY CurveBCoreNotCritical

PTCurve PTCurve

Temperature Pressure

°F psi

60.0 0.0

60.0 141.2

94.1 181.0

115.1 220.8

129.5 260.6

143.0 310.3

153.5 360.1

162.1 409.8

169.3 459.6

175.6 509.3

181.3 559.1

189.7 606.9

196.9 654.7

203.2 702.5

208.8 750.3

213.9 798.1

218.4 845.9

222.6 893.7

226.5 941.5

230.1 989.3

233.4 1037.1

236.6 1084.9

239.5 1132.7

242.3 1180.5

245.0 1228.4

247.5 1276.2

249.8 1324.0

252.1 1371.8

254.3 1419.6

256.4 1467.4

258.4 1515.2

260.3 1563.0

 

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Table 11. MNGP Beltline Region, Curve C, for 72 EFPY CurveCCoreCritical

PTCurve PTCurve

Temperature Pressure

°F psi

70.0 0.0

70.0 122.2

128.5 170.8

155.2 219.3

172.2 267.8

184.7 316.4

194.6 364.9

202.8 413.4

209.8 462.0

215.9 510.5

221.3 559.1

229.7 606.9

236.9 654.7

243.2 702.5

248.8 750.3

253.9 798.1

258.4 845.9

262.6 893.7

266.5 941.5

270.1 989.3

273.4 1037.1

276.6 1084.9

279.5 1132.7

282.3 1180.5

285.0 1228.4

287.5 1276.2

289.8 1324.0

292.1 1371.8

294.3 1419.6

296.4 1467.4

298.4 1515.2

300.3 1563.0

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Table 12. MNGP Bottom Head Region, Curve C, for 72 EFPY CurveCCoreCritical

PTCurve PTCurve

Temperature Pressure

°F psi

70.0 0.0

70.0 376.2

81.5 425.6

90.9 475.1

98.7 524.5

105.5 574.0

111.5 623.4

116.9 672.9

121.7 722.3

126.1 771.8

130.2 821.2

133.9 870.7

137.4 920.1

140.6 969.6

143.7 1019.0

146.6 1068.5

149.3 1117.9

151.9 1167.4

154.3 1216.8

156.7 1266.3

158.9 1315.7

161.1 1365.2

163.1 1414.6

165.1 1464.1

167.0 1513.5

168.8 1563.0

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Table 13. MNGP Non-Beltline Region, Curve C, for 72 EFPY CurveCCoreCritical

PTCurve PTCurve

Temperature Pressure

°F psi

70.0 0.0

70.0 276.4

83.7 312.6

203.0 312.6

203.0 1563.0

 

 

 

 

 

 

 

 

 

 

File No.: 2200284.303 Page 29 of 34 Revision: 0 F0306-01R4 13 Structural Integrity Associates, Inc.* info@structint.com ~ 1-877-451-POWER C., structint.com (@)

Table 14. MNGP Overall Composite Curve, Curve C, for 72 EFPY CurveCCoreCritical

PTCurve PTCurve

Temperature Pressure

°F psi

70.0 0.0

70.0 122.2

127.8 169.8

154.4 217.4

171.4 265.0

183.9 312.6

203.0 312.6

203.0 414.7

209.9 462.8

215.9 510.9

221.3 559.1

229.7 606.9

236.9 654.7

243.2 702.5

248.8 750.3

253.9 798.1

258.4 845.9

262.6 893.7

266.5 941.5

270.1 989.3

273.4 1037.1

276.6 1084.9

279.5 1132.7

282.3 1180.5

285.0 1228.4

287.5 1276.2

289.8 1324.0

292.1 1371.8

294.3 1419.6

296.4 1467.4

298.4 1515.2

300.3 1563.0

 

 

File No.: 2200284.303 Page 30 of 34 Revision: 0 F0306-01R4 13 Structural Integrity Associates, Inc.* info@structint.com ~ 1-877-451-POWER C., structint.com (@)

MNGP P-T Curve A - Pressure Test, Composite Curves

-Beltline - - - Bottom Head - - Non-Beltline - overall 300 I I I I

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Figure 1. MNGP P-T Curve A (Hydrostatic Pressure and Leak Test), 72 EFPY File No.: 2200284.303 Page 31 of 34 Revision: 0 F0306-01R4

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MNGP P-T Curve B - Core Not Critical, Composite Curves

- Beltline - - - Bottom Head - - Non-Beltline Overall 300 COMPLIANCE REQUIRES OPERATION ABOVE THE CURVES

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Figure 2. MNGP P-T Curve B (Normal Operation - Core Not Critical), 72 EFPY File No.: 2200284.303 Page 32 of 34 Revision: 0 F0306-01R4

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MNGP P-T Curve C - Core Critical, Composite Curves

- Beltline - - - Bottom Head - - Non-Beltline Overall 300 COMPLIANCE REQUIRES OPERATION ABOVE THE CURVES 250

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Figure 3. MNGP P-T Curve C (Normal Operation - Core Critical), 72 EFPY File No.: 2200284.303 Page 33 of 34 Revision: 0 F0306-01R4

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MNGP P-T MNGP - Composite Curves A, B, and C

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Figure 4. MNGP Overall Composite Curves A, B, and C, 72 EFPY File No.: 2200284.303 Page 34 of 34 Revision: 0 F0306-01R4

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APPENDIX A P-T CURVE INPUT LISTING File No.: 2200284.303 Page A-1 of A-3 Revision: 0 F0306-01R4 (j info@structint.com ~ 1-877-4S1-POWER e structint.com ~

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Table A-1: MNGP Nozzle Stress Intensity Factors Thermal, Nozzle Applied Pressure, KIp-app Reference KIt 70.59 for 1,000 psi Feedwater 10.37 [11]

pressure Recirculation 75.20 for 1010 psi 25.28 [12]

Inlet (N2) pressure KI in units of ksi-in0.5 Table A-2: MNGP Unit 1 P-T Curve Input Listing General Parameters Values

Unit System for Tables and Plots English

Temperature Instrument Uncertainty Adjustment (°F) 0

Pressure Instrument Uncertainty Adjustment (psig) 0

Water Density (lbm/ft3) 62.4

Full-Vessel Water Height (in) 758

Safety Factor for Curve A 1.5

Safety Factor for Curves B and C 2

Bolt-up Temperature (°F) 60

ART of Closure Flange Region (°F) 10

Default Temperature Increment for Tables (°F) 10

Default Pressure Increment for Composite Tables (psig) 50

Starting Pressure for Curves (psig) 0

Atmospheric Pressure Adjustment (psi) 14.7

Preservice hydrotest pressure (psi) 1563

In-service hydrotest pressure (psi) 1025 Minimum in-service hydrotest temperature (°F) 203 Beltline Parameters Values

Adjusted Reference Temperature (°F) 178.1

Vessel Radius (in) 103.1875

Vessel Thickness (in) 5.0625

Heat-up / Cool-down Rate (°F/hr) 100

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Additional Beltline Nozzle Parameters (N2 Recirculation Inlet) Values

Adjusted Reference Temperature (°F) 116.6

Applied Pressure Stress Intensity Factor (ksi*in^0.5) 75.2

Applied Thermal Stress Intensity Factor (ksi*in^0.5) 25.28 Scale KIT based on Saturation Temperature? Yes

Minimum Transient Temperature (°F) 100

Maximum Transient Temperature (°F) 549

Reference Pressure for Thermal Transient (psig) 1010

BottomHeadParameters Values

AdjustedReferenceTemperature(°F) 26

VesselRadius(in) 103.1875

VesselThickness(in) 5.9375

Heatup/CooldownRate(°F/hr) 100

StressConcentrationFactor 3

UpperVessel(FeedwaterNozzle)Parameters Values

AdjustedReferenceTemperature(°F) 40

AppliedPressureStressIntensityFactor(ksi*in^0.5) 70.59

AppliedThermalStressIntensityFactor(ksi*in^0.5) 10.37



MinimumThermalStressIntensityFactor(ksi*in^0.5)

ScaleKITbasedonSaturationTemperature? Yes

MinimumTransientTemperature(°F) 100

MaximumTransientTemperature(°F) 548

ReferencePressureforThermalTransient(psig) 1000

File No.: 2200284.303 Page A-3 of A-3 Revision: 0 F0306-01R4 e

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APPENDIX B SUPPORTING CALCULATIONS File No.: 2200284.303 Page B-1 of B-15 Revision: 0 F0306-01R4 t>

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Table B-1: MNGP Beltline Region, Curve A Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 35.2 23.4 60.0 0.0

60.0 35.2 23.4 60.0 524.5

70.0 35.6 23.7 70.0 531.3

80.0 36.1 24.1 80.0 539.6

90.0 36.8 24.5 90.0 549.7

100.0 37.5 25.0 100.0 562.1

110.0 38.5 25.7 110.0 577.2

120.0 39.7 26.5 120.0 595.6

130.0 41.1 27.4 130.0 618.2

140.0 42.9 28.6 140.0 645.7

150.0 45.0 30.0 150.0 679.4

160.0 47.6 31.8 160.0 720.4

170.0 50.8 33.9 170.0 770.6

180.0 54.7 36.5 180.0 831.9

190.0 59.5 39.7 190.0 906.8

200.0 65.3 43.6 200.0 998.2

210.0 72.4 48.3 210.0 1109.9

220.0 81.1 54.1 220.0 1246.3

230.0 91.7 61.2 230.0 1412.9

240.0 104.7 69.8 240.0 1616.3

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Table B-2: MNGP Recirculation Inlet Nozzle, Beltline Region, Curve A Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 39.9 26.6 60.0 0.0

60.0 39.9 26.6 60.0 329.7

70.0 41.4 27.6 70.0 343.0

80.0 43.2 28.8 80.0 359.2

90.0 45.4 30.3 90.0 379.0

100.0 48.1 32.1 100.0 403.1

110.0 51.4 34.2 110.0 432.6

120.0 55.4 36.9 120.0 468.6

130.0 60.3 40.2 130.0 512.6

140.0 66.3 44.2 140.0 566.3

150.0 73.6 49.1 150.0 632.0

160.0 82.6 55.1 160.0 712.1

170.0 93.5 62.4 170.0 810.1

180.0 106.9 71.3 180.0 929.6

190.0 123.2 82.1 190.0 1075.7

200.0 143.1 95.4 200.0 1254.1

210.0 167.5 111.6 210.0 1472.0

220.0 197.2 131.5 220.0 1738.2

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Table B-3: MNGP Bottom Head Region, Curve A Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 74.1 49.4 60.0 0.0

60.0 74.1 49.4 60.0 812.8

70.0 83.2 55.5 70.0 915.5

80.0 94.3 62.8 80.0 1040.9

90.0 107.8 71.8 90.0 1194.1

100.0 124.3 82.9 100.0 1381.2

110.0 144.4 96.3 110.0 1609.8

Table B-4: MNGP FW Nozzle / Non-Beltline, Curve A Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 64.1 42.8 60.0 0.0

60.0 64.1 42.8 60.0 578.3

70.0 71.0 47.3 70.0 643.0

80.0 79.3 52.9 80.0 722.0

90.0 89.6 59.7 90.0 818.5

100.0 102.0 68.0 100.0 936.3

110.0 117.3 78.2 110.0 1080.2

120.0 135.9 90.6 120.0 1256.1

130.0 158.6 105.8 130.0 1470.8

140.0 186.4 124.3 140.0 1733.1

File No.: 2200284.303 Page B-4 of B-15 Revision: 0 F0306-01R4 e

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Table B-5: MNGP Beltline Region, Curve B Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 35.2 14.8 60.0 0.0

60.0 35.2 14.8 60.0 321.8

70.0 35.6 15.0 70.0 326.9

80.0 36.1 15.3 80.0 333.1

90.0 36.8 15.6 90.0 340.7

100.0 37.5 16.0 100.0 350.0

110.0 38.5 16.5 110.0 361.3

120.0 39.7 17.1 120.0 375.2

130.0 41.1 17.8 130.0 392.1

140.0 42.9 18.7 140.0 412.8

150.0 45.0 19.8 150.0 438.0

160.0 47.6 21.1 160.0 468.8

170.0 50.8 22.7 170.0 506.4

180.0 54.7 24.6 180.0 552.4

190.0 59.5 27.0 190.0 608.5

200.0 65.3 29.9 200.0 677.1

210.0 72.4 33.5 210.0 760.9

220.0 81.1 37.8 220.0 863.1

230.0 91.7 43.1 230.0 988.1

240.0 104.7 49.6 240.0 1140.7

250.0 120.5 57.5 250.0 1327.1

260.0 139.9 67.2 260.0 1554.8

270.0 163.5 79.0 270.0 1832.8

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Table B-6: MNGP Recirculation Inlet Nozzle, Beltline Region, Curve B Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 39.9 22.8 60.0 0.0

60.0 39.9 12.6 60.0 141.2

70.0 41.4 13.2 70.0 149.8

80.0 43.2 14.0 80.0 160.2

90.0 45.4 14.9 90.0 172.9

100.0 48.1 16.1 100.0 188.5

110.0 51.4 17.5 110.0 207.7

120.0 55.4 19.3 120.0 231.3

130.0 60.3 21.4 130.0 260.5

140.0 66.3 24.1 140.0 296.5

150.0 73.6 27.4 150.0 341.0

160.0 82.6 31.5 160.0 395.9

170.0 93.5 36.5 170.0 463.5

180.0 106.9 42.8 180.0 546.9

190.0 123.2 50.4 190.0 649.6

200.0 143.1 59.8 200.0 776.1

210.0 167.5 71.4 210.0 931.6

220.0 197.2 85.6 220.0 1122.8

230.0 233.5 103.1 230.0 1357.7

240.0 277.8 124.6 240.0 1646.1

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Table B-7: MNGP Bottom Head Region, Curve B Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 74.1 33.0 60.0 0.0

60.0 74.1 33.0 60.0 533.1

70.0 83.2 37.5 70.0 610.2

80.0 94.3 43.0 80.0 704.2

90.0 107.8 49.8 90.0 819.2

100.0 124.3 58.0 100.0 959.5

110.0 144.4 68.1 110.0 1130.9

120.0 169.1 80.4 120.0 1340.3

130.0 199.2 95.5 130.0 1596.0

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Table B-8: MNGP FW Nozzle / Non-Beltline, Curve B Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

60.0 64.1 32.1 60.0 0.0

60.0 64.1 28.1 60.0 370.9

70.0 71.0 31.4 70.0 417.6

80.0 79.3 35.4 80.0 474.8

90.0 89.6 40.4 90.0 544.8

100.0 102.0 46.5 100.0 630.7

110.0 117.3 53.9 110.0 735.9

120.0 135.9 63.0 120.0 864.8

130.0 158.6 74.1 130.0 1022.7

140.0 186.4 87.8 140.0 1215.9

150.0 220.3 104.5 150.0 1452.5

160.0 261.8 124.9 160.0 1742.0

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Table B-9: MNGP Beltline Region, Curve C Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

30.0 34.3 14.4 70.0 0.0

30.0 34.3 14.4 70.0 311.4

40.0 34.5 14.5 80.0 314.2

50.0 34.8 14.7 90.0 317.6

60.0 35.2 14.8 100.0 321.8

70.0 35.6 15.0 110.0 326.9

80.0 36.1 15.3 120.0 333.1

90.0 36.8 15.6 130.0 340.7

100.0 37.5 16.0 140.0 350.0

110.0 38.5 16.5 150.0 361.3

120.0 39.7 17.1 160.0 375.2

130.0 41.1 17.8 170.0 392.1

140.0 42.9 18.7 180.0 412.8

150.0 45.0 19.8 190.0 438.0

160.0 47.6 21.1 200.0 468.8

170.0 50.8 22.7 210.0 506.4

180.0 54.7 24.6 220.0 552.4

190.0 59.5 27.0 230.0 608.5

200.0 65.3 29.9 240.0 677.1

210.0 72.4 33.5 250.0 760.9

220.0 81.1 37.8 260.0 863.1

230.0 91.7 43.1 270.0 988.1

240.0 104.7 49.6 280.0 1140.7

250.0 120.5 57.5 290.0 1327.1

260.0 139.9 67.2 300.0 1554.8

270.0 163.5 79.0 310.0 1832.8

File No.: 2200284.303 Page B-9 of B-15 Revision: 0 F0306-01R4 e

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Table B-10: MNGP Recirculation Inlet Nozzle, Beltline Region, Curve C Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

30.0 36.9 21.2 70.0 0.0

30.0 36.9 11.1 70.0 122.2

40.0 37.7 11.6 80.0 127.8

50.0 38.7 12.0 90.0 134.0

60.0 39.9 12.6 100.0 141.2

70.0 41.4 13.2 110.0 149.8

80.0 43.2 14.0 120.0 160.2

90.0 45.4 14.9 130.0 172.9

100.0 48.1 16.1 140.0 188.5

110.0 51.4 17.5 150.0 207.7

120.0 55.4 19.3 160.0 231.3

130.0 60.3 21.4 170.0 260.5

140.0 66.3 24.1 180.0 296.5

150.0 73.6 27.4 190.0 341.0

160.0 82.6 31.5 200.0 395.9

170.0 93.5 36.5 210.0 463.5

180.0 106.9 42.8 220.0 546.9

190.0 123.2 50.4 230.0 649.6

200.0 143.1 59.8 240.0 776.1

210.0 167.5 71.4 250.0 931.6

220.0 197.2 85.6 260.0 1122.8

230.0 233.5 103.1 270.0 1357.7

240.0 277.8 124.6 280.0 1646.1

File No.: 2200284.303 Page B-10 of B-15 Revision: 0 F0306-01R4 e

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Table B-11: MNGP Bottom Head Region, Curve C Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

30.0 55.7 23.7 70.0 0.0

30.0 55.7 23.7 70.0 376.2

40.0 60.6 26.2 80.0 418.5

50.0 66.7 29.3 90.0 470.1

60.0 74.1 33.0 100.0 533.1

70.0 83.2 37.5 110.0 610.2

80.0 94.3 43.0 120.0 704.2

90.0 107.8 49.8 130.0 819.2

100.0 124.3 58.0 140.0 959.5

110.0 144.4 68.1 150.0 1130.9

120.0 169.1 80.4 160.0 1340.3

130.0 199.2 95.5 170.0 1596.0

File No.: 2200284.303 Page B-11 of B-15 Revision: 0 F0306-01R4 e

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Table B-12: MNGP FW Nozzle / Non-Beltline, Curve C Calculations, 72 EFPY GageFluid PTCurve PTCurve

KIc KIp

Temperature Temperature Pressure

°F ksi*in1/2 ksi*in1/2 °F psig

30.0 50.2 25.1 70.0 0.0

30.0 50.2 21.4 70.0 276.4

40.0 53.9 23.2 80.0 301.8

50.0 58.5 25.4 90.0 332.8

60.0 64.1 28.1 100.0 370.9

70.0 71.0 31.4 110.0 417.6

80.0 79.3 35.4 120.0 474.8

90.0 89.6 40.4 130.0 544.8

100.0 102.0 46.5 140.0 630.7

110.0 117.3 53.9 150.0 735.9

120.0 135.9 63.0 160.0 864.8

130.0 158.6 74.1 170.0 1022.7

140.0 186.4 87.8 180.0 1215.9

150.0 220.3 104.5 190.0 1452.5

160.0 261.8 124.9 200.0 1742.0

File No.: 2200284.303 Page B-12 of B-15 Revision: 0 F0306-01R4 e

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MNGP P-T Curve A - Pressure Test, All Components

-BL * * * * *

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0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Pressure Limit in Reactor Vessel (psig)

Figure B-1: MNGP P-T Curve A (Hydrostatic Pressure and Leak Test), 72 EFPY Note: BL is Beltline, BH is Bottom Head, FWN is Feedwater Nozzle, BN is Recirculation Inlet Nozzle File No.: 2200284.303 Page B-13 of B-15 Revision: 0 F0306-01R4

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MNGP P-T Curve B - Core Not Critical, All Components

-BL ****** BN - -BH FWN 10CFRSO 3 00 I I I I I I I I I

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0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Pressure Limit in Reactor Vessel {psig)

Figure B-2: MNGP P-T Curve B (Normal Operation - Core Not Critical), 72 EFPY Note: BL is Beltline, BH is Bottom Head, FWN is Feedwater Nozzle, BN is Recirculation Inlet Nozzle File No.: 2200284.303 Page B-14 of B-15 Revision: 0 F0306-01R4

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MNGP P-T Curve C - Core Critical, All Components

-BL * * * * *

• BN - -BH FWN 10C:FRSO 300 I I I I I I I I I I I I I I

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+

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+ + +

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0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Pressure Limit in Reactor Vessel (psig)

Figure B-3: MNGP P-T Curve C (Normal Operation - Core Critical), 72 EFPY Note: BL is Beltline, BH is Bottom Head, FWN is Feedwater Nozzle, BN(N2) is Recirculation Inlet Nozzle File No.: 2200284.303 Page B-15 of B-15 Revision: 0 F0306-01R4

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APPENDIXC DEVELOPMENT OF SATURATION STEAM CURVE FITS File No.: 2200284.303 Page C-1 of C-4 Revision: 0 F0306-01R4 t>

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Curve Fit for Saturated Steam

Reference:

Steam Table data obtained from "Steam Tables, Properties of Saturated and Superheated Steam," CE Power Systems, 7th Printing.

Curve Fit: Tsat = 119.3*(0.7987)(1/Psat)

• Psat0.2198 Pressure Temperature Curve Fit Psat Tsat Tsat Difference Error (psia) (°F) (°F) (°F) (%)

14.696 212.00 212.10 0.10 0.05%

15 213.03 213.13 0.10 0.05%

20 227.96 227.89 -0.07 -0.03%

30 250.34 250.07 -0.27 -0.11%

40 267.25 266.89 -0.36 -0.13%

50 281.02 280.62 -0.40 -0.14%

60 292.71 292.32 -0.39 -0.13%

70 302.93 302.55 -0.38 -0.13%

80 312.04 311.69 -0.35 -0.11%

90 320.28 319.96 -0.32 -0.10%

100 327.82 327.54 -0.28 -0.09%

110 334.79 334.54 -0.25 -0.07%

120 341.27 341.06 -0.21 -0.06%

130 347.33 347.16 -0.17 -0.05%

140 353.04 352.91 -0.13 -0.04%

150 358.43 358.34 -0.09 -0.03%

160 363.55 363.49 -0.06 -0.02%

170 368.42 368.40 -0.02 -0.01%

180 373.08 373.08 0.00 0.00%

190 377.53 377.57 0.04 0.01%

200 381.80 381.87 0.07 0.02%

210 385.91 386.01 0.10 0.03%

220 389.88 390.00 0.12 0.03%

230 393.70 393.84 0.14 0.04%

240 397.39 397.56 0.17 0.04%

250 400.97 401.16 0.19 0.05%

260 404.44 404.65 0.21 0.05%

270 407.80 408.03 0.23 0.06%

280 411.07 411.32 0.25 0.06%

290 414.25 414.51 0.26 0.06%

300 417.35 417.62 0.27 0.07%

350 431.73 432.06 0.33 0.08%

400 444.60 444.97 0.37 0.08%

450 456.28 456.67 0.39 0.08%

500 467.01 467.39 0.38 0.08%

550 476.94 477.30 0.36 0.08%

600 486.20 486.54 0.34 0.07%

650 494.89 495.19 0.30 0.06%

700 503.08 503.33 0.25 0.05%

750 510.84 511.03 0.19 0.04%

800 518.21 518.34 0.13 0.03%

850 525.24 525.31 0.07 0.01%

900 531.95 531.95 0.00 0.00%

950 538.39 538.32 -0.07 -0.01%

1000 544.58 544.43 -0.15 -0.03%

File No.: 2200284.303 Page C-2 of C-4 Revision: 0 F0306-01R4 e

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Curve Fit for Saturated Steam

Reference:

Steam Table data obtained from "Steam Tables, Properties of Saturated and Superheated Steam," CE Power Systems, 7th Printing.

Curve Fit: Tsat = 119.3*(0.7987)(1/Psat)

• Psat0.2198 Pressure Temperature Curve Fit Psat Tsat Tsat Difference Error (psia) (°F) (°F) (°F) (%)

1050 550.53 550.31 -0.22 -0.04%

1100 556.28 555.97 -0.31 -0.06%

1150 561.82 561.43 -0.39 -0.07%

1200 567.19 566.71 -0.48 -0.08%

1250 572.38 571.83 -0.55 -0.10%

1300 577.42 576.78 -0.64 -0.11%

1350 582.32 581.59 -0.73 -0.13%

1400 587.07 586.26 -0.81 -0.14%

1450 591.70 590.80 -0.90 -0.15%

1500 596.20 595.22 -0.98 -0.16%

1550 600.59 599.53 -1.06 -0.18%

1600 604.87 603.73 -1.14 -0.19%

1650 609.05 607.83 -1.22 -0.20%

1700 613.13 611.84 -1.29 -0.21%

1750 617.12 615.75 -1.37 -0.22%

1800 621.02 619.58 -1.44 -0.23%

1850 624.83 623.32 -1.51 -0.24%

1900 628.56 626.99 -1.57 -0.25%

1950 632.22 630.58 -1.64 -0.26%

2000 635.80 634.10 -1.70 -0.27%

2100 642.76 640.94 -1.82 -0.28%

2200 649.45 647.53 -1.92 -0.30%

2300 655.89 653.89 -2.00 -0.30%

2400 662.11 660.04 -2.07 -0.31%

2500 668.11 665.99 -2.12 -0.32%

Maximum = 0.39 0.08%

Minimum = -2.12 -0.32%

Average = -0.41 -0.07%

Std. Deviation 0.71 0.12%

File No.: 2200284.303 Page C-3 of C-4 Revision: 0 F0306-01R4 e

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Curve Fit for Saturated Steam Conditions 800 700 600 Saturation Temperature (°F) 500 400 300 Steam Tables Data Curve Fit: Tsat = 119.3*(0.7987)^(1/Psat)

• Psat^0.2198 200 100 0

0 500 1,000 1,500 2,000 2,500 Saturation Pressure (psia)

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APPENDIX D SUPPORTING FILES File No.: 2200284.303 Page D-1 of D-2 Revision: 0 F0306-01R4 tr Structural Integrity info@structint.com m 1-877-451-POWER e structint.com @

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Supporting Files Comment

1. 2200284.303P R0.xlsx Excel file contains the detailed P-T curve calculations for MNGP (File name remains, although information is not proprietary)

File No.: 2200284.303 Page D-2 of D-2 Revision: 0 F0306-01R4 SJ Structural Integrity info@structint.com mi 1-877-451-POWER" structint.com ~

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