ML25091A301

Enclosure 4, Attachment 5 - WCAP-18939-NP, H.B. Robinson Unit 2 Subsequent License Renewal: Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports, Revision 0, Dated November 2024

ML25091A301
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Site:	Robinson
Issue date:	04/01/2025
From:	Duke Energy Progress
To:	Office of Nuclear Reactor Regulation
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References
RA-25-0067 WCAP-18939-NP, Rev 1
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ENCLOSURE 4 ATTACHMENT 5 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NUMBER 2 Westinghouse WCAP-18939-NP, Revision 1, H.B. Robinson Unit 2 Subsequent License Renewal: Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports, January 2025

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 WCAP-18939-NP January 2025 Revision 1 H.B. Robinson Unit 2 Subsequent License Renewal:

Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2025 Westinghouse Electric Company LLC All Rights Reserved WCAP-18939-NP Revision 1 H.B. Robinson Unit 2 Subsequent License Renewal:

Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports Maria A. Rizzilli*

RV/CV Design and Analysis Alexandria M. Scott*

RV/CV Design and Analysis January 2025 Verifier:

Joshua A. Coleman*

RV/CV Design and Analysis Gordon Z. Hall*

Structural Design and Analysis Reviewer:

Anees Udyawar*

Stephen K. Longwell*

Ian T. Senior*

Sylena E. Smith*

Brian J. Hall*

Approved: Remington W. Iddings*, Manager RV/CV Design and Analysis

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii WCAP-18939-NP January 2025 Revision 1 FOREWORD This report contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets. Coding (a,c,e) associated with the brackets sets forth the basis on which the information is considered proprietary.

The proprietary information and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(c)

Its use by a competitor would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

The document herein is bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both proprietary and non-proprietary versions by means of lower-case letters (a)

(c) and (e) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified above.

The proprietary information in the brackets is provided in the proprietary version of this report (WCAP-18939-P Revision 1).

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii WCAP-18939-NP January 2025 Revision 1 RECORD OF REVISIONS Revision Date Revision Description 0

November 2024 Original Final Issue.

1 January 2025 This report is revised from Revision 0 to Revision 1 to incorporate updated forces for the leveling bolt component in Table 4-7. In addition, the stress intensity factors provided in Table 7-7 and Table 8-1 are also updated based on the changes in Table 4-7. Note that the final conclusions in this report remain unchanged and all revisions are marked by revision bars in the left hand margin.

Trademark Notes:

ANSYS, ANSYS Workbench, and any and all ANSYS, Inc. product and service names are registered trademarks or trademarks of ANSYS, Inc. or its subsidiaries located in the United States or other countries.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv WCAP-18939-NP January 2025 Revision 1 TABLE OF CONTENTS FOREWORD................................................................................................................................................ ii LIST OF TABLES........................................................................................................................................ v LIST OF FIGURES..................................................................................................................................... vi EXECUTIVE

SUMMARY

......................................................................................................................... vii 1

INTRODUCTION........................................................................................................................ 1-1 2

GENERAL METHODOLOGY.................................................................................................... 2-1 3

SUPPORT CONFIGURATION, MATERIAL, AND GEOMETRY............................................ 3-1 4

LOADING CONDITIONS AND STRESS ANALYSIS.............................................................. 4-1 4.1 SUPPORT SHOE STRESSES......................................................................................... 4-3 4.2 TOP PLATE STRESSES................................................................................................. 4-4 4.3 BOTTOM PLATE STRESSES........................................................................................ 4-5 4.4 DIAPHRAGM PLATE STRESSES................................................................................ 4-6 4.5 I-BEAM STRESSES....................................................................................................... 4-7 4.6 LEVELING BOLT LOADS............................................................................................ 4-8 4.7 ANCHOR BOLT LOADS............................................................................................... 4-9 5

FRACTURE MECHANICS METHODOLOGY......................................................................... 5-1 5.1 FRACTURE TOUGHNESS DETERMINATION.......................................................... 5-1 5.1.1 Fracture Toughness for the H.B. Robinson Unit 2 RPV Supports................... 5-2 5.1.2 Bulk Metal Temperature................................................................................ 5-10 5.1.3 Neutron Embrittlement and Nil-Ductility Transition Temperature............... 5-10 5.2 STRESS INTENSITY FACTORS AND POSTULATED FLAWS............................... 5-24 5.2.1 Plate Model (Support Shoe, Plates, and I-beams)......................................... 5-24 5.2.2 Round Bar Model (for Bolts)......................................................................... 5-26 6

ALLOWABLE FLAW SIZES....................................................................................................... 6-1 7

FRACTURE MECHANICS RESULTS....................................................................................... 7-1 7.1 SUPPORT SHOE............................................................................................................. 7-4 7.2 TOP PLATE..................................................................................................................... 7-5 7.3 BOTTOM PLATE............................................................................................................ 7-6 7.4 DIAPHRAGM PLATES.................................................................................................. 7-7 7.5 I-BEAMS......................................................................................................................... 7-9 7.6 LEVELING BOLTS...................................................................................................... 7-10 7.7 ANCHOR BOLTS......................................................................................................... 7-11 8

DISCUSSION OF CONCLUSIONS............................................................................................ 8-1 8.1 EVALUATION OF CURRENT CONDITIONS............................................................. 8-1 8.2 FRACTURE MECHANICS CONCLUSIONS............................................................... 8-3 9

REFERENCES............................................................................................................................. 9-1

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 v

WCAP-18939-NP January 2025 Revision 1 LIST OF TABLES Table 4-1: Stresses for the Support Shoes.................................................................................................. 4-3 Table 4-2: Stresses for the Top Plates........................................................................................................ 4-4 Table 4-3: Stresses for the Bottom Plates.................................................................................................. 4-5 Table 4-4: Stresses for the Diaphragm Plates............................................................................................ 4-6 Table 4-5: Limiting Stresses for the Diaphragm Plate near a Hole............................................................ 4-6 Table 4-6: Stresses for the I-Beams........................................................................................................... 4-7 Table 4-7: Forces and Moments for the Leveling Bolts............................................................................. 4-8 Table 4-8: Forces and Moments for the Anchor Bolts............................................................................... 4-9 Table 5-1: Fracture Toughness for the H.B. Robinson Unit 2 RPV Support Components........................ 5-7 Table 5-2: H.B. Robinson Unit 2 Iron Displacement per Atom at the RPV Support Structure (without Analytical Uncertainties)............................................................................................... 5-22 Table 5-3: H.B. Robinson Unit 2 Iron Displacement per Atom at the RPV Support Structure (with Analytical Uncertainties)............................................................................................... 5-22 Table 5-4: H.B. Robinson Unit 2 Iron Displacement Per Atom and NDTT (with Analytical Uncertainties)................................................................................................................. 5-23 Table 5-5: Mode I Stress Intensity Factor Model Description................................................................. 5-28 Table 6-1: Allowable Flaw Size or Lengths............................................................................................... 6-3 Table 7-1: Summary of Support Shoe Stress Intensity Factors versus Fracture Toughness...................... 7-4 Table 7-2: Summary of Top Plate Stress Intensity Factors versus Fracture Toughness............................. 7-5 Table 7-3: Summary of Bottom Plate Stress Intensity Factors versus Fracture Toughness....................... 7-6 Table 7-4: Summary of Diaphragm Plate Stress Intensity Factors versus Fracture Toughness................. 7-8 Table 7-5: Summary of Diaphragm Plate with Hole Stress Intensity Factors versus Fracture Toughness 7-8 Table 7-6: Summary of I-Beam Stress Intensity Factors versus Fracture Toughness................................ 7-9 Table 7-7: Summary of Leveling Bolt Stress Intensity Factors versus Fracture Toughness.................... 7-10 Table 7-8: Summary of Anchor Bolt Stress Intensity Factors versus Fracture Toughness...................... 7-11 Table 8-1: Summary of H.B. Robinson Limiting Stress Intensity Factors versus Fracture Toughness..... 8-5

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi WCAP-18939-NP January 2025 Revision 1 LIST OF FIGURES Figure 2-1: Fracture Mechanics Approach Flowchart............................................................................... 2-2 Figure 3-1: H.B. Robinson Unit 2 Support Location and Orientation [9.b].............................................. 3-4 Figure 3-2: H.B. Robinson Unit 2 Support System Plan View.................................................................. 3-5 Figure 3-3: H.B. Robinson Unit 2 RPV Support Type 1 Model Front Face.............................................. 3-6 Figure 3-4: H.B. Robinson Unit 2 RPV Support Type 2 Model Front Face.............................................. 3-6 Figure 3-5: H.B. Robinson Unit 2 RPV Support Type 1 Full Modeled View............................................ 3-7 Figure 3-6: H.B. Robinson Unit 2 RPV Support Type 2 Full Modeled View............................................ 3-8 Figure 3-7: H.B. Robinson Unit 2 RPV Support Type 1 Back Modeled View.......................................... 3-9 Figure 3-8: H.B. Robinson Unit 2 RPV Supports Type 2 Back Modeled View...................................... 3-10 Figure 4-1: Cut Orientation for the Support Shoe...................................................................................... 4-3 Figure 4-2: Postulated Flaws and Loads in the Leveling Bolts and Anchor Bolts..................................... 4-9 Figure 5-1: [

]a,c,e........... 5-8 Figure 5-2: [

]a,c,e 5-9 Figure 5-3: [

]a,c,e................ 5-9 Figure 5-4: Change in Nil-Ductility Transition Temperature, NDTT as a Function of dpa per Figure 3-1 of NUREG-1509 [4]...................................................................................................... 5-13 Figure 5-5: View of the Model Geometry with Part-Length Shield Assemblies at 0 Degrees................ 5-14 Figure 5-6: View of the Model Geometry with Part-Length Shield Assemblies at 10 Degrees.............. 5-15 Figure 5-7: View of the Model Geometry (no Part-Length Shield Assemblies) at 80 Degrees............... 5-16 Figure 5-8: Plan View of the Model Geometry at the Core Midplane..................................................... 5-17 Figure 5-9: Plan View of the Model Geometry 190 cm above the Core Midplane................................. 5-18 Figure 5-10: Dimensional View of the Bioshield and Pressure Vessel Support...................................... 5-19 Figure 5-11: Iron Atom Displacement through the RPV Support at 70 EFPY........................................ 5-20 Figure 5-12: Iron Atom Displacement through the RPV Support at 70 EFPY with +20% Bias Incorporated to Reflect Analytical Uncertainty................................................................................... 5-21 Figure 5-13: Postulated Surface Flaw in a Plate Model........................................................................... 5-25 Figure 5-14: Postulated Single Edge Through-Wall Flaw in a Plate Model............................................ 5-26 Figure 5-15: Postulated Flaws in a Round Bar Model............................................................................. 5-27

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vii WCAP-18939-NP January 2025 Revision 1 EXECUTIVE

SUMMARY

As part of the H.B. Robinson Unit 2 80-year subsequent license renewal (SLR) application, fracture mechanics analyses are performed to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) supports due to extended plant operation past 60 years. The United States (U.S.) Nuclear Regulatory Commission (NRC) has released Generic Aging Lessons Learned (GALL) SLR Report NUREG-2191 [1] which provides information regarding license renewal for 80 years.

NUREG-2191 does not include guidance for analysis of loss of fracture toughness due to irradiation embrittlement for structural steel support components in the vicinity of the RPV; however, recent U.S. NRC meetings and presentations have provided preliminary irradiation embrittlement of RPV steel supports evaluation guidance [2]. In addition, the U.S. NRC has recently provided draft guidance for the assessment of the RPV structural steel supports based on a fracture mechanics evaluation to account for neutron embrittlement (radiation effects) in the draft revision of NUREG-2191 [2.e]. Note that the radiation effects on RPV supports were previously investigated and resolved as part of GSI-15 in NUREG-0933, Revision 3 [3], NUREG-1509 (published in May 1996) [4] and NUREG/CR-5320 (published in 1989) [5].

The conclusions in NUREG-0933, Revision 3 stated that the supports were acceptable for continued operation to 60 years of life and GSI-15 was resolved.

Therefore, in this report, a detailed linear elastic fracture mechanics (LEFM) evaluation is performed following the general guidance of ASME Section XI [6] to investigate brittle fracture of the structural steel supports per NUREG-0933, Revision 3 and NUREG-1509 for 80 years of plant life. There are two types of support designs within the H.B. Robinson Unit 2 RPV support system, see Section 3, Figure 3-1 through Figure 3-8. Seven separate structural steel components within both support designs were evaluated which included the support shoe, top plate, bottom plate, diaphragm plates, I-beams, leveling bolts, and anchor bolts.

The LEFM evaluation is completed by calculating the stress intensity factor and comparing it to the fracture toughness of the various analyzed components. The stress intensity factors are determined for the various loading combinations at H.B. Robinson Unit 2 RPV supports based on a [

]a,c,e The fracture toughness considered for the RPV supports are based on the ASME Section XI Code Case N-830 [21] 95% lower bound KJc Master Curve. For conservatism, the minimum fracture toughness from the KJc Master Curve of 22.9 ksiin is used for the bottom plate, diaphragm plates, and leveling bolts RPV support components. The use of the minimum lower bound fracture toughness value represents an infinite amount of neutron embrittlement on the RPV structural steel supports to demonstrate that the structural steel supports are acceptable for 80 years (70 effective full power years [EFPY]). However, the plate with hole flaw model considered for the diaphragm plates required a material-specific fracture toughness. For consistency with the top plate and diaphragm plates, a material-specific fracture toughness was also calculated for the bottom plate component. Thus, the material-specific fracture toughness is calculated for the support shoe, top plate, bottom plate, diaphragm plates, I-beams, and anchor bolt components considering the impact of neutron embrittlement at 80 calendar years (70 EFPY), to provide adequate margin in the RPV support structural integrity assessment. The material-specific fracture toughness at 70 EFPY is calculated using the KJc Master Curve as follows: support shoe is 45.5 ksiin, top plate is 38.0 ksiin, bottom plate is 30.5 ksiin, diaphragm plates are 30.8 ksiin, I-beams are 30.8 ksiin, and anchor bolts are 27.7 ksiin.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 viii WCAP-18939-NP January 2025 Revision 1 Based on the magnitude of the calculated stress intensity factors for the RPV components discussed in Section 8 of this report, it can be demonstrated that the stress intensity factors remain less than the fracture toughness values while considering 80 years (70 EFPY) of neutron radiation embrittlement. [

]a,c,e As discussed in Section 8.1, as part of VT-3 examinations performed in RO-32 (2020), it was observed that borated water from a historical reactor cavity seal leakage as well as a historical leakage from the reactor cavity seal access (sand plugs) during past outages had resulted in dry inactive boric acid deposit build ups on the RPV A, B, and C Nozzle supports. During RO-28 (Fall 2013), a permanent cavity seal plate was installed to eliminate the source of the leakage. A corrosion evaluation was performed in [35] due to the presence of boric acid and evaluated through RO-33 (Fall 2022). In order to perform the VT-3 visual examination in RO-33, the support component areas have been cleaned to the best ability possible with a steam cleaning process. The RO-33 VT-3 visual examination concluded that there is no visual evidence of structural deformation or degradation that impedes the current integrity of the components support to perform the intended function. Furthermore, there is no visual evidence of detached, broken loosened component support items, misalignment of existing support component items, and material wastage. It was also visually noted during the VT-3 visual examination that there was light/moderate boron identified throughout, and general surface corrosion exists which does not reduce the load bearing capacity of the support (no visual evidence of material wastage). Furthermore, a reduction of wall thickness due to past corrosion per [35] is conservatively considered in the fracture mechanics evaluation for all seven RPV support components herein. Nevertheless, the RPV support maintains the current structural integrity and continues to perform the intended function.

For the seven analyzed components within the RPV support system, it was concluded that there was adequate margin between the stress intensity factor and the fracture toughness. Thus, based on the detailed conclusions in Section 8, the RPV support system at H.B. Robinson Unit 2 is structurally stable (i.e., flaw tolerant) considering 80 years of radiation embrittlement effects on the supports. Therefore, no additional inspections or enhancements are required for ageing management of the RPV supports, and the current ASME Section XI inspection requirements are sufficient.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 WCAP-18939-NP January 2025 Revision 1 1

INTRODUCTION As plants apply for 80-year licensure (Subsequent License Renewal-SLR), the United States (U.S.) Nuclear Regulatory Commission (NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of GSI-15 in NUREG-0933, Revision 3 [3], NUREG-1509 (published in May 1996) [4]

and NUREG/CR-5320 (published in 1989) [5]. The conclusions in NUREG-0933, Revision 3 stated that the supports were acceptable for continued operation and GSI-15 was resolved, as follows:

The preliminary conclusion indicated that the potential problem [embrittlement of supports due to radiation effects] did not pose an immediate threat to public health and safety. The above tentative results indicated that plant safety could be maintained despite RVSS [reactor vessel support structures] radiation damage. In order to encompass the uncertainties in the various analyses and provide an overall conservative assessment, several structural analyses conducted demonstrated the following:

1.

Postulating that one of the four RPV supports was broken in a typical PWR, the remaining supports would carry the reactor vessel load even under SSE [safe-shutdown earthquake]

seismic loads;

2.

If all supports were assumed to be totally removed (i.e., broken), the short span of piping between the vessel and the shield wall would support the load of the vessel.

The results of the analyses virtually eliminated the concern for both radiation embrittlement and significant structural damage from a postulated RPV failure. Based on the staffs regulatory analysis, the issue was RESOLVED with no new requirements. Consideration of a license renewal period of 20 years did not change this conclusion.

Based on conclusions in NUREG-0933 and U.S. NRC Memorandums on GSI-15 [7], it was concluded that the RPV supports were not a concern for the entirety of its plant life (i.e., 40 and 60 years); even in the extreme case where all the supports were totally removed (i.e., broken), the piping has acceptable margin to carry the load of the vessel. However, for plants applying for 80-year life licensure, the U.S. NRC has requested a re-assessment of the RPV structural steel supports based on a fracture mechanics evaluation to account for neutron embrittlement [2]. Thus, in the report herein, a detailed fracture mechanics structural integrity evaluation is performed on H.B. Robinson Unit 2 RPV structural steel supports to support the U.S.

NRC request for re-assessment of the RPV structural steel supports to account for 80 years of radiation effects. The GALL-SLR report NUREG-2191 [1] for 80-year license renewal does not include guidance for analysis of loss of fracture toughness due to irradiation embrittlement for structural steel support components in the vicinity of the RPV. However, recent U.S. NRC meetings, presentations, and the draft revision of NUREG-2191 Volume 2 [2] have provided preliminary irradiation embrittlement of RPV steel supports evaluation guidance, which states that NUREG-1509 provides general guidance and an acceptable approach to evaluate the loss of fracture toughness of the RPV supports due to radiation effects for long term operation.

There are two potential fracture mechanics strategies that are identified to resolve the radiation embrittlement concern based on NUREG-0933 and NUREG-1509. One approach was to compare the lowest service temperature (LST) with the material adjusted reference temperature, with incorporation of irradiation effects. If the LST is higher than the material adjusted reference temperature, then the RPV

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-2 WCAP-18939-NP January 2025 Revision 1 supports are acceptable. Historically, it was determined that the LST method would not provide sufficient margin, hence this methodology will not be completed for H.B. Robinson structural steel supports.

The second approach is to investigate brittle fracture of the structural steel supports per NUREG-1509 to perform a detailed fracture mechanics evaluation in order to calculate the stress intensity factor and compare it to fracture toughness similar to the general guidance of ASME Section XI Appendix A. Based on the magnitude of the stress intensity factor, the analysis could be used to demonstrate that the stress intensity factor values remain sufficiently small as compared to the fracture toughness values based on 80 years (70 EFPY) of neutron embrittlement. The stress intensity factors are calculated based on [

]a,c,e Note that current ASME Section XI IWF in-service inspection for supports require only a visual examination (VT-3). The objective of the analysis is to demonstrate sufficient level of flaw tolerance to justify continuing the current visual examination (VT-3).

It should be noted, as part of the overall resolution of GSI-15 in NUREG-1509, a detailed fracture mechanics evaluation was performed in 1989 for one of the pilot plants in NUREG/CR-5320, namely, Turkey Point Unit 3 (the other plant was Trojan). The conclusions in NUREG/CR-5320 indicated that for the most severe credible loading (deadweight plus large break LOCA) at 32 effective full power years (EFPY) (i.e., 40 calendar years), the best-estimated minimum critical flaw size depth is 0.3 inches for the structural support beams. Furthermore, the study had concluded that calculated flaw size is insensitive to reactor operating time after ~10 EFPY, and at startup (0 EFPY) the size is 0.6 inches. The study in NUREG/CR-5320 also demonstrated that considering uncertainties in the fracture toughness, initial nil-ductility transition temperature (NDTTo), and the operating temperature of the support components, the

+/-1 (one standard deviation) values of the critical flaw size at 32 EFPY are ~0.2 and 0.6 inches. For the loading case of deadweight plus safe shutdown earthquake, the critical flaw size is substantially larger (1.1 inches as compared with 0.3 inches) at 32 EFPY.

Even though the fracture mechanics study performed for Turkey Point Unit 3 and Trojan in NUREG/CR-5320 (conducted in 1989) had calculated small critical flaw sizes that could be of a source of concern for brittle fracture, the U.S. NRC staff in 1996 had reviewed several other structural analyses, in addition to the fracture mechanics evaluation. Based on the U.S. NRCs review of other structural consequence analysis, a final conclusion (as stated in NUREG-0933 of GSI-15) was reached that even if one of the RPV supports were broken in a PWR, the remaining supports would safely carry the RPV load under seismic events. The structural analyses also concluded that even if all the RPV supports were broken, the short span of piping between the vessel and the shield wall would support the load of the RPV.

As a result, the fracture mechanics evaluation in NUREG/CR-5320 can be considered a defense-in-depth study of the structural integrity of the RPV supports, as supplemented by the structural analysis which demonstrated that the piping can withstand the load of the RPV after failure of all vessel supports. It should be noted that the flaw sizes postulated in the NUREG/CR-5320 would have been identified during original fabrication (pre-service inspection) of the support welds either by dye penetrant testing, magnetic particle or even ultrasonic equipment as required by ASME (American Society of Mechanical Engineers), AISC (American Institute of Steel Construction), and AWS (American Welding Society).

The goal of the analysis herein for H.B Robinson Unit 2 for the 80-year license renewal is to keep consistent with the overall methodology that had been previously accepted by the industry and U.S. NRC in NUREG-0933, NUREG-1509, and NUREG/CR-5320, while at the same time demonstrate that the RPV supports are structurally stable with consideration of neutron embrittlement for plant life extension to

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-3 WCAP-18939-NP January 2025 Revision 1 80 years. The fracture mechanics flaw tolerance analysis methodology provided in this report has been previously reviewed and accepted by the U.S. NRC for other plants subsequent license renewal applications, such as for Point Beach [41] and St. Lucie [42].

Thus, the assessment in this report for H.B. Robinson Unit 2 RPV structural steel supports determines the stress intensity factors and compares them to fracture toughness to investigate the impact of neutron embrittlement (radiation effects) for an operating life of 80 years. The general methodology of the fracture mechanics evaluation is described in Section 2. H.B. Robinson Unit 2 support configuration, materials, and geometry are provided in Section 3 of this report. Section 4 describes the plant-specific loading conditions and the plant-specific stresses used in the evaluation, while Section 5 will provide information regarding fracture toughness, plant-specific neutron embrittlement, methodology for stress intensity factors and postulated flaw sizes. Section 6 provides additional information regarding the postulated allowable flaw sizes. Section 7 provides the fracture mechanics results, i.e., stress intensity factors, that were determined for the supports to demonstrate structural stability based on the linear elastic fracture mechanics evaluations.

Section 8 provides the final conclusions of the fracture mechanics evaluation. All cited references are provided in Section 9.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-1 WCAP-18939-NP January 2025 Revision 1 2

GENERAL METHODOLOGY The goal of the fracture mechanics evaluation is to demonstrate that brittle fracture is not a concern for the RPV structural steels at H.B. Robinson Unit 2 based on 80 years (70 EFPY) of neutron embrittlement.

Linear elastic fracture mechanics (LEFM) will be used as a conservative methodology to evaluate the structural integrity of the supports. The LEFM methodology is illustrated in a flow chart format (see Figure 2-1) based on the guidance provided in NUREG-1509 [4] for a fracture mechanics approach to account for radiation effects on RPV support steels. The LEFM methodology is briefly described in the following paragraphs.

The LEFM evaluation is completed by calculating a stress intensity factor and comparing it to a fracture toughness value. The limiting component for the fracture mechanics analysis is based on a combination of component geometry, operating condition, stress, material property, and neutron embrittlement. H.B.

Robinson Unit 2 support configurations are provided in Section 3 of this report. The H.B. Robinson Unit 2 specific loading conditions (i.e., normal and accident) are considered in this report. Consideration of all applicable loading conditions, such as deadweight, seismic, loss-of-coolant accident, welding residual stresses, and thermal stresses are accounted for in the analysis as described in Section 4 of this report.

[

]a,c,e Stress intensity factors will be considered as described in Section 5.2 of this report based on the component geometry. [

]a,c,e are also included in this report for informational purposes, see Section 5.2 for additional discussion. These stress intensity factors are then compared to the fracture toughness. The fracture toughness for the RPV support components are based on the 95% lower bound KJc Master Curve. The minimum fracture toughness from the KJc Master Curve of 22.9 ksiin is used for the bottom plate, diaphragm plates, and leveling bolts. The use of the minimum lower bound fracture toughness value represents an infinite amount of neutron embrittlement on the RPV structural steel supports. Material-specific fracture toughness for the support shoe, top plate, bottom plate, diaphragm plates, I-beams, and anchor bolts are calculated using the KJc Master Curve to provide additional margin in the fracture mechanics assessment for these components.

The material-specific fracture toughness for the support shoe, top plates, bottom plate, diaphragm plates, I-beams, and anchor bolts will be [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-2 WCAP-18939-NP January 2025 Revision 1 Geometry

• Plate

• Bar Postulated Flaw Size

[

]a,c,e Stress Intensity Factor (KI)

Temperature

• Plant-specific operating bulk material temperature is used for calculating material-specific fracture toughness.

Material-Specific or Minimum Lower Bound Fracture Toughness based on KJc per Code Case N-830 Loads/Stresses

• Deadweight, Thermal, Seismic, Pipe Break

• Loading Condition: Normal and Accident.

Neutron Embrittlement

• Material-specific iron dpa based on end of license, 70 EFPY (80 calendar years)

• Iron dpa (E > 0.1 MeV) to determine NDTT per Figure 3-1 of NUREG-1509 is used for calculating material-specific fracture toughness.

Stress intensity factors are compared to a material-specific or minimum lower bound fracture toughness for each RPV support component.

If the acceptance criteria are met, RPV structural steel support component is flaw tolerant for 70 EFPY (80 calendar years)

Continue current ASME Section XI 10-year Inspection Figure 2-1: Fracture Mechanics Approach Flowchart Acceptance Criteria for RPV Support Components Stress intensity factor is less than the fracture toughness

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-1 WCAP-18939-NP January 2025 Revision 1 3

SUPPORT CONFIGURATION, MATERIAL, AND GEOMETRY This section of the report describes the general H.B. Robinson RPV support configuration, material designation, and geometry.

H.B. Robinson Unit 2 is a 3-loop layout where two types of supports are shown (i.e., support type 1 at 80° and 320° locations and support type 2 at the 200° location), see Figure 3-1. Both the RPV support structures (short-columns) consists of a support shoe attached to the top of the support by leveling bolts. The three RPV supports are positioned under the inlet nozzles, see Figure 3-2. The inlet nozzle support pad weld build-up rest on top of the support shoe. The support pad weld build-up is part of the nozzle component and not included in the RPV support evaluation herein.

Both types of supports (i.e., support types 1 and 2 per Figure 3-1) contain a support shoe, that is a structural member that transmits the support loads to the supporting structure. The support configurations are designed to restrain vertical, lateral, and rotational movement of the RPV but to allow for thermal growth by permitting radial sliding on the shim plates. The shim plates are designed to allow for thermal expansion of the RPV nozzle by providing an interface surface between the nozzle and the support shoe. The top plate, which is fastened to the bottom of the support shoe using leveling bolts, is then fillet welded to the I-beams that sit below the top plate. The I-beams for support types 1 and 2 have slightly different configurations.

Both supports 1 and 2 contain three I-beams that are fixed under the top plate. However, support type 1 has two I-beams that are perpendicular to three inner I-beams, while support 2 does not have the two additional perpendicular I-beams in the support configuration. See Figure 3-3 through Figure 3-8 for both support configurations. The bottom plate, which sits on the concrete, is welded to the bottom of the I-beams for both support types 1 and 2. In addition, diaphragm plates are positioned between the top plate, bottom plate, and I-beam components, for both support types 1 and 2. These diaphragm plates are fillet welded to the surrounding components. Lastly, both RPV support structure types contain anchor bolts that span through the top plate, I-beams, and bottom plate, into the concrete, which then connect to anchor plates that are embedded into the concrete. The vertical loadings are transmitted from the reactor vessel nozzle pads into the concrete, through the steel shim plates, forged shoe, top plate, I-beams, base plate, anchor bolts, and the grout pad. The horizontal loadings are transmitted from the reactor vessel nozzle pads into the concrete, through the steel shim plates, forged shoe, leveling bolts, top plate, fillet welds on the top of the I-beams, through the I-beams, fillet welds on the bottom of the I-beams, base plate, and the anchor bolts.

The H.B. Robinson supports are designed and fabricated in accordance with the Ebasco Specification for Structural Steel Equipment Supports [8]. The following seven RPV support steel components are considered for the fracture mechanics evaluation: support shoe, top plate, bottom plate, diaphragm plates, I-beams, leveling bolts, and anchor bolts. These components are chosen as these locations could experience tensile stresses and/or high embrittlement effects. Stress intensity factors are calculated for each component and compared to an appropriate fracture toughness value. The geometry of interest for the previously mentioned RPV support components are based on drawings [9] and [10], and the type of material of the components are provided in the following paragraphs as confirmed per [11]. A short discussion regarding the shim plates is also provided in the last paragraph of this section.

In the stress intensity factor calculations, the thickness for each component is reduced by 0.046 [35] due to the consideration of corrosion. Corrosion is conservatively considered in the fracture mechanics evaluation for each component cut path based on the current conditions of the RPV support structures.

Further discussion on corrosion is provided in Section 8.1. Note that the thickness provided in the short

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-2 WCAP-18939-NP January 2025 Revision 1 description for each component as shown below are the nominal thicknesses based on the RPV support drawing geometry without the reduction of corrosion.

The information provided for each component in the following paragraphs are applicable for support types 1 and 2.

Support Shoe (Figure 3-3 and Figure 3-4)

[

]a,c,e Top Plate (Figure 3-3 and Figure 3-4)

[

]a,c,e Bottom Plates (Figure 3-3 and Figure 3-4)

[

]a,c,e Diaphragm Plates (Figure 3-3, Figure 3-4, Figure 3-7, and Figure 3-8)

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-3 WCAP-18939-NP January 2025 Revision 1 I-Beams (Figure 3-3 and Figure 3-4)

[

]a,c,e Leveling Bolts (Figure 3-5 and Figure 3-6)

[

]a,c,e Anchor Bolts (Figure 3-3, Figure 3-4, Figure 3-5, and Figure 3-6)

[

]a,c,e Shim Plates (Figure 3-3 and Figure 3-4)

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-4 WCAP-18939-NP January 2025 Revision 1 Figure 3-1: H.B. Robinson Unit 2 Support Location and Orientation [9.b]

a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-5 WCAP-18939-NP January 2025 Revision 1 Figure 3-2: H.B. Robinson Unit 2 Support System Plan View a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-6 WCAP-18939-NP January 2025 Revision 1 Figure 3-3: H.B. Robinson Unit 2 RPV Support Type 1 Model Front Face Figure 3-4: H.B. Robinson Unit 2 RPV Support Type 2 Model Front Face a,c,e a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-7 WCAP-18939-NP January 2025 Revision 1 Figure 3-5: H.B. Robinson Unit 2 RPV Support Type 1 Full Modeled View a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-8 WCAP-18939-NP January 2025 Revision 1 Figure 3-6: H.B. Robinson Unit 2 RPV Support Type 2 Full Modeled View a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-9 WCAP-18939-NP January 2025 Revision 1 Figure 3-7: H.B. Robinson Unit 2 RPV Support Type 1 Back Modeled View a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-10 WCAP-18939-NP January 2025 Revision 1 Figure 3-8: H.B. Robinson Unit 2 RPV Supports Type 2 Back Modeled View a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-1 WCAP-18939-NP January 2025 Revision 1 4

LOADING CONDITIONS AND STRESS ANALYSIS The stress intensity factors are calculated using H.B. Robinson Unit 2 plant-specific stresses, component geometry, and [

]a,c,e, which are then compared to an applicable fracture toughness. The stress intensity factors are calculated for the following seven components for the support types 1 and 2: support shoe, top plate, bottom plate, diaphragm plates, I-beams, leveling bolts, and anchor bolts. The design basis loading combinations for the H.B Robinson Unit 2 RPV supports are as follows:

[

]a,c,e The normal inward and normal outward loading combinations consider a seismic load, where inward and outward refer to the radial load direction applied to the reactor pressure vessel support. These loading combinations are consistent with the analysis of record.

The RPV support components analyzed represent the locations of highest stresses and/or locations subjected to high neutron embrittlement within the RPV support system. The normal inward, normal outward, and accident stresses or loads used to evaluate the seven components are provided in Table 4-1 through Table 4-8. These stresses are plant-specific to H.B. Robinson Unit 2. [

]a,c,e The typical stress components considered [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-2 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e The stress intensity factor methodology for each of the seven support components is described in Section 5.2 of this report.

Welding residual stress (WRS) is also considered for the welded components within the RPV support structure, which includes the top plate, bottom plate, diaphragm plates, and I-beams. [

]a,c,e The following sections in this report provide the actual plant-specific stress on each of the seven components for support types 1 and 2:

Section 4.1: Support Shoes Section 4.2: Top Plates Section 4.3: Bottom Plates Section 4.4: Diaphragm Plates Section 4.5: I-beams Section 4.6: Leveling Bolts Section 4.7: Anchor Bolts

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-3 WCAP-18939-NP January 2025 Revision 1 4.1 SUPPORT SHOE STRESSES The plant-specific stresses on the support shoes are provided for both the horizontally oriented cuts and vertically oriented cuts on support types 1 and 2 in Table 4-1. The stresses are combined based on [

]a,c,e The stresses in Table 4-1 are used to calculate the stress intensity factors provided in Section 7.1. An illustration of the support shoe and the cut orientations are shown in Figure 4-1.

Table 4-1: Stresses for the Support Shoes Figure 4-1: Cut Orientation for the Support Shoe a,c,e a,c,e

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Westinghouse Non-Proprietary Class 3 4-4 WCAP-18939-NP January 2025 Revision 1 4.2 TOP PLATE STRESSES The plant-specific stresses on the top plates are provided for support types 1 and 2 in Table 4-2. The stresses are combined based on [

]a,c,e The stresses in Table 4-2 are used to calculate the stress intensity factors provided in Section 7.2.

Table 4-2: Stresses for the Top Plates a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-5 WCAP-18939-NP January 2025 Revision 1 4.3 BOTTOM PLATE STRESSES The plant-specific stresses on the bottom plates are provided for support types 1 and 2 in Table 4-3. The stresses are combined based on [

]a,c,e The stresses in Table 4-3 are used to calculate the stress intensity factors provided in Section 7.3.

Table 4-3: Stresses for the Bottom Plates a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-6 WCAP-18939-NP January 2025 Revision 1 4.4 DIAPHRAGM PLATE STRESSES The plant-specific stresses on the diaphragm plates are provided for support types 1 and 2 in Table 4-4. The stresses are combined based on [

]a,c,e The stresses in Table 4-4 are used to calculate the stress intensity factors provided in Section 7.4.

In addition to the cut paths analyzed through the body of the diaphragm plates, cut paths were considered near the holes in the diaphragm plates, see Figure 3-3 and Figure 3-4. The limiting stresses for a cut near a hole in a diaphragm plate are provided in Table 4-5; note, that support type 2 stresses bound those for support type 1. As previously stated, the stresses are combined based on [

]a,c,e The stresses in Table 4-5 are used to calculate a stress intensity factor provided in Section 7.4.

Table 4-4: Stresses for the Diaphragm Plates Table 4-5: Limiting Stresses for the Diaphragm Plate near a Hole a,c,e a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-7 WCAP-18939-NP January 2025 Revision 1 4.5 I-BEAM STRESSES The plant-specific stresses on the I-beams are provided for support types 1 and 2 in Table 4-6. The stresses are combined based on [

]a,c,e The stresses in Table 4-6 are used to calculate the stress intensity factors provided in Section 7.5.

Table 4-6: Stresses for the I-Beams a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-8 WCAP-18939-NP January 2025 Revision 1 4.6 LEVELING BOLT LOADS The plant-specific forces and moments on both the leveling bolts located in the support shoe are provided for support types 1 and 2 in Table 4-7. Each of the leveling bolts were evaluated individually in the fracture mechanics analysis herein. The forces and moments are shown in Figure 4-2. [

]a,c,e The loads in Table 4-7 are used to calculate the stress intensity factors provided in Section 7.6.

Table 4-7: Forces and Moments for the Leveling Bolts a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-9 WCAP-18939-NP January 2025 Revision 1 4.7 ANCHOR BOLT LOADS The plant-specific forces and moments for the anchor bolts, located through the top plate, I-beams, and bottom plate of the RPV supports, are provided for the normal inward, normal outward, and accident combinations in Table 4-8, and shown in Figure 4-2. Each of the thirty-four anchor bolts for support type 1 and twenty-four anchor bolts for support type 2 were evaluated individually in the fracture mechanics analysis herein. [

]a,c,e The limiting anchor bolt loads in Table 4-8 are used to calculate the stress intensity factors provided in Section 7.7.

Table 4-8: Forces and Moments for the Anchor Bolts Figure 4-2: Postulated Flaws and Loads in the Leveling Bolts and Anchor Bolts a,c,e a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-1 WCAP-18939-NP January 2025 Revision 1 5

FRACTURE MECHANICS METHODOLOGY As discussed in Section 2, the linear elastic fracture mechanics methodology is used to evaluate the structural integrity of the H.B. Robinson RPV structural steel supports. The goal of the evaluation in this report is to demonstrate that the calculated stress intensity factors for the RPV support components remain below the applicable fracture toughness. The fracture toughness values analyzed in this report are either based on the minimum value of the 95% lower bound Master Curve (KJc = 22.9 ksiin) per [21], which inherently assumes infinite embrittlement, or a calculated fracture toughness based on 80 years of neutron embrittlement when additional margin is required between the stress intensity factor and fracture toughness.

The use of the 95% lower bound Master Curve for the fracture mechanics evaluation of the RPV supports has been previously approved by the U.S. NRC for other plants subsequent license renewal applications (i.e., St. Lucie [42]). The stress intensity factors are determined by [

]a,c,e The determination of material-specific fracture toughness values are discussed in Section 5.1, while the discussion for stress intensity factor methodology is provided in Section 5.2.

5.1 FRACTURE TOUGHNESS DETERMINATION For pressure vessel steels, it would be appropriate to use the KIc or KIR fracture toughness curves in ASME Section XI [6] and ASME Section III Appendix G-2000 [20], respectively; however, not all of the RPV support materials (some of which are high strength materials) are comparable to the steels which were tested to generate the ASME Section XI and Section III curves (i.e., SA-533 Grade B Class 1 and SA-508 Class 1, 2, and 3). [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-2 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e 5.1.1 Fracture Toughness for the H.B. Robinson Unit 2 RPV Supports The fracture toughness used in the comparison to stress intensity factors for the RPV support components, as well as the inclusion of embrittlement and strain rate effects is discussed in the following sections. The material-specific fracture toughness is calculated for the support shoe, top plate, diaphragm plates, I-beams, and anchor bolts considering neutron embrittlement for 70 EFPY. The minimum value of the 95% lower bound Master Curve (KJc = 22.9 ksiin), which inherently assumes infinite embrittlement, is conservatively used for the bottom plate, diaphragm plates, and leveling bolts since there is sufficient margin between stress intensity factor and fracture toughness. However, material-specific fracture toughness is also calculated for the bottom plate, since it is the same material as the top plate, and diaphragm plates. A summary of the fracture toughness for each component is provided in Table 5-1.

5.1.1.1

[

]a,c,e Fracture Toughness (Support Shoe)

The support shoe is [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-3 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e 5.1.1.2

[

]a,c,e Fracture Toughness (Top, Bottom, and Diaphragm Plates)

The top plates, bottom plates, and diaphragm plates are made [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-4 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e 5.1.1.3

[

]a,c,e Fracture Toughness (I-Beams)

The I-beams are of [

]a,c,e 5.1.1.4

[

]a,c,e Fracture Toughness (Leveling Bolts)

The leveling bolts that are in the support shoe are made of [

]a,c,e 5.1.1.5

[

]a,c,e Fracture Toughness (Anchor Bolts)

The anchor bolts are made of [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-5 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e 5.1.1.6 Weld The welded plates within the structural portions of the H.B. Robinson Unit 2 RPV supports structure are

[

]a,c,e 5.1.1.7 Strain Rate Effects Per the guidance in Section 4.3.3.1 of NUREG-1509 [4] strain rates associated with dynamic loading for earthquake or pipe break scenarios should be addressed (i.e., the rate of load application). Per

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-6 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-7 WCAP-18939-NP January 2025 Revision 1 Table 5-1: Fracture Toughness for the H.B. Robinson Unit 2 RPV Support Components a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-8 WCAP-18939-NP January 2025 Revision 1 Figure 5-1: [

]a,c,e

[

]a,c,e a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-9 WCAP-18939-NP January 2025 Revision 1 Figure 5-2: [

]a,c,e Figure 5-3: [

]a,c,e a,c,e a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-10 WCAP-18939-NP January 2025 Revision 1 5.1.2 Bulk Metal Temperature The bulk material temperature (T) is required to determine fracture toughness; however, bulk material temperature of the H.B. Robinson Unit 2 RPV support components is unavailable and/or not measured.

[

]a,c,e 5.1.3 Neutron Embrittlement and Nil-Ductility Transition Temperature A Westinghouse Owners Group (WOG) (now known as Pressurized Water Reactor Owners Group

[PWROG]) program was performed in WCAP-14422, Revision 2-A [29] during the late 1990s and completed in the year 2000 that reassessed the aging effects from neutron embrittlement on the RPV supports for the first license renewal program (60 years). The assessment in WCAP-14422 referenced the extensive industry research and plant-specific evaluations performed in NUREG-1509, NUREG/CR-5320 and the resolution of GSI-15 in NUREG-0933 to conclude that aging management is not a concern for the RPV supports for 60 years of plant life operation.

In the final U.S. NRC safety evaluation for WCAP-14422, the U.S. NRC staff concluded the staff considers that neutron embrittlement is not a concern for the supports and does not warrant an aging management program [29]. The conclusion was based on an evaluation that shows that if all the supports failed, the short span of piping between the vessel and the shield wall would support the load of the vessel. This eliminated the staffs concern with RPV support embrittlement.

The embrittlement prediction models developed in NUREG/CR-5320 and used in the fracture mechanics analysis of Trojan and Turkey Point, were discussed in Section 2.3 of NUREG-1509. The major issue of the embrittlement curve in NUREG/CR-5320 was that only fast neutron fluence (E >1.0 MeV) was considered to cause embrittlement damage. Sections 2 and 3 of NUREG-1509 concluded that low-energy neutron irradiation (below 1 MeV) could potentially make a significant contribution to the observed embrittlement.

Therefore, the guidance provided in NUREG-0933 GSI-15 resolution is to utilize Figure 3-1 of NUREG-1509 (reproduced herein as Figure 5-4) to calculate the change in nil-ductility transition temperature (NDTT) based on dpa for the energy spectrum E > 0.1 MeV.

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-11 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e The impact of neutron embrittlement (radiation effects) on the H.B. Robinson Unit 2 RPV supports was determined for 70 EFPY (80 calendar years of operation). The neutron embrittlement is defined as a function of iron displacement per atom (dpa) consistent with Figure 3-1 of NUREG-1509 (reproduced herein as Figure 5-4) to determine change in nil-ductility transition temperature (NDTT). The neutron transport methodology used to generate the iron atom displacement data followed the guidance of U.S. NRC Regulatory Guide 1.190 [31], and was consistent with the U.S. NRC approved methodology described in WCAP-18124-NP-A [32] and WCAP-18124-NP-A Revision 0 Supplement 1-NP-A [33]. Although this methodology has not been approved by the U.S. NRC for the RPV supports, the methodology has been generically approved for calculations of exposure of the RPV beltline (generally, RPV materials opposite the active fuel) [32] and extended beltline [33]. The following paragraphs describe the neutron transport methodology for H.B. Robinson Unit 2.

Discrete ordinates transport calculations were performed on a fuel-cycle-specific basis to determine the neutron and gamma ray environment within the reactor geometry. The transport calculations were carried out using the three-dimensional discrete ordinates code RAPTOR-M3G and the BUGLE-96 cross-section library. The BUGLE-96 library provides a coupled 47-meutron-, 20-gamma-ray-group cross-section data set produced specifically for light water reactor applications. Energy-and space-dependent core power distributions as well as system operating temperature were treated on a fuel-cycle-specific basis. The core power distributions were processed to produce spatial-and energy-dependent source distributions for use in the discrete ordinates transport calculations. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S16 order of angular quadrature.

Figure 5-5 shows a slice of the model geometry zones at 0°. This geometry features zones for the part length shield assemblies. Figure 5-6 shows a similar view of the geometry at the 10° position, which is where the outlet nozzle geometry was centered. Figure 5-7 shows a slice of the model geometry zones at 80°, which is where the inlet nozzle geometry was centered. This geometry was applied for fuel cycles that did not use part length shield assemblies. Also featured in Figure 5-7 is the zone for the reactor vessel support.

Figure 5-8 shows a plan view of the model geometry sliced at the core midplane. This figure features the instrument wells in the concrete bioshield. Figure 5-9 shows a plan view of the model geometry sliced near the bottom of the reactor vessel support. Figure 5-10 shows a 3-dimensional view of the concrete bioshield and the pressure vessel support.

In addition to the core, reactor vessel internals, RPV, and concrete bioshield, the RAPTOR-M3G model developed for this quadrant geometry includes explicit representations of the surveillance capsules, RPV cladding, and the insulation located external to the RPV.

The components of the RPV supports for which radiation exposure was calculated are the top plate, bottom plate, I-beams (also applicable to the diaphragm plates), leveling bolts, support shoe, and anchor bolts. Note the top plate and bottom plate are also referred to as the upper plate and base plate. The neutron exposure

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-12 WCAP-18939-NP January 2025 Revision 1 was calculated at the RPV support structure inner surface (closest to the RPV). Figure 5-11 and Figure 5-12 show the neutron exposure in terms of iron atom displacements through the reactor geometry at the RPV support. The RPV supports evaluation in this report conservatively considers iron atom displacement from all energy levels - the iron atom displacement values include contributions of fast neutrons with energies above 0.1 MeV as well as energies below 0.1 MeV.

The impact of neutron embrittlement (radiation effects), defined as iron displacement per atom (dpa) values, on the RPV supports for 70 EFPY are provided (without analytical uncertainties on embrittlement) in Table 5-2. The RPV extended beltline uncertainty analysis in WCAP-18124-NP-A Revision 0 Supplement 1-NP-A [33] quantified the analytical uncertainty associated with calculated fast neutron (E > 1.0 MeV) fluence rates at the RPV inner and outer surfaces at various elevations above and below the active fuel. The uncertainty analysis results determined at the RPV outer surface and elevations of 60 cm and 30 cm above the top of the active fuel were used as the starting point for estimating the uncertainty associated with the RPV support structure exposures with additional weight allocated to concrete composition uncertainty to account for streaming effects. Table 5-3 shows the calculated exposures to the RPV support subcomponents after application of the analytical uncertainty to the values reported in Table 5-2. The analytical uncertainty associated with the fast neutron (E > 1.0 MeV) fluence and iron atom displacement results was estimated to be 20%.

The iron atom displacement values that were used to determine NDTT at 70 EFPY are provided in Table 5-3 (with analytical uncertainties). The resulting NDTT calculated based on Figure 3-1 of NUREG-1509 (reproduced herein as Figure 5-4) using the upper bound curve is provided in Table 5-4. Note that the leveling bolts use a conservative lower bound fracture toughness; thus, NDTT is not included in Table 5-4 for the leveling bolt component. The iron atom displacement values provided in Table 5-2 and Table 5-3 for EFPYs other than 70 EFPY are included for informational purposes.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-13 WCAP-18939-NP January 2025 Revision 1 Figure 5-4: Change in Nil-Ductility Transition Temperature, NDTT as a Function of dpa per Figure 3-1 of NUREG-1509 [4]

Upper Bound Curve Medium Curve

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-14 WCAP-18939-NP January 2025 Revision 1 Figure 5-5: View of the Model Geometry with Part-Length Shield Assemblies at 0 Degrees

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-15 WCAP-18939-NP January 2025 Revision 1 Figure 5-6: View of the Model Geometry with Part-Length Shield Assemblies at 10 Degrees

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-16 WCAP-18939-NP January 2025 Revision 1 Figure 5-7: View of the Model Geometry (no Part-Length Shield Assemblies) at 80 Degrees

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-17 WCAP-18939-NP January 2025 Revision 1 Figure 5-8: Plan View of the Model Geometry at the Core Midplane

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-18 WCAP-18939-NP January 2025 Revision 1 Figure 5-9: Plan View of the Model Geometry 190 cm above the Core Midplane

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-19 WCAP-18939-NP January 2025 Revision 1 Figure 5-10: Dimensional View of the Bioshield and Pressure Vessel Support

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-20 WCAP-18939-NP January 2025 Revision 1 Figure 5-11: Iron Atom Displacement through the RPV Support at 70 EFPY

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-21 WCAP-18939-NP January 2025 Revision 1 Figure 5-12: Iron Atom Displacement through the RPV Support at 70 EFPY with +20% Bias Incorporated to Reflect Analytical Uncertainty

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-22 WCAP-18939-NP January 2025 Revision 1 Table 5-2: H.B. Robinson Unit 2 Iron Displacement per Atom at the RPV Support Structure (without Analytical Uncertainties)

Cumulative Operating Time (EFPY)

Iron Atom Displacements (dpa)(1)

Base Plate(2)

I-Beams &

Diaphragm Plates(3)

Upper Plate(2)

& Leveling Bolts Support Shoe Anchor Bolt 42 1.63E-03 1.56E-03 7.63E-04 6.12E-04 1.73E-03 48 1.89E-03 1.81E-03 8.85E-04 7.09E-04 1.98E-03 54 2.15E-03 2.06E-03 1.01E-03 8.06E-04 2.24E-03 60 2.41E-03 2.31E-03 1.13E-03 9.04E-04 2.49E-03 66 2.67E-03 2.55E-03 1.25E-03 1.00E-03 2.75E-03 70 2.84E-03 2.72E-03 1.33E-03 1.07E-03 2.92E-03 72 2.93E-03 2.80E-03 1.37E-03 1.10E-03 3.00E-03 Notes:

(1) Iron atom displacement values were calculated for energy levels above and below 0.1 MeV (i.e., all energy levels).

(2) The base plate and upper plates are also referred to as bottom plate and top plate, respectively.

(3) The diaphragm plates are positioned between the RPV support I-beams. Thus, the iron atom displacement of the I-beams is used in the diaphragm plate evaluation.

Table 5-3: H.B. Robinson Unit 2 Iron Displacement per Atom at the RPV Support Structure (with Analytical Uncertainties)

Cumulative Operating Time (EFPY)

Iron Atom Displacements (dpa)(1) +20% for uncertainties Base Plate(2)

I-Beams &

Diaphragm Plates(3)

Upper Plate(2)

& Leveling Bolts Support Shoe Anchor Bolt 42 1.96E-03 1.87E-03 9.16E-04 7.34E-04 2.08E-03 48 2.27E-03 2.17E-03 1.06E-03 8.51E-04 2.38E-03 54 2.58E-03 2.47E-03 1.21E-03 9.67E-04 2.69E-03 60 2.89E-03 2.77E-03 1.36E-03 1.08E-03 2.99E-03 66 3.20E-03 3.06E-03 1.50E-03 1.20E-03 3.30E-03 70 3.41E-03 3.26E-03 1.60E-03 1.28E-03 3.50E-03 72 3.52E-03 3.36E-03 1.64E-03 1.32E-03 3.60E-03 Notes:

(1) Iron atom displacement values were calculated for energy levels above and below 0.1 MeV (i.e., all energy levels).

(2) The base plate and upper plate are also referred to as bottom plate and top plate, respectively.

(3) The diaphragm plates are positioned between the RPV support I-beams. Thus, the iron atom displacement of the I-beams is used in the diaphragm plate evaluation.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-23 WCAP-18939-NP January 2025 Revision 1 Table 5-4: H.B. Robinson Unit 2 Iron Displacement Per Atom and NDTT (with Analytical Uncertainties)

Component(1)

Iron Atom Displacement at 70 EFPY Upper Bound Curve NDTT at 70 EFPY, F Support Shoe 1.28E-03 93 Top Plate 1.60E-03 108 Bottom Plate 3.41E-03 172 I-Beams &

Diaphragm Plates 3.26E-03 168 Anchor Bolts 3.50E-03 174 Note:

(1) The support shoe, top plate, bottom plate, I-beams, diaphragm plates, and anchor bolts consider the embrittlement shift fracture toughness methodology. The leveling bolts use a conservative minimum lower bound fracture toughness value that represents infinite embrittlement and does not require determination of iron atom displacement values or NDTT.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-24 WCAP-18939-NP January 2025 Revision 1 5.2 STRESS INTENSITY FACTORS AND POSTULATED FLAWS A range of stress intensity factor methodologies were considered in the analysis of linear elastic fracture mechanics to account for the support geometry (bolt and flat plate models) and flaw shapes (semi-elliptical flaws and through-wall single edge flaws). The prevalent crack-opening stress components are the stresses normal to the crack face; however, shear stresses are also present in the supports. [

]a,c,e A general description of the stress intensity factor methodologies are provided in Table 5-5 for each of the seven RPV support components and the following sections provide more detail for each stress intensity factor correlations.

The crack tip stress intensity factors are determined based on the stress intensity factor expressions from API-579 2016 Edition [18]. The API-579 stress intensity factor database includes industry accepted solutions and have been used frequently for previous fracture mechanics evaluations. [

]a,c,e A variety of flaw shapes are considered in the fracture mechanics analysis based on the API-579 stress intensity factor database. The flaw shapes which are appropriate for the various support geometry (bolt and flat plate models) are described in Table 5-5. [

]a,c,e Note the semi-elliptical flaws in the support plates and I-beams are subjected to welding residual stress as described in Section 4. The bar components are analyzed with a semi-circular front flaw shape as discussed in Section 5.2.2. Note that the 360° continuous circumferential flaw shape is included for informational purposes in this report, see Section 5.2.2 for a more detailed discussion.

5.2.1 Plate Model (Support Shoe, Plates, and I-beams)

The support shoe, top plates, bottom plates, diaphragm plates, and I-beams can all be modeled as a flat plate.

[

]a,c,e The stress provided in Section 4.1 through Section 4.5 for the support shoe, plates, and I-beams is used to calculate a stress intensity factor which is then compared to an applicable fracture toughness. The semi-elliptical flaws in the top plate, bottom plate, diaphragm plates, and I-beams included welding residual stress since the flaws are postulated near the welds of the components.

The stress intensity factor correlation for the semi-elliptical flaw with pure membrane stress is provided in API-579 [18] Section 9B.3.4 as follows:

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-25 WCAP-18939-NP January 2025 Revision 1

= M

Where:

KI = stress intensity factor (ksi-in0.5)

Mm = factor to account for flaw size, aspect ratio, and geometry m = membrane stress (ksi) a = flaw depth or size (inch)

Q = 1 + 1.464(a/c)1.65 for a/c < 1 c = half flaw length (inch)

= elliptic angle (0° for surface, 90° for deepest point)

W = distance from the center of the flaw to the free edge of the plate (inch) t = thickness (inch)

Figure 5-13: Postulated Surface Flaw in a Plate Model Similarly, the stress intensity factor correlation for the plate with hole single edge through-wall flaw with through-wall membrane and bending stress is provided in API-579 [18] Section 9B.4.1 as follows:

= M{}+ M(+ )

Where:

KI = stress intensity factor (ksi-in0.5)

Mm = factor to account for flaw size, aspect ratio, and geometry due to membrane stress m = membrane stress (ksi)

Mb = factor to account for flaw size, aspect ratio, and geometry due to bending stress b = bending stress (ksi) c = flaw depth or size (inch)

Rh = radius of the hole (inch)

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-26 WCAP-18939-NP January 2025 Revision 1 Figure 5-14: Postulated Single Edge Through-Wall Flaw in a Plate Model 5.2.2 Round Bar Model (for Bolts)

The leveling bolts and anchor bolts were analyzed using the semi-circular flaw crack model in API-579 [18].

A 360° continuous circumferential flaw is also postulated for informational purposes only. Per Section XI,

[

]a,c,e In addition, a 360° flaw in the leveling bolts or anchor bolts would have been replaced during initial fabrication; thus, the stress intensity factors for the continuous circumferential flaw is for information only.

A graphical representation of the two flaw types in a bar model is provided in Figure 5-15. The axial force, shear force, and moments on the leveling bolt and anchor bolt components are provided in Section 4.6 and Section 4.7. The stress intensity factor for the bolts is based on API-579 [18] Sections 9B.11.1 and 9B.11.3 for round bar, surface circumferential crack - 360° and semi-circular front crack, respectively with through-wall membrane and bending stresses as follows:

= (M+ M)

Where:

KI = stress intensity factor (ksi-in0.5)

Mm = influence factor to account for flaw size and geometry for membrane stress m = membrane stress (ksi)

Mb = influence factor to account for flaw size and geometry for bending stress b = bending stress (ksi) a = flaw depth or size (inch)

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-27 WCAP-18939-NP January 2025 Revision 1 (a) Semi-Circular Surface Crack Figure 5-15: Postulated Flaws in a Round Bar Model a

a Ro Ro

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-28 WCAP-18939-NP January 2025 Revision 1 Table 5-5: Mode I Stress Intensity Factor Model Description Component Model Shape Flaw Configuration Stress Intensity Factor Reference and Section No.

Figure of Postulated Flaw Shape Support Shoe, Top Plates, Bottom Plates, Diaphragm Plates, & I-Beams Plate (1)

Semi-Elliptical (AR = 6:1)

API-579 Section 9B.3.4

[18]

Figure 5-13 Diaphragm Plates Plate (1)

Through-Wall Single Edge Plate with a Hole API-579 Section 9B.4.1

[18]

Figure 5-14 Leveling Bolts Anchor Bolts Bar Semi-Circular Front API-579 Section 9B.11.3 [18]

Figure 5-15 360° Continuous Circumferential API -579 Section 9B.11.1 [18]

Note:

a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-1 WCAP-18939-NP January 2025 Revision 1 6

ALLOWABLE FLAW SIZES The LEFM analysis to evaluate the structural integrity of the H.B. Robinson Unit 2 RPV supports with 80 years (70 EFPY) of neutron embrittlement is completed using the acceptance criteria as described in Section 2 and illustrated in a flow chart format in Figure 2-1. The acceptance criteria is as follows: the stress intensity factors are calculated based on postulated allowable flaw sizes, plant-specific component stresses, and geometry; the stress intensity factors are then compared to an applicable fracture toughness value, where the goal of the LEFM analysis is to demonstrate that the calculated stress intensity factors remain below the applicable fracture toughness values. [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-2 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-3 WCAP-18939-NP January 2025 Revision 1 Table 6-1: Allowable Flaw Size or Lengths a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-1 WCAP-18939-NP January 2025 Revision 1 7

FRACTURE MECHANICS RESULTS As discussed in several previous sections of this report, the goal of the fracture mechanics analysis for H.B. Robinson Unit 2 RPV supports is to justify that plant life extension to 80 years does not cause a structural integrity concern based on radiation embrittlement. An assessment of the structural stability of the RPV supports for continued operation past 60 years can be demonstrated through a detailed linear elastic fracture mechanics. The LEFM methodology is illustrated in a flow chart format (see Figure 2-1) based on the guidance provided in NUREG-1509 [4] for a fracture mechanics approach to account for radiation effects on RPV support steels. The LEFM evaluation is completed by calculating stress intensity factor and comparing it to an applicable fracture toughness value for the H.B. Robinson RPV support components.

Per Section 4.2.4 of NUREG-1509, if the following criteria are met, the supports would demonstrate safe operation with consideration of neutron radiation embrittlement:

1. The initial nil-ductility transition temperature of the RPV supports is well below the minimum operating temperature.

2. The radiation exposure at the supports is low.

3. The peak tensile stresses are 6 ksi, or less.

However, as shown in Section 4, [

]a,c,e thus, a fracture mechanics evaluation is completed for the support shoes, top plates, bottom plates, diaphragm plates, I-beams, leveling bolts, and anchor bolts.

Section 7.1 through Section 7.7 of this report provides the limiting stress intensity factor values for the support shoe, top plate, bottom plate, diaphragm plates, I-beams, leveling bolts, and anchor bolts. The stress intensity factors are based on API-579 [18] methodology and equations, and use the Section XI allowable flaw sizes, plant-specific stresses, and component geometry. The limiting stress intensity factors are then compared to an applicable fracture toughness value.

The fracture toughness for the RPV support components is either based on the [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-2 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-3 WCAP-18939-NP January 2025 Revision 1

[

]a,c,e Even with the previously mentioned conservatism, the fracture mechanics analysis demonstrates that there is margin between the stress intensity factors and the fracture toughness for the H.B. Robinson Unit 2 RPV support structural steel components. The stress intensity factor results, as compared to fracture toughness, are provided in Section 7.1 through Section 7.7.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-4 WCAP-18939-NP January 2025 Revision 1 7.1 SUPPORT SHOE The support shoes for both support types 1 and 2 are located under the inlet nozzles, as shown in Figure 3-1. To avoid over conservatism in the support shoe analysis, material-specific fracture toughness is calculated considering the neutron embrittlement effects at 70 EFPY. The support shoe is subjected to a moderate amount of embrittlement resulting in a material-specific fracture toughness value of 45.5 ksiin at 70 EFPY (see Table 5-1).

Two cut orientations are analyzed for both support type 1 and 2 support shoes. One orientation is in the vertical direction, and the other is in the horizontal direction. Multiple cut paths along the vertical and horizontal cut orientations are evaluated, and the limiting stresses are provided in Table 4-1. Figure 4-1 shows the cut orientations for the support shoe. Stress intensity factors are then calculated using these plant-specific stresses and provided in Table 7-1. [

]a,c,e The stress intensity factors remain below the fracture toughness value; thus, the support shoe continues to be structurally stable considering 80 years of radiation embrittlement effects on the supports. The support shoe is considered to be flaw tolerant for 80 years and there is no concern of structural stability due to radiation embrittlement.

Table 7-1: Summary of Support Shoe Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness(1)

Vertical Horizontal Loading Combination Stress Intensity Factor (KI)

[ksiin]

Loading Combination Stress Intensity Factor (KI)

[ksiin]

1 45.5 ksiin at 70 EFPY Normal Outward 19.5 Normal Outward 35.3 Normal Inward 32.7 Normal Inward 38.2 Accident 11.0 Accident 9.0 2

45.5 ksiin at 70 EFPY Normal Outward 20.4 Normal Outward 36.7 Normal Inward 34.0 Normal Inward 39.9 Accident 11.5 Accident 9.4 Note:

(1) A fracture toughness of 45.5 ksiin at 70 EFPY can be used for the support shoe to demonstrate margin between the stress intensity factor and fracture toughness.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-5 WCAP-18939-NP January 2025 Revision 1 7.2 TOP PLATE The RPV top plates for both support type 1 and 2 includes [

]a,c,e The top plates and associated welds conservatively considered the ASME Code Case N-830 95% lower bound minimum fracture toughness value of 22.9 ksiin, with the exception of support type 2 top plate limiting cut path location. The 95% lower bound minimum fracture toughness value of 22.9 ksiin represents infinite embrittlement. However, the support type 2 top plate limiting cut path location required material-specific fracture toughness to be calculated considering the neutron embrittlement effects at 70 EFPY. The support type 2 top plate is subjected to a moderate amount of embrittlement resulting in a material-specific fracture toughness value of 38.0 ksiin at 70 EFPY (see Table 5-1).

In addition to the plant-specific stresses on the top plate (see Table 4-2), welding residual stress is considered per Section 4 for the postulated semi-elliptical flaw in the welded portion of the plate. Based on Section 4, [

]a,c,e The stress intensity factors are calculated and provided for the limiting top plates cut paths in Table 7-2.

[

]a,c,e The stress intensity factors remain below the fracture toughness; thus, the top plates continue to be structurally stable considering 80 years of radiation embrittlement effects on the supports. The top plate is considered to be flaw tolerant for 80 years and there is no concern of structural stability due to radiation embrittlement.

Table 7-2: Summary of Top Plate Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness Loading Combination Stress Intensity Factor (KI)

[ksiin]

1 22.9 ksiin (1) (2)

Normal Outward 19.7 Normal Inward 21.1 Accident 12.0 2

38.0 ksiin at 70 EFPY Normal Outward 23.0 Normal Inward 20.3 Accident 15.5 Notes:

(1) The minimum fracture toughness is based on the 95% lower bound Master Curve.

(2) A fracture toughness of 38.0 ksiin at 70 EFPY can be used for the support type 1 top plate in addition to the minimum lower bound fracture toughness to demonstrate additional margin between the stress intensity factor and fracture toughness.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-6 WCAP-18939-NP January 2025 Revision 1 7.3 BOTTOM PLATE The RPV bottom plate for both support type 1 and 2 includes [

]a,c,e The bottom plate and associated welds conservatively considered an ASME Code Case N-830 95% lower bound minimum fracture toughness value of 22.9 ksiin, which represents infinite embrittlement (see Table 5-1). Note that for information, material-specific fracture toughness is calculated considering the neutron embrittlement effects at 70 EFPY in Table 5-1 for the bottom plate.

The bottom plate material-specific fracture toughness in Table 5-1 can be used to demonstrate additional margin between the stress intensity factor and fracture toughness.

In addition to the plant-specific stresses on the bottom plate (see Table 4-3), welding residual stress is considered per Section 4 for the postulated semi-elliptical flaw in the welded portion of the plate. Based on Section 4, [

]a,c,e The stress intensity factors are calculated and provided for the limiting bottom plates cut paths in Table 7-3.

[

]a,c,e The stress intensity factors remain below the fracture toughness; thus, the bottom plates continue to be structurally stable considering 80 years of radiation embrittlement effects on the supports.

Table 7-3: Summary of Bottom Plate Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness Loading Combination Stress Intensity Factor (KI)

[ksiin]

1 22.9 ksiin(1)(2)

Normal Outward 12.1 Normal Inward 13.0 Accident 10.8 2

22.9 ksiin(1)(2)

Normal Outward 12.2 Normal Inward 15.7 Accident 11.6 Note:

(1) The minimum fracture toughness is based on the 95% lower bound Master Curve.

(2) A fracture toughness of 30.5 ksiin at 70 EFPY can be used for the bottom plate in addition to the lower bound fracture toughness to demonstrate additional margin between the stress intensity factor and fracture toughness.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-7 WCAP-18939-NP January 2025 Revision 1 7.4 DIAPHRAGM PLATES The diaphragm plates for both support type 1 and 2 includes [

]a,c,e The diaphragm plates and associated welds conservatively considered an ASME Code Case N-830 95% lower bound minimum fracture toughness value of 22.9 ksiin, with the exception of support type 2 limiting cut path location (cut path near a hole in the diaphragm plate). The 95% lower bound minimum fracture toughness value of 22.9 ksiin represents infinite embrittlement. However, the support type 2 diaphragm plate limiting cut path location required material-specific fracture toughness to be calculated considering the neutron embrittlement effects at 70 EFPY. The support type 2 diaphragm plate is subjected to a moderate amount of embrittlement resulting in a material-specific fracture toughness value of 30.8 ksiin at 70 EFPY (see Table 5-1).

In addition to the plant-specific stresses on the diaphragm plate (see Table 4-4), welding residual stress is considered per Section 4 for the postulated semi-elliptical flaw in the welded portion of the plate. Based on Section 4, [

]a,c,e The stress intensity factors are calculated and provided for the limiting diaphragm plates cut paths in Table 7-4 and Table 7-5. Note, that the locations for support type 2 were considered for a plate with a hole single edge through-wall flaw, as it is more limiting than that of support type 1. Thus, only support type 2 was included in Table 7-5. [

]a,c,e The stress intensity factors remain below the fracture toughness; thus, the diaphragm plates continue to be structurally stable considering 80 years of radiation embrittlement effects on the supports.

The diaphragm plates are considered to be flaw tolerant for 80 years and there is no concern of structural stability due to radiation embrittlement.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-8 WCAP-18939-NP January 2025 Revision 1 Table 7-4: Summary of Diaphragm Plate Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness Loading Combination Stress Intensity Factor (KI)

[ksiin]

1 22.9 ksiin(1) (2)

Normal Outward 11.2 Normal Inward 10.8 Accident 5.8 2

22.9 ksiin(1) (2)

Normal Outward 16.3 Normal Inward 18.5 Accident 11.4 Notes:

(1) The minimum fracture toughness is based on the 95% lower bound Master Curve.

(2) A fracture toughness of 30.8 ksiin at 70 EFPY and can be used for the diaphragm plate in addition to the lower bound fracture toughness to demonstrate additional margin between the stress intensity factor and fracture toughness.

Table 7-5: Summary of Diaphragm Plate with Hole Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness(1)

Loading Combination Stress Intensity Factor (KI)

[ksiin]

2 30.8 ksiin at 70 EFPY Normal Outward 23.2 Normal Inward 23.4 Accident 6.9 Note:

(1) A fracture toughness of 30.8 ksiin at 70 EFPY can be used for the diaphragm plate with hole to demonstrate margin between the stress intensity factor and fracture toughness.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-9 WCAP-18939-NP January 2025 Revision 1 7.5 I-BEAMS The I-beams for both support type 1 and 2 are [

]a,c,e The limiting cut paths for the I-beams required material-specific fracture toughness to be calculated considering the neutron embrittlement effects at 70 EFPY. The I-beams are subjected to a moderate amount of embrittlement resulting in a material-specific fracture toughness value of 30.8 ksiin at 70 EFPY (see Table 5-1).

In addition to the plant-specific stresses on the I-beams (see Table 4-6), welding residual stress is considered per Section 4 for the postulated semi-elliptical flaw in the welded portion of the I-beam. Based on Section 4,

[

]a,c,e The stress intensity factors are calculated and provided for the limiting I-beam cut paths in Table 7-6. [

]a,c,e The stress intensity factors remain less the fracture toughness; thus, the I-beams continue to be structurally stable considering 80 years of radiation embrittlement effects on the supports. The I-beams are considered to be flaw tolerant for 80 years and there is no concern of structural stability due to radiation embrittlement.

Table 7-6: Summary of I-Beam Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness(1)

Loading Combination Stress Intensity Factor (KI)

[ksiin]

1 30.8 ksiin at 70 EFPY Normal Outward 23.7 Normal Inward 21.6 Accident 11.2 2

30.8 ksiin at 70 EFPY Normal Outward 24.5 Normal Inward 23.6 Accident 10.2 Note:

(1) A fracture toughness of 30.8 ksiin at 70 EFPY can be used for the I-beams to demonstrate margin between the stress intensity factor and fracture toughness.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-10 WCAP-18939-NP January 2025 Revision 1 7.6 LEVELING BOLTS There are [

]a,c,e The leveling bolts conservatively considered an ASME Code Case N-830 95% lower bound minimum fracture toughness value of 22.9 ksiin (see Table 5-1), which represents infinite embrittlement.

The [

]a,c,e The calculated stress intensity factors remain below the fracture toughness; thus, the leveling bolts continue to be structurally stable considering 80 years of radiation embrittlement effects on the supports.

Table 7-7: Summary of Leveling Bolt Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness Loading Combination Stress Intensity Factor (KI)

[ksiin]

Semi-Circular 360 Circumferential(2) 1 22.9 ksiin(1)

Normal Outward 6.57 19.22 Normal Inward 6.88 20.12 Accident 2.66 7.73 2

22.9 ksiin(1)

Normal Outward 6.44 18.07 Normal Inward 6.66 18.68 Accident 2.57 7.17 Notes:

(1) The minimum fracture toughness is based on the 95% lower bound Master Curve.

(2) The Section XI allowable flaw size, based on surface examination of bolts, is limited to [

]a,c,e thus, a 360° fully circumferential flaw is inherently not allowed. However, the stress intensity factors are provided in this table for informational purposes only.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-11 WCAP-18939-NP January 2025 Revision 1 7.7 ANCHOR BOLTS There are [

]a,c,e The anchor bolts required a material-specific fracture toughness to be calculated considering the neutron embrittlement effects at 70 EFPY. The anchor bolts are subject to a moderate amount of embrittlement resulting in a material-specific fracture toughness value of 27.7 ksiin at 70 EFPY (see Table 5-1).

The anchor bolts [

]a,c,e The anchor bolts are considered to be flaw tolerant for 80 years and there is no concern of structural stability due to radiation embrittlement.

Table 7-8: Summary of Anchor Bolt Stress Intensity Factors versus Fracture Toughness Support Type Fracture Toughness(1)

Loading Combination Stress Intensity Factor (KI)

[ksiin]

Semi-Circular 360 Circumferential(2) 1 27.7 ksiin at 70 EFPY Normal Outward 11.59 23.24 Normal Inward 12.24 24.61 Accident 3.96 7.98 2

27.7 ksiin at 70 EFPY Normal Outward 6.52 12.24 Normal Inward 13.04 24.66 Accident 4.67 8.83 Notes:

(1) A fracture toughness of 27.7 ksiin at 70 EFPY can be used for the anchor bolts to demonstrate margin between the stress intensity factor and fracture toughness.

(2) The Section XI allowable flaw size, based on surface examination of bolts, is limited to

[

]a,c,e thus, a 360° flaw is inherently not allowed.

However, the stress intensity factors are provided in this table for informational purposes only.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-1 WCAP-18939-NP January 2025 Revision 1 8

DISCUSSION OF CONCLUSIONS 8.1 EVALUATION OF CURRENT CONDITIONS Per Section 4.3.1 of NUREG-1509 [4], physical examination of the structural components is essential to the evaluation completed herein and an assessment of the overall condition of the RPV support structure.

Based on the RPV support Ebasco specification [8], the structural steel components shall be in accordance with ASTM specification A-441 unless otherwise specified in the drawings, and shall be designed, detailed and fabricated in accordance with AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings. The weld details, qualifications, and procedures shall conform with AWS Specification D2.0 and shall be dye penetrant inspected. During initial fabrication, any unsatisfactory conditions were to be removed, re-welded, and re-examined. Thus, it is expected that the analyzed components are free from cracks after initial fabrication and after an extended period of time since crack growth mechanisms are not present at the RPV supports.

During refueling outage RO-32 (2020) [38], inactive boric acid deposit build ups on the RPV A, B, and C Nozzle supports have been identified due to borated water from historical reactor cavity seal leakage as well as historical leakage from the reactor cavity seal access ports (sand plugs) during previous outages.

Per [35], the leakage from the cavity was initially identified in RO-20 (2001); in response to the leakage, a corrosion assessment was performed based on chemistry of the cavity effluent and the containment environmental conditions. The assessment concluded that no significant material wastage would occur on the vessel supports as the support ventilation and surrounding air temperature would evaporate the cavity water during normal plant operation. Therefore, recommendations to inspect the supports during refueling outage RO-22 (2004) were made and completed as planned.

Subsequently, during refueling outage RO-26 (2010), significant cavity seal leakage was noted at the support on primary loop A, where standing water was present and of concern. Corrosion was also observed on the B and C support, but to a lesser degree than support A, due to the lower build up of boric acid crystals and better overall material condition of the steel. It was again concluded that while continued corrosion had persisted during refueling outages (when leakage can occur) that subsequently during online operations, active corrosion is arrested due to the RPV support environment. Furthermore, a review of historical cavity leakage revealed that cavity leakage continued during refueling outage RO-27 (2012), and visual examinations were performed during RO-27, where all locations were determined to be acceptable for continued operation. Thus, in an effort to eliminate seal leakage, a new cavity design was evaluated and implemented, where a permanent cavity seal plate was installed in refueling outage RO-28. Subsequently, refueling outages RO-29 (2015) and RO-31 (2018) went on to identify no cavity leakage caused by the cavity seal.

It should be noted that during refueling outage RO-30 (2017), leakage was identified at the C cold leg vessel support due to a missing O-ring at the C RCP cold leg sand plug cover. Leakage has not been identified in subsequent outages and as a result this area has been allowed to completely dry out. Support C is also judged to be in a better overall condition than support A.

In order to perform the VT-3 visual examination in RO-33 (2022) [36], the support component areas have been cleaned to the best ability possible with a steam cleaning process. The RO-33 VT-3 visual examination concluded that there is no visual evidence of structural deformation or degradation that impedes the current integrity of the components support to perform the intended function. There is no visual evidence of detached, broken loosened component support items, misalignment of existing support component items,

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-2 WCAP-18939-NP January 2025 Revision 1 and material wastage. It was also visually noted during the VT-3 visual examination that there was light/moderate boron identified throughout, and general surface corrosion exists which does not reduce the load bearing capacity of the support (no visual evidence of material wastage). The corrosion was calculated from the beginning of life through RO-33 in [35], and was determined to be 0.046 in. Note that the 0.046 in.

reduction is used in the structural integrity evaluation considered in [35]. Therefore, this reduction in thickness is also considered in the stress intensity factor calculation herein by decreasing the wall thickness by 0.046 in. for all seven RPV support components.

In conclusion per [36], the RPV supports are within the ASME Section XI ISI program and any further leaks and conditions that would affect the supports will be identified and periodically monitored. Thus, the H.B. Robinson Unit 2 RPV supports continues to maintain the current structural integrity and perform its intended function.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-3 WCAP-18939-NP January 2025 Revision 1 8.2 FRACTURE MECHANICS CONCLUSIONS Based on conclusions in NUREG-0933, it was determined that the RPV supports were not a concern for the entirety of its plant life (i.e., 40 or 60 years), even if all the supports were totally removed (i.e., broken),

the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for long term life extensions (i.e., beyond 60 years), the U.S. NRC has recently been requesting a re-assessment of the RPV structural steel supports based on a fracture mechanics evaluation to account for neutron embrittlement [2].

The goal of the LEFM analysis in this report for H.B. Robinson Unit 2 was to show structural stability and flaw tolerance of the RPV supports by demonstrating that the calculated stress intensity factors remain below the applicable fracture toughness values for the RPV support structural steel components. The stress intensity factors were calculated based on the [

]a,c,e There are seven components within the H.B. Robinson RPV support structure which were analyzed via the fracture mechanics approach described in NUREG-1509. These components include the support shoe, top plates, bottom plates, diaphragm plates, I-beams, leveling bolts, and anchor bolts. All seven of these locations could experience tensile stresses and high embrittlement effects. Thus, the stress intensity factors for all seven components were calculated and compared to an applicable fracture toughness. The LEFM analysis was based on the latest plant-specific stresses, PWHT welding residual stress, and latest stress intensity factors correlations used in the industry. In addition, the RPV support component thicknesses were reduced based on the corrosion evaluated in [35].

The fracture toughness for the RPV support components is either based on the minimum 95% lower bound Master Curve KJc fracture toughness of 22.9 ksiin or a material-specific fracture toughness when additional margin was needed. The minimum fracture toughness from the KJc Master Curve of 22.9 ksiin is used for the bottom plate, and leveling bolts. The use of the minimum lower bound fracture toughness value represents an infinite amount of neutron embrittlement on the RPV structural steel supports.

[

]a,c,e Section 7.1 through Section 7.7 contains the detailed results of the fracture mechanics evaluation for H.B. Robinson Unit 2 RPV supports (support type 1 and support type 2). The limiting fracture mechanics results and corresponding margins for each component are summarized in Table 8-1. As shown in

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-4 WCAP-18939-NP January 2025 Revision 1 Table 8-1, the stress intensity factors remain less than the fracture toughness; thereby concluding that the H.B. Robinson Unit 2 RPV supports continue to be structurally stable (i.e., flaw tolerant) considering 80 years of radiation embrittlement effects on the supports.

[

]a,c,e Based on the previous discussions and the results provided in this report, it is concluded that the RPV supports at H.B. Robinson Unit 2 are structurally stable (i.e., flaw tolerant) considering 80 calendar years (70 EFPY) of radiation embrittlement effects, and a sufficient level of flaw tolerance is demonstrated to justify continuing the current visual examination (VT-3) of the RPV structural steel supports. In conclusion, the loss of fracture toughness due to neutron embrittlement for 80 years is not significant and, therefore, the H.B. Robinson Unit 2 RPV structural steel supports are flaw tolerant for 80 years of operation.

Therefore, continued periodic visual inspections of the RPV supports per ASME Section XI, Subsection IWF Program [see pg. 3-399 of Reference 40] is required.

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-5 WCAP-18939-NP January 2025 Revision 1 Table 8-1: Summary of H.B. Robinson Limiting Stress Intensity Factors versus Fracture Toughness (ksiin)

Loading Condition Support Shoe

[ Fracture Toughness 45.5 ksiin(1) ]

Top Plate(2)

[ Fracture Toughness 38.0 ksiin(1) ]

Bottom Plate(2)

[ Fracture Toughness 22.9 ksiin(1) ]

Diaphragm Plate(2)

[ Fracture Toughness 30.8 ksiin(1) ]

I-Beam(2)

[ Fracture Toughness 30.8 ksiin(1) ]

Leveling Bolt(3)

[ Fracture Toughness 22.9 ksiin (1)]

Anchor Bolt(3)

[ Fracture Toughness 27.7 ksiin (1) ]

KI Margin(4)

KI Margin(4)

KI Margin(4)

KI Margin(4)

KI Margin(4)

KI Margin(4)

KI Margin(4)

Normal Outward 36.7 1.2 23.0 1.6 12.2 1.8 23.2 1.3 24.5 1.2 6.6 3.4 11.6 2.3 Normal Inward 39.9 1.1 21.1 1.8 15.7 1.4 23.4 1.3 23.6 1.3 6.9 3.3 13.0 2.1 Accident 11.5 3.9 15.5 2.4 11.6 1.9 11.4 2.7 11.2 2.7 2.7 8.4 4.7

5.8 Notes

(1) The stress intensity factors remain below the fracture toughness values. Thus the H.B. Robinson Unit 2 RPV supports continue to be structurally stable (i.e., flaw tolerant) considering 70 EFPY (80 years) of radiation or embrittlement effects on the supports.

(2) This location considers PWHT welding residual stress.

(3) The stress intensity factors are based on the semi-circular postulated flaw sizes.

(4) Margin = fracture toughness (KJc) / stress intensity factor (KI).

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-1 WCAP-18939-NP January 2025 Revision 1 9

REFERENCES

1. NUREG-2191, Volume 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report - Final Report, July 2017.

2. U.S. NRC Lessons Learned for SLR and Draft Report:

a. U.S. NRC Public Meeting: SLR Change - DRAFT New FE SRP-SLR Section 3.5.2.2.2.7 and AMR for Irradiation of RV Steel Supports, Meeting Held 02/20/20. [ADAMS Accession Number ML20049H359]

b. U.S. NRC Report, Summary of April 7, 2020 Teleconference on Topics Related to Lessons Learned from the Review of the First Subsequent License Renewal Applications, April 27, 2020.

[ADAMS Accession Number ML20107F699]

c. U.S. NRC Presentation during 2022 Materials Technical Exchange Public Meeting, Reactor Vessel Steel Supports for Subsequent License Renewal, May 25, 2022. [ADAMS Accession Number ML22143A838]

d. U.S. NRC Public Meeting - GALL-SLR, SLR Guidance Revision Focused Technical Sessions, September 7, 2022. [ADAMS Accession Number ML22243A268]

e. U.S. NRC, NUREG-2191, Volume 2, Draft Revision 1, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report Draft Report for Comment, July 2023.

[ADAMS Accession Number ML23180A188]

3. U.S. NRC, NUREG-0933, Resolution of Generic Safety Issues (Formerly entitled A Prioritization of Generic Safety Issues), Main Report with Supplements 1-34. Generic Issue No: 15, Radiation Effects on Reactor Vessel Supports (Rev. 3).

4. U.S. NRC, NUREG-1509, Radiation Effects on Reactor Pressure Vessel Supports, May 1996.

5. NUREG/CR-5320, Impact of Radiation Embrittlement on Integrity of Pressure Vessel Supports for Two PWR Plants, January 1989.

6. ASME Boiler & Pressure Vessel Code Section XI:

a. ASME Boiler & Pressure Vessel Code, 2007 Edition with 2008 Addenda Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

b. ASME Boiler & Pressure Vessel Code, 2019 Edition Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

7. Memorandums by the U.S. NRC Regarding Resolution of GSI-15 as cited in NUREG-0933. [ADAMS Accession Number: ML003749352]

a. Memorandum for James M. Taylor from David L. Morrison, Resolution of Generic Safety Issue 15, 'Radiation Effects on Reactor Vessel Supports', May 29, 1996. (3 pages).

b. Memorandum for John T. Larkins from Joseph A. Murphy, Proposed Resolution of GSI-15,

'Radiation Effects on Reactor Pressure Vessel Supports', June 22, 1994 (3 pages) (

Enclosures:

206 pages).

8. [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-2 WCAP-18939-NP January 2025 Revision 1

9. Westinghouse Electric Corporation Drawings:

a. [

]a,c,e

b. [

]a,c,e

10. Pittsburgh Bridge Iron Works Drawings:

a. 5370-01420, Revision 0, Reactor Supports, June 4, 1968.

b. 5370-01421, Revision 0, Reactor Supports, June 4, 1968.

c. 5370-01422, Revision 0, Reactor Supports, June 4, 1968.

11. Duke Design Input Request Responses:

a. WEC-DUKE-RNP-SLR-22-007, Design Input Request (DIR) RPV Supports Evaluations (WS07), August 2022.

b. WEC-DUKE-RNP-SLR-24-004, Design Input Request (DIR) RPV Supports Evaluations (WS07), January 2024.

12. ASTM A-508-64, Standard Specification for Quenched and Tempered Vacuum-Treated Carbon and Alloy Steel Forgings for Pressure Vessels.

13. ASTM A-242-81, Standard Specification for High-Strength Low-Alloy Structural Steel

14. ASTM A-441-70, Standard Specification for High-Strength Low-Alloy Structural Manganese Vanadium Steel.

15. ASTM A-434-76, Standard Specification for Steel Bars, Alloy, Hot-Rolled or Cold-Finished, Quenched and Tempered.

16. ASTM A-490-76, Standard Specification for Structural Bolts, Alloy Steel, Heat Treated, 150 Minimum Tensile Strength.

17. ASTM A-354-78, Standard Specification for Quenched and Tempered Alloy Steel Bolts, Studs, and other Externally Threaded Fasteners.

18. American Petroleum Institute, API-579-1/ASME FFS-1, Fitness-For-Service, June 2016.

19. American Welding Society AWS D2.0-66, Specifications for Welded Highway and Railway Bridges, 1966.

20. ASME Code Section III, Division 1 - Appendices, Rules for Construction of Nuclear Power Plant Components, 1989 Edition.

21. [

]a,c,e

22. [

]a,c,e

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-3 WCAP-18939-NP January 2025 Revision 1

23. [

]a,c,e

24. [

]a,c,e

25. ASME Code,Section II, Part D, Properties (Customary) Materials, 2007 Edition with 2008 Addenda.

26. ASME Code,Section III, Division 1-Appendicies, Rules for Construction of Nuclear Power Plant Components, 1971 Edition Winter Addenda.

27. [

]a,c,e

28. Toll Bridge Program Oversight Committee, Report on the A354 Grade BD High-Strength Steel Rods on the New East Span of the San Francisco-Oakland Bay Bridge, With Findings and Decisions, California Department of Transportation, 2013.

29. WCAP-14422, Revision 2-A, Licensing Renewal Evaluation: Aging Management for Reactor Coolant System Supports, December 2000.

30. ORNL/TM-2009/152, Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels, Oak Ridge National Laboratory, May 2011.

31. U.S. NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Office of Nuclear Regulatory Research, March 2001.

32. WCAP-18124-NP-A, Rev. 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018.

33. WCAP-18124-NP-A Revision 0 Supplement 1-NP-A, Rev. 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, June 2022.

34. Nuclear Generation Group RNP-M/HVAC-1076, Reactor Support Temperature Following Loss of HVH-6A and B.

35. Duke Energy EC0000418725, Reactor Vessel Support Evaluation (R2R32), December 1, 2020.

[Includes Attachment A, B, and C of EC0000418725]

36. Duke Energy VT-22-020, Visual Examination of Pipe Hanger, Support or Restraint (VT-3), RV Vessel Cold Leg Supports, November 2022.

37. American Institute of Steel Construction ANSI/AISC A303-22, Code of Standard Practice for Steel Buildings and Bridges, May 9, 2022.

38. Duke Energy Corporation, Inservice Inspection (ISI) Final Outage Report Robinson Unit 2 Fall 2020 Refueling Outage, November 2020.

39. U.S. NRC Docket No. 50-261, H.B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 279 Regarding Application of Leak-Before-Break Methodology for Auxiliary Reactor Coolant System Piping (EPID L-2023-LLA-0122), June 3, 2024. [ADAMS Accension Number ML24114A015]

40. U.S. NRC, NUREG-1785, Docket No. 50-261 Safety Evaluation Report Related to the License Renewal of H.B. Robinson Steam Electric Plant, Unit 2, March 2004. [ADAMS Accension Number ML040990702]

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-4 WCAP-18939-NP January 2025 Revision 1

41. U.S. NRC Safety Evaluation Report, Related to the Subsequent License Renewal of Point Beach Nuclear Plant, Units 1 and 2, Docket Nos. 50-266 and 50-301, NextEra Energy Point Beach, LLC, Issued February 2022. [ADAMS Accession Number ML22054A108]

42. U.S. NRC Safety Evaluation Report, Related to the Subsequent License Renewal of St. Lucie Plant, Units 1 and 2, Docket Nos. 50-335 and 50-389, Florida Power & Light Company, Revision 1, Issued September 2023. [ADAMS Accession Number ML23219A003]

• • • This record was final approved on 01/24/2025 13:49:27. (This statement was added by the PRIME system upon its validation)

WCAP-18939-NP Revision 1 Non-Proprietary Class 3

• • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Rizzilli Maria A Jan-22-2025 16:19:39 Author Approval Scott Alexandria M Jan-23-2025 08:13:35 Verifier Approval Coleman Joshua A Jan-23-2025 09:08:45 Verifier Approval Hall Gordon Z Jan-23-2025 09:30:16 Reviewer Approval Smith Sylena E Jan-23-2025 09:47:01 Reviewer Approval Udyawar Anees Jan-23-2025 10:21:07 Reviewer Approval Hall J Brian Jan-23-2025 13:30:52 Reviewer Approval Senior Ian T Jan-23-2025 14:18:45 Reviewer Approval Longwell Stephen K Jan-23-2025 16:51:28 Manager Approval Iddings Remington Jan-24-2025 13:49:27 Files approved on Jan-24-2025