ML25091A299
| ML25091A299 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 04/01/2025 |
| From: | Duke Energy Progress |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25091A290 | List: |
| References | |
| RA-25-0067 WCAP-15628-NP, Rev 1 | |
| Download: ML25091A299 (1) | |
Text
ENCLOSURE 4 ATTACHMENT 3 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NUMBER 2 Westinghouse WCAP-15628-NP, Revision 1, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the H.B. Robinson Unit 2 Nuclear Power Plant for the Subsequent License Renewal Program, May 2024
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15628-NP May 2024 Revision 1 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the H.B. Robinson Unit 2 Nuclear Power Plant for the Subsequent License Renewal Program
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA
© 2024 Westinghouse Electric Company LLC All Rights Reserved WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15628-NP Revision 1 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the H.B. Robinson Unit 2 Nuclear Power Plant for the Subsequent License Renewal Program May 2024 Author:
Alexandria M. Scott*
Reactor Vessel and Containment Vessel Design and Analysis Timothy J. Nowicki*
Reactor Vessel and Containment Vessel Design and Analysis Reviewer:
Momo Wiratmo*
Operating Plants Piping and Supports Approved: Remington Iddings*, Manager Reactor Vessel and Containment Vessel Design and Analysis
- Electronically approved records are authenticated in the electronic document management system.
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii WCAP-15628-NP May 2024 Revision 1 TABLE OF CONTENTS 1.0 Introduction................................................................................................................................... 1-1 1.1 Purpose............................................................................................................................ 1-1 1.2 Scope and Objective........................................................................................................ 1-1 1.3 Background Information.................................................................................................. 1-2 1.4 References........................................................................................................................ 1-4 2.0 Operation and Stability of the Reactor Coolant System............................................................... 2-1 2.1 Stress Corrosion Cracking............................................................................................... 2-1 2.2 Water Hammer................................................................................................................. 2-2 2.3 Low Cycle and High Cycle Fatigue................................................................................. 2-3 2.4 References........................................................................................................................ 2-3 3.0 Pipe Geometry and Loading......................................................................................................... 3-1 3.1 Introduction to Methodology........................................................................................... 3-1 3.2 Calculation of Loads and Stresses................................................................................... 3-2 3.3 Loads for Leak Rate Evaluation...................................................................................... 3-2 3.4 Load Combination for Crack Stability Analyses............................................................. 3-3 3.5 References........................................................................................................................ 3-3 4.0 Material Characterization.............................................................................................................. 4-1 4.1 Primary Loop Pipe and Fittings Materials....................................................................... 4-1 4.2 Tensile Properties............................................................................................................. 4-1 4.3 Fracture Toughness Properties......................................................................................... 4-2 4.4 Reference......................................................................................................................... 4-7 5.0 Critical Location and Evaluation Criteria..................................................................................... 5-1 5.1 Critical Locations............................................................................................................. 5-1 5.2 Fracture Criteria............................................................................................................... 5-1 6.0 Leak Rate Predictions................................................................................................................... 6-1 6.1 Introduction...................................................................................................................... 6-1 6.2 General Considerations.................................................................................................... 6-1 6.3 Calculation Method.......................................................................................................... 6-1 6.4 Leak Rate Calculations.................................................................................................... 6-2 6.5 References........................................................................................................................ 6-2 7.0 Fracture Mechanics Evaluation..................................................................................................... 7-1 7.1 Local Failure Mechanism................................................................................................ 7-1 7.2 Global Failure Mechanism............................................................................................... 7-1 7.3 Results of Crack Stability Evaluation.............................................................................. 7-2 7.4 RPV Nozzle Alloy 82/182 Welds..................................................................................... 7-4 7.5 References........................................................................................................................ 7-5 8.0 Fatigue Crack Growth Analysis.................................................................................................... 8-1 8.1 References........................................................................................................................ 8-5 9.0 Assessment of Margins................................................................................................................. 9-1 10.0 Conclusions................................................................................................................................. 10-1 APPENDIX A: Limit Moment................................................................................................................. A-1
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) iv WCAP-15628-NP May 2024 Revision 1 LIST OF TABLES Table 3-1: Dimensions, Normal Loads and Normal Stresses for H.B. Robinson Unit 2........................... 3-4 Table 3-2: Dimensions, Faulted Loads and Faulted Stresses for H.B. Robinson Unit 2........................... 3-5 Table 4-1: Measured Tensile Properties for H.B. Robinson Unit 2 Primary Loop Piping (ASTM A-376 TP316)..................................................................................................................................... 4-8 Table 4-2: Measured Tensile Properties for H.B. Robinson Unit 2 Primary Loop Fittings - Elbows (ASTM A-351 CF8M).......................................................................................................................... 4-9 Table 4-3: Mechanical Properties for H.B. Robinson Unit 2 Materials at Operating Temperatures........ 4-10 Table 4-4: Chemistry and Fracture Toughness Properties of the A-351 CF8M Material Heats of H.B.
Robinson Unit 2..................................................................................................................... 4-11 Table 4-5: Fracture Toughness Properties of A-351 CF8M for H.B. Robinson Unit 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations............................................................ 4-15 Table 6-1: Flaw Sizes Yielding a Leak Rate of 10 gpm at the Critical Locations..................................... 6-3 Table 7-1: H.B. Robinson Unit 2 Stability Results Based on Elastic-Plastic J-Integral Evaluations for A-351 CF8M Material........................................................................................................................ 7-6 Table 7-2: H.B. Robinson Unit 2 Stability Results Based on Limit Load for A-351 CF8M, A-376 TP316, and Alloy 82/182 Materials..................................................................................................... 7-6 Table 8-1 Summary of Reactor Vessel Transients for H.B. Robinson Unit 2............................................ 8-6 Table 8-2: Fatigue Crack Growth at Inlet Nozzle Safe-End Region.......................................................... 8-7 Table 8-3: Fatigue Crack Growth at Outlet Nozzle Safe-End Region....................................................... 8-7 Table 9-1: Leakage Flaw Sizes, Critical Flaw Sizes and Margins for H.B. Robinson Unit 2.................... 9-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) v WCAP-15628-NP May 2024 Revision 1 LIST OF FIGURES Figure 3-1: Hot Leg Coolant Pipe............................................................................................................ 3-6 Figure 3-2: Schematic Diagram of H.B. Robinson Unit 2 Primary Loop Showing Weld Locations....... 3-7 Figure 4-1: Pre-Service J vs. a for SA-351 CF8M Cast Stainless Steel at 600°F................................ 4-16 Figure 6-1: Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures.............................. 6-4 Figure 6-2: [
]a,c,e Pressure Ratio as a Function of L/D.............................................. 6-5 Figure 6-3: Idealized Pressure Drop Profile Through a Postulated Crack............................................... 6-6 Figure 7-1: [
]a,c,e Stress Distribution................................................................................... 7-7 Figure 8-1: Typical Cross-Section of RPV Inlet and Outlet Nozzle Safe-End......................................... 8-8 Figure 8-2: Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels........... 8-9 Figure 8-3: Reference Fatigue Crack Growth Law for Stainless Steel in Air Environments................. 8-10 Figure 8-4: Reference Fatigue Crack Growth Law for [
]a,c,e in a Water Environment at 600°F...................................................................................................................................... 8-11 Figure A-1: Pipe with a Through-Wall Crack in Bending........................................................................ A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) vi WCAP-15628-NP May 2024 Revision 1 EXECUTIVE
SUMMARY
The original structural design basis of the reactor coolant system for the H.B. Robinson Unit 2 Nuclear Power Plant required consideration of dynamic effects resulting from pipe break and that protective measures for such breaks be incorporated into the design. Subsequent to the original H.B Robinson Unit 2 design, an additional concern of Asymmetric Blowdown loads was raised as described in Unresolved Safety Issue A-2 (Asymmetric Blowdown Loads on the Reactor Coolant System). H.B. Robinson Unit 2 Nuclear Power Plant was part of the utilities which sponsored Westinghouse to resolve the A-2 issue. Generic analyses by Westinghouse to resolve the A-2 issue was approved by the NRC and documented in Generic Letter 84-04 (Reference 1-1).
The approved Westinghouse Generic Analyses were indicated to be directly applicable to H.B. Robinson Unit 2 in Reference 1-1.
Research by the NRC and industry coupled with operating experience determined that safety could be negatively impacted by placement of pipe whip restraints on certain systems. As a result, NRC and industry initiatives resulted in demonstrating that Leak-before-break (LBB) criteria can be applied to reactor coolant system piping based on fracture mechanics technology and material toughness.
Subsequently, the NRC modified 10CFR50 General Design Criterion 4, and published in the Federal Register (Vol. 52, No, 207) on October 27, 1987 its final rule, Modification of General Design Criterion 4 Requirements for Protections against Dynamic Effects of Postulated Pipe Ruptures (Reference 1-8). This change to the rule allows use of leak-before-break technology for excluding from the design basis the dynamic effects of postulated ruptures in primary loop piping in pressurized water reactors (PWRs).
WCAP-15628 Revision 0:
This report was issued to demonstrate compliance with LBB technology for the H.B. Robinson Unit 2 reactor coolant system based on plant specific analysis for the License Renewal Application.
Revision 1:
The purpose of Revision 1 of WCAP-15628 is to document the LBB evaluations for the H.B. Robinson Unit 2 primary reactor loop piping due to the subsequent license renewal (SLR) program for the plant operation extension into 80 years.
For the SLR program, this report demonstrates that the conclusions reached in WCAP-15628, Revision 0 remain applicable in the structural design basis for the 80-year plant life.
In addition, this report reviews the dissimilar metal (DM) weld locations at the reactor pressure vessel nozzles, which are susceptible to primary water stress corrosion cracking (PWSCC) effect to confirm that those locations have been appropriately evaluated for LBB. The reactor pressure vessel outlet nozzle (RPVON) and reactor pressure vessel inlet nozzle (RPVIN) Alloy 82/182 DM weld locations have not been mitigated from PWSCC effect. Thus, the updated LBB evaluations completed in this report includes conservatisms to account for the potential effects of PWSCC.
All critical locations are evaluated, including the unmitigated RPVON and RPVIN safe-end locations, to reconfirm that the LBB evaluation conclusions remain valid for 80-year plant life SLR program. As shown in the LBB evaluation, the presence of unmitigated Alloy 82/182 DM welds at the RPVON (Location 1) and RPVIN (Location 14) is also acceptable, since all the recommended LBB margins are satisfied.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 1-1 Introduction May 2024 WCAP-15628-NP Revision 1
1.0 INTRODUCTION
1.1 PURPOSE This report applies to the H.B. Robinson Unit 2 Reactor Coolant System (RCS) primary loop piping only and does not apply to any branch lines connected to the primary loop piping (e.g., surge, accumulator, RHR and safety injection branch lines). It is intended to demonstrate that for H.B. Robinson Unit 2, the RCS primary loop pipe breaks need not be considered in the structural design basis for the 80-year plant life subsequent license renewal (SLR) program. The approach taken has been accepted by the U.S. NRC per Generic Letter 84-04 (Reference 1-1).
1.2 SCOPE AND OBJECTIVE The purpose of this investigation is to demonstrate Leak-Before-Break (LBB) for the H.B. Robinson Unit 2 primary loops piping for 80-years of plant service. The recommendations and criteria proposed in SRP 3.6.3 (Reference 1-2 and Reference 1-3) are used in this evaluation. The criteria and resulting steps of the evaluation procedure can be briefly summarized as follows:
1.
Calculate the applied loads. Identify the locations at which the highest stress occurs.
2.
Identify the materials and the associated material properties.
3.
Postulate a through-wall flaw at the governing locations. The size of the flaw should be large enough so that the leakage is assured of detection with margin using the installed leak detection equipment when the pipe is subjected to normal operating loads. A margin of 10 is demonstrated between the calculated leak rate and the leak detection capability.
4.
Using maximum faulted loads, demonstrate that there is a margin of at least 2 between the leakage flaw size and the critical flaw size.
5.
Review the operating history to ascertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer, or low and high cycle fatigue.
6.
For the materials actually used in the plant, provide representative material properties including toughness and tensile test data. Evaluate long term effects such as thermal aging.
7.
Demonstrate margin on the calculated applied load value; margin of 1.4 using algebraic summation of faulted loads or margin of 1.0 using absolute summation of faulted loads.
8.
Perform an assessment of fatigue crack growth. Show that a through-wall crack will not result.
This report provides a fracture mechanics demonstration of primary loop integrity for the H.B. Robinson Unit 2 consistent with the NRC position for exemption from consideration of dynamic effects.
It should be noted that the terms flaw and crack have the same meaning and are used interchangeably.
Governing location and critical location are also used interchangeably throughout the report.
Note that there are several locations in this report where proprietary information has been identified and bracketed. For each of the bracketed locations, the reason for the proprietary classification is given using a WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 1-2 Introduction May 2024 WCAP-15628-NP Revision 1 standardized system. The proprietary brackets are labeled with three different letters, to provide this information, and the explanation for each letter is given below:
a.
The information reveals the distinguishing aspects of a process or component, structure, tool, method, etc., and the prevention of its use by Westinghouses competitors, without license from Westinghouse, gives Westinghouse a competitive economic advantage.
c.
The information, if used by a competitor, would reduce the competitors expenditure of resources or improve the competitors advantage in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
e.
The information reveals aspects of past, present, or future Westinghouse or customer-funded development plans and programs of potential commercial value to Westinghouse.
The proprietary information in the brackets has been deleted from the proprietary version of this report (WCAP-15628, Revision 1).
1.3 BACKGROUND
INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that RCS primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in the RCS primary loop was first presented to the NRC in 1978 in WCAP-9283 (Reference 1-4). That topical report employed a deterministic fracture mechanics evaluation and a probabilistic analysis to support the elimination of RCS primary loop pipe breaks. That approach was then used as a means of addressing Generic Issue A-2 and Asymmetric Loss of Coolant Accident (LOCA) loads.
Westinghouse performed additional testing and analysis to justify the elimination of RCS primary loop pipe breaks. This material was provided to the NRC along with Letter Report NS-EPR-2519 (Reference 1-5).
The NRC funded research through Lawrence Livermore National Laboratory (LLNL) to address this same issue using a probabilistic approach. As part of the LLNL research effort, Westinghouse performed extensive evaluations of specific plant loads, material properties, transients, and system geometries to demonstrate that the analysis and testing previously performed by Westinghouse and the research performed by LLNL applied to all Westinghouse plants (Reference 1-6 and Reference 1-7). The results from the LLNL study were released at a March 28, 1983, Advisory Committee on Reactor Safeguards (ACRS) Subcommittee meeting.
These studies which are applicable to all Westinghouse plants east of the Rocky Mountains determined the mean probability of a direct LOCA (RCS primary loop pipe break) to be 4.4 x 10-12 per reactor year and the mean probability of an indirect LOCA to be 10-7 per reactor year. Thus, the results previously obtained by Westinghouse (Reference 1-4) were confirmed by an independent NRC research study.
Based on the studies by Westinghouse, LLNL, the ACRS, and the Atomic Industrial Forum (AIF), the NRC completed a safety review of the Westinghouse reports submitted to address asymmetric blowdown loads that result from a number of discrete break locations on the pressurized water reactor (PWR) primary systems. The NRC Staff evaluation (Reference 1-1) concludes that an acceptable technical basis has been provided so that asymmetric blowdown loads need not be considered for those plants that can demonstrate the applicability of the modeling and conclusions contained in the Westinghouse response or can provide an equivalent fracture mechanics demonstration of the primary coolant loop integrity. In a more formal WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 1-3 Introduction May 2024 WCAP-15628-NP Revision 1 recognition of LBB methodology applicability for PWRs, the NRC appropriately modified 10 CFR 50, General Design Criterion 4, Requirements for Protection Against Dynamic Effects of Postulated Pipe Rupture (Reference 1-8).
This report provides a fracture mechanics demonstration of primary loop integrity for the H.B. Robinson Unit 2 consistent with the NRC position for exemption from consideration of dynamic effects. The re-evaluations were performed to ensure that the LBB evaluation conclusions remain valid for 80-year plant life in the SLR program.
Several computer codes are used in the evaluations. The LBB computer programs are under Configuration Control which has requirements conforming to Standard Review Plan 3.9.1. The computer codes used in this evaluation for leak rate and fracture mechanics calculations have been validated and used for all the LBB applications by Westinghouse.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 1-4 Introduction May 2024 WCAP-15628-NP Revision 1
1.4 REFERENCES
1-1 U.S. NRC Generic Letter 84-04, Subject Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops, February 1, 1984.
1-2 Standard Review Plan; public comments solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, February 28, 1987/Notices, pp. 32626-32633.
1-3 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
1-4 WCAP-9283, The Integrity of Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events, March 1978.
1-5 Letter Report NS-EPR-2519, Westinghouse (E. P. Rahe) to NRC (D. G. Eisenhut), Westinghouse Proprietary Class 2, November 10, 1981.
1-6 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston), April 25, 1983.
1-7 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston), July 25, 1983.
1-8 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 2-1 Operation and Stability of the Reactor Coolant System May 2024 WCAP-15628-NP Revision 1 2.0 OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM 2.1 STRESS CORROSION CRACKING The H.B. Robinson Unit 2 RCS primary loop piping contains Alloy 82/182 DM welds at the Reactor Pressure Vessel Outlet Nozzle (RPVON) and Reactor Pressure Vessel Inlet Nozzle (RPVIN) locations which are susceptible to PWSCC. A detailed evaluation of Alloy 82/182 welds is documented in Sections 6.0, 7.4, 8.0, and 9.0. While there are currently no plans to implement mitigation measures for the PWSCC concerns (via inlay, mechanical stress improvement process, peening, etc), H.B. Robinson Unit 2 have implemented the necessary Alloy 600 management plans with use of the inspection per ASME Code Case N-770-5. For the hot leg, this includes visual inspection each refueling outage and ultrasonic inspection every 5 years. For the cold leg, this includes visual inspection once per interval and ultrasonic inspection once per interval not to exceed 13 years. Additionally, a zinc injection program was implemented during fuel cycle 28 via the chemical volume control system. Zinc addition does provide some mitigation with respect to slowing the initiation and propagation of PWSCC. Finally ongoing work associated with the Electric Power Research Institute (EPRI) eXtremely Low Probability of Rupture (xLPR) program along with the U.S. NRC review of the program with respect to PWSCC concerns is still in process. However, Reference 2-3 concludes Successful application of the xLPR code in this study to demonstrate that the rupture probabilities of PWR piping systems that contain DMWs and were previously approved for LBB remain extremely low when subject to PWSCC reinforces the role of probabilistic fracture mechanics in making the demonstrations required by GDC 4 as originally envisioned by the Commission. Therefore, based on the conservative PWSCC considerations as described in Section 6.0, 7.4, 8.0 and 9.0, as well as the Alloy 600 management plan, Zinc addition, and current xLPR work in process, it is concluded that the PWSCC concerns have been addressed in this report and that LBB methodology and results for H.B.
Robinson Unit 2 are still valid at these locations.
For other stress corrosion cracking mechanisms, the Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design.
This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking (IGSCC)). This operating history totals over 1400 reactor-years, including 16 plants each having over 30 years of operation, 10 other plants each with over 25 years of operation, 11 plants each with over 20 years of operation, and 12 plants each with over 15 years of operation.
In 1978, the United States Nuclear Regulatory Commission (U.S. NRC) formed the second Pipe Crack Study Group. (The first Pipe Crack Study Group (PCSG), established in 1975, addressed cracking only in boiling water reactors). One of the objectives of the second PCSG was to include a review of the potential for stress corrosion cracking in Pressurized Water Reactors (PWRs). The results of the study performed by the PCSG were presented in NUREG-0531 (Reference 2-1) entitled Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants. In that report the PCSG stated:
The PCSG has determined that the potential for stress corrosion cracking in PWR primary system piping is extremely low because the ingredients that produce IGSCC are not all present. The use of hydrazine additives and a hydrogen overpressure limit the oxygen in the coolant to very low levels. Other impurities that might cause stress-corrosion cracking, such as halides or caustic, WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 2-2 Operation and Stability of the Reactor Coolant System May 2024 WCAP-15628-NP Revision 1 are also rigidly controlled. Only for brief periods during reactor shutdown when the coolant is exposed to the air and during the subsequent startup are conditions even marginally capable of producing stress-corrosion cracking in the primary systems of PWRs. Operating experience in PWRs supports this determination. To date, no stress corrosion cracking has been reported in the primary piping or safe ends of any PWR.
During 1979, several instances of cracking in PWR feedwater piping led to the establishment of the third PCSG. The investigations of the PCSG reported in NUREG-0691 (Reference 2-2) further confirmed that no occurrences of IGSCC have been reported for PWR primary coolant systems.
The discussion below further qualifies the PCSGs findings.
For stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. The potential for stress corrosion is minimized by properly selecting a material immune to SCC as well as preventing of a corrosive environment. The material specifications consider compatibility with the systems operating environment (both internal and external) as well as other material in the system, applicable American Society of Mechanical Engineers (ASME) Code rules, fracture toughness, welding, fabrication, and processing.
The elements of a water environment known to increase the susceptibility of austenitic stainless steel to stress corrosion are oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides, sulfites, and thionates). Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes and preoperational testing, water chemistry is controlled in accordance with written specifications.
Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping.
During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. For example, during normal power operation, oxygen concentration in the RCS is expected to be in the ppb range by controlling charging flow chemistry and maintaining hydrogen in the reactor coolant at specified concentrations. Halogen concentrations are also stringently controlled by maintaining concentrations of chlorides and fluorides within the specified limits.
Thus, during plant operation, the likelihood of stress corrosion cracking is minimized.
2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. The reactor coolant system, including piping and primary components, is designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. Other valve and pump actuations are relatively slow transients with no significant effect on the WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 2-3 Operation and Stability of the Reactor Coolant System May 2024 WCAP-15628-NP Revision 1 system dynamic loads. To ensure dynamic system stability, reactor coolant parameters are stringently controlled. Temperature during normal operation is maintained within a narrow range by control rod position; pressure is controlled by pressurizer heaters and pressurizer spray also within a narrow range for steady-state conditions. The flow characteristics of the system remain constant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics, are controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping are such that no significant water hammer can occur.
2.3 LOW CYCLE AND HIGH CYCLE FATIGUE An assessment of the low cycle fatigue loadings was carried out as part of this study in the form of fatigue crack growth analysis, as discussed in Section 8.0.
High cycle fatigue loads in the system would result primarily from pump vibrations. These are minimized by restrictions placed on shaft vibrations during hot functional testing and operation. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing, including plants similar to H.B. Robinson Unit 2. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for fatigue crack growth.
2.4 REFERENCES
2-1 Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants, NUREG-0531, U.S. Nuclear Regulatory Commission, February 1979.
2-2 Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.
2-3 U.S. NRC Technical Letter Report TLR-RES/DE/REB-2021-14, Probabilistic Leak-Before-Break Evaluations of Pressurized-Water Reactor Piping Systems using the Extremely Low Probability of Rupture Code, (ADAMS Accession Number ML21266A045).
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-1 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 3.0 PIPE GEOMETRY AND LOADING
3.1 INTRODUCTION
TO METHODOLOGY The general approach is discussed first. As an example, a segment of the primary coolant hot leg pipe is shown in Figure 3-1. The as-built outside diameter and minimum wall thickness of the pipe are 34.00 in and 2.40 in, respectively, as shown in the figure. The normal stresses at the weld locations are from the load combination procedure discussed in Section 3.3 whereas the faulted loads are developed as described in Section 3.4. The components for normal loads are pressure, deadweight, and thermal expansion. An additional component, safe shutdown earthquake (SSE), is considered for faulted loads. The enveloping loads for H.B. Robinson Unit 2 are shown in Table 3-1 and Table 3-2. As seen from Table 3-2, the highest faulted stress is at the reactor vessel outlet nozzle to pipe weld, Location 1. This highest stressed location is a load critical location and is one of the locations at which, as an enveloping location, Leak-Before-Break is to be established. Location 1 is also the critical location for the Alloy 82/182 welds. Essentially a circumferential flaw is postulated to exist at this location which is subjected to both the normal loads and faulted loads to assess leakage and stability, respectively. The loads (developed below) at this location are also given in Figure 3-1.
Since the elbows are made of different materials than the pipe, locations other than the highest stressed pipe location are examined taking into consideration both fracture toughness and stress. The elbows are cast stainless steel, and therefore thermal aging must be considered (see Section 4.0). Thermal aging of cast stainless steel (CASS) material results in lower fracture toughness; thus, locations other than the load critical locations must be examined taking into consideration both fracture toughness and stress. These enveloping locations are called toughness critical locations. The most critical locations are apparent only after the full analysis is completed. Once loads (this section) and fracture toughness values (see Section 4.0) are available, the load critical and toughness critical locations are determined (see Section 5.0). At these locations, leak rate evaluations (see Section 6.0) and fracture mechanics evaluations (see Section 7.0) are performed per the guidance of References 3-1 and 3-2. Fatigue crack growth (see Section 8.0) and stability margins are also evaluated (see Section 9.0). All the weld locations considered for the LBB evaluation are those shown in Figure 3-2.
Figure 3-2 summarizes the operating conditions considered in the analyses for the hot legs, cold legs, and crossover legs. The temperatures considered in the evaluation bound the actual normal operating conditions for H.B. Robinson Unit 2, which is 7.5°F lower for the hot legs and 7.7°F lower for the cold leg and crossover legs. The lower operating temperatures of H.B. Robinson Unit 2 is a negligible reduction to the normal operating thermal expansion loads (less than 2%), as well a negligible impact to the material properties, fracture toughness values and results in the analysis. Therefore, the operating conditions summarized in Figure 3-2 are determined to be acceptable for use in H.B. Robinson Unit 2.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-2 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 3.2 CALCULATION OF LOADS AND STRESSES The stresses due to axial loads and total moments are calculated by the following equation:
(3-1)
Where:
=
stress, ksi F
=
axial load, kips M
=
total moment, in-kips A
=
pipe cross-sectional area, in2 Z
=
section modulus, in3 The total moments for the desired loading combinations are calculated by the following equation:
M M M
M (3-2)
- Where, M
=
total moment for required loading MX =
X component of bending moment MY =
Y component of bending moment MZ
=
Z component of bending moment The axial load and total moment for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.3 and 3.4.
3.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:
F
=
FDW + FTH + FP (3-3)
MX
=
(MX)DW + (MX)TH (3-4)
MY
=
(MY)DW + (MY)TH (3-5)
MZ
=
(MZ)DW + (MZ)TH (3-6)
The subscripts of the above equations represent the following loading cases:
=
deadweight TH
=
normal thermal expansion.
P
=
load due to internal pressure This method of combining loads is often referred to as the algebraic sum method (References 3-1 and 3-2).
The loads based on this method of combination are provided in Table 3-1 at all the weld locations identified in Figure 3-2. The as-built dimensions are also given in Table 3-1.
Z M
A F
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-3 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 3.4 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the margin in terms of applied loads needs to be demonstrated by crack stability analysis. Margin on loads of 1.4 can be demonstrated if normal plus Safe Shutdown Earthquake (SSE) are applied algebraically and increased by 1.4. The 1.4 margin can be reduced to 1.0 if the deadweight, thermal expansion, internal pressure, Safe Shutdown Earthquake (SSE) inertia and seismic anchor motion (SAM) loads are combined based on individual absolute values as shown in the following equations:
F = FDW + FTH + FP + FSSEINERTIA + FSSEAM (3-7)
MX = (MX)DW + (MX)TH + (MX)SSEINERTIA + (MX)SSEAM (3-8)
MY = (MY)DW + (MY)TH + (MY)SSEINERTIA + (MY)SSEAM (3-9)
MZ = (MZ)DW + (MZ)TH + (MZ)SSEINERTIA + (MZ)SSEAM (3-10)
Where subscript SSEINERTIA refers to safe shutdown earthquake inertia and SSEAM is safe shutdown earthquake anchor motion.
The loads so determined are used in the fracture mechanics evaluations (Section 7.0) to demonstrate the LBB margins at the locations established to be the governing locations. These loads at all the weld locations (see Figure 3-2) are given in Table 3-2.
3.5 REFERENCES
3-1 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.
3-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-4 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 Table 3-1: Dimensions, Normal Loads and Normal Stresses for H.B. Robinson Unit 2 Location (a)
Outside Diameter (in)
Minimum Thickness (in)
Axial Load (b)
(kip)
Bending Moment (in-kip)
Total Stress (ksi) 1 34.00 2.40 1492 22386 18.99 2
34.00 2.40 1492 4327 8.72 3
34.00 2.40 1492 9062 11.41 4
37.75 3.275 1849 16204 10.97 5
37.63 3.21 1665 4563 6.45 6
36.31 2.56 1654 4273 8.10 7
36.31 2.56 1651 4351 8.12 8
36.31 2.56 1706 1084 6.80 9
36.31 2.56 1706 2779 7.59 10 37.63 3.21 1801 7465 7.90 11 32.25 2.28 1354 7526 11.33 12 32.25 2.28 1354 4968 9.63 13 32.25 2.28 1354 5841 10.21 14 33.56 2.93 1339 7260 8.39 Notes:
a See Figure 3-2 for piping layout and weld locations.
b Axial force includes pressure.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-5 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 Table 3-2: Dimensions, Faulted Loads and Faulted Stresses for H.B. Robinson Unit 2 Location (a)
Outside Diameter (in)
Minimum Thickness (in)
Axial Load (b)
(kip)
Bending Moment (in-kip)
Total Stress (ksi) 1 34.00 2.40 1639 22824 19.85 2
34.00 2.40 1638 5075 9.76 3
34.00 2.40 1638 9714 12.40 4
37.75 3.275 1956 18794 12.19 5
37.63 3.21 1882 14302 10.61 6
36.31 2.56 1840 9572 11.26 7
36.31 2.56 1837 6079 9.62 8
36.31 2.56 1841 6355 9.76 9
36.31 2.56 1839 7233 10.17 10 37.63 3.21 1846 15456 10.92 11 32.25 2.28 1405 13288 15.41 12 32.25 2.28 1405 10660 13.66 13 32.25 2.28 1406 9465 12.87 14 33.56 2.93 1401 11305 10.65 Notes:
a See Figure 3-2 for piping layout and weld locations.
b Axial force includes pressure.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-6 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 Location 1 ODa = 34.00 in ta = 2.40 in Normal Loadsa Faulted Loadsb Axial Forcec: 1492 kips Axial Forcec: 1639 kips Total Moment: 22386 in-kip Total Moment: 22824 in-kips Notes:
a See Table 3-1.
b See Table 3-2.
c Includes the force due to operating pressure of 2250 psia.
Figure 3-1: Hot Leg Coolant Pipe WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 3-7 Pipe Geometry and Loading May 2024 WCAP-15628-NP Revision 1 HOT LEG Analyzed Operating Temperature 612°F, Normal Operating Pressure 2250 psia CROSSOVER LEG Analyzed Operating Temperature 554°F, Normal Operating Pressure 2250 psia COLD LEG Analyzed Operating Temperature 554°F, Normal Operating Pressure 2250 psia Figure 3-2: Schematic Diagram of H.B. Robinson Unit 2 Primary Loop Showing Weld Locations Critical Locations Critical Location Critical Location WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-1 Material Characterization May 2024 WCAP-15628-NP Revision 1 4.0 MATERIAL CHARACTERIZATION 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS The primary loop pipe materials are A-376 TP316 and the elbow fittings are A-351 CF8M.
4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) for H.B. Robinson Unit 2 were used to establish the tensile properties for the Leak-Before-Break analyses. The CMTRs include tensile properties at room temperature and/or at 650°F for each of the heats of material. These properties for H.B. Robinson Unit 2 primary loop piping and fittings are given in Table 4-1 and Table 4-2, respectively. The average yield strength and the minimum yield and ultimate strengths are also identified.
Piping: For the A-376 TP316 material, the properties at operating temperatures were established from the tensile properties at room temperature and at 650°F given in Table 4-1 by utilizing Section III of the ASME Boiler and Pressure Vessel (B&PVC) Code 1989 Edition (Reference 4-1). Code tensile properties at the normal operating temperatures (612°F for Hot Leg and 554°F for Crossover leg and Cold leg) were obtained by a linear interpolation of the tensile properties provided in the Code. The ratios of the tensile properties at the applicable operating temperatures to the corresponding tensile properties at room temperature were then applied to the room temperature values given in Table 4-1 to obtain the plant specific properties for A-376 TP316 at normal operating temperatures. Similarly, the ratios of the tensile properties at the applicable operating temperatures to the corresponding tensile properties at 650°F were applied to the 650°F values given in Table 4-1 to obtain the plant specific properties at normal operating temperature. The average yield strength of 22.7 ksi at 650°F from Table 4-1 is used to calculate the average yield strength at operating temperature as shown in Table 4-3. The calculated value in Table 4-3 is used in the leakage analysis for the A-376 TP316 pipe material since it results in a larger leakage flaw size.
Fittings (Elbows): For the A-351 CF8M material, the properties at operating temperature were established from the tensile properties at room temperature given in Table 4-2. The representative tensile properties for A-351 CF8M at operating temperatures (612°F for Hot Leg and 554°F for Crossover leg and Cold leg) were obtained by utilizing Section III of the ASME B&PVC Code 1989 Edition (Reference 4-1). Code tensile properties at the applicable operating temperatures considered in this LBB analysis were obtained by a linear interpolation of the tensile properties provided in the Code. To obtain the plant specific properties for A-351 CF8M at operating temperatures of 612°F and 554°F as shown in Table 4-3, the Code minimum properties at the applicable operating temperatures were adjusted to account for the actual yield strength and ultimate tensile strength from the CMTR values at room temperature given in Table 4-2.
Weld Material: For Alloy 82/182 DM welds CMTR data was not available, and the typical tensile properties from Westinghouse source for the 82/182 weld material at the applicable operating temperatures as listed in Table 4-3 are used in the LBB evaluation.
The LBB evaluation considers the normal operating temperature of 612°F for Hot leg and 554°F for Crossover and Cold legs for material property interpolation.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-2 Material Characterization May 2024 WCAP-15628-NP Revision 1 The average and lower bound yield strengths and ultimate strengths at operating temperatures of 612°F and 554°F which are used in the LBB evaluation are summarized in Table 4-3. The ASME Code values for modulus of elasticity at the applicable operating temperatures are also provided. Poissons ratio was taken as 0.3.
4.3 FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast austenitic stainless steel (CASS) that are of interest are in terms of JIc (J at Crack Initiation) and have been found to be very high at 600F. [
]a,c,e However, cast stainless steel is susceptible to thermal aging at the reactor operating temperature, that is, about 550°F. Thermal aging of cast stainless steel results in embrittlement, which means a decrease in the ductility, impact strength, and fracture toughness of the material. Depending on the material composition, the Charpy impact energy of a cast stainless steel component could decrease to a small fraction of its original value after exposure to reactor temperatures during service.
The susceptibility of the material to thermal aging increases with increasing ferrite and molybdenum contents.
In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials (Reference 4-2). The ANL research program measured mechanical properties of cast stainless steel materials after they had been heated in controlled ovens for long periods of time. ANL compiled a data base, both from data within ANL and from international sources, of about 85 compositions of cast stainless steel exposed to a temperature range of 290°-400°C (550°-750°F) for up to 58,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (6.5 years). In 2015, the work done by ANL was augmented, and the fracture toughness database for CASS materials was aged to 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 290°-
350°C (554°-633°F). The methodology for estimating fracture properties has been extended to cover CASS materials with a ferrite content of up to 40%. From this database (NUREG/CR-4513, Revision 2),
ANL developed correlations for estimating the extent of thermal aging of cast stainless steel (Reference 4-4).
ANL developed the fracture toughness estimation procedures by correlating data in the database conservatively. After developing the correlations, ANL validated the estimation procedures by comparing the estimated fracture toughness with the measured value for several cast stainless steel plant components removed from actual plant service. The procedure developed by ANL in Revision 1 and Revision 2 of NUREG/CR-4513 (References 4-3 and 4-4) was used to calculate the end of life limiting fracture toughness values of the CASS elbows for the cold leg, crossover leg and hot leg locations. Note that LBB analyses have acceptable margins when performing the elastic-plastic J-integral evaluations with the use of lower bound fracture toughness properties from NUREG/CR-4513, Revision 1 and Revision 2. Furthermore, this report used saturated toughness approved by NRC in NUREG/CR-4513 Revision 1 and Revision 2.
Therefore, the LBB analysis is acceptable from a saturated toughness perspective.
The method described below was used to calculate the toughness properties for the cast material, A-351 CF8M, of the H.B. Robinson Unit 2 primary coolant loop elbows.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-3 Material Characterization May 2024 WCAP-15628-NP Revision 1 The JIc, Jmax and Tmat values for each material heat for CASS elbows was calculated using both Revision 1 and Revision 2 of NUREG/CR-4513 (References 4-3 and 4-4) as summarized in Table 4-4. The most limiting JIc and Jmax values for the Hot leg, Crossover leg, and Cold leg reported in Table 4-5 are based on Revision 2 of NUREG/CR-4513.
Based on Reference 4-3 and 4-4, the lower bound fully aged fracture toughness correlations are used for the A-351 CF8M material.
The chemical compositions of the H.B. Robinson Unit 2 primary loop elbow fitting material (A-351 CF8M) are available from CMTRs and are provided in Table 4-4. The following equations are taken from References 4-3 and 4-4:
Creq = Cr + 1.21(Mo) + 0.48(Si) - 4.99 = (Chromium equivalent)
(4-1)
Nieq = (Ni) + 0.11(Mn) - 0.0086(Mn)2 + 18.4(N) + 24.5(C) + 2.77 = (Nickel equivalent)
(4-2) c =100.3(Creq / Nieq )2 - 170.72(Creq / Nieq ) + 74.22 = (Ferrite Content)
(4-3) where the elements are in percent weight and c is ferrite in percent volume.
The saturation room temperature (RT) impact energies of the cast stainless steel materials were determined from the chemical compositions available from CMTRs and provided in Table 4-4.
The following equations are taken from NUREG/CR-4513 Revision 1 (Reference 4-3):
For CF8M steel with <10% Ni, the saturation value of RT impact energy Cvsat (J/cm2) is the lower value determined from:
log10Cvsat = 1.10 + 2.12 exp (-0.041)
(4-4.a)
Where the material parameter is expressed as:
= c (Ni + Si +Mn)2(C + 0.4N)/5.0 (4-5.a) and from:
log10Cvsat = 7.28 - 0.011c - 0.185Cr - 0.369Mo - 0.451Si- 0.007Ni - 4.71(C + 0.4N)
(4-6.a)
For CF8M steel with 10% Ni, the saturation value of RT impact energy Cvsat (J/cm2) is the lower value determined from:
log10Cvsat = 1.10 + 2.64 exp (-0.064)
(4-7.a)
Where the material parameter is expressed as:
= c (Ni + Si +Mn)2(C + 0.4N)/5.0 (4-8.a)
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-4 Material Characterization May 2024 WCAP-15628-NP Revision 1 and from:
log10Cvsat = 7.28 - 0.011c - 0.185Cr - 0.369Mo - 0.451Si - 0.007Ni - 4.71(C + 0.4N)
(4-9.a)
The saturation J-R curve at RT, for static-cast CF8M steel is given by:
Jd = 16 (Cvsat)0.67(a)n (4-10.a) n = 0.23 + 0.08 log10 (Cvsat)
(4-11.a)
Where Jd is the deformation J in kJ/m2 and a is the crack extension in mm.
The saturation J-R curve at 290C (554F), for static-cast CF8M steel is given by:
Jd = 49 (Cvsat)0.41(a)n (4-12.a) n = 0.23 + 0.06 log10 (Cvsat)
(4-13.a)
Where Jd is the deformation J in kJ/m2 and a is the crack extension in mm.
The following equations are taken from NUREG/CR-4513 Revision 2 (Reference 4-4):
For CF8M steel with <10% Ni, the saturation value of RT impact energy Cvsat (J/cm2) is the lower value determined from:
log10Cvsat = 0.27 + 2.81 exp (-0.022)
(4-4.b)
Where the material parameter is expressed as:
= c (Ni + Si + Mn)2(C + 0.4N)/5.0 (4-5.b) and from:
log10Cvsat = 7.28 - 0.011c - 0.185Cr - 0.369Mo - 0.451S- 0.007Ni - 4.71(C + 0.4N)
(4-6.b)
For CF8M steel with 10% Ni, the saturation value of RT impact energy Cvsat (J/cm2) is the lower value determined from:
log10Cvsat = 0.84 + 2.54 exp (-0.047)
(4-7.b)
Where the material parameter is expressed as:
= c (Ni + Si + Mn)2(C + 0.4N)/5.0 (4-8.b) and from:
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-5 Material Characterization May 2024 WCAP-15628-NP Revision 1 log10Cvsat = 7.28 - 0.011c - 0.185Cr - 0.369Mo - 0.451Si - 0.007Ni - 4.71(C + 0.4N)
(4-9.b)
The saturation J-R curve at RT, for static-cast CF8M steel is given by:
Jd = 1.44 (Cvsat)1.35(a)n for Cvsat < 35 J/cm2 (4-10.1.b)
Jd = 16 (Cvsat)0.67(a)n for Cvsat 35 J/cm2 (4-10.2.b) n = 0.20 + 0.08 log10 (Cvsat)
(4-11.b)
Where Jd is the deformation J in kJ/m2 and a is the crack extension in mm.
The saturation J-R curve at 290-320C (554-608F), for static-cast CF8M steel is given by:
Jd = 5.5 (Cvsat)0.98(a)n for Cvsat < 46 J/cm2 (4-12.1.b)
Jd = 49 (Cvsat)0.41(a)n for Cvsat 46 J/cm2 (4-12.2.b) n = 0.19 + 0.07 log10 (Cvsat)
(4-13.b)
Where Jd is the deformation J in kJ/m2 and a is the crack extension in mm.
[
]a,c,e The critical heats with the most limiting allowable fracture values (lowest fracture toughness properties values and lowest tearing modulus) for H.B. Robinson Unit 2 primary loop elbows from Table 4-4 are selected as shown below and provided in Table 4-5:
[
]a,c,e Toughness properties from these material heats are conservatively used to evaluate the critical locations on the Hot leg, Crossover leg, and Cold leg.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-6 Material Characterization May 2024 WCAP-15628-NP Revision 1 JIc and Jmax Calculations:
[
]a,c,e Tmat Calculations:
The material tearing modulus, Tmat, is calculated as follows:
Tmat = dJ/da x E/(fa)2 The results from the ANL Research Program indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). In addition, historic testing done on representative plants documented in References 4-5 and 4-6, has shown that the wrought and cast stainless steel piping exhibits more limiting (unaged) fracture toughness properties than the weld metal.
Since the CASS materials aged lower bound fracture toughness values are similar to that of Submerged Arc Welds (SAWs), and since SAWs are considered to be the most limiting of welding processes (with respect to GTAW and SMAW), it is concluded that the aged fracture toughness of the wrought and cast base metal is more limiting than the aged fracture toughness of the stainless-steel weld metal. Therefore, the stainless-steel weld regions are less limiting than the cast material, and the applied value of the J-integral for a flaw in the weld regions will be lower than that in the base metal because the yield stress for the stainless steel weld materials is much higher at operating temperature1.
Forged stainless steel piping such as A-376 TP316 does not degrade due to thermal aging. Thus, fracture toughness values well in excess of that established for the cast material exist for this material throughout service life and are not limiting.
Alloy 82/182 weld materials have high toughness values and do not degrade due to thermal aging. As discussed in Reference 4-7, the fracture resistance of Ni Alloys (Alloys 82) and their welds have been investigated by conducting fracture toughness J-R curve tests at 24-338°C in deionized water. The results indicated that Alloy 690 welds exhibit excellent fracture toughness in air and high-temperature water
(>93°C).
Since nickel alloys are known to have high toughness properties and because CF8M CASS base metal is susceptible to thermal aging degradation of the fracture toughness, it is determined that the CF8M CASS base metal presents the most limiting condition. Therefore, in the fracture mechanics analyses that follow, the thermally aged fracture toughness allowables of the CASS material given in Table 4-5 will be used as the criteria against which the calculated applied fracture toughness values will be compared.
1 In the report, all the applied J values were conservatively determined by using base metal strength properties.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-7 Material Characterization May 2024 WCAP-15628-NP Revision 1 4.4 REFERENCE 4-1 ASME Boiler and Pressure Vessel Code Section III, Rules for Construction of Nuclear Power Plant Components, 1989 Edition.
4-2 O. K. Chopra and W. J. Shack, Assessment of Thermal Embrittlement of Cast Stainless Steels, NUREG/CR-6177, U.S. Nuclear Regulatory Commission, Washington, DC, May 1994.
4-3 O. K. Chopra, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, NUREG/CR-4513, Revision 1, U.S. Nuclear Regulatory Commission, Washington, DC, August 1994.
4-4 O. K. Chopra, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, NUREG/CR-4513, Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, May 2016 including Errata, March 15, 2021.
4-5 Westinghouse Report, WCAP-9787, Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, May 1981.
4-6 Westinghouse Report, WCAP-9558, Revision 2, Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack, May 1981.
4-7 NUREG/CR-6721, Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds, April 2001.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-8 Material Characterization May 2024 WCAP-15628-NP Revision 1 Table 4-1: Measured Tensile Properties for H.B. Robinson Unit 2 Primary Loop Piping (ASTM A-376 TP316)
Component Heat Number Yield Strength (psi) at Room Temperature Ultimate Strength (psi) at Room Temperature Yield Strength (psi) at 650°F Ultimate Strength (psi) at 650°F Hot Leg F0190 42000 88800 21300 58200 43000 86000 Hot Leg V0126 40500 83000 23400 65200 46100 90200 Hot Leg D8774 36000 79200 24000 67400 37000 79700 Hot Leg 52152 37100 77400 20800 61400 36500 78600 Hot Leg F0214 42500 82300 22400 62300 44500 77300 Crossover Leg D8777 36100 78200 20800 63700 38500 77800 Crossover Leg D8915 38500 77400 24200 62300 38600 77200 Crossover Leg D8785 36100 74200 20400 57200 39700 79800 Crossover Leg F0189 37700 80600 25200 70000 44100 91000 Crossover Leg D8775 36100 77800 20500 64100 39300 79000 Crossover Leg D8915 37700 79600 22800 62400 39300 80200 Cold Leg F0216 40900 83000 21300 66600 42500 83500 Cold Leg 52263 34200 75100 23100 63700 37800 75200 Cold Leg D8768 36000 83000 23800 71700 39300 82700 Cold Leg 52152 44400 89700 21600 58800 34899 75200 Cold Leg V0342 35100 75200 25600 52500 36100 75100 Cold Leg D8913 35100 78400 24300 68500 41100 84800 Minimum 34200 74200 20400 52500 Average 38950 22676 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-9 Material Characterization May 2024 WCAP-15628-NP Revision 1 Table 4-2: Measured Tensile Properties for H.B. Robinson Unit 2 Primary Loop Fittings - Elbows (ASTM A-351 CF8M)
Component Heat Number Yield Strength (psi) at Room Temperature Ultimate Strength (psi) at Room Temperature Hot Leg 04204 43500 87500 Hot Leg 07896 43500 87500 Hot Leg 08066 49500 88500 Crossover Leg 03327-7 51000 90000 Crossover Leg 06079-1 54000 93800 Crossover Leg 06185-4 43500 87500 Crossover Leg 09390A 45000 88500 Crossover Leg 09517 48000 89000 Crossover Leg 09436 45000 88500 Crossover Leg 09476 43500 85500 Crossover Leg 09964 48000 90000 Crossover Leg 10165 55500 96500 Crossover Leg 09305A 51000 91500 Crossover Leg 09640 48000 88000 Crossover Leg 09720 45750 88000 Crossover Leg 09760 45500 89500 Crossover Leg 09841 45000 88500 Crossover Leg 09882 45000 85500 Cold Leg 04589 45000 89000 Cold Leg 05065 45000 87500 Cold Leg 05529 54000 96000 Minimum 43500 85500 Average 47345 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-10 Material Characterization May 2024 WCAP-15628-NP Revision 1 Table 4-3: Mechanical Properties for H.B. Robinson Unit 2 Materials at Operating Temperatures a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-11 Material Characterization May 2024 WCAP-15628-NP Revision 1 Table 4-4: Chemistry and Fracture Toughness Properties of the A-351 CF8M Material Heats of H.B. Robinson Unit 2 a,c,e
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-12 Material Characterization May 2024 WCAP-15628-NP Revision 1 a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-13 Material Characterization May 2024 WCAP-15628-NP Revision 1 a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-14 Material Characterization May 2024 WCAP-15628-NP Revision 1 a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-15 Material Characterization May 2024 WCAP-15628-NP Revision 1 Table 4-5: Fracture Toughness Properties of A-351 CF8M for H.B. Robinson Unit 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 4-16 Material Characterization May 2024 WCAP-15628-NP Revision 1 Figure 4-1: Pre-Service J vs. a for SA-351 CF8M Cast Stainless Steel at 600°F Note: This plot is shown for demonstration purposes.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 5-1 Critical Location and Evaluation Criteria May 2024 WCAP-15628-NP Revision 1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA 5.1 CRITICAL LOCATIONS The Leak-Before-Break (LBB) evaluation margins are to be demonstrated for the limiting locations (governing locations). Candidate locations are designated load critical locations or toughness critical locations as discussed in Section 3.0. Such locations are established considering the loads (Section 3.0) and the material properties established in Section 4.0. These locations are defined below for H.B. Robinson Unit 2 as well as shown in Figure 3-2.
Load Critical Locations The highest stressed location for the entire primary loop is at A-376 TP316 straight pipe Location 1 (in the Hot leg) at the reactor vessel outlet nozzle to pipe weld. Location 1 is the critical weld location for the pipe.
Location 1 is also the critical location for the stainless steel and Alloy 82/182 welds.
Since the elbows are made of cast materials, the critical weld locations for the elbows are as follows. For the Hot leg (HL), the highest stressed location is at weld Location 3. For the Crossover leg (XL), the highest stressed location is at weld Location 6. For the Cold leg (CL), the highest stressed location is at weld Location 13.
Thus, it is concluded that the enveloping locations in H.B. Robinson Unit 2 for which LBB methodology is to be applied are Locations 1, 3, 6, and 13.
Toughness Critical Location Low toughness locations are at the end of each elbow since the elbows are made of cast materials and can be susceptible to thermal aging. Per Section 4.3, the critical material location for the elbows is [
]a,c,e, due to low toughness. As identified above the highest faulted stresses at the elbow locations are 3, 6 and 13. The limiting toughness values determined in Section 4.3 is conservatively used in evaluating of these locations.
For the critical locations, the tensile properties are shown in Table 4-3, and the allowable fracture toughness properties are shown in Table 4-5.
5.2 FRACTURE CRITERIA As will be discussed later, fracture mechanics analyses are made based on loads and postulated flaw sizes related to leakage. The stability criteria against which the calculated J (i.e., Japp) and tearing modulus (Tapp) are compared are:
(1)
If Japp < JIc, then an existing crack is stable (or a crack will not initiate);
(2)
If Japp > JIc; and Tapp < Tmat and Japp < Jmax, then the crack is stable.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 5-2 Critical Location and Evaluation Criteria May 2024 WCAP-15628-NP Revision 1 Where:
Japp
=
Applied J JIc
=
J at Crack Initiation Tapp =
Applied Tearing Modulus Tmat =
Material Tearing Modulus Jmax =
Maximum J value of the material For critical locations, the limit load method discussed in Section 7.0 was also used.
For global failure mechanism, the stability analysis is performed using limit load method based on loads and postulated flaw sizes related to leakage, with the criteria as follows:
Margin of 10 on the Leak Rate
Margin of 2.0 on Flaw Size
Margin of 1.0 on Loads (using the absolute summation method for faulted load combination).
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 6-1 Leak Rate Predictions May 2024 WCAP-15628-NP Revision 1 6.0 LEAK RATE PREDICTIONS
6.1 INTRODUCTION
The purpose of this section is to discuss the method which is used to predict the flow through postulated through-wall cracks and present the leak rate calculation results for through-wall circumferential cracks.
6.2 GENERAL CONSIDERATIONS The flow of hot pressurized water through an opening to a lower back pressure causes flashing which can result in choking. For long channels where the ratio of the channel length, L, to hydraulic diameter, DH, (L/DH) is greater than [
]a,c,e 6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [
]a,c,e The flow rate through a crack was calculated in the following manner. Figure 6-1 (from Reference 6-2) was used to estimate the critical pressure, Pc, for the primary loop enthalpy condition and an assumed flow.
Once Pc was found for a given mass flow, the [
]a,c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [
]a,c,e Therefore, this method will yield the two-phase pressure drop due to momentum effects as illustrated in Figure 6-3, where Po is the operating pressure. Now using the assumed flow rate, G, the frictional pressure drop can be calculated using:
Pf (6-1) where the friction factor f is determined using the [
]a,c,e The crack relative roughness,,
was obtained from fatigue crack data on stainless steel samples. The relative roughness value used in these calculations was [
]a,c,e The frictional pressure drop using Equation 6-1 is then calculated for the assumed flow rate and added to the [
]a,c,e to obtain the total pressure drop from the primary system to the atmosphere.
a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 6-2 Leak Rate Predictions May 2024 WCAP-15628-NP Revision 1 That is, for the primary loop:
Absolute Pressure - 14.7 = [
]a,c,e (6-2) for a given assumed flow rate G. If the right-hand side of Equation 6-2 does not agree with the pressure difference between the primary loop and the atmosphere, then the procedure is repeated until Equation 6-2 is satisfied to within an acceptable tolerance which in turn leads to flow rate value for a given crack size.
6.4 LEAK RATE CALCULATIONS Leak rate calculations were made as a function of crack length at the governing locations previously identified in Section 5.1. The normal operating loads of Table 3-1 were applied in these calculations. The crack opening areas were estimated using the method of Reference 6-3, and the leak rates were calculated using the two-phase flow formulation described above. The average material properties of Section 4.0 (see Table 4-3) were used for these calculations.
The flaw sizes to yield a leak rate of 10 gpm for H.B. Robinson Unit 2 were calculated at the governing locations with pipe material A-376 TP316, elbow material A-351 CF8M, and weld material Alloy 82/182 are given in Table 6-1. The flaw sizes, so determined, are called leakage flaw sizes. Based on the PWSCC crack morphology, an increase factor of 1.69 between the PWSCC and fatigue crack morphologies (Reference 6-4) is applied to the leakage flaw sizes for the Alloy 82/182 DM welds as shown in Table 6-1.
The H.B. Robinson Unit 2 RCS pressure boundary leak detection system meets the intent of Regulatory Guide 1.45 (Reference 6-5), and the plant leak detection capability is 1 gpm. Thus, to satisfy the margin of 10 on the leak rate, the flaw sizes (leakage flaw sizes) are determined which yield a leak rate of 10 gpm.
6.5 REFERENCES
6-1
[
]a,c,e 6-2 M. M, El-Wakil, Nuclear Heat Transport, International Textbook Company, New York, N.Y.,
1971.
6-3 Tada, H., The Effects of Shell Corrections on Stress Intensity Factors and the Crack Opening Area of Circumferential and a Longitudinal Through-Crack in a Pipe, Section II-1, NUREG/CR-3464, September 1983.
6-4 D. Rudland, R. Wolterman, G. Wilkowski, R. Tregoning, Impact of PWSCC and Current Leak Detection on Leak-Before-Break, proceedings of Conference on Vessel Head Penetration, Inspection, Cracking, and Repairs, Sponsored by USNRC, Marriot Washingtonian Center, Gaithersburg, MD, September 29 to October 2, 2003.
6-5 Regulator Guide 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, 2008.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 6-3 Leak Rate Predictions May 2024 WCAP-15628-NP Revision 1 Table 6-1: Flaw Sizes Yielding a Leak Rate of 10 gpm at the Critical Locations a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 6-4 Leak Rate Predictions May 2024 WCAP-15628-NP Revision 1 Figure 6-1: Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 6-5 Leak Rate Predictions May 2024 WCAP-15628-NP Revision 1 a,c,e Figure 6-2: [
]a,c,e Pressure Ratio as a Function of L/D WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 6-6 Leak Rate Predictions May 2024 WCAP-15628-NP Revision 1 Figure 6-3: Idealized Pressure Drop Profile Through a Postulated Crack WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-1 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1 7.0 FRACTURE MECHANICS EVALUATION 7.1 LOCAL FAILURE MECHANISM The local mechanism of failure is primarily dominated by the crack tip behavior in terms of crack-tip blunting, initiation, extension, and final crack instability. The local stability will be assumed if the crack does not initiate at all. It has been accepted that the initiation toughness measured in terms of JIc from a J-integral resistance curve is a material parameter defining the crack initiation. If, for a given load, the calculated J-integral value is shown to be less than the JIc of the material, then the crack will not initiate. If the initiation criterion is not met, one can calculate the tearing modulus as defined by the following relation:
2 f
app E
x da dJ T
(7-1)
Where:
Tapp
=
applied tearing modulus E
=
modulus of elasticity f
=
0.5 (y + u) = flow stress a
=
crack length y, u
=
yield and ultimate strength of the material, respectively Stability is said to exist when ductile tearing does not occur if Tapp is less than Tmat, the experimentally determined tearing modulus. Since a constant Tmat is assumed, a further restriction is placed in Japp. Japp must be less than Jmax; where Jmax is the maximum value of J for which the experimental Tmat is greater than or equal to the Tapp used.
As discussed in Section 5.2 the local crack stability criteria is a two-step process:
(1)
If Japp < JIc, then an existing crack is stable (or a crack will not initiate);
(2)
If Japp > JIc; and Tapp < Tmat and Japp < Jmax, then the crack is stable.
7.2 GLOBAL FAILURE MECHANISM Determination of the conditions which lead to failure in stainless steel should be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. One method for predicting the failure of ductile material is the plastic instability method, based on traditional plastic limit load concepts, but accounting for strain hardening and taking into account the presence of a flaw. The flawed pipe is predicted to fail when the remaining net section reaches a stress level at which a plastic hinge is formed. The stress level at which this occurs is termed as the flow stress. The flow stress is generally taken as the average of the yield and ultimate tensile strength of the material at the temperature of interest.
This methodology has been shown to be applicable to ductile piping through a large number of experiments and will be used here to predict the critical flaw size in the primary coolant piping. The failure criterion has been obtained by requiring equilibrium of the section containing the flaw (Figure 7-1) when loads are applied. The detailed development is provided in Appendix A for a through-wall circumferential flaw in a WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-2 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1 pipe with internal pressure, axial force, and imposed bending moments. The limit moment for such a pipe is given by:
]
[
a,c,e Where:
[
The analytical model described above accurately accounts for the piping internal pressure as well as imposed axial force as they affect the limit moment. Good agreement was found between the analytical predictions and the experimental results (Reference 7-1).
For application of the limit load methodology, the material, including consideration of the configuration, must have a sufficient ductility and ductile tearing resistance to sustain the limit load.
7.3 RESULTS OF CRACK STABILITY EVALUATION As discussed in Sections 7.1 and 7.2, the LBB evaluation for H.B. Robinson Unit 2 consists of evaluating two failure mechanisms. Stability analyses were performed at the critical locations established in Section 5.1.
The elastic-plastic fracture mechanics (EPFM) J-integral analyses for through-wall circumferential cracks in a cylinder were performed using the procedure in the EPRI fracture mechanics handbook (Reference 7-2).
The more limiting lower-bound tensile properties for base metal for A-351 CF8M elbow material from Section 4.0 were applied (see Table 4-3). The fracture toughness properties established in Section 4.3, and the normal plus SSE loads given in Table 3-2 were used for the EPFM calculations. Evaluations were performed at the toughness critical locations identified in Section 5.1. Note that one bounding J-integral evaluation was performed for the Hot Leg, Crossover Leg, and Cold Leg based on the fracture toughness properties provided in Table 4-5. The results of the elastic-plastic fracture mechanics J-integral evaluations are given in Table 7-1. The associated leakage flaw sizes from Table 6-1 are also presented in Table 7-1.
a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-3 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1 A stability analysis based on limit load as described in Section 7.2 was performed for critical Locations 1, 3, 6, and 13. Table 7-1 summarize the results of the stability analyses based on the limit load for piping A-376 TP316 and A-351 CF8M elbow materials.
The limit load analyses consider material properties (yield and ultimate strength) of the base metal, and not the material properties of the weld metal. The base metal (piping) is considered to have more limiting material properties than the weld metal. Therefore, in the limit load evaluation the faulted loads (include both the axial loads (including pressure) and the moment loads) from Table 3-2 were increased by the Z-correction factors to account for reduction of the material toughness due to the welding process used during construction consistently with the methodology of SRP 3.6.3. It is confirmed that the limit load analysis in this report bounds both the weld metal and base metal since the more limiting material properties of the base metals were used in combination with additional penalty Z-correction factor for the stainless-steel weld.
The welding process implemented at field weld Location 1 is a combination of Gas Tungsten Arc Welding (GTAW) and Shielded Metal Arc Welding (SMAW). A Z-correction factor is not applicable for the GTAW welding process; therefore, the SMAW process governs these evaluations at Location 1. The shop welds are assumed to be made of GTAW, SMAW, or SAW combination weld. Shop welds are at critical Locations 3, 6, and 13. For LBB evaluations, SAW is more limiting for crack stability analysis compared to SMAW process. Therefore, a conservative approach is taken assuming a SAW process at Locations 3, 6 and 13.
The Z-correction factor for the SMAW and SAW welding processes (References 7-3 and 7-5) are as follows:
Location 1:
Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW Locations 3, 6 and 13: Z = 1.30 [1.0 + 0.010 (OD-4)] for SAW Where OD is the outer diameter of the pipe in inches.
The Z-correction factors were calculated for the critical locations using the dimensions given in Table 3-1.
The Z-correction factors are 1.60 for Location 1, 1.69 for Location 3, 1.72 for Location 6 and 1.67 for Location 13.
In the J-integral evaluation, Japp was calculated based on the faulted loads in Table 3-2 without any Z-correction factors to account for reduction in fracture toughness. This is because the calculation of JIc, as part of the J-integral evaluations, already considers reduction in fracture toughness due to thermal aging of the CASS materials at normal operating temperature over extended operating periods. This reduction in fracture toughness is based on correlations in NUREG/CR-4513 Revisions 1 and 2 (References 7-6 and 7-7), which have determined lower bound fracture toughness as discussed in Section 4.3. Therefore, no additional Z-factors are necessary because the reduction in fracture toughness is already captured with the consideration of end-of-life (saturated) fracture toughness values from NUREG/CR-4513.
Therefore, the limit load analysis for CASS materials in this report considered the reduced fracture toughness of the weld (Z-correction factor), and the J-applied analysis considered the reduced fracture toughness of the thermally aged CASS material per References 7-6 and 7-7.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-4 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1 7.4 RPV NOZZLE ALLOY 82/182 WELDS Alloy 82/182 or Alloy 82 welds which are susceptible to PWSCC are present at the RPVINs and RPVONs for all loops. As discussed in Section 2.1, for RPVINs and RPVONs the potential PWSCC have not been mitigated. The limit load evaluation for the unmitigated weld locations (Location 1 is performed and envelopes Location 14).
The typical material properties of the Alloy 82/182 DM weld material from Table 4-3 were considered for the limit load analysis at Location 1.
[
]a,c,e The Z-multiplication factor of 1.21 for the Alloy 82/182 material was calculated at Location 1. Note that in the limit load calculation for these locations, the applicable Z-correction factors for SMAW were conservatively used instead of the Z-multiplication factor of 1.21 for Alloy 82/182 material.
As discussed in Section 6.4, an increased factor of 1.69 to account for the PWSCC as applicable is applied to the leakage flaw size calculation.
Table 7-2 provides summary results for Alloy 82/182 DM weld material including associated leakage flaw sizes from Table 6-1.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-5 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1
7.5 REFERENCES
7-1 Kanninen, M. F., et al., Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks, EPRI NP-192, September 1976.
7-2 Kumar, V., German, M. D. and Shih, C. P., An Engineering Approach for Elastic-Plastic Fracture Analysis, EPRI Report NP-1931, Project 1237-1, Electric Power Research Institute, July 1981.
7-3 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.
7-4 ASME Pressure Vessel and Piping Division Conference Paper PVP2008-61840, Technical Basis for Revision to Section XI Appendix C for Alloy 600/82/182/132 Flaw Evaluation in Both PWR and BWR Environments, July 28-31, Chicago IL, USA.
7-5 NUREG-0800, Revision 1, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures, March 2007.
7-6 O. K. Chopra, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, NUREG/CR-4513, Revision 1, U.S. Nuclear Regulatory Commission, Washington, DC, August 1994.
7-7 O. K. Chopra, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, NUREG/CR-4513, Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, May 2016 including Errata, March 15, 2021.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-6 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1 Table 7-1: H.B. Robinson Unit 2 Stability Results Based on Elastic-Plastic J-Integral Evaluations for A-351 CF8M Material Table 7-2: H.B. Robinson Unit 2 Stability Results Based on Limit Load for A-351 CF8M, A-376 TP316, and Alloy 82/182 Materials a,c,e a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 7-7 Fracture Mechanics Evaluation May 2024 WCAP-15628-NP Revision 1 Figure 7-1: [
]a,c,e Stress Distribution WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-1 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 8.0 FATIGUE CRACK GROWTH ANALYSIS To determine the sensitivity of the primary coolant system to the presence of small cracks, a plant specific fatigue crack growth (FCG) analysis was carried out for the [
]a,c,e of Figure 3-2). These region were selected because crack growth calculated here will be typical (i.e., the design transient thermal and pressure stresses will be representative) of that in the entire primary loop. The crack growth at these regions will demonstrate that small surface flaws would not develop to through-wall flaws during the plant design life. Crack growths calculated at other locations can be expected to show less than 10% variation.
The methods used in the fatigue crack growth analysis reported here are the same as those suggested by Section XI of the ASME Code. The analysis procedures involves postulating an initial flaw at specific regions and predicting the growth of that flaw due to an imposed series of loading transients. The input required for a fatigue crack growth analysis is basically the information necessary to calculate the parameter KI which depends on crack and structure geometry and the range of applied stresses in the area where the crack exists. Once KI is calculated, the growth due to that particular stress cycle can be calculated. This increment of growth is then added to the original crack size, and the analysis proceeds to the next transient.
The procedure is continued in this manner until all the transients predicted to occur in the period of evaluation have been analyzed.
The transients used for the fatigue crack growth of the H.B. Robinson Unit 2 plant are listed in Table 8-1.
The transients used in this evaluation are not those contained in the original equipment specification (Reference 8-1); instead, the latest transient specification available has been used. The number of cycles listed within Table 8-1 are still applicable and bound H.B. Robinson Unit 2, with respect to the 80-year transient cycle projections, for the subsequent period of extended operation.
All normal, upset, and test conditions were considered. A summary of the applied transients is provided in Table 8-1. Circumferentially oriented surface flaws were postulated in the region, assuming the flaw was located in three different locations, as shown in Figure 8-1. Specifically, these were:
Cross Section A: Inconel Cross Section B: SA 508 Class 2 or 3 Low Alloy Steel Cross Section C: Stainless Steel CRACK GROWTH RATE REFERENCE CURVES - FERRITIC STEEL The crack growth rate curves used in the analyses were taken directly from Appendix A of Section XI of the ASME code (Reference 8-7). Water environment curves were used for all inside surface flaws, and the air environment curve was used for embedded flaws and outside surface flaws.
For water environments crack growth rate is a function of both the applied stress intensity factor range, and the R ratio (Kmin/Kmax) for the transient.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-2 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 For R 0.25 19,
1.0210 (8-1) 19,
1.0110 Where, da/dN = crack growth rate, micro-inches/cycle.
For R 0.65 12,
1.2010 (8-2) 12,
2.5210 For R ratio between these two extremes, interpolation is recommended.
The crack growth rate reference curve for air environments is a single curve, with growth rate being only a function of applied K. This reference curve is also shown in Figure 8-2.
0.026710 (8-3)
- Where, da/dN =
crack growth rate, micro-inches/cycle.
KI
=
stress intensity factor range, ksiin
=
(KImax - KImin)
FATIGUE CRACK GROWTH RATE REFERENCE CURVES - STAINLESS STEEL The reference crack growth law used for the stainless steel portions of the system was taken from that developed by the Metal Properties Council - Pressure Vessel Research Committee Task Force in Crack Propagation Technology. The reference curve has the equation:
(8-4)
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-3 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1
]a,c,e This equation appears in Section XI, Appendix C (1989 Addendum) for air environments and its basis is provided in Reference 8-2, and is shown in Figure 8-3. For water environments, and environmental factor of 2 was used, based on the crack growth tests in PWR environments reported in Reference 8-3.
FATIGUE CRACK GROWTH RATE REFERENCE CURVES - ALLOY 600, 182, AND 82 MATERIALS The crack growth rate reference curves for these nickel based alloys have not been developed for the ASME Code, so information was obtained from the literature. [
]a,c,e The crack growth rate is a function of both R Ratio (Kmin/Kmax) and the range of applied stress intensity factor. [
]a,c,e The resulting equation is shown below.
2.2310/1.0 0.5.
(8-5)
This crack growth rate law is slightly steeper than that for stainless steel. Fatigue crack growth results for the Inconel 182 welds are expected to be about the same as Inconel 600 weld.
RESULTS AND CONCLUSIONS The transients and cycles for the H.B. Robinson Unit 2 plant for 80 years are the same as those of 40 years.
It is therefore concluded that the fatigue crack growth analysis shown in Table 8-2 and Table 8-3 is applicable for 80 years. The results show that fatigue crack growth is not a concern for the H.B Robinson Unit 2 primary loop piping.
The intent of FCG in the LBB analysis was not to use initial flaw depths that are larger than the Acceptance Tables of ASME Section XI IWB-3410-1, but rather to show a defense in-depth fatigue crack growth based on small flaw sizes that are detectable based on NDE examination techniques, which would not become through-wall flaws over the design life of the plant.
Additionally, the fatigue crack growth evaluation is considered a defense in depth review. FCG is no longer a requirement for the Leak-Before-Break (LBB) analysis, since the LBB analysis is based on the postulation WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-4 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 of through-wall flaw, whereas the FCG analysis is performed based on the surface flaw. Furthermore, Reference 8-6 has indicated that the Commission deleted the fatigue crack growth analysis in the proposed rule. This requirement was found to be unnecessary because it was bounded by the crack stability analysis.
Nevertheless, the fatigue crack growth analysis is retained herein for information purposes and to demonstrate that small surface flaws do not result in through-wall flaws over the life of the plant.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-5 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1
8.1 REFERENCES
8-1 Westinghouse Equipment Specification Number 676413 - Rev. 4, and Addendum 952542, 1973, 8-2 James, L. A., and Jones, D. P., Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, in Predictive Capabilities in Environmentally Assisted Cracking, ASME publication PVP-99, Dec. 1985.
8-3 Bamford, W. H., Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment, Trans ASME, Journal of Pressure Vessel Technology, Feb. 1979, Engineering Development Labs Report HEDL-TME-76-43, May 1976.
8-4 James, L. A, Fatigue Crack Propagation Behavior of Inconel 600, in Hanford Engineering Labs Report HEDL-TME-76-43, May 1976.
8-5 Hale, D. A. et al, Fatigue Crack Growth in Piping and RPV Steels in Simulated BWR Water Environment.
8-6 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Rupture, Final Rule, Federal Register/Vol. 52, No. 207/Tuesday, October 27, 1987/ Rules and Regulations, pp. 41288-41295.
8-7 ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 1983 Edition, July 1, 1983.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-6 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 Table 8-1 Summary of Reactor Vessel Transients for H.B. Robinson Unit 2 Typical Transient Identification Number of Cycles Heatup/Cooldown 200 Unit Loading and Unloading Between 15% And 100 % @ 5% Full Power 18300 Unit Loading and Unloading Between 0% And 15% Of Full Power 500 Step Load Decrease/Increase 2000 Large Step Decrease with Steam Dump 200 Steady State Fluctuation 150000 Random Fluctuation 3000000 Feedwater Cycling 2000 Refueling 80 Loss of Load 80 Loss of Power 40 Loss of Flow 80 Reactor Trip with No Cooldown 230 Reactor Trip with Cooldown, No SI 160 Reactor Trip with Cooldown, And SI 10 Inadvertent RCS Depressurization 60 Inadvertent Startup of an Inactive Loop 20 Inadvertent SI Actuation 60 Control Rod Drop 80 Excessive Feedwater Flow 30 Boron Concentration 26400 Loop Out-Of-Service, Normal Loop Startup 70 Loop Out-Of-Service, Normal Loop Shutdown 80 Primary Side Leak Test 200 OBE 200 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-7 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 Table 8-2: Fatigue Crack Growth at Inlet Nozzle Safe-End Region Initial Flaw (in)
Final Flaw (in)
Ferritic Steel (Nozzle)
Stainless Inconel 0.305 0.3069 0.3066 0.3053 0.458 0.4644 0.4609 0.4590 0.610 0.6194 0.6141 0.6123 Table 8-3: Fatigue Crack Growth at Outlet Nozzle Safe-End Region Initial Flaw (in)
Final Flaw (in)
Ferritic Steel (Nozzle)
Stainless Inconel 0.25 0.2761 0.2677 0.2525 0.375 0.4664 0.4119 0.3840 0.500 0.6128 0.5505 0.5160 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-8 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 Figure 8-1: Typical Cross-Section of RPV Inlet and Outlet Nozzle Safe-End WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-9 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 Figure 8-2: Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-10 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 Figure 8-3: Reference Fatigue Crack Growth Law for Stainless Steel in Air Environments WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 8-11 Fatigue Crack Growth Analysis May 2024 WCAP-15628-NP Revision 1 Figure 8-4: Reference Fatigue Crack Growth Law for [
]a,c,e in a Water Environment at 600°F WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 9-1 Assessment of Margins May 2024 WCAP-15628-NP Revision 1
9.0 ASSESSMENT
OF MARGINS The results of the leak rates of Section 6.4 and the corresponding stability and fracture toughness evaluations of Sections 7.1, 7.2, 7.3 and 7.4 are used in performing the assessment of margins. Margins are shown for piping A-376 TP316 material, for elbow A-351 CF8M material and for Alloy 82/182 weld material in Table 9-1. All of the LBB recommended margins are satisfied. Also, the existence of Alloy 82 DM welds at the RPVIN and RPVON to safe-end are acceptable for the SLR program for 80-years of plant operation.
In summary, at all the critical locations relative to:
1.
Flaw Size - Using faulted loads obtained by the absolute sum method, a margin of 2 or more exists between the critical flaw and the flaw having a leak rate of 10 gpm (the leakage flaw).
2.
Leak Rate - A margin of 10 exists between the calculated leak rate from the leakage flaw and the plant leak detection capability of 1 gpm.
3.
Loads - At the critical locations the leakage flaw was shown to be stable using the faulted loads obtained by the absolute sum method (i.e., a flaw twice the leakage flaw size is shown to be stable; hence the leakage flaw size is stable). A margin of 1 on loads using the absolute summation of faulted load combinations is satisfied.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 9-2 Assessment of Margins May 2024 WCAP-15628-NP Revision 1 Table 9-1: Leakage Flaw Sizes, Critical Flaw Sizes and Margins for H.B. Robinson Unit 2 a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation) 10-1 Conclusions May 2024 WCAP-15628-NP Revision 1
10.0 CONCLUSION
S This report justifies the elimination of RCS primary loop pipe breaks from the structural design basis for the H.B. Robinson Unit 2 for the 80-year license renewal period (SLR) as follows:
a.
Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation.
Alloy 82/182 welds are present at the reactor pressure vessel outlet nozzle (RPVON) and reactor pressure vessel inlet nozzle (RPVIN). The Alloy 82/182 welds are susceptible to PWSCC (Primary Water Stress Corrosion Cracking).
b.
The LBB has been reevaluated for 80-year plant life SLR program at the unmitigated DM weld locations. The LBB evaluation has been performed considering Alloy 82/182 material properties which includes appropriate PWSCC crack morphology parameter.
c.
Evaluation of the RCS piping considering the thermal aging effects for the 80-year plant life period of the SLR program, and also the use of the most limiting fracture toughness properties ensures that each materials profile is appropriately bounded by the LBB results presented in this report. As stated in Section 7.0, for local and global failure mechanisms, all locations are evaluated using the cast stainless steel material properties (A-351 CF8M) which present a limiting condition due to the thermal aging effects.
d.
Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.
e.
The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.
f.
Adequate margin exists between the leak rate of small stable flaws and the capability of the H.B.
Robinson Unit 2 reactor coolant system pressure boundary Leakage Detection System.
g.
Ample margin exists between the small stable flaw sizes of item (f) and larger stable flaws.
h.
Ample margin exists in the material properties used to demonstrate end-of-service life (fully aged) stability of the critical flaws.
For the critical locations, postulated flaws will be stable because of the ample margins described in f, g, and h above.
Based on the discussion above, the Leak-Before-Break conditions and margins are satisfied for the H.B.
Robinson Unit 2 primary loop piping. All the recommended margins are satisfied. It is therefore concluded that dynamic effects of RCS primary loop pipe breaks need not be considered in the structural design basis for H.B. Robinson Unit 2 for the 80-year plant life (subsequent license renewal program).
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
A-1 Limit Moment May 2024 WCAP-15628-NP Revision 1 APPENDIX A: LIMIT MOMENT
[
] a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
A-2 Limit Moment May 2024 WCAP-15628-NP Revision 1 Figure A-1: Pipe with a Through-Wall Crack in Bending WESTINGHOUSE NON-PROPRIETARY CLASS 3
- This record was final approved on 06/11/2024 12:41:25. (This statement was added by the PRIME system upon its validation)
WCAP-15628-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 Nowicki Timothy J Jun-11-2024 10:09:14 Author Approval Scott Alexandria M Jun-11-2024 10:17:34 Reviewer Approval Wiratmo Momo Jun-11-2024 10:50:34 Approver Approval Iddings Remington Jun-11-2024 12:41:25 Files approved on Jun-11-2024