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A System-Level Model for Estimating Residual Strain and Life of Nuclear Reactor Coolant System Components Under Connected-System-Thermal–Mechanical Boundary Conditions

  • Sp Iss: Advances in Residual Stress Technology
  • Published:
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Abstract

Background

Environmental-assisted fatigue (EAF) is a major issue for the long-term survival of nuclear power plant fleets in the U.S. and worldwide. Multi-material welded regions (e.g., nozzles) and other high-stress regions of reactor coolant system (RCS) components are prone to EAF-related damage.

Objective

The discussed work describes a system-level finite element (FE) model of RCS components of a pressurized water reactor (PWR). This is with the goal of predicting the stress hotspots, strain residuals, strain amplitudes and the resulting fatigue lives.

Methods

The FE model was developed considering system-level loading conditions (under connected system thermal–mechanical boundary conditions). Thermal–mechanical stress analysis was performed considering thermal stratification and a design-basis reactor loading cycle. Based on the FE model results, the strain residuals, strain amplitudes and resulting fatigue lives of RCS components were predicted.

Results

The results show that some of the RCS components can have significantly different strain amplitudes, residual strain, and fatigue lives, despite having similar geometry and material. Higher residual strain can lead to accelerated cyclic hardening of material and the associated effect of EAF. The simulated component-level strain profile (under realistic multi-axial-multi-physics loading cycle) can guide the selection of appropriate test inputs for conducting laboratory-scale EAF tests, which is a focus of future works.

Conclusions

Despite similar geometry and material the RCS component can have significantly different strain profiles and resulting fatigue lives. The discussed approach can help to identify and prioritize the RCS components for conducting expensive nondestructive evaluation (NDE) inspections.

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Abbreviations

CL:

Cold Leg

DMW:

Dissimilar Metal Weld

EAF:

Environmental Assisted Fatigue

HL:

Hot Leg

LAS:

Low Alloy Steel

LTO:

Long Term Operation

NDE:

Non-Destructive Evaluation

NPP:

Nuclear Power Plant

PRZ:

Pressurizer

PWR:

Pressurized Water Reactor

RCS:

Reactor Cooling System

RPV:

Reactor Pressure Vessel

SG:

Steam Generator

SL:

Surge Line

SMW:

Similar Metal Weld

SS:

Stainless Steel

References

  1. The Enduring Value of Nuclear Energy Assets (2020) World Nuclear News. https://www.world-nuclear-news.org/Articles/LTO-essential-to-cost-effective-climate-benefits-o

  2. Chopra O, Stevens GL (2018) Effect of LWR Water Environments on the Fatigue Life of Reactor Materials. The U.S. Nuclear Regulatory Commission (NUREG/CR-6909, Revision 1) – Final Report. https://www.nrc.gov/docs/ML1631/ML16319A004.pdf

  3. Environmental Fatigue Evaluation Method for Nuclear Power Plants (2011) Nuclear Energy System Safety Division, Japan Nuclear Energy Safety Organization, JNES-SS-105, (March) https://www.nrc.gov/docs/ML1130/ML113010189.pdf

  4. Brust FW et al (2010) Summary of Weld Residual Stress Analyses for Dissimilar Metal Weld Nozzles. In Pressure Vessels and Piping Conference (Vol. 49255:1521–1529). https://doi.org/10.1115/PVP2010-26106

  5. Song TK et al (2014) Prediction of welding residual stress profile in dissimilar metal nozzle butt weld of nuclear power plant. Procedia Mater Sci 3:784–789. https://doi.org/10.1016/j.mspro.2014.06.128

    Article  Google Scholar 

  6. Koo DY et al (2017) Estimation of residual stress in welding of dissimilar metals at nuclear power plants using cascaded support vector regression. Nucl Eng Technol 49(4):817–824. https://doi.org/10.1016/j.net.2017.02.003

    Article  Google Scholar 

  7. Kim JS, Seo JH (2012) A study on welding residual stress analysis of a small-bore nozzle with dissimilar metal welds. Int J Press Vessels Pip 90:69–76. https://doi.org/10.1016/j.ijpvp.2011.10.010

    Article  Google Scholar 

  8. Maekawa A et al (2013) Residual stress study in dissimilar metal welds of a pwr pressurizer surge nozzle: validation of developed fast analysis method and examination of safe-end length effect. In Pressure Vessels and Piping Conference (Vol. 55713, Paper No. V06BT06A060). Am Soc Mech Eng. https://doi.org/10.1115/PVP2013-97176

  9. Liu RF, Wang JC (2021) Finite element analyses of the effect of weld overlay sizing on residual stresses of the dissimilar metal weld in PWRs. Nucl Eng Des 372:110959. https://doi.org/10.1016/j.nucengdes.2020.110959

    Article  Google Scholar 

  10. Lee KS et al (2009) Finite element analysis and measurement for residual stress of dissimilar metal weld in pressurizer safety nozzle mockup. J Mech Sci Technol 23(11):2948–2955. https://doi.org/10.1007/s12206-009-0809-2

    Article  Google Scholar 

  11. Hill MR, Olson MD (2013) Biaxial Residual Stress Mapping in a PWR Dissimilar Metal Weld. Pressure Vessels and Piping Conference (Vol. 55713, p. V06BT06A064). Am Soc Mech Eng. https://doi.org/10.1115/PVP2013-97246

  12. Hill MR et al (2016) Biaxial residual stress mapping for a dissimilar metal welded nozzle. J Press Vessel Technol 138(1). https://doi.org/10.1115/1.4031504

  13. Olson MD et al (2015) Residual stress measurements in dissimilar weld metal. Exp Mech 55(6):1093–1103. https://doi.org/10.1007/s11340-015-0010-8

    Article  Google Scholar 

  14. Barua B et al (2018) Methodology for stress-controlled fatigue test under in-air and pressurized water reactor coolant water condition and to evaluate the effect of pressurized water reactor water and loading rate on ratcheting. J Press Vessel Technol 140(3). https://doi.org/10.1115/1.4039345

  15. Mohanty S et al (2017) Modeling the cycle-dependent material hardening behavior of 508 low alloy steel. Exp Mech 57(6):847–855. https://doi.org/10.1007/s11340-017-0278-y

    Article  Google Scholar 

  16. Barua B et al (2018) A cyclic-plasticity-based mechanistic approach for fatigue evaluation of 316 stainless steel under arbitrary loading. J Press Vessel Technol 140(1). https://doi.org/10.1115/1.4038525

  17. Mohanty S et al (2015) Tensile And 17. Fatigue Testing and Material Hardening Model Development For 508 LAS Base Metal and 316 SS Similar Metal Weld Under In-Air and PWR Primary Loop Water Conditions. U.S. Department of Energy Office of Scientific and Technical Information, Argonne National Laboratory (No. ANL/LWRS-15/02). https://doi.org/10.2172/1224989

  18. Mohanty S et al (2016) Thermal-mechanical stress analysis of PWR pressure vessel and nozzles under grid load-following mode: Interim report on the effect of cyclic hardening material properties and pre-existing cracks on stress analysis results. U.S. Department of Energy Office of Scientific and Technical Information, Argonne National Laboratory (No. ANL/LWRS-16/01). https://doi.org/10.2172/1249554

  19. Mohanty S et al (2017) Final Report on CFD and Thermal-Mechanical Stress Analysis of PWR Surge Line under Transient Condition Thermal Stratification and an Evolutionary Cyclic Plasticity Based Transformative Fatigue Evaluation Approach without Using S~N Curve. U.S. Department of Energy Office of Scientific and Technical Information, Argonne National Laboratory (No. ANL/LWRS-17/03, Rev-1). https://doi.org/10.2172/1480513

  20. ASME Boiler and Pressure Vessel Code (2017) Design by Analysis Sections III and VIII, Division 2. Am Soc Mech Eng. https://www.asme.org

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Acknowledgements and Funding

This research was supported through the U.S. Department of Energy’s Light Water Reactor Sustainability program under the work package of environmental fatigue study with program manager Dr. Thomas. M. Rosseel and deputy program manager Dr. Xiang (Frank) Chen.

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Authors and Affiliations

  1. Nuclear Science and Engineering Division, Argonne National Laboratory, Lemont, IL, 60439, USA

    S. Mohanty

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  1. S. Mohanty
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Correspondence to S. Mohanty.

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Appendix

Appendix

Expansion coefficients for 316SS and 508LAS base, 316SS-316SS SW filler, and 316SS-508LAS DW filler and butter welds.

Temp (°C)

316SS base metals

508LAS base metals

316SS-316SS SW filler (E316-16) welds

316SS-508LAS DW filler (In 82) welds

316SS-508LAS DW butter (In 182) welds

Expansion coeff (1/°C)

Expansion coeff (1/°C)

Expansion coeff (1/°C)

Expansion coeff (1/ °C)

Expansion coeff (1/ °C)

21.11

2.8385E-06

8.0805e-06

4.4273e-07

3.3078e-06

7.9477e-06

31.11

2.8385E-06

8.2561e-06

4.4273e-07

4.067e-06

7.8772e-06

41.11

3.4739E-06

8.7299e-06

8.7463e-07

4.7487e-06

7.8772e-06

51.11

4.6378E-06

9.1746e-06

2.2e-06

5.2931e-06

7.8772e-06

61.11

5.7242E-06

9.5914e-06

3.4306e-06

5.7206e-06

7.8772e-06

71.11

6.7361E-06

9.9815e-06

4.5703e-06

6.0495e-06

8.8574e-06

81.11

7.6763E-06

1.0346e-05

5.6232e-06

6.2969e-06

9.9318e-06

91.11

8.5479E-06

1.0686e-05

6.5934e-06

6.4781e-06

1.0949e-05

101.11

9.3536E-06

1.1003e-05

7.4849e-06

6.6072e-06

1.1855e-05

111.11

1.0096E-05

1.1298e-05

8.3017e-06

6.6965e-06

1.2624e-05

121.11

1.0779E-05

1.1572e-05

9.0479e-06

6.7574e-06

1.3251e-05

131.11

1.1405E-05

1.1826e-05

9.7275e-06

6.7998e-06

1.375e-05

141.11

1.1976E-05

1.2062e-05

1.0344e-05

6.8322e-06

1.414e-05

151.11

1.2496E-05

1.2281e-05

1.0903e-05

6.8621e-06

1.4445e-05

161.11

1.2967E-05

1.2483e-05

1.1407e-05

6.896e-06

1.4693e-05

171.11

1.3393E-05

1.267e-05

1.186e-05

6.9392e-06

1.4906e-05

181.11

1.3776E-05

1.2843e-05

1.2267e-05

6.9961e-06

1.5104e-05

191.11

1.4119E-05

1.3004e-05

1.2632e-05

7.0699e-06

1.5301e-05

201.11

1.4425E-05

1.3153e-05

1.2958e-05

7.1633e-06

1.5506e-05

211.11

1.4697E-05

1.3292e-05

1.325e-05

7.278e-06

1.572e-05

221.11

1.4938E-05

1.3422e-05

1.3512e-05

7.415e-06

1.5941e-05

231.11

1.5151E-05

1.3544e-05

1.3747e-05

7.5745e-06

1.6162e-05

241.11

1.5338E-05

1.3659e-05

1.396e-05

7.7563e-06

1.6372e-05

251.11

1.5503E-05

1.3768e-05

1.4155e-05

7.9594e-06

1.6558e-05

261.11

1.5648E-05

1.3873e-05

1.4336e-05

8.1823e-06

1.6709e-05

271.11

1.5777E-05

1.3974e-05

1.4506e-05

8.4232e-06

1.6816e-05

281.11

1.5891E-05

1.4073e-05

1.4671e-05

8.6799e-06

1.6872e-05

291.11

1.5995E-05

1.4172e-05

1.4833e-05

8.9497e-06

1.6878e-05

301.11

1.6091E-05

1.427e-05

1.4998e-05

9.2298e-06

1.6842e-05

311.11

1.6181E-05

1.437e-05

1.5168e-05

9.5171e-06

1.6783e-05

321.11

1.6269E-05

1.4473e-05

1.5348e-05

9.8083e-06

1.6728e-05

331.11

1.6358E-05

1.4579e-05

1.5543e-05

1.01e-05

1.672e-05

341.11

1.6450E-05

1.469e-05

1.5755e-05

1.0389e-05

1.6811e-05

350

1.6538E-05

1.4794e-05

1.5963e-05

1.0642e-05

1.7028e-05

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Mohanty, S. A System-Level Model for Estimating Residual Strain and Life of Nuclear Reactor Coolant System Components Under Connected-System-Thermal–Mechanical Boundary Conditions. Exp Mech 62, 1501–1517 (2022). https://doi.org/10.1007/s11340-022-00847-5

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