Abstract
Computational design for fission reactor materials is ready to accelerate the development and qualification of nuclear materials. This review is primarily aimed at computational materials scientists that seek to apply ICME to the development of fission reactor materials. We summarize reactor materials and technology, discuss reactor material development and qualification today, show how ICME is being applied to the unique requirements of reactor materials, and provide a future vision.
Similar content being viewed by others
Data availability
No datasets were generated or analyzed during the current study.
References
Hanna, B. N., Abou-Jaoude, A., Guaita, N., Talbot, P. & Lohse, C. Navigating economies of scale and multiples for nuclear-powered data centers and other applications with high service availability needs. Energies 17, 5073 (2024).
Soto Gonzalez, G. J. et al. Powering Data Centers with Clean Energy: A Techno-economic Case Study of Nuclear And Renewable Energy Dependability. Technical Report. (Idaho National Laboratory (INL), Idaho Falls, ID (United States) 2024).
Chant, I. & Murty, K. Structural materials issues for the next generation fission reactors. JOM 62, 67–74 (2010).
Allen, T., Busby, J., Meyer, M. & Petti, D. Materials challenges for nuclear systems. Mater. Today 13, 14–23 (2010).
Olander, D. Nuclear fuels–present and future. J. Nucl. Mater. 389, 1–22 (2009).
Zinkle, S. J. & Busby, J. T. Structural materials for fission & fusion energy. Mater. Today 12, 12–19 (2009).
Wachs, D. M. Transient testing of nuclear fuels and materials in the United States. JOM 64, 1396–1402 (2012).
Crawford, D. C. et al. An approach to fuel development and qualification. J. Nucl. Mater. 371, 232–242 (2007).
Terrani, K. A. et al. Accelerating nuclear fuel development and qualification: modeling and simulation integrated with separate-effects testing. J. Nucl. Mater. 539, 152267 (2020).
Mika, M., Probert, A. & Aitkaliyeva, A. Nuclear fuel qualification: history, current state, and future. Prog. Nucl. Energy 177, 105460 (2024).
Hafner, J., Wolverton, C. & Ceder, G. Toward computational materials design: the impact of density functional theory on materials research. MRS Bull. 31, 659–668 (2006).
Kuehmann, C. J. & Olson, G. B. Computational materials design and engineering. Mater. Sci. Technol. 25, 472–478 (2009).
Curtarolo, S. et al. The high-throughput highway to computational materials design. Nat. Mater. 12, 191–201 (2013).
Butler, T. et al. Computational materials design of crystalline solids. Chem. Soc. Rev. 45, 6138–6146 (2016).
Shevlin, S., Castro, B. & Li, X. Computational materials design. Nat. Mater. 20, 727–727 (2021).
Horstemeyer, M. F. Integrated Computational Materials Engineering (ICME) for Metals: Concepts and Case Studies (John Wiley & Sons, 2018). Google-Books-ID: OkJLDwAAQBAJ.
Ghosh, S., Woodward, C. & Przybyla, C. Integrated Computational Materials Engineering (ICME): Advancing Computational and Experimental Methods (Springer Nature, 2020). Google-Books-ID: Xi3YDwAAQBAJ.
Wang, W. Y. et al. A brief review of data-driven ICME for intelligently discovering advanced structural metal materials: Insight into atomic and electronic building blocks. J. Mater. Res. 35, 872–889 (2020).
Aguiar, J. A., Jokisaari, A. M., Kerr, M. & Allen Roach, R. Bringing nuclear materials discovery and qualification into the 21st century. Nat. Commun. 11, 2556 (2020).
Fundamental Aspects of Nuclear Reactor Fuel Elements: Solutions to Problems. Technical Report. (Department of Nuclear Engineering, California University, Berkeley (USA), 1976).
Fink, J. K. Thermophysical properties of uranium dioxide. J. Nucl. Mater. 279, 1–18 (2000).
Manara, D. et al. Thermodynamic and thermophysical properties of the actinide carbides (2012).
Parrish, R. & Aitkaliyeva, A. A review of microstructural features in fast reactor mixed oxide fuels. J. Nucl. Mater. 510, 644–660 (2018).
Gaiduchenko, A. B. Thermophysical properties of irradiated uranium-zirconium fuel. Energy 104, 5–10 (2008).
Ahn, S., Irukuvarghula, S. & McDeavitt, S. M. Thermophysical investigations of the uranium-zirconium alloy system. J. Alloy. Compd. 611, 355–362 (2014).
Janney, D. E. et al. Metallic Fuels Handbook. Technical Report. (Idaho National Lab. (INL), Idaho Falls, ID (United States), 2017).
Meyer, M. K. et al. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel. J. Nucl. Mater. 304, 221–236 (2002).
Ajantiwalay, T., Smith, C., Keiser, D. D. & Aitkaliyeva, A. A critical review of the microstructure of U-Mo fuels. J. Nucl. Mater. 540, 152386 (2020).
Vasudevamurthy, G. & Nelson, A. T. Uranium carbide properties for advanced fuel modeling—a review. J. Nucl. Mater. 558, 153145 (2022).
Watkins, J. K., Wagner, A. R., Gonzales, A., Jaques, B. J. & Sooby, E. S. Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium diboride and uranium carbide. J. Nucl. Mater. 560, 153502 (2022).
Chiotti, P., Robinson, W. C. & Kanno, M. Thermodynamic properties of uranium oxycarbides. J. Less Common Met. 10, 273–289 (1966).
Phillips, J. A., Nagley, S. G. & Shaber, E. L. Fabrication of uranium oxycarbide kernels and compacts for HTR fuel. Nucl. Eng. Des. 251, 261–281 (2012).
Wojtaszek, D. T. & Bromley, B. P. Physics evaluation of alternative uranium-based oxy-carbide annular fuel concepts for potential use in compact high-temperature gas-cooled reactors. J. Nucl. Eng. Rad. Sci. 9, 011504 (2023).
Watkins, J. K., Gonzales, A., Wagner, A. R., Sooby, E. S. & Jaques, B. J. Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium mononitride. J. Nucl. Mater. 553, 153048 (2021).
Sajdova, A. Accident-Tolerant Uranium Nitride. PhD Thesis, Chalmers University of Technology (2017).
Hania, P. R. et al. Irradiation and post-irradiation examination of uranium-free nitride fuel. J. Nucl. Mater. 466, 597–605 (2015).
Birtcher, R. & Wang, L. Stability of uranium silicides during high energy ion irradiation. MRS Online Proc. Libr. 235, 467 (1991).
Finlay, M. R., Hofman, G. L. & Snelgrove, J. L. Irradiation behaviour of uranium silicide compounds. J. Nucl. Mater. 325, 118–128 (2004).
Gonzales, A., Watkins, J. K., Wagner, A. R., Jaques, B. J. & Sooby, E. S. Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium silicide. J. Nucl. Mater. 553, 153026 (2021).
Greenspan, E. et al. Hydride fuel for LWRs—project overview. Nucl. Eng. Des. 239, 1374–1405 (2009).
Demkowicz, P. A., Liu, B. & Hunn, J. D. Coated particle fuel: historical perspectives and current progress. J. Nucl. Mater. 515, 434–450 (2019).
Petti, D. A., Buongiorno, J., Maki, J. T., Hobbins, R. R. & Miller, G. K. Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance. Nucl. Eng. Des. 222, 281–297 (2003).
Zhou, X. W. & Tang, C. H. Current status and future development of coated fuel particles for high temperature gas-cooled reactors. Prog. Nucl. Energy 53, 182–188 (2011).
Brown, N. R. A review of in-pile fuel safety tests of TRISO fuel forms and future testing opportunities in non-HTGR applications. J. Nucl. Mater. 534, 152139 (2020).
Brown, N. R., Hernandez, R. & Nelson, A. T. High volume packing fraction TRISO-based fuel in light water reactors. Prog. Nucl. Energy 146, 104151 (2022).
Petti, D. A. NGNP Fuel Qualification White Paper. Technical Report. (Idaho National Lab (INL), Idaho Falls, ID (United States), 2010).
Karoutas, Z. et al. The maturing of nuclear fuel: past to accident tolerant fuel. Prog. Nucl. Energy 102, 68–78 (2018).
Chapman, R., Crowley, J., Longest, A. & Hofmann, G. Zircaloy Cladding Deformation in a Steam Environment with Transient Heating. Technical Report. (Oak Ridge National Lab., TN (USA), 1978).
Moalem, M. & Olander, D. R. Oxidation of Zircaloy by steam. J. Nucl. Mater. 182, 170–194 (1991).
Kim, H. H. et al. High-temperature oxidation behavior of Zircaloy-4 and Zirlo in steam ambient. J. Mater. Sci. Technol. 26, 827–832 (2010).
Hayes, T. A. & Kassner, M. E. Creep of zirconium and zirconium alloys. Metall. Mater. Trans. A 37, 2389–2396 (2006).
Massih, A. High-temperature creep and superplasticity in zirconium alloys. J. Nucl. Sci. Technol. 50, 21–34 (2013).
Kim, H.-G. et al. Adhesion property and high-temperature oxidation behavior of cr-coated zircaloy-4 cladding tube prepared by 3d laser coating. J. Nucl. Mater. 465, 531–539 (2015).
Lee, Y., Lee, J. I. & No, H. C. Mechanical analysis of surface-coated zircaloy cladding. Nucl. Eng. Technol. 49, 1031–1043 (2017).
Rebak, R. B. Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants. EPJ Nucl. Sci. Technol. 3, 34 (2017).
Charit, I. Accident tolerant nuclear fuels and cladding materials. JOM 70, 173–175 (2018).
Yin, L. et al. Uniform corrosion of FeCrAl cladding tubing for accident tolerant fuels in light water reactors. J. Nucl. Mater. 554, 153090 (2021).
Wagih, M., Spencer, B., Hales, J. & Shirvan, K. Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings. Ann. Nucl. Energy 120, 304–318 (2018).
Braun, J. et al. Chemical compatibility between UO2 fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels). J. Nucl. Mater. 487, 380–395 (2017).
Pham, H. V., Kurata, M. & Steinbrueck, M. Steam oxidation of silicon carbide at high temperatures for the application as accident tolerant fuel cladding, an overview. Thermo 1, 151–167 (2021).
Azevedo, C. R. F. A review on neutron-irradiation-induced hardening of metallic components. Eng. Fail. Anal. 18, 1921–1942 (2011).
Allen, T., Burlet, H., Nanstad, R. K., Samaras, M. & Ukai, S. Advanced structural materials and cladding. MRS Bull. 34, 20–27 (2009).
Keiser, D. D. 3.15-Metal Fuel-Cladding Interaction (Elsevier, 2012).
Was, I. & Beam, H. F. I. 3.3 Advanced Reactor Cladding. Advanced Fuels Campaign Fy 2015 Accomplishments Report 96 (2015).
Sencer, B. H., Kennedy, J. R., Cole, J. I., Maloy, S. A. & Garner, F. A. Microstructural stability of an HT-9 fuel assembly duct irradiated in FFTF. J. Nucl. Mater. 414, 237–242 (2011).
Roberts, J. T. A. Structural Materials in Nuclear Power Systems (Springer Science & Business Media, 2013). Google-Books-ID: 36jhBwAAQBAJ.
Yvon, P., Le Flem, M., Cabet, C. & Seran, J. L. Structural materials for next generation nuclear systems: challenges and the path forward. Nucl. Eng. Des. 294, 161–169 (2015).
Murty, K. L. & Charit, I. Structural materials for Gen-IV nuclear reactors: challenges and opportunities. J. Nucl. Mater. 383, 189–195 (2008).
Yvon, P. & Carré, F. Structural materials challenges for advanced reactor systems. J. Nucl. Mater. 385, 217–222 (2009).
Yvon, P. Structural Materials for Generation IV Nuclear Reactors (Woodhead Publishing, 2016). Google-Books-ID: i3W0CwAAQBAJ.
Pomaro, B. A review on radiation damage in concrete for nuclear facilities: from experiments to modeling. Model. Simul. Eng. 2016, 4165746 (2016).
Kurtis, K. E. et al. Can we design concrete to survive nuclear environments. Concr. Int 39, 53–59 (2017).
Marsden, B., Jones, A., Hall, G., Treifi, M. & Mummery, P. Graphite as a Core Material for Generation IV Nuclear Reactors (Elsevier, 2017).
Zhou, X. -w, Tang, Y. -p, Lu, Z. -m, Zhang, J. & Liu, B. Nuclear graphite for high temperature gas-cooled reactors. New Carbon Mater. 32, 193–204 (2017).
Kurpaska, L. et al. Structural and mechanical properties of different types of graphite used in nuclear applications. J. Mol. Struct. 1217, 128370 (2020).
Zhou, L. et al. Research progress of steels for nuclear reactor pressure vessels. Materials 15, 8761 (2022).
Davies, L. M. A comparison of Western and Eastern nuclear reactor pressure vessel steels. Int. J. Press. Vessels Pip. 76, 163–208 (1999).
Phythian, W. J. & English, C. A. Microstructural evolution in reactor pressure vessel steels. J. Nucl. Mater. 205, 162–177 (1993).
Odette, G. R. Radiation induced microstructural evolution in reactor pressure vessel steels. MRS Online Proc. Libr. 373, 137 (1994).
Odette, G. R. & Lucas, G. E. Recent progress in understanding reactor pressure vessel steel embrittlement. Radiat. Eff. Defects Solids 144, 189–231 (1998).
Shivprasad, A. P. et al. Advanced Moderator Material Handbook. Technical Report. (LA-UR-20-27683, Los Alamos National Laboratory (LANL), Los Alamos, NM (United States), https://doi.org/10.2172/1671020. https://www.osti.gov/biblio/1671020 (2020).
Vetrano, J. B. Hydrides as neutron moderator and reflector materials. Nucl. Eng. Des. 14, 390–412 (1971).
Gustafson, J. L. Space nuclear propulsion fuel and moderator development plan conceptual testing reference design. Nucl. Technol. 0, 1–3 (2021).
Cameron, I. R. The heavy-water-moderated reactor. in Nuclear Fission Reactors (ed Cameron, I. R.) 269–282 (Springer US, 1982).
Baker, D. E. Graphite as a neutron moderator and reflector material. Nucl. Eng. Des. 14, 413–444 (1971).
Beeston, J. M. Beryllium metal as a neutron moderator and reflector material. Nucl. Eng. Des. 14, 445–474 (1971).
Miao, Y., Stauff, N., Bhattacharya, S., Yacout, A. & Kim, T. K. Advanced Moderation Module for High-Temperature Micro-Reactor Applications. Technical Report. (ANL/CFCT-20/19, Argonne National Lab. (ANL), Argonne, IL (United States), https://doi.org/10.2172/1656612. https://www.osti.gov/biblio/1656612 (2020).
Ehrlich, K., Konys, J. & Heikinheimo, L. Materials for high performance light water reactors. J. Nucl. Mater. 327, 140–147 (2004).
Terricabras, A. J. et al. Performance and properties evolution of near-term accident tolerant fuel: Cr-doped uo2. J. Nucl. Mater. 594, 155022 (2024).
Pieńkowski, L. Competitiveness strategies and technical innovations in light-water small modular reactor projects. Energies 18, 1268 (2025).
Pasanen, A. A. Fundamentals of CANDU reactor nuclear design. 131–214. https://inis.iaea.org/records/e866m-7kj92 (1982).
Morad, C. M., Stefani, G. L. d., Santos, T. A. d. & Associação Brasileira de Energia Nuclear (ABEN), R. d. J. CANDU: Study and Review. Technical Report. (INIS-BR–19726, Associação Brasileira de Energia Nuclear (ABEN), Rio de Janeiro, RJ (Brazil) https://inis.iaea.org/records/j1qfj-fpw08 (2017).
Lineberry, M. J. & Allen, T. R. The Sodium-Cooled Fast Reactor (SFR). Technical Report. (ANL/NT/CP-108933, Argonne National Lab., IL (US) https://www.osti.gov/biblio/803901 (2002).
Ohshima, H. & Kubo, S. 5 - Sodium-cooled fast reactor. In Handbook of Generation IV Nuclear Reactors (ed Pioro, I. L.) Woodhead Publishing Series in Energy, 97–118 (Woodhead Publishing, 2016).
Aoto, K. et al. A summary of sodium-cooled fast reactor development. Prog. Nucl. Energy 77, 247–265 (2014).
The Natrium technology: Providing reliable, carbon-free energy to complement wind and solar.
Natrium™ Reactor and Integrated Energy Storage. https://www.ans.org/news/article-2782/the-natrium-technology-providing-reliable-carbonfree-energy-to-complement-wind-and-solar/ (2022).
McTiernan, N. Arc-100 sodium fast reactor pressure boundary classification and material selection. In Proc. Canadian Nuclear Society (2023).
Kittel, J. H. et al. History of fast reactor fuel development. J. Nucl. Mater. 204, 1–13 (1993).
Burkes, D. E., Fielding, R. S., Porter, D. L., Meyer, M. K. & Makenas, B. J. A US perspective on fast reactor fuel fabrication technology and experience. Part II: Ceramic fuels. J. Nucl. Mater. 393, 1–11 (2009).
Raj, B., Vasudeva Rao, P., Puthiyavinayagam, P. & Ananthasivan, K. Advanced ceramic fuels for sodium-cooled fast reactors. Handbook of Advanced Ceramics and Composites: Defense, Security, Aerospace and Energy Applications 667–702 (Springer, 2020).
Sommer, C. M. et al. Transmutation fuel cycle analyses of the SABR fission-fusion hybrid burner reactor for transuranic and minor actinide fuels. Nucl. Technol. 182, 274–285 (2013).
Kataeva, N. et al. Comparative analysis of the long-term strength of russian ferritic-martensitic reactor steels. J. Nucl. Mater. 605, 155575 (2025).
Saraev, O. M. et al. BN-800 design validation and construction status. Energy 108, 248–253 (2010).
Tashlykov, O. L., Sesekin, A. N., Klimova, V. A. & Mahmoud, K. A. Optimization of the route of the refueling machine to reduce the refueling time: Case study of the BN-800 reactor. Nucl. Eng. Des. 431, 113725 (2025).
Coz, P. L., Sauvage, J.-F. & Serpantie, J.-P. Sodium-Cooled Fast Reactors: the ASTRID Plant Project. Revue Générale Nucléaire 39–44. https://doi.org/10.1051/rgn/20115039 (2011).
Chanteclair, F., Rodriguez, G., Hamy, J.-M. & Dupraz, R.ASTRID reactor: design overview and main innovative options for Basic Design. https://cea.hal.science/cea-02435077 (2017).
Varaine, F. et al. The collaboration of Japan and France on the design of Astrid sodium fast reactor. In International Congress on Advances in Nuclear Power Plants (ICAPP - 2017) (International Congress on Advances in Nuclear Power Plants (ICAPP—2017), 2017).
Goodjohn, A. J. Summary of gas-cooled reactor programs. Energy 16, 79–106 (1991).
Iwatsuki, J. et al. 1 - Overview of high temperature gas-cooled reactor. In Takeda, T. & Inagaki, Y. (eds.) High Temperature Gas-Cooled Reactors, Vol. 5 of JSME Series in Thermal and Nuclear Power Generation 1–16 (Academic Press, 2021).
Beck, J. M. & Pincock, L. F. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant. Technical Report. INL/EXT-10-19329 (Idaho National Lab. (INL), Idaho Falls, ID (United States), 2011).
Zhang, P., Xu, J., Shi, L. & Zhang, Z. Nuclear Hydrogen Production Based on High Temperature Gas Cooled Reactor in China. Strateg. Study Chin. Acad. Eng. 21, 20–28 (2019).
Ritcher, Z., Davidson, E., Skutnik, S. & Munk, M. Modeling and Simulation of Xe-100-type Pebble Bed Gas-Cooled Reactor with SCALE. Technical Report. ORNL/TM-2023/2959 (Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States), 2023).
Ueta, S. et al. Development of high temperature gas-cooled reactor (HTGR) fuel in Japan. Prog. Nucl. Energy 53, 788–793 (2011).
Shibata, T. et al. Present status of JAEA’s R&D toward HTGR deployment. Nucl. Eng. Des. 398, 111964 (2022).
LeBlanc, D. Molten salt reactors: a new beginning for an old idea. Nucl. Eng. Des. 240, 1644–1656 (2010).
Mochizuki, H. Summary of researches on operational characteristics and safety of molten salt fast reactors based on neutronics and thermal-hydraulics coupling analysis. Nucl. Eng. Des. 435, 113941 (2025).
Kelleher, B. C., Gagnon, S. F. & Mitchell, I. G. Thermal gradient mass transport corrosion in NaCl-MgCl2 and MgCl2-NaCl-KCl molten salts. Mater. Today Commun. 33, 104358 (2022).
Young, G. A., Hacket, M. J. & Sham, T. L. Materials challenges for molten salt reactors. http://inis.iaea.org/Search/search.aspx?orig_q=RN:51012440.
Rykhlevskii, A. Chapter 26 - Kairos power pebble bed reactor. In Molten Salt Reactors and Thorium Energy 2nd edn (eds Dolan, T. J., Pázsit, I., Rykhlevskii, A. & Yoshioka, R.) Woodhead Publishing Series in Energy, 945–952 (Woodhead Publishing, 2024).
Khakim, A., Rhoma, F., Waluyo, A. & Suharyana, S. The neutronic characteristics of thermal molten salt reactor. In AIP Conference Proceedings Vol. 2374, 020028 (AIP Publishing LLC, 2021).
Dunn, M. Molten salt reactors: current technology status and the challenges for maritime applications. In Conference Proceedings of INEC Vol. 2024 (Institute of Marine Engineering, Science and Technology (IMarEST), 2024).
Gabriel-Ohanu, E. et al. Optimization of a primary heat exchanger for flibe molten salt nuclear reactor with SCO2 power system. In Turbo Expo: Power for Land, Sea, and Air Vol. 85048, V010T30A023 (American Society of Mechanical Engineers, 2021).
Locatelli, G., Bingham, C. & Mancini, M. Small modular reactors: a comprehensive overview of their economics and strategic aspects. Prog. Nucl. Energy 73, 75–85 (2014).
Ingersoll, D. T. & Carelli, M. D. Handbook of Small Modular Nuclear Reactors (Elsevier, 2014). Google-Books-ID: rJlzAwAAQBAJ.
Schlegel, J. P. & Bhowmik, P. K. Chapter 14 - Small modular reactors. In Nuclear Power Reactor Designs (eds Wang, J., Talabi, S. & Leon, S. B. Y.) 283–308 (Academic Press, 2024).
Peakman, A., Hodgson, Z. & Merk, B. Advanced micro-reactor concepts. Prog. Nucl. Energy 107, 61–70 (2018).
Testoni, R., Bersano, A. & Segantin, S. Review of nuclear microreactors: status, potentialities and challenges. Prog. Nucl. Energy 138, 103822 (2021).
Black, G. et al. Prospects for nuclear microreactors: a review of the technology, economics, and regulatory considerations. Nucl. Technol. 209, S1–S20 (2023).
Yan, B. H., Wang, C. & Li, L. G. The technology of micro heat pipe cooled reactor: a review. Ann. Nucl. Energy 135, 106948 (2020).
Aldebie, F., Fernandez-Cosials, K. & Hassan, Y. Thermal-mechanical safety analysis of heat pipe micro reactor. Nucl. Eng. Des. 420, 113003 (2024).
Duderstadt, J. J. & Hamilton, L. J. Nuclear Reactor Analysis (John Wiley and Sons, Inc., 1975).
Drzewiecki, T., Schmidt, J., n Wert, C. V. & Clifford, P. Fuel Qualification For Advanced Reactors, Final (Nureg-2246) (United States Nuclear Regulatory Commission, 2022).
Sridhar, N., Dunn, D. S., Cragnolino, G. & Yang, N. Nureg/cr-7299: An Integrated Structure and Environment Dependent Corrosion Model for Used Nuclear Fuel Dry Storage Systems. Technical Report. NUREG/CR-7299 (U.S. Nuclear Regulatory Commission, 2010).
Faibish, R. Accelerated Fuel Qualifcation White Paper. Technical Report. ML21287A646. Available at https://www.nrc.gov/docs/ML2128/ML21287A646.pdf (General Atomics, 2021).
White, A. The materials genome initiative: one year on. MRS Bull. 37, 715–716 (2012).
Lemaignan, C. & Motta, A. T. Zirconium alloys in nuclear applications. Mater. Sci. Technol. 10, 1–51 (1994).
Woods, C. Properties of Zircaloy-4 Tubing. Technical Report. (Bettis Atomic Power Lab.(BAPL), Pittsburgh, PA (United States), 1966).
Sabol, G. P. Zirlo™—an alloy development success, j. ASTM Int 2, 3–24 (2005).
Mardon, J.-P., Charquet, D. & Senevat, J. Influence of composition and fabrication process on out-of-pile and in-pile properties of m5 alloy. In Zirconium in the Nuclear Industry: Twelfth International Symposium (ASTM International, 2000).
Tang, C., Stüber, M., Seifert, H. J. & Steinbrück, M. Metallic and ceramic coatings for enhanced accident tolerant fuel cladding. Compr. Nucl. Mater. 4, 490–514 (2020).
Ko, J., Kim, J. W., Min, H. W., Kim, Y. & Yoon, Y. S. Review of manufacturing technologies for coated accident tolerant fuel cladding. J. Nucl. Mater. 561, 153562 (2022).
Mankins, W. L., Hosier, J. C. & Bassford, T. H. Microstructure and phase stability of INCONEL alloy 617. Metall. Trans. 5, 2579–2590 (1974).
McCoy, H. E. & King, J. F. Mechanical Properties of Inconel 617 and 618. Technical Report. ORNL/TM-9337 (Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States) https://doi.org/10.2172/711763. https://www.osti.gov/biblio/711763 (1985).
Nickel, H., Schubert, F. & Schuster, H. Evaluation of alloys for advanced high-temperature reactor systems. Nucl. Eng. Des. 78, 251–265 (1984).
Ren, W. & Swindeman, R. A Review of Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications 489–500 (American Society of Mechanical Engineers Digital Collection, 2008).
Ren, W. & Swindeman, R. A review on current status of alloys 617 and 230 for Gen IV nuclear reactor internals and heat exchangers 1. J Pressure Vessel Technol. 131, 044002 (2009).
Dayton, R., Oxley, J. & Townley, C. Ceramic coated particle nuclear fuels. J. Nucl. Mater. 11, 1–31 (1964).
Goeddel, W. Coated-particle fuels in high-temperature reactors: a summary of current applications. Nucl. Appl. 3, 599–614 (1967).
Sowder, A. & Marciulescu, C. Uranium Oxycarbide (UCO) Tristructural Isotropic (triso) Coated Particle Fuel Performance. Technical Report. Topical Report EPRI-AR-1 (NP), Technical Report (2019).
Zhang, F.-Y., Zhu, G.-F., Zou, Y., Yan, R. & Xu, H.-J. Conceptual design and neutronic analysis of a megawatt-level vehicular microreactor based on TRISO fuel particles and S-CO2 direct power generation. Nucl. Sci. Tech. 33, 69 (2022).
Skerjanc, W. F. & Youinou, G. J. High uranium loading TRISO particle for microreactor applications. Nucl. Eng. Des. 414, 112628 (2023).
DeHart, M. D., Schunert, S. & Labouré, V. M. Nuclear thermal propulsion. In Nuclear Reactors-Spacecraft Propulsion, Research Reactors, and Reactor Analysis Topics (IntechOpen, 2022).
Kiegiel, K., Herdzik-Koniecko, I., Fuks, L. & Zakrzewska-Kołtuniewicz, G. Management of radioactive waste from HTGR reactors including spent TRISO fuel—state of the art. Energies 15, 1099 (2022).
Guittonneau, F., Abdelouas, A. & Grambow, B. Htr fuel waste management: triso separation and acid-graphite intercalation compounds preparation. J. Nucl. Mater. 407, 71–77 (2010).
Council, N. R., on Engineering, D., Sciences, P., Board, N. M. A. & on Integrated Computational Materials Engineering, C. Integrated Computational Materials Engineering: A Transformational Discipline for Improved Competitiveness and National Security (National Academies Press, 2008).
de Pablo, J. J., Jones, B., Kovacs, C. L., Ozolins, V. & Ramirez, A. P. The materials genome initiative, the interplay of experiment, theory and computation. Curr. Opin. Solid State Mater. Sci. 18, 99–117 (2014).
de Pablo, J. J. et al. New frontiers for the materials genome initiative. npj Comput. Mater. 5, 1–23 (2019).
Attari, V. & Arroyave, R. Machine learning-assisted high-throughput exploration of interface energy space in multi-phase-field model with CALPHAD potential. Mater. Theory 6, 5 (2022).
Zhao, Y.-H. Editorial: phase field method and integrated computing materials engineering. Front. Mater. 10, https://doi.org/10.3389/fmats.2023.1145833 (2023).
Otis, R. A. & Liu, Z.-K. High-throughput thermodynamic modeling and uncertainty quantification for ICME. Jom 69, 886–892 (2017).
Wong, T. T. & Paramsothy, M. Icme after one decade: success and challenges. JOM 70, 1642–1643 (2018).
The Minerals, M. M. S. T. Accelerating the Broad Implementation of Verification & Validation in Computational Models of the Mechanics of Materials and Structures (TMS, 2020).
Olson, G. B. Genomic materials design: the ferrous frontier. Acta Mater. 61, 771–781 (2013).
Olson, G. B. & Kuehmann, C. J. Materials genomics: from CALPHAD to flight. Scr. Mater. 70, 25–30 (2014).
Taskin, K. High Performance Ferrium Steels for Aerospace Gearing and Bearing Applications. https://doi.org/10.1520/STP162320190055 (2020).
Josephson, R. Development of ferrium S53 high-strength, corrosion-resistant steel (2009).
Grabowski, J., Sebastian, J., Olson, G., Asphahani, A. & Genellie Jr, R. Integrated computational materials engineering helps successfully develop aerospace alloys. AMP Tech. Artic. 171, 17–19 (2013).
Sebastian, J. & Olson, G. Examples of QuesTek Innovations’ Application ICME to Materials Design, Development, and Rapid Qualification. In 55th AIAA/ASME/ASCE/AHS/ASC Structures, Structural Dynamics, and Materials Conference (American Institute of Aeronautics and Astronautics).
Integrated Computational Materials Design: From Genome o Flight. https://doi.org/10.2514/6.2013-1847. https://arc.aiaa.org/doi/10.2514/6.2013-1847.
Warren, J. A. The Materials Genome Initiative and the Metals Industry. Bridge 54, 25–30 (2024).
Questek. FERRIUM S53: ultrahigh-strength stainless steel. Alloy Digest 57, SS–942 (2008).
Questek. FERRIUM M54: martensitic secondary hardening steel. Alloy Digest 67, SA–822 (2018).
Questek. FERRIUM C64: martensitic secondary hardening steel. Alloy Digest 67, SA–827 (2018).
Cao, W. & Kennedy, R. Role of chemistry in 718-type alloys—allvac 718plus alloy development. Superalloys 2004, 91–99 (2004).
Kennedy, R. L., Cao, W., Bayha, T. & Jeniski, R. Developments in wrought Nb containing superalloys (718+ 100 f), the mineral. Metals & Materials Society (2003).
Xie, X. et al. Structure stability study on a newly developed nickel-base superalloy—allvac® 718plus™. Superalloys 718, 625–706 (2005).
Kushan, M. C., Uzgur, S. C., Uzunonat, Y. & Diltemiz, F. Allvac 718 plus™ superalloy for aircraft engine applications. Recent Adv. Aircr. Technol. 4, 75–96 (2012).
Zinkle, S. J. & Was, G. S. Materials challenges in nuclear energy. Acta Mater. 61, 735–758 (2013).
Tonks, M. R. et al. Mechanistic materials modeling for nuclear fuel performance. Ann. Nucl. Energy 105, 11–24 (2017).
Bhowmik, P. K. et al. Integral and Separate effects test facilities to support water cooled small modular reactors: a review. Prog. Nucl. Energy 160, 104697 (2023).
Nelson, A. T., Adorno Lopes, D., Capps, N. A. & Petrie, C. M. Role of accelerated burnup irradiation testing in support of accelerated fuel qualification. Nucl. Technol. 0, 1–30 (2025).
Choi, H. et al. Accelerated fuel qualification of fast modular reactor fuel in a thermal reactor: modeling and simulation paired with irradiation testing. Nucl. Technol. 0, 1–16 (2025).
Zhou, W. et al. Proton irradiation-decelerated intergranular corrosion of Ni-Cr alloys in molten salt. Nat. Commun. 11, 3430 (2020).
Was, G. S. Challenges to the use of ion irradiation for emulating reactor irradiation. J. Mater. Res. 30, 1158–1182 (2015).
Taller, S., Chen, Y., Song, R., Chen, W.-Y. & Jokisaari, A. An approach to combine neutron and ion irradiation data to accelerate material qualification for nuclear reactors. J. Nucl. Mater. 603, 155385 (2025).
Jokisaari, A. M., Taller, S., Chen, Y., Chen, W.-Y. & Song, R. Promoting regulatory acceptance of combined ion and neutron irradiation testing of nuclear reactor materials: Modeling and software considerations. Prog. Nucl. Energy 178, 105518 (2025).
Chernatynskiy, A., Phillpot, S. R. & LeSar, R. Uncertainty quantification in multiscale simulation of materials: a prospective. Annu. Rev. Mater. Res. 43, 157–182 (2013).
Dienstfrey, A. et al. Uncertainty quantification in materials modeling. JOM 66, 1342–1344 (2014).
Honarmandi, P. & Arróyave, R. Uncertainty quantification and propagation in computational materials science and simulation-assisted materials design. Integr. Mater. Manuf. Innov. 9, 103–143 (2020).
Matthews, C. et al. Mechanistic multiscale uncertainty propagation in support of accelerated fuel qualification. Nucl. Technol. 0, 1–13 (2025).
Notley, M. J. F. & Hastings, I. J. A microstructure-dependent model for fission product gas release and swelling in UO2 fuel. Nucl. Eng. Des. 56, 163–175 (1980).
Lewis, B. J., Iglesias, F. C., Cox, D. S. & Gheorghiu, E. A model for fission gas release and fuel oxidation behavior for defected UO2 fuel elements. Nucl. Technol. 92, 353–362 (1990).
Lee, B.-H., Koo, Y.-H. & Sohn, D.-S. Modeling of the rim effect on thermal behavior of high-burnup UO2 fuel. Nucl. Technol. 127, 151–159 (1999).
Jernkvist, L. O. A continuum model for cracked UO2 fuel. Nucl. Eng. Des. 176, 273–284 (1997).
Loewen, E. P., Wilson, R. D., Hohorst, J. K. & Kumar, A. S. Preliminary FRAPCON-3Th Steady-state fuel analysis of ThO2 and UO2 fuel mixtures. Nucl. Technol. 136, 261–277 (2001).
Lee, B.-H., Koo, Y.-H., Oh, J.-Y., Cheon, J.-S. & Sohn, D.-S. Improvement of fuel performance code COSMOS with recent in-pile data for MOX and UO2 fuels. Nucl. Technol. 157, 53–64 (2007).
Rossiter, G. Development of the enigma fuel performance code for whole core analysis and dry storage assessments. Nucl. Eng. Technol. 43, 489–498 (2011).
Demarco, G. L. & Marino, A. C. 3D finite elements modelling for design and performance analysis of UO2 pellets. Sci. Technol. Nucl. Install. 2011, 843491 (2011).
Rashid, J. Y. R., Yagnik, S. K. & Montgomery, R. O. Light water reactor fuel performance modeling and multi-dimensional simulation. JOM 63, 81–88 (2011).
Walker, J. & Catlow, C. Structural and dynamic properties of UO2 at high temperatures. J. Phys. C: Solid State Phys. 14, L979 (1981).
Yamada, K., Kurosaki, K., Uno, M. & Yamanaka, S. Evaluation of thermal properties of uranium dioxide by molecular dynamics. J. Alloy. Compd. 307, 10–16 (2000).
Morelon, N.-D., Ghaleb, D., Delaye, J.-M. & Van Brutzel, L. A new empirical potential for simulating the formation of defects and their mobility in uranium dioxide. Philos. Mag. 83, 1533–1555 (2003).
Basak, C. B., Sengupta, A. K. & Kamath, H. S. Classical molecular dynamics simulation of UO2 to predict thermophysical properties. J. Alloy. Compd. 360, 210–216 (2003).
Arima, T., Yamasaki, S., Inagaki, Y. & Idemitsu, K. Evaluation of thermal properties of UO2 and PuO2 by equilibrium molecular dynamics simulations from 300 to 2000 K. J. Alloy. Compd. 400, 43–50 (2005).
Meis, C. & Chartier, A. Calculation of the threshold displacement energies in UO2 using ionic potentials. J. Nucl. Mater. 341, 25–30 (2005).
Skomurski, F. N., Ewing, R. C., Rohl, A. L., Gale, J. D. & Becker, U. Quantum mechanical vs. empirical potential modeling of uranium dioxide (UO2) surfaces: (111), (110), and (100). Am. Mineral. 91, 1761–1772 (2006).
Yakub, E., Ronchi, C. & Staicu, D. Molecular dynamics simulation of premelting and melting phase transitions in stoichiometric uranium dioxide. J. Chem. Phys. 127, 094508 (2007).
Goel, P., Choudhury, N. & Chaplot, S. L. Atomistic modeling of the vibrational and thermodynamic properties of uranium dioxide, UO2. J. Nucl. Mater. 377, 438–443 (2008).
Govers, K., Lemehov, S., Hou, M. & Verwerft, M. Comparison of interatomic potentials for UO2. Part I: Static calculations. J. Nucl. Mater. 366, 161–177 (2007).
Govers, K., Lemehov, S., Hou, M. & Verwerft, M. Comparison of interatomic potentials for UO2. J. Nucl. Mater. 376, 66–77 (2008).
Watanabe, T. et al. Thermal transport properties of uranium dioxide by molecular dynamics simulations. J. Nucl. Mater. 375, 388–396 (2008).
Williamson, R. L. et al. Multidimensional multiphysics simulation of nuclear fuel behavior. J. Nucl. Mater. 423, 149–163 (2012).
Williamson, R. L. et al. BISON: a flexible code for advanced simulation of the performance of multiple nuclear fuel forms. Nucl. Technol. 207, 954–980 (2021).
Tonks, M. R., Zhang, Y., Bai, X. & Millett, P. C. Demonstrating the temperature gradient impact on grain growth in UO2 using the phase field method. Mater. Res. Lett. 2, 23–28 (2014).
Tonks, M. R., Zhang, Y., Butterfield, A. & Bai, X.-M. Development of a grain boundary pinning model that considers particle size distribution using the phase field method. Model. Simul. Mater. Sci. Eng. 23, 045009 (2015).
Tonks, M. R., Simon, P.-C. A. & Hirschhorn, J. Mechanistic grain growth model for fresh and irradiated UO2 nuclear fuel. J. Nucl. Mater. 543, 152576 (2021).
Andersson, D. A., Uberuaga, B. P., Nerikar, P. V., Unal, C. & Stanek, C. R. U and Xe transport in UO2: Density functional theory calculations. Phys. Rev. B 84, 054105 (2011).
Millett, P. C., Zhang, Y., Tonks, M. R. & Biner, S. B. Consideration of grain size distribution in the diffusion of fission gas to grain boundaries. J. Nucl. Mater. 440, 435–439 (2013).
Andersson, D. A. et al. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in : Implications for nuclear fuel performance modeling. J. Nucl. Mater. 451, 225–242 (2014).
Millett, P. C., Tonks, M. R. & Biner, S. B. Grain boundary percolation modeling of fission gas release in oxide fuels. J. Nucl. Mater. 424, 176–182 (2012).
Millett, P. C. et al. Phase-field simulation of intergranular bubble growth and percolation in bicrystals. J. Nucl. Mater. 425, 130–135 (2012).
Aagesen, L. K., Schwen, D., Tonks, M. R. & Zhang, Y. Phase-field modeling of fission gas bubble growth on grain boundaries and triple junctions in UO2 nuclear fuel. Comput. Mater. Sci. 161, 35–45 (2019).
Phillpot, S. R., El-Azab, A., Chernatynskiy, A. & Tulenko, J. S. Thermal conductivity of UO2 fuel: Predicting fuel performance from simulation. JOM 63, 73–79 (2011).
Chockalingam, K., Millett, P. C. & Tonks, M. R. Effects of intergranular gas bubbles on thermal conductivity. J. Nucl. Mater. 430, 166–170 (2012).
Tonks, M. R. et al. Multiscale development of a fission gas thermal conductivity model: Coupling atomic, meso and continuum level simulations. J. Nucl. Mater. 440, 193–200 (2013).
Tonks, M. R. et al. Development of a multiscale thermal conductivity model for fission gas in UO2. J. Nucl. Mater. 469, 89–98 (2016).
Chakraborty, P., Tonks, M. R. & Pastore, G. Modeling the influence of bubble pressure on grain boundary separation and fission gas release. J. Nucl. Mater. 452, 95–101 (2014).
Chakraborty, P., Zhang, Y. & Tonks, M. R. Multi-scale modeling of microstructure dependent intergranular brittle fracture using a quantitative phase-field based method. Comput. Mater. Sci. 113, 38–52 (2016).
Jiang, W., Hu, T., Aagesen, L. K. & Zhang, Y. Three-dimensional phase-field modeling of porosity dependent intergranular fracture in UO2. Comput. Mater. Sci. 171, 109269 (2020).
Neilson, W. D., Galvin, C. O. T., Dillon, S. J., Cooper, M. W. D. & Andersson, D. A. Irradiation-induced creep in UO2: The role of grain boundaries. Phys. Rev. Mater. 8, 103602 (2024).
Galvin, C. O. T., Andersson, D. A., Sweet, R. T., Capolungo, L. & Cooper, M. W. D. Diffusional creep model in UO2 informed by lower-length scale simulations. J. Nucl. Mater. 607, 155659 (2025).
Xiao, H., Long, C. & Chen, H. Model for evolution of grain size in the rim region of high burnup UO2 fuel. J. Nucl. Mater. 471, 74–79 (2016).
Bai, X.-M., Tonks, M. R., Zhang, Y. & Hales, J. D. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels. J. Nucl. Mater. 470, 208–215 (2016).
Veshchunov, M. & Tarasov, V. Modelling of pore coarsening in the high burn-up structure of uo2 fuel. J. Nucl. Mater. 488, 191–195 (2017).
Abdoelatef, M. G. et al. Mesoscale modeling of high burn-up structure formation and evolution in UO2. JOM 71, 4817–4828 (2019).
Barani, T. et al. Modeling high burnup structure in oxide fuels for application to fuel performance codes. part I: High burnup structure formation. J. Nucl. Mater. 539, 152296 (2020).
Brinkley, W. C., Capps, N. & Wirth, B. An initial microstructurally informed model of high burnup structure formation in UO2 Fuel. Nucl. Sci. Eng. 0, 1–16 (2025).
Biswas, S. & Aagesen, L. K. Mesoscale modeling of restructuring in high burnup UO2 fuel. Comput. Mater. Sci. 258, 114052 (2025).
Capps, N., Schappel, D. & Nelson, A. Initial development of an RIA envelope for dispersed nuclear fuel. Ann. Nucl. Energy 148, 107719 (2020).
Khvostov, G. Analytical criteria for fuel fragmentation and burst FGR during a LOCA. Nucl. Eng. Technol. 52, 2402–2409 (2020).
Capps, N., Sweet, R., Harp, J. & Petrie, C. M. High-burnup fuel stress analysis prior to and during a local transient. J. Nucl. Mater. 556, 153194 (2021).
Jiang, W., Hu, T., Aagesen, L. K., Biswas, S. & Gamble, K. A. A phase-field model of quasi-brittle fracture for pressurized cracks: application to UO2 high-burnup microstructure fragmentation. Theor. Appl. Fract. Mech. 119, 103348 (2022).
Zhang, S., Jiang, W., Gamble, K. A. & Tonks, M. R. Comparing the impact of thermal stresses and bubble pressure on intergranular fracture in UO2 using 2D phase field fracture simulations. J. Nucl. Mater. 574, 154158 (2023).
Greenquist, I., Wysocki, A., Hirschhorn, J. & Capps, N. Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility-Part 1: overview and code coupling strategies. Ann. Nucl. Energy 191, 109913 (2023).
Hirschhorn, J., Greenquist, I., Wysocki, A. & Capps, N. Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility–part 2: High-burnup steady-state operating and fuel performance conditions. Ann. Nucl. Energy 192, 109952 (2023).
Gencturk, M. et al. Thermo-mechanical phase-field modeling of fracture in high-burnup UO2 fuels under transient conditions. Materials 18, https://doi.org/10.3390/ma18051162 (2025).
Pastore, G. et al. Modelling of transient fission gas behaviour in oxide fuel and application to the BISON Code. In Enlarged Halden Programme Group Meeting, Røros, Norway (2014).
Capps, N. et al. Empirical and mechanistic transient fission gas release model for high-burnup LOCA conditions. J. Nucl. Mater. 584, 154557 (2023).
Lieou, C. K. C., Capps, N. A., Cooper, M. W. D., Simon, P.-C. A. & Wirth, B. D. An integrated statistical-thermodynamic model for fission gas release and swelling in nuclear fuels. J. Nucl. Mater. 589, 154869 (2024).
ARBORELIUS, J. et al. Advanced Doped UO2 Pellets in LWR Applications. J. Nucl. Sci. Technol. 43, 967–976 (2006).
Cooper, M. W. D., Stanek, C. R. & Andersson, D. A. The role of dopant charge state on defect chemistry and grain growth of doped UO2. Acta Mater. 150, 403–413 (2018).
Owen, M. W. et al. Diffusion in undoped and Cr-doped amorphous UO2. J. Nucl. Mater. 576, 154270 (2023).
Kocevski, V., Cooper, M. W. D. & Andersson, D. A. Modeling of fission gas diffusion and release for doped. J. Nucl. Mater. 584, 154575 (2023).
Greenquist, I., Tonks, M., Cooper, M., Andersson, D. & Zhang, Y. Grand potential sintering simulations of doped UO2 accident-tolerant fuel concepts. J. Nucl. Mater. 532, 152052 (2020).
Gamble, K. A., Pastore, G. & Cooper, M. W. D. BISON Development and Validation for Priority LWR-ATF Concepts. Technical Report. INL/EXT-20-59969-Rev000 (Idaho National Lab. (INL), Idaho Falls, ID (United States) https://doi.org/10.2172/1717844. https://www.osti.gov/biblio/1717844 (2020).
Cooper, M. W. D., Andersson, A. D. R. & Stanek, C. R. Mn-Doped Oxide Nuclear Fuel. Technical Report. 10,847,271 (Los Alamos National Laboratory (LANL), Los Alamos, NM (United States) https://www.osti.gov/biblio/1771664 (2020).
Badry, F., Brito, R., Abdoelatef, M. G., McDeavitt, S. & Ahmed, K. An experimentally validated mesoscale model of thermal conductivity of a UO2 and BeO composite nuclear fuel. JOM 71, 4829–4838 (2019).
Hilty, F. W. & Tonks, M. R. Development and application of a microstructure dependent thermal resistor model for UO2 reactor fuel with high thermal conductivity additives. J. Nucl. Mater. 540, 152334 (2020).
Sweidan, F., Costa, D. R., Liu, H. & Olsson, P. Temperature-dependent thermal conductivity and fuel performance of UN-UO2 and UN-X-UO2 (X=Mo, W) composite nuclear fuels by finite element modeling. J. Mater. 10, 937–946 (2024).
Hilty, F. W., Kim, D.-U. & Tonks, M. R. Impact of fission gas bubbles on thermal conductivity of UO2 fuels with high thermal conductivity additives. J. Nucl. Mater. 546, 152779 (2021).
Cheniour, A. et al. Development of a grain growth model for U3Si2 using experimental data, phase field simulation and molecular dynamics. J. Nucl. Mater. 532, 152069 (2020).
Aagesen, L. K. et al. Phase-field simulations of intergranular fission gas bubble behavior in U3Si2 nuclear fuel. J. Nucl. Mater. 541, 152415 (2020).
Gamble, K. A. et al. Improvement of the BISON U3Si2 modeling capabilities based on multiscale developments to modeling fission gas behavior. J. Nucl. Mater. 555, 153097 (2021).
Andersson, D. A., Stanek, C. R., Matthews, C. & Uberuaga, B. P. The past, present, and future of nuclear fuel. MRS Bull. 48, 1154–1162 (2023).
Ducher, R., Dubourg, R., Barrachin, M. & Pasturel, A. First-principles study of defect behavior in irradiated uranium monocarbide. Phys. Rev. B 83, 104107 (2011).
Freyss, M. First-principles study of uranium carbide: accommodation of point defects and of helium, xenon, and oxygen impurities. Phys. Rev. B 81, 014101 (2010).
Basak, C. B. Classical molecular dynamics simulation of uranium monocarbide (UC). Comput. Mater. Sci. 40, 562–568 (2007).
Chartier, A. & Van Brutzel, L. Modeling of point defects and rare gas incorporation in uranium mono-carbide. Nucl. Instrum. Methods Phys. Res. Sect. B Beam Interact. Mater. 255, 146–150 (2007).
Mankad, V. H. & Jha, P. K. Thermodynamic properties of nuclear material uranium carbide using density functional theory. J. Therm. Anal. Calorim. 124, 11–20 (2016).
Mei, Z.-G., Ye, B., Yacout, A. M., Beeler, B. & Gao, Y. First-principles study of the surface properties of uranium carbides. J. Nucl. Mater. 542, 152257 (2020).
Kocevski, V., Cooper, M. W. D., Claisse, A. J. & Andersson, D. A. Development and application of a uranium mononitride (UN) potential: thermomechanical properties and Xe diffusion. J. Nucl. Mater. 562, 153553 (2022).
AbdulHameed, M., Beeler, B., Galvin, C. O. T. & Cooper, M. W. D. Assessment of uranium nitride interatomic potentials. J. Nucl. Mater. 600, 155247 (2024).
Kocevski, V., Rehn, D. A., Cooper, M. W. D. & Andersson, D. A. First-principles investigation of uranium mononitride (UN): Effect of magnetic ordering, spin-orbit interactions and exchange correlation functional. J. Nucl. Mater. 559, 153401 (2022).
Cooper, M. W. D. et al. Simulations of self- and Xe diffusivity in uranium mononitride including chemistry and irradiation effects. J. Nucl. Mater. 587, 154685 (2023).
Schneider, A. et al. Radiation induced athermal diffusivity in uranium mononitride. J. Nucl. Mater. 601, 155313 (2024).
Kocevski, V. et al. Finite temperature properties of uranium mononitride. J. Nucl. Mater. 576, 154241 (2023).
Kocevski, V. et al. CALPHAD modeling of uranium nitride (UN) fabrication routes enabled by first-principles calculations. Calphad 79, 102463 (2022).
Rizk, J. T. et al. Mechanistic nuclear fuel performance modeling of uranium nitride. J. Nucl. Mater. 606, 155604 (2025).
Billone, M. C., Liu, Y. Y., Gruber, E. E., Hughes, T. H. & Kramer, J. M. Status of Fuel Element Modeling Codes for Metallic Fuels. In Proc. American Nuclear Society International Conference on Reliable Fuels for Liquid Metal Reactors (American Nuclear Society (ANS), Tucson, Arizona, 1968).
Kobayashi, T. et al. Development of the SESAME metallic fuel performance code. Nucl. Technol. 89, 183–193 (1990).
Hwang, W., Nam, C., Byun, T. S. & Kim, Y. C. MACSIS: a metallic fuel performance analysis code for simulating in-reactor behavior under steady-state conditions. Nucl. Technol. 123, 130–141 (1998).
Ogata, T. & Yokoo, T. Development and Validation of ALFUS: an irradiation behavior analysis code for metallic fast reactor fuels. Nucl. Technol. 128, 113–123 (1999).
Galloway, J., Unal, C., Carlson, N., Porter, D. & Hayes, S. Modeling constituent redistribution in U-Pu-Zr metallic fuel using the advanced fuel performance code BISON. Nucl. Eng. Des. 286, 1–17 (2015).
Hou, Y. et al. U-50Zr helical cruciform fuel performance analysis based on MOOSE framework. Ann. Nucl. Energy 204, 110529 (2024).
Chen, X., Xie, Z., Mao, X. & Ding, S. Modeling of zirconium atom redistribution and phase transformation coupling behaviors in U-10Zr-based helical cruciform fuel rods under irradiation. Metals 14, 745 (2024).
Beeler, B. et al. First principles calculations for defects in U. J. Phys. Condens. Matter 22, 505703 (2010).
Huang, G.-Y. & Wirth, B. D. First-principles study of diffusion of interstitial and vacancy in α U-Zr. J. Phys. Condens. Matter 23, 205402 (2011).
Beeler, B., Deo, C., Baskes, M. & Okuniewski, M. Atomistic properties of γ uranium. J. Phys. Condens. Matter 24, 075401 (2012).
Beeler, B., Deo, C., Baskes, M. & Okuniewski, M. Atomistic Investigations of Intrinsic and Extrinsic Point Defects in bcc Uranium. In Atomistic Investigations of Intrinsic and Extrinsic Point Defects in bcc Uranium (2013).
Xiong, W., Xie, W., Shen, C. & Morgan, D. Thermodynamic modeling of the U-Zr system—a revisit. J. Nucl. Mater. 443, 331–341 (2013).
Moore, A. P., Beeler, B., Deo, C., Baskes, M. I. & Okuniewski, M. A. Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics. J. Nucl. Mater. 467, 802–819 (2015).
Moore, A. P., Deo, C., Baskes, M. I. & Okuniewski, M. A. Atomistic mechanisms of morphological evolution and segregation in U-Zr alloys. Acta Mater. 115, 178–188 (2016).
Hirschhorn, J., Tonks, M., Aitkaliyeva, A. & Adkins, C. Development and verification of a phase-field model for the equilibrium thermodynamics of U-Pu-Zr. Ann. Nucl. Energy 124, 490–502 (2019).
Hirschhorn, J., Aitkaliyeva, A., Adkins, C. & Tonks, M. The microstructure and thermodynamic behavior of as-cast U-24Pu-15Zr: unexpected results and recommendations for U-Pu-Zr fuel research methodology. J. Nucl. Mater. 518, 80–94 (2019).
Hirschhorn, J., Tonks, M., Aitkaliyeva, A. & Adkins, C. Reexamination of a U-Zr diffusion couple experiment using quantitative phase-field modeling and sensitivity analysis. J. Nucl. Mater. 529, 151929 (2020).
Hirschhorn, J., Tonks, M. R., Aitkaliyeva, A. & Adkins, C. A study of constituent redistribution in U-Zr fuels using quantitative phase-field modeling and sensitivity analysis. J. Nucl. Mater. 523, 143–156 (2019).
Hirschhorn, J., Tonks, M. & Matthews, C. A CALPHAD-informed approach to modeling constituent redistribution in Zr-based metallic fuels using BISON. J. Nucl. Mater. 544, 152657 (2020).
Jung, W., Kim, J.-S. & Chang, K. Investigating constituent redistribution in U-Zr metallic fuels: A phase-field approach incorporating porosity and sodium infiltration. Ann. Nucl. Energy 224, 111679 (2025).
Zhou, S. et al. Combined ab initio and empirical model of the thermal conductivity of uranium, uranium-zirconium, and uranium-molybdenum. Phys. Rev. Mater. 2, 083401 (2018).
Ortega, L. H., Blamer, B., Stern, K. M., Vollmer, J. & McDeavitt, S. M. Thermal conductivity of uranium metal and uranium-zirconium alloys fabricated via powder metallurgy. J. Nucl. Mater. 531, 151982 (2020).
Zhou, S., Zhang, Y. & Morgan, D. An ab-initio based semi-empirical thermal conductivity model for multiphase uranium-zirconium alloys. J. Nucl. Mater. 553, 153044 (2021).
Chen, W. & Bai, X.-M. Mesoscale modeling of microstructure-dependent thermal conductivity in U-Zr fuels. J. Nucl. Mater. 562, 153593 (2022).
Huang, G.-Y. & Wirth, B. D. First-principles study of bubble nucleation and growth behaviors in $\char"03B1$ U-Zr. J. Phys. Condens. Matter 24, 415404 (2012).
Smirnova, D. E. et al. A ternary EAM interatomic potential for U-Mo alloys with xenon. Model. Simul. Mater. Sci. Eng. 21, 035011 (2013).
Smirnova, D. E., Kuksin, A. Y., Starikov, S. V. & Stegailov, V. V. Atomistic modeling of the self-diffusion in γ-U and γ-U-Mo. Phys. Met. Metallogr. 116, 445–455 (2015).
Smirnova, D. E., Kuksin, A. Y. & Starikov, S. V. Investigation of point defects diffusion in bcc uranium and U-Mo alloys. J. Nucl. Mater. 458, 304–311 (2015).
Kolotova, L. N., Starikov, S. V. & Ozrin, V. D. Atomistic simulation of the fission-fragment-induced formation of defects in a uranium-molybdenum alloy. J. Exp. Theor. Phys. 129, 59–65 (2019).
Xu, Z. et al. Modeling the homogenization kinetics of as-cast U-10wt% Mo alloys. J. Nucl. Mater. 471, 154–164 (2016).
Wang, X. et al. Modeling early-stage processes of U-10 Wt.%Mo alloy using integrated computational materials engineering concepts. JOM 69, 2532–2537 (2017).
Chakraborty, S., Choudhuri, G., Somayajulu, P. S., Agarwal, R. & Khan, K. B. Microstructure characterization and phase field analysis of dendritic crystal growth of γ-U and BCC-Mo dendrite in U-33 at.% Mo fast reactor fuel. J. Mater. Res. 33, 225–238 (2018).
Lu, Y. et al. A phase-field study of spinodal decomposition impeded by irradiation in U-Mo and U-Mo-Zr alloys. Materials 16, 7546 (2023).
Liang, L., Mei, Z.-G., Soo Kim, Y., Anitescu, M. & Yacout, A. M. Three-dimensional phase-field simulations of intragranular gas bubble evolution in irradiated U-Mo fuel. Comput. Mater. Sci. 145, 86–95 (2018).
Liang, L., Kim, Y. S., Mei, Z.-G., Aagesen, L. K. & Yacout, A. M. Fission gas bubbles and recrystallization-induced degradation of the effective thermal conductivity in U-7Mo fuels. J. Nucl. Mater. 511, 438–445 (2018).
Hu, S., Setyawan, W., Beeler, B. W., Gan, J. & Burkes, D. E. Defect cluster and nonequilibrium gas bubble associated growth in irradiated UMo fuels—a cluster dynamics and phase field model. J. Nucl. Mater. 542, 152441 (2020).
Lan, X., Jiang, Y., La, Y. & Liu, W. Calculation of the effective thermal conductivity of porous polycrystalline U-10Mo alloy based on phase-field simulation. Nucl. Eng. Des. 413, 112552 (2023).
Jiang, Y., Gao, S., La, Y. & Liu, W. Formation of gas bubble superlattice in U-Mo alloys: A phase-field study. Nucl. Eng. Des. 435, 113912 (2025).
Beeler, B., Hu, S., Zhang, Y. & Gao, Y. A improved equation of state for Xe gas bubbles in γ U-Mo fuels. J. Nucl. Mater. 530, 151961 (2020).
Beeler, B., Casagranda, A., Aagesen, L., Zhang, Y. & Novascone, S. Atomistic calculations of the surface energy as a function of composition and temperature in γ U-Zr to inform fuel performance modeling. J. Nucl. Mater. 540, 152271 (2020).
Beeler, B., Cooper, M. W. D., Mei, Z.-G., Schwen, D. & Zhang, Y. Radiation driven diffusion in γU-Mo. J. Nucl. Mater. 543, 152568 (2021).
Iasir, A. R. M. & Hammond, K. D. Xenon mobility in γ-uranium and uranium-molybdenum alloys. J. Appl. Phys. 131, 025105 (2022).
Potter, P. E. Over forty years of ‘Thermodynamics of Nuclear Materials’. J. Nucl. Mater. 389, 29–44 (2009).
Besmann, T. M., McMurray, J. W. & Simunovic, S. Application of thermochemical modeling to assessment/evaluation of nuclear fuel behavior. Calphad 55, 47–51 (2016).
Matthews, C., Perriot, R., Cooper, M. W. D., Stanek, C. R. & Andersson, D. A. Cluster dynamics simulation of uranium self-diffusion during irradiation in UO2. J. Nucl. Mater. 527, 151787 (2019).
Matthews, C., Perriot, R., Cooper, M. W. D., Stanek, C. R. & Andersson, D. A. Cluster dynamics simulation of xenon diffusion during irradiation in UO2. J. Nucl. Mater. 540, 152326 (2020).
Liu, X. Y. et al. Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels. J. Nucl. Mater. 561, 153539 (2022).
McMurray, J. W. & Besmann, T. M. Thermodynamic Modeling of Nuclear Fuel Materials. In Handbook of Materials Modeling 1–29 (Springer, 2018).
Corcoran, E. C., Kaye, M. H. & Piro, M. H. A. An overview of thermochemical modelling of CANDU fuel and applications to the nuclear industry. Calphad 55, 52–62 (2016).
Piro, M. H. A., Besmann, T. M., Simunovic, S., Lewis, B. J. & Thompson, W. T. Numerical verification of equilibrium thermodynamic computations in nuclear fuel performance codes. J. Nucl. Mater. 414, 399–407 (2011).
Piro, M. H. A. Thermodynamically informed nuclear fuel codes—a review and perspectives. Thermo 1, 262–285 (2021).
Guéneau, C. et al. TAF-ID: an international thermodynamic database for nuclear fuels applications. Calphad 72, 102212 (2021).
Wooding, S. J. & Bacon, D. J. A molecular dynamics study of displacement cascades in α-zirconium. Philos. Mag. A 76, 1033–1051 (1997).
Wooding, S. J., Howe, L. M., Gao, F., Calder, A. F. & Bacon, D. J. A molecular dynamics study of high-energy displacement cascades in α-zirconium. J. Nucl. Mater. 254, 191–204 (1998).
Rossi, M. L. & Taylor, C. D. Atomistic simulations of formation of elementary Zr-I systems. Open J. Phys. Chem. 1, 104–108 (2011).
Rossi, M. L., Taylor, C. D. & van Duin, A. C. Reduced yield stress for zirconium exposed to iodine: reactive force field simulation. Adv. Model. Simul. Eng. Sci. 1, 19 (2014).
Rossi, M. L. & Taylor, C. D. First-principles insights into the nature of zirconium-iodine interactions and the initiation of iodine-induced stress-corrosion cracking. J. Nucl. Mater. 458, 1–10 (2015).
Taylor, C. D. & Rossi, M. L. Multiphysics modeling of the role of iodine in environmentally assisted cracking of zirconium via pellet-clad interaction. Corrosion 72, 978–988 (2016).
Christensen, M. et al. Diffusion of point defects, nucleation of dislocation loops, and effect of hydrogen in hcp-Zr: ab initio and classical simulations. J. Nucl. Mater. 460, 82–96 (2015).
Arjhangmehr, A. & Feghhi, S.aH. Irradiation deformation near different atomic grain boundaries in α-Zr: an investigation of thermodynamics and kinetics of point defects. Sci. Rep. 6, 23333 (2016).
Sahi, Q. -u-a & Kim, Y.-S. Molecular dynamics simulations of the coupled effects of strain and temperature on displacement cascades in α-zirconium. Nucl. Eng. Technol. 50, 907–914 (2018).
Christiaen, B., Domain, C., Thuinet, L., Ambard, A. & Legris, A. A new scenario for <c> vacancy loop formation in zirconium based on atomic-scale modeling. Acta Mater. 179, 93–106 (2019).
Maxwell, C., Pencer, J. & Torres, E. Atomistic simulation study of clustering and evolution of irradiation-induced defects in zirconium. J. Nucl. Mater. 531, 151979 (2020).
Sakaël, C., Domain, C., Ambard, A., Thuinet, L. & Legris, A. Modelling of zirconium growth under irradiation and annealing conditions. Int. J. Plast. 168, 103699 (2023).
Ghorbani, A., Luo, Y., Saidi, P. & Béland, L. K. Anisotropic diffusion of radiation-induced self-interstitial clusters in HCP zirconium: a molecular dynamics and rate-theory assessment. Scr. Mater. 238, 115755 (2024).
Jia, X., Bao, Y., Cao, S., Su, Y. & Qian, P. Temperature effects on radiation damage in HCP-zirconium: a molecular dynamics study using a fine-tuned machine-learned potential. J. Nucl. Mater. 616, 156025 (2025).
Bai, X.-M., Zhang, Y. & R. Tonks, M. Strain effects on oxygen transport in tetragonal zirconium dioxide. Phys. Chem. Chem. Phys. 15, 19438–19449 (2013).
Siripurapu, R. K., Szpunar, B. & Szpunar, J. A. Molecular dynamics study of hydrogen in α-zirconium. Int. J. Nucl. Energy 2014, 912369 (2014).
Glazoff, M. V., Tokuhiro, A., Rashkeev, S. N. & Sabharwall, P. Oxidation and hydrogen uptake in zirconium, Zircaloy-2 and Zircaloy-4: computational thermodynamics and ab initio calculations. J. Nucl. Mater. 444, 65–75 (2014).
Chakraborty, P., Moitra, A. & Saha-Dasgupta, T. Effect of hydrogen on degradation mechanism of zirconium: a molecular dynamics study. J. Nucl. Mater. 466, 172–178 (2015).
Zhang, Y. et al. Homogeneous hydride formation path in α-Zr: Molecular dynamics simulations with the charge-optimized many-body potential. Acta Mater. 111, 357–365 (2016).
Wimmer, E. et al. Hydrogen in zirconium: atomistic simulations of diffusion and interaction with defects using a new embedded atom method potential. J. Nucl. Mater. 532, 152055 (2020).
Brimbal, D., Chaari, N., Barberis, P. & Bourlier, F. Zirconium-applied anisotropic cluster dynamics for irradiation-induced defect modeling in presence of hydrogen. In Zirconium in the Nuclear Industry: 19th International Symposium Vol. STP1622-EB, (eds Motta, A. T. & Yagnik, S. K.) (ASTM International, 2021).
Yu, F., Xiang, X., Zu, X. & Hu, S. Hydrogen diffusion in zirconium hydrides from on-the-fly machine learning molecular dynamics. Int. J. Hydrog. Energy 56, 1057–1066 (2024).
Mamivand, M., Asle Zaeem, M., El Kadiri, H. & Chen, L.-Q. Phase field modeling of the tetragonal-to-monoclinic phase transformation in zirconia. Acta Mater. 61, 5223–5235 (2013).
Asle Zaeem, M. & El Kadiri, H. An elastic phase field model for thermal oxidation of metals: Application to zirconia. Comput. Mater. Sci. 89, 122–129 (2014).
Dykhuis, A. F. & Short, M. P. Phase field modeling of irradiation-enhanced corrosion of Zircaloy-4 in PWRs. Corros. Sci. 146, 179–191 (2019).
Lin, C., Ruan, H. & Shi, S.-Q. Mechanical-chemical coupling phase-field modeling for inhomogeneous oxidation of zirconium induced by stress-oxidation interaction. npj Mater. Degrad. 4, 22 (2020).
Xin, T., Kharchenko, D. O., Kharchenko, V. O. & Lu, W. Phase field modeling of sub-parabolic oxidation kinetics and in-plane stress in zircaloy under neutron/proton irradiation. Phys. Scr. 100, 035941 (2025).
Ma, X. Q., Shi, S. Q., Woo, C. H. & Chen, L. Q. Simulation of γ-hydride precipitation in bi-crystalline zirconium under uniformly applied load. Mater. Sci. Eng.: A 334, 6–10 (2002).
Ma, X. Q., Shi, S. Q., Woo, C. H. & Chen, L. Q. Effect of applied load on nucleation and growth of γ-hydrides in zirconium. Comput. Mater. Sci. 23, 283–290 (2002).
Ma, X. Q., Shi, S. Q., Woo, C. H. & Chen, L. Q. Phase-field simulation of hydride precipitation in bi-crystalline zirconium. Scr. Mater. 47, 237–241 (2002).
Guo, X., Shi, S.-Q., Zhang, Q. & Ma, X. An elastoplastic phase-field model for the evolution of hydride precipitation in zirconium. Part I: smooth specimen. J. Nucl. Mater. 378, 110–119 (2008).
Guo, X. H., Shi, S. Q., Zhang, Q. M. & Ma, X. Q. An elastoplastic phase-field model for the evolution of hydride precipitation in zirconium. Part II: specimen with flaws. J. Nucl. Mater. 378, 120–125 (2008).
Zhao, Z. et al. Characterization of Zirconium Hydrides and Phase Field Approach to a Mesoscopic-Scale Modeling of Their Precipitation. In Zirconium in the Nuclear Industry: 15th International Symposium (eds Kammenzind, B. & Limbäck, M.) 29–29–22 (ASTM International, 2009).
Thuinet, L., De Backer, A. & Legris, A. Phase-field modeling of precipitate evolution dynamics in elastically inhomogeneous low-symmetry systems: application to hydride precipitation in Zr. Acta Mater. 60, 5311–5321 (2012).
Thuinet, L., Legris, A., Zhang, L. & Ambard, A. Mesoscale modeling of coherent zirconium hydride precipitation under an applied stress. J. Nucl. Mater. 438, 32–40 (2013).
Shi, S.-Q. & Xiao, Z. A quantitative phase field model for hydride precipitation in zirconium alloys: Part I. Development of quantitative free energy functional. J. Nucl. Mater. 459, 323–329 (2015).
Xiao, Z., Hao, M., Guo, X., Tang, G. & Shi, S.-Q. A quantitative phase field model for hydride precipitation in zirconium alloys: Part II. Modeling of temperature dependent hydride precipitation. J. Nucl. Mater. 459, 330–338 (2015).
Jokisaari, A. M. & Thornton, K. General method for incorporating CALPHAD free energies of mixing into phase field models: application to the α-zirconium/δ-hydride system. Calphad 51, 334–343 (2015).
Bair, J., Zaeem, M. A. & Tonks, M. A phase-field model to study the effects of temperature change on shape evolution of$\char"x03B3$-hydrides in zirconium. J. Phys. D Appl. Phys. 49, 405302 (2016).
Bair, J., Asle Zaeem, M. & Schwen, D. Formation path of δ hydrides in zirconium by multiphase field modeling. Acta Mater. 123, 235–244 (2017).
Simon, P. C. A. et al. Investigation of δ zirconium hydride morphology in a single crystal using quantitative phase field simulations supported by experiments. J. Nucl. Mater. 557, 153303 (2021).
Lin, J. -l & Heuser, B. J. Modeling hydrogen solvus in zirconium solution by the mesoscale phase-field modeling code Hyrax. Comput. Mater. Sci. 156, 224–231 (2019).
Welland, M. J. & Hanlon, S. M. Prediction of the zirconium hydride precipitation barrier with an anisotropic 3D phase-field model incorporating bulk thermodynamics and elasticity. Comput. Mater. Sci. 171, 109266 (2020).
Shin, W. & Chang, K. Phase-field modeling of hydride reorientation in zirconium cladding materials under applied stress. Comput. Mater. Sci. 182, 109775 (2020).
Toghraee, A., Bair, J. & Asle Zaeem, M. Effects of applied load on formation and reorientation of zirconium hydrides: a multiphase field modeling study. Comput. Mater. Sci. 192, 110367 (2021).
Wu, S. et al. Phase-field model of hydride blister growth kinetics on zirconium surface. Front. Mater. 9, https://doi.org/10.3389/fmats.2022.916593 (2022).
Wei, M. et al. A study of δ-hydride precipitation behavior in Zr alloys by phase-field method. J. Mater. Res. Technol. 31, 2618–2628 (2024).
Patel, M., Reali, L., Sutton, A. P., Balint, D. S. & Wenman, M. R. A fast efficient multi-scale approach to modelling the development of hydride microstructures in zirconium alloys. Comput. Mater. Sci. 190, 110279 (2021).
Simon, P.-C., Chen, L.-Q., Daymond, M., Tonks, M. & Motta, A. Mechanisms of Mesoscale Hydride Morphology and Reorientation in a Polycrystal Investigated Using Phase-Field Modeling. In Zirconium in the Nuclear Industry: 20th International Symposium 807–830 (ASTM International, 2023).
Couet, A., Motta, A. T. & Ambard, A. The coupled current charge compensation model for zirconium alloy fuel cladding oxidation: I. Parabolic oxidation of zirconium alloys. Corros. Sci. 100, 73–84 (2015).
Courty, O., Motta, A. T. & Hales, J. D. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding. J. Nucl. Mater. 452, 311–320 (2014).
Lebensohn, R. A. & Tomé, C. A self-consistent anisotropic approach for the simulation of plastic deformation and texture development of polycrystals: application to zirconium alloys. Acta Metall. Mater. 41, 2611–2624 (1993).
Onimus, F. & Béchade, J.-L. A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys. J. Nucl. Mater. 384, 163–174 (2009).
Erinosho, T. O. & Dunne, F. P. E. Strain localization and failure in irradiated zircaloy with crystal plasticity. Int. J. Plast. 71, 170–194 (2015).
Ahn, D.-H., Lee, G.-G., Moon, J., Kim, H. S. & Chun, Y.-B. Analysis of texture and grain shape effects on the yield anisotropy of Zr-2.5wt%Nb pressure tube alloy using crystal plasticity finite element method. J. Nucl. Mater. 555, 153112 (2021).
Jose, N. M., Samal, M. K., Durgaprasad, P. V., Alankar, A. & Dutta, B. K. Crystal Plasticity Modelling of Neutron Irradiation Effects on the Flow and Damage Behaviour of Zircaloy-4. In Advances in Structural Integrity (eds Jonnalagadda, K., Alankar, A., Balila, N. J. & Bhandakkar, T.) 255-265 (Springer, 2022).
Hardie, C., Thomas, R., Liu, Y., Frankel, P. & Dunne, F. Simulation of crystal plasticity in irradiated metals: A case study on Zircaloy-4. Acta Mater. 241, 118361 (2022).
Frydrych, K. Modelling irradiation effects in metallic materials using the crystal plasticity theory—a review. Crystals 13, 771 (2023).
Zan, X. D. et al. Anisotropic deformation mechanisms of rolling-textured Zircaloy-4 alloy by a crystal plasticity model. Comput. Mater. Sci. 229, 112424 (2023).
Patra, A., Tomé, C. N. & Golubov, S. I. Crystal plasticity modeling of irradiation growth in Zircaloy-2. Philos. Mag. 97, 2018–2051 (2017).
Ahn, D.-H., Lee, G.-G., Chun, Y.-B. & Jung, J. Y. Prediction of the in-reactor deformation of Zr-2.5wt%Nb pressure tubes using the crystal plasticity finite element method framework. J. Nucl. Mater. 570, 153947 (2022).
Roy, R., Long, F. & Daymond, M. R. Evaluation of deformation fields associated with irradiation-induced growth and grain boundary interactions in zirconium. Materialia 39, 102325 (2025).
Dupin, N. et al. A thermodynamic database for zirconium alloys. J. Nucl. Mater. 275, 287–295 (1999).
Qi, F. et al. Pellet-cladding mechanical interaction analysis of Cr-coated Zircaloy cladding. Nucl. Eng. Des. 367, 110792 (2020).
Ma, Z., Shirvan, K., Wu, Y. & Su, G. H. Numerical investigation of ballooning and burst for chromium coated zircaloy cladding. Nucl. Eng. Des. 383, 111420 (2021).
Wei, J., Xu, Z., Li, J., Liu, Y. & Wang, B. Crack safety analysis of coating with plastic behavior in surface-coated Zircaloy cladding. Eng. Fract. Mech. 280, 109134 (2023).
Capps, N., Mai, A., Kennard, M. & Liu, W. PCI analysis of Zircaloy coated clad under LWR steady state and reactor startup operations using BISON fuel performance code. Nucl. Eng. Des. 332, 383–391 (2018).
Dunbar, C. et al. Fuel performance analysis of Cr-coated Zircaloy-4 cladding during a prototypical LOCA event using BISON. Ann. Nucl. Energy 200, 110411 (2024).
Aragón, P., Feria, F., Herranz, L. E., Schubert, A. & Van Uffelen, P. Fuel performance modelling of Cr-coated Zircaloy cladding under DBA/LOCA conditions. Ann. Nucl. Energy 211, 110950 (2025).
Yang, L., Liu, R., Qiu, C. & Liu, S. Multiphysics analysis of thermal mechanical behavior and tritium migration in FeCrAl and Cr-coated Zircaloy cladding under PWR normal operating and transient conditions. Prog. Nucl. Energy 166, 104942 (2023).
Kim, D. & Lee, Y. Diffusion of chromium of Cr-coated Zircaloy accident tolerant fuel cladding: Model development and experimental validation. Surf. Coat. Technol. 468, 129698 (2023).
Zhang, W. et al. A crystal plasticity finite element-based model for predicting the fatigue life of Cr-coated Zr4 alloy. Int. J. Fatigue 200, 109091 (2025).
Gamble, K. A. et al. An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions. J. Nucl. Mater. 491, 55–66 (2017).
Sweet, R. T., George, N. M., Maldonado, G. I., Terrani, K. A. & Wirth, B. D. Fuel performance simulation of iron-chrome-aluminum (FeCrAl) cladding during steady-state LWR operation. Nucl. Eng. Des. 328, 10–26 (2018).
Liu, R., Zhou, W. & Cai, J. Multiphysics modeling of accident tolerant fuel-cladding U3Si2-FeCrAl performance in a light water reactor. Nucl. Eng. Des. 330, 106–116 (2018).
Chen, P. et al. A comparative study of in-pile behaviors of FeCrAl cladding under normal and accident conditions with updated FROBA-ATF code. Nucl. Eng. Des. 371, 110889 (2021).
Aragón, P., Feria, F. & Herranz, L. E. Modelling FeCrAl cladding thermo-mechanical performance. Part I: Steady-state conditions. Prog. Nucl. Energy 153, 104417 (2022).
Aragón, P., Feria, F. & Herranz, L. E. Modelling FeCrAl cladding thermo-mechanical performance. Part II: comparative analysis with Zircaloy under LOCA conditions. Prog. Nucl. Energy 163, 104838 (2023).
Lee, S. K. et al. BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions. J. Nucl. Mater. 564, 153676 (2022).
Deng, C. et al. Finite element based fuel performance investigation of U3Si2-FeCrAl design under normal and RIA conditions. Prog. Nucl. Energy 149, 104265 (2022).
Chang, K. et al. Theory-guided bottom-up design of the FeCrAl alloys as accident tolerant fuel cladding materials. J. Nucl. Mater. 516, 63–72 (2019).
Bigdeli, S. et al. Strategies for High-Temperature Corrosion Simulations of Fe-Based Alloys Using the Calphad Approach: Part I. J. Phase Equilib. Diffus. 42, 403–418 (2021).
Roy, I. et al. Optimizing chemistry for designing oxidation resistant FeCrAl alloys. MRS Adv. 8, 21–26 (2023).
Roy, I. et al. Data-driven predictive modeling of FeCrAl oxidation. Mater. Lett.: X 17, 100183 (2023).
Ye, T. et al. Primary radiation damage characteristics in displacement cascades of FeCrAl alloys. J. Nucl. Mater. 549, 152909 (2021).
Yao, H. et al. Atomistic observation of defect generation and microstructural evolution in polycrystalline FeCrAl alloys under different irradiation conditions. Nanomaterials 15, 988 (2025).
Zhang, J. & Ding, S. Mesoscale simulation research on the homogenized elasto-plastic behavior of FeCrAl alloys. Mater. Today Commun. 22, 100718 (2020).
Kumagai, T., Pachaury, Y., Maccione, R., Wharry, J. & El-Azab, A. An atomistic investigation of dislocation velocity in body-centered cubic FeCrAl alloys. Materialia 18, 101165 (2021).
Yao, H. et al. Atomic-scale investigation of creep behavior and deformation mechanism in nanocrystalline FeCrAl alloys. Mater. Des. 206, 109766 (2021).
Dai, H. et al. Effect of Cr and Al on elastic constants of FeCrAl alloys investigated by molecular dynamics method. Metals 12, 558 (2022).
Yao, H. et al. Structural evolution and transitions of mechanisms in creep deformation of nanocrystalline FeCrAl alloys. Nanomaterials 13, 631 (2023).
Pachaury, Y., Warren, G., Wharry, J. P., Po, G. & El-Azab, A. Plasticity in irradiated FeCrAl nanopillars investigated using discrete dislocation dynamics. Int. J. Plast. 167, 103676 (2023).
Li, C., Chen, S., Du, S., Yu, J. & Zhang, Y. Development of constitutive relationship for thermomechanical processing of FeCrAl alloy to predict hot deformation behavior. Materials 18, 3007 (2025).
Wei, Q. et al. Crystal structures, mechanical properties, and electronic structure analysis of ternary FeCrAl alloys. Phys. Lett. A 533, 130228 (2025).
Massey, C. P. et al. Microstructure dependent burst behavior of oxide dispersion-strengthened FeCrAl cladding. Mater. Des. 234, 112307 (2023).
Sweet, R. T., Pastore, G. & Wirth, B. D. Analysis of FeCrAl cladding performance under loss-of-coolant accident conditions. Nucl. Eng. Des. 414, 112556 (2023).
Sweet, R. T., Massey, C. P., Hirschhorn, J. A., Bell, S. B. & Kane, K. A. Wrought FeCrAl alloy (C26M) cladding behavior and burst under simulated loss-of-coolant accident conditions. Nucl. Eng. Des. 431, 113712 (2025).
Ravi, S. K. et al. Elucidating precipitation in FeCrAl alloys through explainable AI: a case study. Comput. Mater. Sci. 230, 112440 (2023).
Wang, J., Zhang, H., Huang, W. & Lu, Z. Effects of aluminum diffusion on the oxide of the FeCrAl alloys surface: a first-principles study. Mater. Today Commun. 33, 104594 (2022).
Wang, J., Yan, K., Huang, W. & Lu, Z. Mechanisms of Al2O3 and Cr2O3 formation during FeCrAl alloy oxidation: a first-principles study. Appl. Surf. Sci. 644, 158782 (2024).
Ni, Y. et al. A first-principle study of the effect of yttrium on the oxidation resistance of FeCrAl alloys. Mater. Technol. 40, 2502957 (2025).
Kharchenko, D. O. et al. Modeling phase separation and composition patterning in FeCrAl alloys at neutron irradiation. Phys. Scr. 99, 075921 (2024).
Deck, C. P., Khalifa, H. E., Sammuli, B., Hilsabeck, T. & Back, C. A. Fabrication of SiC-SiC composites for fuel cladding in advanced reactor designs. Prog. Nucl. Energy 57, 38–45 (2012).
Stempien, J. D., Carpenter, D. M., Kohse, G. & Kazimi, M. S. Characteristics of composite silicon carbide fuel cladding after irradiation under simulated PWR conditions. Nucl. Technol. 183, 13–29 (2013).
Stone, J. G. et al. Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding. J. Nucl. Mater. 466, 682–697 (2015).
Angelici Avincola, V., Guenoun, P. & Shirvan, K. Mechanical performance of SiC three-layer cladding in PWRs. Nucl. Eng. Des. 310, 280–294 (2016).
Singh, G. et al. Parametric evaluation of SiC/SiC composite cladding with UO2 fuel for LWR applications: fuel rod interactions and impact of nonuniform power profile in fuel rod. J. Nucl. Mater. 499, 155–167 (2018).
Li, W. & Shirvan, K. ABAQUS analysis of the SiC cladding fuel rod behavior under PWR normal operation conditions. J. Nucl. Mater. 515, 14–27 (2019).
Wang, H., Liu, S., Li, Y., Li, W. & Wu, J. Modeling of the thermomechanical behavior of braided SiCf/SiC composite cladding tube during irradiation. J. Nucl. Mater. 608, 155723 (2025).
Lee, Y., No, H. C. & Lee, J. I. Design optimization of multi-layer Silicon Carbide cladding for light water reactors. Nucl. Eng. Des. 311, 213–223 (2017).
Cozzo, C. & Rahman, S. SiC cladding thermal conductivity requirements for normal operation and LOCA conditions. Prog. Nucl. Energy 106, 278–283 (2018).
Alabdullah, M. & Ghoniem, N. M. Damage mechanics modeling of the non-linear behavior of SiC/SiC ceramic matrix composite fuel cladding. J. Nucl. Mater. 524, 296–311 (2019).
Spilker, K., Lebensohn, R. A., Jacobsen, G. & Capolungo, L. A mean field homogenization model for the mechanical response of ceramic matrix composites. Compos. Struct. 352, 118630 (2025).
Maximenko, A., Izhvanov, O. & Olevsky, E. A. Modeling of fuel-cladding stresses in porous UC/SIC fuel pins. Nucl. Eng. Des. 359, 110455 (2020).
Kocevski, V., Lopes, D. A., Claisse, A. J. & Besmann, T. M. Understanding the interface interaction between U3Si2 fuel and SiC cladding. Nat. Commun. 11, 2621 (2020).
Bonny, G., Buongiorno, L., Bakaev, A. & Castin, N. Models and regressions to describe primary damage in silicon carbide. Sci. Rep. 10, 10483 (2020).
Yin, C. et al. A multi-scale simulation study of irradiation swelling of silicon carbide. Materials 15, 3008 (2022).
Pan, C. et al. Atomistic simulation of brittle-to-ductile transition in silicon carbide embedded with nano-sized helium bubbles. J. Phys. D: Appl. Phys. 56, 485301 (2023).
Rabiee, H., Hassanzadeh, A., Sakhaeinia, H. & Alahyarizadeh, G. Effect of the point defect of silicon carbide cladding on mechanical properties: a molecular-dynamics study. Chem. Pap. 78, 3815–3830 (2024).
Kobayashi, K. & Alam, S. B. Physics-regularized neural networks for predictive modeling of silicon carbide swelling with limited experimental data. Sci. Rep. 14, 30666 (2024).
Feng, Y. et al. Multiscale modeling of SiCf/SiC nuclear fuel cladding based on FE-simulation of braiding process. Front. Mater. 7, https://doi.org/10.3389/fmats.2020.634112 (2021).
Yan, Z. & Fan, X. Study on mechanical properties of SiCf/SiC nuclear cladding tube based on braiding process simulation and multi-scale FE analysis. J. Ceram. 45, 191–198 (2024).
Hu, C., Labuz, J. F., Koyanagi, T. & Le, J.-L. Mechanistic modeling of lifetime distribution of SiC/SiC composite claddings. J. Am. Ceram. Soc. 106, 3066–3077 (2023).
Xiao, C. et al. Study on frictional behavior of SiCf/SiC composite clad tube clamping condition under nuclear irradiation. Friction 12, 919–938 (2024).
Singh, G., Yu, J., Xu, F., Yao, T. & Xu, P. Multiscale modeling of silicon carbide cladding for nuclear applications: thermal performance modeling. Energies 17, 6124 (2024).
Jacobsen, G. M. et al. Multiscale modeling of the mechanical response of silicon carbide composite within the accelerated fuel qualification framework. Nucl. Technol. 0, 1–19 (2025).
Xue, J., Wang, S., Chen, Z. & Yang, Z. A multiscale model for the thermomechanical behavior of SiC composite cladding subjected to thermo-mechanical irradiation coupling. J. Nucl. Mater. 615, 155948 (2025).
Boltax, A., Murray, P. & Biancheria, A. Fast reactor fuel performance model development. Nucl. Appl. Technol. 9, 326–337 (1970).
Kramer, J. M. & Dimelfi, R. J. Modeling deformation and failure of fast reactor cladding during simulated accident transients. Nucl. Eng. Des. 63, 47–54 (1981).
Kramer, J. M., Liu, Y. Y., Billone, M. C. & Tsai, H. C. Modeling the behavior of metallic fast reactor fuels during extended transients. J. Nucl. Mater. 204, 203–211 (1993).
Sobolev, V., Malambu, E. & Abderrahim, H. A. Design of a fuel element for a lead-cooled fast reactor. J. Nucl. Mater. 385, 392–399 (2009).
Karahan, A. Modelling of thermo-mechanical and irradiation behavior of metallic and oxide fuels for sodium fast reactors. Thesis, Massachusetts Institute of Technology. http://dspace.mit.edu/handle/1721.1/57693 (2009).
Karahan, A. & Buongiorno, J. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors. J. Nucl. Mater. 396, 283–293 (2010).
Karahan, A. & Buongiorno, J. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors. J. Nucl. Mater. 396, 272–282 (2010).
Miao, Y. et al. Metallic fuel cladding degradation model development and evaluation for BISON. Nucl. Eng. Des. 385, 111531 (2021).
Ryu, H. J., Kim, Y. S. & Yacout, A. M. Thermal creep modeling of HT9 steel for fast reactor applications. J. Nucl. Mater. 409, 207–213 (2011).
Clausen, B., Brown, D. W., Bourke, M. A. M., Saleh, T. A. & Maloy, S. A. In situ neutron diffraction and Elastic-Plastic Self-Consistent polycrystal modeling of HT-9. J. Nucl. Mater. 425, 228–232 (2012).
Tallman, A. E. et al. Data-driven constitutive model for the inelastic response of metals: Application to 316h steel. Integr. Mater. Manuf. Innov. 9, 339–357 (2020).
Tallman, A. E., Arul Kumar, M., Matthews, C. & Capolungo, L. Surrogate modeling of viscoplasticity in steels: application to thermal, irradiation creep and transient loading in HT-9 cladding. JOM 73, 126–137 (2021).
Zhou, F., Hu, C., Zeng, X. & Kang, M. Simulation of steel corrosion and iron yield in LFR with lead coolant. Energy Rep. 9, 243–253 (2023).
Wang, G., Wang, Z. & Yun, D. Cladding failure modelling for lead-based fast reactors: a review and prospects. Metals 13, 1524 (2023).
Odette, G. R. & Lucas, G. E. Embrittlement of nuclear reactor pressure vessels. JOM 53, 18–22 (2001).
Odette, G. & Lucas, G. Irradiation Embrittlement of Reactor Pressure Vessel Steels: Mechanisms, Models, and Data Correlations. In Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review (Second Volume) Vol. STP909-EB (ed. Steele, L.) (ASTM International, 1986).
Murakami, S., Miyazaki, A. & Mizuno, M. Modeling of irradiation embrittlement of reactor pressure vessel steels. J. Eng. Mater. Technol. 122, 60–66 (1999).
Kwon, J., Kwon, S. C. & Hong, J.-H. Prediction of radiation hardening in reactor pressure vessel steel based on a theoretical model. Ann. Nucl. Energy 30, 1549–1559 (2003).
Margolin, B. Z. & Kostylev, V. I. Modeling for ductile-to-brittle transition under ductile crack growth for reactor pressure vessel steels. Int. J. Press. Vessels Pip. 76, 309–317 (1999).
Margolin, B. Z., Kostylev, V. I., Ilyin, A. V. & Minkin, A. I. Simulation of JR-curves for reactor pressure vessels steels on the basis of a ductile fracture model. Int. J. Press. Vessels Pip. 78, 715–725 (2001).
Margolin, B. Z., Kostylev, V. I., Minkin, A. I. & Il’in, A. V. Modeling of ductile crack growth in reactor pressure-vessel steels and determination of JR curves. Strength Mater. 34, 120–130 (2002).
Chhibber, R., Singh, H., Arora, N. & Dutta, B. K. Micromechanical modelling of reactor pressure vessel steel. Mater. Des. (1980-2015) 36, 258–274 (2012).
Chakraborty, P. & Biner, S. B. A unified cohesive zone approach to model the ductile to brittle transition of fracture toughness in reactor pressure vessel steels. Eng. Fract. Mech. 131, 194–209 (2014).
Odette, G. R., Wirth, B. D., Bacon, D. J. & Ghoniem, N. M. Multiscale-multiphysics modeling of radiation-damaged materials: embrittlement of pressure-vessel steels. MRS Bull. 26, 176–181 (2001).
Jumel, S. & Van-Duysen, J. C. RPV-1: a virtual test reactor to simulate irradiation effects in light water reactor pressure vessel steels. J. Nucl. Mater. 340, 125–148 (2005).
Odette, G. R. & Nanstad, R. K. Predictive reactor pressure vessel steel irradiation embrittlement models: issues and opportunities. JOM 61, 17–23 (2009).
Al Mazouzi, A., Alamo, A., Lidbury, D., Moinereau, D. & Van Dyck, S. PERFORM 60: prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime. Nucl. Eng. Des. 241, 3403–3415 (2011).
Kwon, J., Mohamed, H. F. M., Kim, Y. M. & Kim, W. Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels. Nucl. Instrum. Methods Phys. Res. Sect. B Beam Interact. Mater. 262, 255–260 (2007).
Monnet, G., Domain, C., Queyreau, S., Naamane, S. & Devincre, B. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel. J. Nucl. Mater. 394, 174–181 (2009).
Bonny, G., Pasianot, R., Castin, N. & Malerba, L. Ternary Fe-Cu-Ni many-body potential to model reactor pressure vessel steels: first validation by simulated thermal annealing. Philos. Mag. 89, 3531–3546 (2009).
Zhang, Y., Millett, P. C., Tonks, M. R., Bai, X.-M. & Biner, S. B. Preferential Cu precipitation at extended defects in bcc Fe: an atomistic study. Comput. Mater. Sci. 101, 181–188 (2015).
Odette, G. R. & Wirth, B. D. A computational microscopy study of nanostructural evolution in irradiated pressure vessel steels. J. Nucl. Mater. 251, 157–171 (1997).
Messina, L., Chiapetto, M., Olsson, P., Becquart, C. S. & Malerba, L. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels. Phys. Status Solidi (a) 213, 2974–2980 (2016).
Xiong, W. et al. Thermodynamic models of low-temperature Mn-Ni-Si precipitation in reactor pressure vessel steels. MRS Commun. 4, 101–105 (2014).
Jacob, A., Domain, C., Adjanor, G., Todeschini, P. & Povoden-Karadeniz, E. Thermodynamic modeling of G-phase and assessment of phase stabilities in reactor pressure vessel steels and cast duplex stainless steels. J. Nucl. Mater. 533, 152091 (2020).
Biner, S. B., Rao, W. & Zhang, Y. The stability of precepitates and the role of lattice defects in Fe-1at%Cu-1at%Ni-1at%Mn alloy: a phase-field model study. J. Nucl. Mater. 468, 9–16 (2016).
Zhao, Y. Co-precipitated Ni/Mn shell coated nano Cu-rich core structure: a phase-field study. J. Mater. Res. Technol. 21, 546–560 (2022).
Yang, W. et al. Dislocation loop assisted precipitation of Cu-rich particles: A phase-field study. Comput. Mater. Sci. 228, 112338 (2023).
Ke, H. et al. Thermodynamic and kinetic modeling of Mn-Ni-Si precipitates in low-Cu reactor pressure vessel steels. Acta Mater. 138, 10–26 (2017).
Bai, X.-M., Ke, H., Zhang, Y. & Spencer, B. W. Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation. J. Nucl. Mater. 495, 442–454 (2017).
Ke, J.-H. & Spencer, B. W. Cluster dynamics modeling of Mn-Ni-Si precipitates coupled with radiation-induced segregation in low-Cu reactor pressure vessel steels. J. Nucl. Mater. 569, 153910 (2022).
Monnet, G., Vincent, L. & Gélébart, L. Multiscale modeling of crystal plasticity in reactor pressure vessel steels: prediction of irradiation hardening. J. Nucl. Mater. 514, 128–138 (2019).
Shu, S., Wells, P. B., Odette, G. R. & Morgan, D. A kinetic lattice Monte Carlo study of post-irradiation annealing of model reactor pressure vessel steels. J. Nucl. Mater. 524, 312–322 (2019).
Mora, D. F., Niffenegger, M., Qian, G., Jaros, M. & Niceno, B. Modelling of reactor pressure vessel subjected to pressurized thermal shock using 3D-XFEM. Nucl. Eng. Des. 353, 110237 (2019).
Liu, Y. -c. et al. Machine learning predictions of irradiation embrittlement in reactor pressure vessel steels. npj Comput. Mater. 8, 85 (2022).
Liu, Y. -c, Morgan, D., Yamamoto, T. & Odette, G. R. Characterizing the flux effect on the irradiation embrittlement of reactor pressure vessel steels using machine learning. Acta Mater. 256, 119144 (2023).
Jacobs, R., Yamamoto, T., Odette, G. R. & Morgan, D. Predictions and uncertainty estimates of reactor pressure vessel steel embrittlement using Machine learning. Mater. Des. 236, 112491 (2023).
McMurtrey, M. & Messner, M. Qualification Challenges for Additive Manufacturing in High Temperature Nuclear Applications (American Society of Mechanical Engineers Digital Collection, 2021).
Mondal, K., Martinez, O. & Jain, P. Advanced manufacturing and digital twin technology for nuclear energy*. Front. Energy Res. 12, https://doi.org/10.3389/fenrg.2024.1339836 (2024).
Pitts, S. A. et al. Chapter 14 - Modeling and simulation of advanced manufacturing techniques using MOOSE and MALAMUTE. In Risk-Informed Methods and Applications in Nuclear and Energy Engineering (eds Smith, C. L., Le Blanc, K. & Mandelli, D.) 263–286 (Academic Press, 2024).
Xie, Z., Jiang, W., Wang, C. & Wu, X. Bayesian inverse uncertainty quantification of a MOOSE-based melt pool model for additive manufacturing using experimental data. Ann. Nucl. Energy 165, 108782 (2022).
Yaseen, M., Yushu, D., German, P. & Wu, X. Fast and accurate reduced-order modeling of a MOOSE-based additive manufacturing model with operator learning. Int. J. Adv. Manuf. Technol. 129, 3123–3139 (2023).
Roy, A., Swope, A., Devanathan, R. & Van Rooyen, I. J. Chemical composition based machine learning model to predict defect formation in additive manufacturing. Materialia 33, 102041 (2024).
Mantri, S. A., Zhang, X. & Chen, W. Y. Influence of post deposition annealing on the microstructural evolution and tensile behavior of austenitic stainless-steel alloy 709 made by laser powder bed fusion. Mater. Sci. Eng. A 927, 148054 (2025).
Kadambi, S. B., Schwen, D., Ke, J.-H., He, L. & Jokisaari, A. M. Phase-field modeling of radiation-induced composition redistribution: An application to additively manufactured austenitic Fe-Cr-Ni. Comput. Mater. Sci. 255, 113895 (2025).
Hales, J. D. et al. Verification of the BISON fuel performance code. Ann. Nucl. Energy 71, 81–90 (2014).
Pastore, G. et al. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling. J. Nucl. Mater. 456, 398–408 (2015).
Bouloré, A. Importance of uncertainty quantification in nuclear fuel behaviour modelling and simulation. Nucl. Eng. Des. 355, 110311 (2019).
Tonks, M. R., Bhave, C., Wu, X. & Zhang, Y. 10 - Uncertainty quantification of mesoscale models of porous uranium dioxide. In Uncertainty Quantification in Multiscale Materials Modeling, Elsevier Series in Mechanics of Advanced Materials (eds Wang, Y. & McDowell, D. L.) 329–354 (Woodhead Publishing, 2020).
Fish, J., Wagner, G. J. & Keten, S. Mesoscopic and multiscale modelling in materials. Nat. Mater. 20, 774–786 (2021).
Rempe, J. L. et al. Advanced in-pile instrumentation for materials testing reactors. IEEE Trans. Nucl. Sci. 61, 1984–1994 (2014).
Patnaik, S., Lopes, D. A., Besmann, T. M., Spencer, B. W. & Knight, T. W. Experimental system for studying temperature gradient-driven fracture of oxide nuclear fuel out of reactor. Rev. Sci. Instrum. 91, 035101 (2020).
Mengyan, H., Xueyan, Z., Cuiting, P., Yixuan, Z. & Jun, Y. Current status of digital twin architecture and application in nuclear energy field. Ann. Nucl. Energy 202, 110491 (2024).
Chen, M. et al. Development of the environmental assisted fatigue assessment method for nuclear plants in digital twin. Nucl. Eng. Technol. 57, 103402 (2025).
Stewart, R. et al. The AGN-201 Digital Twin: a test bed for remotely monitoring nuclear reactors. Ann. Nucl. Energy 213, 111041 (2025).
Yacout, A. M. et al. Fipd: the SFR metallic fuels irradiation & physics database. Nucl. Eng. Des. 380, 111225 (2021).
Fuel Experimental Data—IAEA Data Platform. https://data.iaea.org/pages/the-iaea-fuel-database.
Advanced Material Database—IAEA Data Platform. https://data.iaea.org/pages/advanced-material-database.
Gain Databases. https://gain.inl.gov/resources/databases/.
Acknowledgements
The authors would like to thank Andrea Jokisaari for preliminary planning and discussions regarding this work. Los Alamos National Laboratory, an affirmative action/equal opportunity employer, is operated by Triad National Security, LLC, for the National Nuclear Security Administration of the U.S. Department of Energy under contract number 89233218CNA000001. M.R.T. and A.A. were supported by the University of Florida. D.A.A. was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program under the U.S. Department of Energy, Office of Nuclear Energy, United States. The funding agencies played no role in the research, planning, or writing of this review.
Author information
Authors and Affiliations
Contributions
M.R.T. wrote the first draft of the manuscript and figures. D.A.A. and A.A. revised the manuscript and added additional text. M.R.T., D.A.A., and A.A. reviewed the manuscript.
Corresponding author
Ethics declarations
Competing interests
The authors declare no competing interests.
Additional information
Publisher’s note Springer Nature remains neutral with regard to jurisdictional claims in published maps and institutional affiliations.
Rights and permissions
Open Access This article is licensed under a Creative Commons Attribution-NonCommercial-NoDerivatives 4.0 International License, which permits any non-commercial use, sharing, distribution and reproduction in any medium or format, as long as you give appropriate credit to the original author(s) and the source, provide a link to the Creative Commons licence, and indicate if you modified the licensed material. You do not have permission under this licence to share adapted material derived from this article or parts of it. The images or other third party material in this article are included in the article’s Creative Commons licence, unless indicated otherwise in a credit line to the material. If material is not included in the article’s Creative Commons licence and your intended use is not permitted by statutory regulation or exceeds the permitted use, you will need to obtain permission directly from the copyright holder. To view a copy of this licence, visit http://creativecommons.org/licenses/by-nc-nd/4.0/.
About this article
Cite this article
Tonks, M.R., Andersson, D.A. & Aitkaliyeva, A. Computational design of materials for nuclear reactors. npj Comput Mater (2026). https://doi.org/10.1038/s41524-026-01980-8
Received:
Accepted:
Published:
DOI: https://doi.org/10.1038/s41524-026-01980-8
