Abstract
A finite-strain homogenization creep model for composite fuels under irradiation conditions is developed and verified, with the irradiation creep strains of the fuel particles and matrix correlated to the macroscale creep responses, excluding the contributions of volumetric strain induced by the irradiation swelling deformations of fuel particles. A finite element (FE) modeling method for uniaxial tensile creep tests is established with the irradiation effects of nuclear materials taken into account. The proposed models and simulation strategy are numerically implemented to a kind of composite nuclear fuel, and the predicted mesoscale creep behaviors and the macroscale creep responses are investigated. The research results indicate that: (1) the macroscale creep responses and the mesoscale stress and strain fields are all greatly affected by the irradiation swelling of fuel particles, owing to the strengthened mechanical interactions between the fuel particles and the matrix. (2) The effective creep rates for a certain case are approximately two constants before and after the critical fission density, which results from the accelerated fission gas swelling after fuel grain recrystallization, and the effects of macroscale tensile stress will be more enhanced at higher temperatures. (3) The macroscale creep contributions from the fuel particles and matrix depend mainly on the current volume fractions varying with fission density. (4) As a function of the macroscale stress, temperature, initial particle volume fraction and particle fission rate, a multi-variable mathematical model for effective creep rates is fitted out for the considered composite fuels, which matches well with the FE predictions. This study supplies important theoretical models and research methods for the multi-scale creep behaviors of various composite fuels and provides a basis for simulation of the thermal–mechanical behavior in related composite fuel elements and assemblies.
Similar content being viewed by others
References
Holden AN. Dispersion fuel elements. New York: Gordon and Breach Science Publishers, Inc.; 1967.
Kloosterman JL, Damen PMG. Reactor physics aspects of plutonium burning in inert matrix fuels. J Nucl Mater. 1999;274:112–9.
Savchenko AM, Vatulin AV, Morozov AV, et al. Inert matrix fuel in dispersion type fuel elements. J Nucl Mater. 2006;352(1–3):372–7.
Neeft EAC, Bakker K, Schram RPC, et al. The EFTTRA-T3 irradiation experiment on inert matrix fuels. J Nucl Mater. 2003;320(1–2):106–16.
Lombardi C, Luzzi L, Padovani E, et al. Thoria and inert matrix fuels for a sustainable nuclear power. Prog Nucl Energy. 2008;50(8):944–53.
Carmack WJ, Todosow M, Meyer MK, Pasamehmetoglu KO. Inert matrix fuel neutronic, thermal–hydraulic, and transient behavior in a light water reactor. J Nucl Mater. 2006;352:276–84.
Gong X, Ding S, Zhao Y, Huo Y, Zhang L, Li Y. Effects of irradiation hardening and creep on the thermo-mechanical behaviors in inert matrix fuel elements. Mech Mater. 2013;65:110–23.
Duyn LV. Evaluation of the mechanical behavior of a metal matrix dispersion fuel for plutonium burning. A Thesis for the Degree Master of Science in Mechanical Engineering, Georgia Institute of Technology, 2003.
Savchenko A, Konovalov L, Vatulin A, Morozov A, Orlov V, Uferov O, Ershov S, Laushkin A, Kulakov G, Maranchak S, Petrova Z. Dispersion type zirconium matrix fuels fabricated by capillary impregnation method. J Nucl Mater. 2007;362(2–3):356–63.
Rodriguez P, Krishnan R, Sundaram CV. Radiation effects in nuclear reactor materials—correlation with structure. Bull Mater Sci. 1984;6(2):339–67.
Adamson RB, Coleman CE, Griffiths M. Irradiation creep and growth of zirconium alloys: a critical review. J Nucl Mater. 2019;521:167–244.
Cochran KM, Solomon AA. Fabrication and creep testing of UO2 tensile specimens. J Nucl Mater. 1983;119:162–9.
Safari M, Aghaie M, Minuchehr A, Allahyarizadeh Gh. Numerical study of hyperstoichiometric fuel creep (UO2+x) in fuel clad interaction of WWER1000. Ann Nucl Energy. 2019;133:950–9.
Zhao Y, Ding S, Zhang X, Wang C, Yang L. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets. J Nucl Mater. 2016;482:278–93.
Jeong GY, Kim YS, Jamison LM, Robinson AB, Lee KH, Sohn D-S. Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel. J Nucl Mater. 2017;487:265–79.
Xing ZH, Ying S. H, Study on the irradiation swelling of U3Si2–Al dispersion fuel. At Energy Sci Technol. 2001;35(1):15–9.
Cui Y, Ding SR, Chen ZT, Huo YZ. Modifications and applications of the mechanistic gaseous swelling model for UMo fuel. J Nucl Mater. 2015;457:157–64.
Tian X, Kong XZ, Yan F, Ding SR. Hydrostatic pressure effect of fission gas swelling in UMo/Al dispersion fuel plate. At Energy Sci Technol. 2017;51(11):2062–8.
Zhao YM, Zhang JY, Ding SR. A new method for solving the fission gas diffusion equations with time varying diffusion coefficient and source term considering recrystallization of fuel grains. Nucl Mater Energy. 2019;20: 100686.
Rest J. A model for fission-gas-bubble behavior in amorphous uranium silicide compounds. J Nucl Mater. 2004;325(2–3):107–17.
Rest J. Application of a mechanistic model for radiation-induced amorphization and crystallization of uranium silicide to recrystallization of UO2. J Nucl Mater. 1997;248:180–4.
Jeong GY, Kim YS, Sohn D. Mechanical analysis of UMo/Al dispersion fuel. J Nucl Mater. 2015;466:509–21.
Cai W, Zhao YM, Gong X, Ding SR, Huo Y. Z, Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel. At Energy Sci Technol. 2015;49(3):502–10.
Rest J. DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements. Nucl Technol. 1998;126:88–101.
Gong X, Zhao YM, Ding SRA. new method to simulate the micro-thermo-mechanical behaviors evolution in dispersion nuclear fuel elements. Mech Mater. 2014;77:14–27.
Wang QM, Yan XQ, Ding SR, Huo YZ. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements. J Nucl Mater. 2010;399(1):41–54.
Wang QM, Cui Y, Ding SR, Huo YZ. Simulation of the coupling behaviors of particle and matrix irradiation swelling and cladding irradiation growth of plate-type dispersion nuclear fuel elements. Mech Mater. 2011;43(4):222–41.
Zhang J, Wang H, Wei H, Zhang J, Tang C, Lu C, Huang C, Ding S, Li Y. Modelling of effective irradiation swelling for inert matrix fuels. Nucl Eng Technol. 2021;53:2616–28.
Zhang J, Zhang JY, Wang HY, Wei HY, Tang CB, Lu C, Ding SR, Li YM. Research on the homogenized postirradiation elastoplastic constitutive relations for composite nuclear fuels. Front Mater. 2021. https://doi.org/10.3389/fmats.2021.651875.
Clough DJ. Creep-properties of oxide and carbide fuels under irradiation. J Nucl Mater. 1977;65(1):24–36.
Brucklacher D, Dienst W. Creep behavior of ceramic nuclear fuels under neutron-irradiation. J Nucl Mater. 1972;42(3):285–96.
Mikheev EN, Fedotov AV, Novikov VV, et al. Methodology and results of investigation of radiation creep in large-grain uranium dioxide based fuel. At Energy. 2014;116(1):20–6.
Ding S, Wang Q, Huo Y. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect. II. Effects of variations of the fuel particle diameters. J Nucl Mater. 2010;397:80–91.
Nakata K, Matsuda T, Kawai M. Multi-scale creep analysis of plain-woven laminates using time-dependent homogenization theory: effects of laminate configuration. Int J Mod Phys B. 2008;22:6173–8.
Yu P, Duan YH, Chen E, Tang SW, Hanif A, Fan YL. Microstructure-based homogenization method for early-age creep of cement paste. Constr Build Mater. 2018;188:1193–206.
Zhao YM, Gong X, Ding SR, Huo YZA. numerical method for simulating the non-homogeneous irradiation effects in full-sized dispersion nuclear fuel plates. Int J Mech Sci. 2014;81:174–83.
Liu ML, Lee YH, Rao DV. Development of effective thermal conductivity model for particle-type nuclear fuels randomly distributed in a matrix. J Nucl Mater. 2018;508:168–80.
Gavrikov AA, Knyazkov D, Melnikov AM, et al. On limits of applicability of the homogenization method to modeling of layered creep media. IFAC-Papers OnLine. 2018;51(2):144–9.
Shamaev AS, Shumilova VV. Homogenization of the equations of state for a heterogeneous layered medium consisting of two creep materials. Proc Steklov Inst Math. 2016;295(1):213–24.
Matsuda T. Homogenized creep behavior of creep laminates at high temperature. Int J Mod Phys B. 2008;22:6161–6.
Matsuda T, Fukuta Y. Multi-scale creep analysis of angle-ply CFRP laminates based on a homogenization theory. J Solid Mech Mater Eng. 2010;4(11):1664–72.
Phan V, Zhang X, Li YM, et al. Microscale modeling of creep deformation and rupture in nickel-based superalloy IN 617 at high temperature. Mech Mater. 2017;114:215–27.
Zhao N, Roy A, Wang W, et al. Coupling crystal plasticity and continuum damage mechanics for creep assessment in Cr-based power-plant steel. Mech Mater. 2019;130:29–38.
Zhao Y, Gong X, Cui Y, Ding S. Simulation of the fission-induced swelling and creep in the CERCER fuel pellets. Mater Des. 2016;89:183–95.
Jung J, Kim Y, Park J, Ryu S. Transfer learning for enhancing the homogenization-theory-based prediction of elasto-plastic response of particle/short fiber-reinforced composites. Compos Struct. 2022;285: 115210.
Jansson S. Homogenized nonlinear constitutive properties and local stress concentrations for composites with periodic internal structure. Int J Solids Struct. 1992;29(17):2181–200.
Atkins SL, Gibeling JC. A finite element model of the effects of primary creep in an Al–SiC metal matrix composite. Metall Mater Trans A. 1995;26A:3067–79.
Zhang JY, Ding SR. Mesoscale simulation research on the homogenized elasto-plastic behavior of FeCrAl alloys. Mater Today Commun. 2020;22: 100718.
Lemes M, Soba A, Denis A. An empirical formulation to describe the evolution of the high burnup structure. J Nucl Mater. 2015;456:174–81.
Spino J, Stalios AD, Santa Cruz H, Baron D. Stereological evolution of the rim structure in PWR-fuels at prolonged irradiation: dependencies with burn-up and temperature. J Nucl Mater. 2006;354:66–84.
Noirot J, Desgranges L, Lamontagne J. Detailed characterisations of high burn-up structures in oxide fuels. J Nucl Mater. 2008;372:318–39.
MATPRO. A handbook of materials properties for use in the analysis of light water reactor fuel rod behavior. US Department of Energy; 1978.
Suzuki M, Saitou H. Light water reactor fuel analysis code FEMAXI-6 (Ver.1) [Z]. US Department of Energy; 2005.
Rest J. A model for the effect of the progression of irradiation-induced recrystallization from initiation to completion on swelling of UO2 and U–10Mo nuclear fuels. J Nucl Mater. 2005;346:226–32.
Spino J, Rest J, Goll W, Walker CT. Matrix swelling rate and cavity volume balance of UO2 fuels at high burn-up. J Nucl Mater. 2005;346:131–44.
Hales JD, Williamson RL, Novascone SR, et al. BISON theory manual the equations behind nuclear fuel analysis. Idaho: Idaho National Laboratory; 2016.
Hayes TA, Kassner ME. Creep of zirconium and zirconium alloys. Metall Mater Trans A. 2006;37:2389–96.
Funding
The authors are very grateful for the supports from the National Natural Science Foundation of China (Nos. 12132005, 12102094 and 12135008) and the Shanghai Sailing Program (21YF1402200), and the foundation from the Science and Technology on Reactor System Design Technology Laboratory.
Author information
Authors and Affiliations
Corresponding author
Ethics declarations
Conflict of interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.
Appendices
Appendix 1
Considering the bi-linear relation between the effective creep strains and irradiation time, the effective creep rate model with a temperature of 523 K and an initial particle volume fraction of 30% is obtained as
where \(\dot{\overline{\varepsilon }}_{{}}^{{{\text{cr}}}}\) is in 1/d, and other parameters are the same as those in Table 1.
To verify the effectiveness of Eq. (19) and the homogenization creep model, the verification method in [38] is adopted. A macroscale homogenized-medium model is established in Fig. 7a, and then the virtual multi-axial compression tests are implemented and the macroscale creep results are obtained, which are compared with the creep results of multi-particle composite model, as shown in Fig. 7b. The model parameters of homogenized-medium model are described in Table 2. Similarly, the mechanical constitutive relation for the homogenized media contains the elastic, plastic, irradiation swelling and creep deformations. The effective swelling model used herein is developed in our previous study [28]. It should be noted that two compression tests with different pressure conditions are implemented to better verify the creep results. The boundary conditions are described in Table 2.
Figure 7c, d give the creep strains along the z-direction of the two models for Test 1, together with the creep strains along the x-direction for Test 2. It can be seen that the absolute values of the homogenized-medium model for Test 1 are a little higher than those of the multi-particle composite model, and the maximum relative error is ~ 6.67%. As for Test 2, the absolute results of the homogenized-medium model are slightly lower than those of the multi-particle composite model, with the maximum relative error of ~ 8.27%. The relative error is acceptable, which indicates that the macroscale creep model is effective, and can be used for the three-dimensional thermal–mechanical behavior simulations of inert matrix composite fuel elements and assemblies.
Appendix 2
2.1 Effects of Fission Rate
Figure 8a displays the effective creep strains under different fission rates for the designated calculation conditions. It can be seen that the effective creep strains give rise to fission density, but are almost the same at different fission rates. Figure 8b displays the creep contributions of each part for different fission rates. It can be seen that the fission rate has a slight effect on the creep contribution proportions. This is induced by the similar current particle volume fractions under different fission rates. These similar values indicate that the mechanical interactions between the fuel particles and the matrix will be almost the same at certain fission densities. It should be stated that the fission rate has a slight influence on effective creep strain, but will affect the effective creep rate, owing to its description with the irradiation time.
Rights and permissions
About this article
Cite this article
Zhang, J., Zhang, J., Wang, H. et al. Modeling of Mesoscale Creep Behaviors and Macroscale Creep Responses of Composite Fuels Under Irradiation Conditions. Acta Mech. Solida Sin. 35, 1040–1054 (2022). https://doi.org/10.1007/s10338-022-00331-6
Received:
Revised:
Accepted:
Published:
Issue Date:
DOI: https://doi.org/10.1007/s10338-022-00331-6