Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U 3 Si 2 -FeCrAl

Nuclear Materials and Energy

Yuan

2020

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The neutronic properties of UO2-BeO fuel with various claddings are investigated through the Monte Carlo methods and the Linear Reactivity Model. A second order polynomial function is suggested to describe the relationship between the difference of the reactivity at the End of Cycle and the two factors, the uranium enrichment and the volume content of BeO in the UO2-BeO fuel. The uranium enrichment is determined to ensure the cycle length for each BeO fraction with the zircaloy, FeCrAl, and SiC claddings. Similar neutronic properties are observed between the zircaloy and SiC claddings for a given BeO fraction, including the infinity multiplication factor, the gas release, the power distribution, and the isotopic concentrations. An important feature of the UO2-BeO fuel is the production of 4 He, which is significant compared with the current UO2 fuel. Less production rates of xenon and krypton are found in the case of UO2-BeO-FeCrAl fuel-cladding system, while those in the UO2-BeO-zircaloy and UO2-BeO-SiC systems are quite similar to the current UO2-zircaloy system. The power distribution in an assembly or in a fuel pellet is flatter for the higher BeO fraction with the same cladding. Higher peak power in a fuel assembly is found with the FeCrAl cladding than that with the other two claddings, while the radial power distribution is quite similar for the three claddings.

Section: Isotopic Concentrationsmentioning

confidence: 96%

Section: Introductionmentioning

confidence: 99%

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confidence: 99%

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confidence: 99%

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Neutronic study of UO2-BeO fuel with various claddings

Nuclear Materials and Energy

Yuan

2020

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“…Therefore, we investigate the importance of high-order Legendre polynomials of angular distribution for DPA calculations in Section 4. The examples on 56 Fe are taken to show the numerical results because of its high abundance in iron-based steels, which are used in RPV in Light Water Reactor (LWR), fuel cladding in Fast Reactor (FR), and candidate fuel cladding in Accident Tolerant Fuel (ATF) [11], [12]. The numerical applications in this paper are mainly based on the nuclear data of JEFF-3.1.1 [7], which is widely used and qualified by CEA, EDF, and FRAMATOME.…”

Section: Introductionmentioning

confidence: 99%

Calculation and verification of neutron irradiation damage with differential cross sections

Nuclear Instruments and Methods in Physics Research Section B:

Bernard

Tamagno

et al. 2019

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The Displacement per Atom (DPA) rate is conventionally computed with DPA cross sections in reactor applications. The method of direct calculation with energyangular distributions given in the Center of Mass (CM) frame is proposed and recommended in the present work. The methods for refining and verifying the calculations of DPA cross sections are proposed: (i) Gauss-Legendre-Quadrature-based Piecewise Integration (GLQPI) for ensuring the numeric convergence of integral over emission angle due to the discontinuity of integrand; (ii) verification of the convergence for trapezoidal integration over the secondary energy; (iii) interpolation of doubledifferential cross sections. For 56 Fe of JEFF-3.1.1, the current numeric integration over emission angle is shown not convergent, whereas the direct trapezoidal over the secondary energy and the direct interpolation of energy-angle-integrated damage are shown accurate. On the other hand, it is shown that the DPA cross sections are overestimated if isotropic angular distributions are assumed. However, the DPA cross section is not sensitive to the high-order Legendre polynomials because the former is an angle-integrated quantity. Numerical results of neutron elastic scattering show that 2 orders of Legendre polynomials can give the DPA rates of 56 Fe within 0.5% overestimation for fission reactors, while 4 orders are required for fusion reactors. For neutron inelastic scatterings-induced DPA, the first order Legendre polynomial is sufficient for both fission and fusion reactors.

Multi‐objective optimization design of U ₃ Si ₂ –FeCrAl accident tolerant fuel elements based on Gaussian process and genetic algorithm

Intl J of Energy Research

Zhou

et al. 2022

Summary Among various accident tolerant fuel (ATF) concepts, U3Si2–FeCrAl fuel systems are considered as ones of the most potential fuel for the next generation light water reactor. In this work, a multi‐objective optimization framework based on multi‐objective genetic algorithm (GA) and Gaussian process (GP) was established for the optimal design of U3Si2–FeCrAl ATF fuel element. The coupling of GA and surrogate GP model could take advantage of the high fidelity of multi‐physics modeling and the global searching capability of GA, therefore allowing to perform a relatively fast and accurate multi‐objective optimization design for the U3Si2–FeCrAl ATF fuel element. Maximum fuel temperature, maximum cladding Von–Mises stress, and maximum plenum pressure are considered as the three objective functions for the optimization problem in this study. According to the Pareto approximation surface obtained in the optimization process, a lowest maximum fuel temperature of 871 K, a lowest maximum cladding stress of 87.9 MPa, and a lowest plenum pressure of 3.1 MPa for the U3Si2–FeCrAl ATF fuel element could be achieved based on different design parameters. The multi‐objective optimization framework developed in this work could perform as an efficient tool for the multi‐objective optimization design of the ATF fuel element.