arXiv subjects

Shengli Chen

Publications and source records attributed to Shengli Chen.

12 recordsLinked to original sources

arXiv2025

B-spline periodization of Fourier pseudo-spectral method for non-periodic problems

Spectral methods are renowned for their high accuracy and efficiency in solving partial differential equations. The Fourier pseudo-spectral method is limited to periodic domains and suffers from Gibbs oscillations in non-periodic problems. The Chebyshev method mitigates this issue but requires edge-clustered grids, which does not match the characteristics of many physical problems. To overcome these restrictions, we propose a B-spline-periodized Fourier (BSPF) method that extends to non-periodic problems while retaining spectral-like accuracy and efficiency. The method combines a B-spline approximation with a Fourier-based residual correction. The B-spline component enforces the smooth matching of boundary values and derivatives, while the periodic residual is efficiently treated by Fourier differentiation/integration. This construction preserves spectral convergence within the domain and algebraic convergence at the boundaries. Numerical tests on differentiation and integration confirm the accuracy of the BSPF method superior to Chebyshev and finite-difference schemes for interior-oscillatory data. Analytical mapping further extends BSPF to non-uniform meshes, which enables selective grid refinement in regions of sharp variation. Applications of the BSPF method to the one-dimensional Burgers' equation and two-dimensional shallow water equations demonstrate accurate resolution of sharp gradients and nonlinear wave propagation, proving it as a flexible and efficient framework for solving non-periodic PDEs with high-order accuracy.

math.NA↗

arXiv2025

Consistency analysis and nuclear data validation for two series of beryllium reflector critical benchmark experiments

Neutron-induced nuclear reaction data on beryllium playing a crucial role in nuclear application. However, discrepancies have been observed in two closely related series of beryllium-reflector fast-spectrum critical benchmark experiments, HMF-058 and HMF-066, which are widely used in current nuclear data validation. In this work, we address these inconsistencies by improving the secondary angular distributions of the (n,n) and (n,2n) reactions of beryllium, thereby making the theoretical calculations (C) and experimental results (E) of these two series more consistent, and reducing the cumulative ${\chi^2}$ value from 7.58 using the ENDF/B-VII.1 to 4.52. All calculations based on the improved nuclear data agree with the experimental measurements within 1${\sigma}$ experimental uncertainty. Based on the latest comprehensive evaluation of uranium nuclear data, this consistency is slightly improved, and the cumulative ${\chi^2}$ value decreases to 4.36 once again. Despite these advances, systematic differences in the expected values of C/E between the two series still exist. The C/E values of the HMF-066 series are generally 230-330 pcm lower than those of the HMF-058 series, comparable to their experimental uncertainties of 200-400 pcm. Therefore, drawing a definitive conclusion about this systematic difference remains challenging. If the current improvement of differential nuclear data based on experimental data of ${^9}$Be is accurate, then the HMF-058 series experiments seem to be more reliable than the HMF-066 series.

physics.app-ph↗

arXiv2024

Unveiling High Selectivity Origin of Pt-Bi Catalysts for Alkaline Methanol Electrooxidation via CO-free pathway

A long-standing puzzle for methanol electrooxidation is how to achieve a CO-free pathway and accurately understand the origin of electrocatalytic selectivity. Herein, we unequivocally demonstrate that the Bi-modified Pt/C follows a CO-free dominated pathway during alkaline methanol electrooxidation, and unveil the formaldehyde (HCHO) intermediate as a critical factor influencing pathway selectivity. These findings are substantiated by kinetic isotope effects, formate Faradaic efficiency, in situ spectroscopy, ab initio molecular dynamic simulations, and density functional theory calculations. Bi modification significantly increases the HCHO dehydrogenation barrier, which facilitates its desorption and subsequent conversion to the H2COOH- anion at the alkaline interface, intrinsically avoiding CO formation. More specifically, the formation of ensemble sites featuring V-shaped Bi-Pt-Bi configuration inhibits the cleavage of C-H bond, and the weak OH binding energy at Bi adatoms effectively prevents blockage of oxygenated species, allowing such ensemble sites to fulfill their functional role. Our study opens up a novel dimension for designing advanced CO-free catalysts.

physics.chem-ph↗

arXiv2020

Neutronic Study of UO2-BeO Fuel with Various Claddings

The neutronic properties of UO2-BeO fuel with various claddings are investigated through the Monte Carlo method 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 two factors, the uranium enrichment and the volume fraction of BeO in the UO2-BeO fuel. The uranium enrichment is determined to ensure the same cycle length for each BeO fraction with zircaloy, FeCrAl, and SiC claddings. Similar neutronic properties are observed between the zircaloy and SiC claddings for a given BeO fraction, including the infinite multiplication factor, the gas release, the power distribution, and the isotopic concentrations. An important feature of UO2-BeO fuel is the production of 4He, which is significant compared with current UO2 fuel. Lower production rates of xenon and krypton are found in the case of the 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 and in a fuel pellet is flatter for a higher BeO fraction with the same cladding. Higher peak power is found in fuel assemblies with FeCrAl compared to the other two claddings, while the radial power distribution is quite similar for the three claddings.

physics.app-ph↗

arXiv2019

Forecasting Implied Volatility Smile Surface via Deep Learning and Attention Mechanism

The implied volatility smile surface is the basis of option pricing, and the dynamic evolution of the option volatility smile surface is difficult to predict. In this paper, attention mechanism is introduced into LSTM, and a volatility surface prediction method combining deep learning and attention mechanism is pioneeringly established. LSTM's forgetting gate makes it have strong generalization ability, and its feedback structure enables it to characterize the long memory of financial volatility. The application of attention mechanism in LSTM networks can significantly enhance the ability of LSTM networks to select input features. The experimental results show that the two strategies constructed using the predicted implied volatility surfaces have higher returns and Sharpe ratios than that the volatility surfaces are not predicted. This paper confirms that the use of AI to predict the implied volatility surface has theoretical and economic value. The research method provides a new reference for option pricing and strategy.

q-fin.CP↗

arXiv2019

Calculation and Verification of Irradiation Damage Cross Section with Energy-Angular Distribution

To complete the computation of Displacements per Atom (DPA) cross sections, the present work shows the methods of calculating DPA cross sections with the nuclear data of energy-angular distribution in both the laboratory and the Center-of-Mass (CM) frames. The method of direct calculation with data in the CM frame is proposed and recommended to decrease the computation burden and keep all information. Theoretical analyses reveal that more than 7-point Gauss-Legendre Quadrature (GLQ) should be used to ensure the convergence of the angular integration for DPA computations. Numerical results show that 8-point GLQ is sufficient for the continuum inelastic neutron scattering, while 64-point GLQ is implemented in NJOY. Because the integrand over secondary energy is not derivable in the whole domain of the secondary energy, the trapezoidal integration is used to perform the numerical integration. The numerical calculations show that the trapezoidal integration is suitable to perform the integration over the secondary energy on the fine grid given by nuclear data files at least for 56Fe. The present work reveals that the direct interpolation of energy-angular-integrated damage can give the same results computed with standard interpolated energy-angular distributions. The DPA cross sections will be overestimated if isotropic angular distributions are assumed. However, the first-order Legendre polynomial can give DPA cross sections within 0.4% deviation, while 12 orders are required to describe the anisotropic angular distribution.

nucl-th↗

arXiv2018

Minor Actinides Transmutation in Candidate Accident Tolerant Fuel-Claddings U3Si2-FeCrAl and U3Si2-SiC

An advanced transmutation method is suggested that the long-lived Minor Actinides (MAs) in the spent fuel can be efficiently transmuted in the candidate Accident Tolerant Fuel (ATF). The transmutation of MAs is investigated through the Monte Carlo simulations in two potential fuel-claddings of ATF, U3Si2-FeCrAl and U3Si2-SiC. The critical loadings of MAs are determined through the Linear Reactivity Model (LRM) in order to keep the same reactivity as the current UO2-zircaloy system at the End of Cycle (EOC). In all cases, excellent transmutation efficiencies are found for the most important three MAs, 237Np, 241Am, and 243Am, of which the total transmutation rates are around 60%, 90%, and 60%, respectively. If only the longest-lived isotope 237Np is considered, one U3Si2-SiC assembly can transmute 237Np from six normal assemblies. The loading of MAs has little influences on the neutronic properties, such as the power distributions inside an assembly and inside a fuel rod. The transmutation of MAs in the ATF assembly is shown to be more efficient and safe comparing with the normal assembly, while other important properties are kept, such as the cycle length and the power distribution.

physics.app-ph↗

arXiv2018

Radial Distributions of Power and Isotopic Concentrations in Candidate Accident Tolerant Fuel U3Si2 and UO2/U3Si2 Fuel Pins with FeCrAl Cladding

Monte Carlo simulations show similarity on radial distributions of power and isotopic concentrations at any effective full power depletion time among five kinds of fuel-cladding combinations with the same cycle length, including the normal UO2-zircaloy combination, the candidate Accident Tolerant Fuel (ATF) UO2/U3Si2-FeCrAl combination, and three kinds of candidate ATF U3Si2-FeCrAl combinations. An analytical formula f(x,s) including the fuel exposure (s) and the relative radial (x) is proposed to describe the radial properties for all five kinds of fuel-cladding combinations. f(x,s) has the form of the second order polynomial term of s with the exponential type of coefficients depending on x. It is shown that the suggested function f(x,s) gives a nice description on the simulation data with rather small deviations and can immediately provide radial distribution of power, burnup, and isotopic concentrations of 235U, 238U, 239Pu, and 241Pu at any fuel exposure and relative radius. It is useful to discuss the fuel temperature through the present analytical formula. The realistic radial power distribution gives flatter radial temperature distribution compared with the uniform power distribution. Because of the different thermal conductivities of fuels and claddings and the different thicknesses of claddings, the present discussed five kinds of fuel-cladding combinations have different radial temperature distributions, although their radial power distributions are quite similar. The present work provides an analytical formula to describe the radial properties of the ATF which is expected to be helpful for further neutronic and multi-physics coupling studies.

physics.app-ph↗

arXiv2018

Transmutation of Minor Actinides and Power Flattening in PWR MOX Fuel

In order to transmute long-lived MAs and flatten power distribution in a PWR MOX fuel assembly, the authors have proposed to replace some high power fuel rods by MAs loaded fuel rods. The present work proves the high efficiency of long-lived MAs transmutation in a PWR MOX assembly with 92 fuel rods with 3% MAs loaded. The power and the burnup distribution have been flatted by using the MAs loading method proposed by authors. The 237Np loading method is expected for the transmutation by comparing with natural decay time to achieve the same reduction. The mixed MAs loading has better behaviors on power flattening and negative contribution of reactivity. In addition, the mixed MAs loading can largely reduce the quantity of 241Am, while the addition of other MAs has no influence on the transmutation efficiency of 237Np.

physics.app-ph↗

arXiv2017

Study of Minor Actinides Transmutation in PWR MOX fuel

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. Partitioning-Transmutation is supposed to be an efficient method to treat the long-lived radionuclides in spent fuel. Some Minor Actinides (MAs) have very long half-lives among the radionuclides in the spent fuel. Accordingly, the study of MAs transmutation is a significant work for the post-processing of spent fuel. In the present work, the transmutations in Pressurized Water Reactor (PWR) mixed oxide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs, $^{237}$Np and five MAs ($^{237}$Np, $^{241}$Am, $^{243}$Am, $^{244}$Cm and $^{245}$Cm) are incorporated homogeneously into the MOX fuel assembly. The transmutation of MAs is simulated with different initial MOX concentrations. The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for the two kinds of MAs primarily generated in the UOX fuel, $^{237}$Np and $^{241}$Am. In addition, the inclusion of $^{237}$Np in MOX has no large influence for other MAs, while the transmutation efficiency of $^{237}$Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be a new, efficient nuclear spent fuel management method for the future nuclear power generation.

physics.app-ph↗

arXiv2017

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U$_3$Si$_2$-FeCrAl

Neutronic performance is investigated for a potential accident tolerant fuel (ATF),which consists of U$_3$Si$_2$ fuel and FeCrAl cladding. In comparison with current UO$_2$-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U$_3$Si$_2$ has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U$_3$Si$_2$-FeCrAl fuel-cladding systemis taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly.These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U$_3$Si$_2$-FeCrAl system is a potential ATF candidate from a neutronic view.

physics.app-ph↗

arXiv2016

A simple formula for local burnup based on constant relative reaction rate per nuclei

A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of $^{235}$U, $^{238}$U, and $^{239}$Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives very similar results as the MC calculation from the starting to high burnup level, but takes just a few minutes. The relative reaction rates are found to be almost independent on the radius (except $(n,γ)$ of $^{238}$U) and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance on the accuracy and the cost on time.

nucl-th↗