[en] Highlights: • A new chord length distribution model is derived and developed for CLS methods. • The new model is specifically for MC analyses of reactors with TRISO-type fuel. • Fuel kernels can be sampled on the fly using the new model in MC simulations. • The model is verified to be accurate in analyzing simple and realistic systems. • The new model shows a higher efficiency than the old model in MC simulations. - Abstract: A new chord length distribution model is proposed to characterize the stochastic distribution of TRISO fuel particles in nuclear reactor systems and is used in the chord length sampling (CLS) method to analyze the neutronic behavior of TRISO fuel systems. In this model, the coating layers of fuel particles are homogenized with the background matrix region. The probability density function (PDF) of the chord length between fuel kernels, instead of fuel particles, is developed and is used in the CLS method. We first apply the new CLS model to solving one-group eigenvalue problems in a simplified 3-D stochastic medium system. Good accuracy is obtained in predicting the multiplication factor and fission density distribution. The relative differences are within 1.0% for both the multiplication factor and the total fission density in all the studied scenarios. We then apply the new CLS model to analyzing three realistic reactor designs: two Very High Temperature Gas-cooled Reactor unit cells and one fuel pin unit cell of an innovative light water reactor design with accident tolerant fuel. Infinite multiplication factor and intra-Dancoff factor are evaluated for the three unit cells respectively. Compared with the reference results, predictions from CLS simulations show a relative difference of less than 0.26% for infinite multiplication factors and less than 1.0% for intra-Dancoff factors in all the studied cases. Meanwhile, a significant improvement in the computational efficiency has been observed for the CLS new model compared with the old model, with a speedup of at least 1.70. The CLS method with the new chord length distribution model is verified to be an efficient Monte Carlo method, superior over the old model and without sacrificing the accuracy

[en] Highlights: • Noble metal particulate transport, deposition, and extraction in the MSRE is analyzed. • Insoluble fission product behavior during the MSRE is reviewed. • Methodology for a noble metal-helium bubble extraction model is presented. • Results and sensitivity analysis of the noble metal-helium bubble model is conducted. • Safety concerns and possible multiphysics effects of noble metal deposition is discussed. This work provides a general overview of the detailed reports of noble metal fission product behavior during the Molten Salt Reactor Experiment (MSRE) and applies a species mass transport analysis to replicate and explain the basic behavior of these fission products in molten salt systems. Specifically, the operational difference between the MSRE experimental runs which used ²³⁵U and ²³³U as the enriched fuel is described and analyzed in light of the different transport behavior of noble metals during these two very different sets of operational runs. The generation of a circulating helium bubble swarm during the MSRE is also analyzed and a helium bubble model is coupled with the noble metal mass transport analysis. This coupled model provides specific insight in how noble metal deposition on the helium bubble swarm affected the reactor behavior and the results are discussed in relation to the general findings of the MSRE.

[en] Highlights: • Method of on-the-fly sampling S(α,β) data at any temperature is proposed. • Temperature dependence of thermal energy and momentum transfer CDFs is studied. • Functional fits are found to the data using a regression analysis. • Fits can construct the CDFs at any temperature at thermal energies. • Storage of fitting coefficients is much less than the current ACE data. - Abstract: Temperature can strongly affect the probabilities of certain neutron interactions (fission, capture, scattering, etc.) with materials. These probabilities are referred to in the nuclear community as ‘cross sections’ and are used as inputs for computer simulations. During the lifetime of a nuclear reactor, the core and its surrounding materials will experience a wide range of temperatures. To simulate the neutronic behavior in a realistic core, it is required to pre-store a large amount of cross section data to encompass the entire temperature range a neutron may experience. In recent years, methods have been developed to reduce data storage and obtain the cross section at the desired temperature ‘on-the-fly’ during radiation transport simulations using Monte Carlo codes. At thermal energies, however, the scattering of neutrons is complicated by their relatively small wavelengths, making molecular binding and lattice effects significant. Current approaches typically require nuclear data file sizes of tens to hundreds of MB per temperature, which can be prohibitive for realistic reactor physics simulations. To reduce the storage burden, a fitting approach in temperature is investigated that allows for the efficient evaluation of the thermal neutron scattering physics at an arbitrary temperature within a predefined range. The physics for thermal neutron scattering in graphite and hydrogen in water are evaluated with this approach. In both cases, the functional fits are able to accurately reproduce the scattering probabilities. The data storage for the fitting approach requires only a few 100 kB, which is a significant memory savings over the existing methods. These data can be used to sample a neutron's outgoing energy and scattered angle at an arbitrary temperature with minimal errors

[en] Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k_eff) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron transport. As high as 1.00% relative difference in k_eff and ∼1.50% relative difference in peak fission power density are observed. As the packing fraction and optical thickness increase, the effect gradually dissipates. Neutron channeling between fuel particles is identified as the effect most responsible for the different neutronic results. Different size distributions result in the difference in the average number of fuel particles and their average size. As a result, different degrees of neutron channeling are produced. The size effect in realistic reactor unit cells is also studied and, from the predicted values of infinite multiplication factors, it is concluded that the fuel particle size distribution effects are not negligible

[en] Highlights: • The Dancoff factors for randomly distributed TRISO fuel particles are evaluated. • A new “dual-sphere” model is proposed to predict Dancoff factors. • The new model accurately accounts for the coating regions of fuel particles. • High accuracy is achieved over a broad range of design parameters. • The new model can be used to analyze reactors with double heterogeneity. - Abstract: A new mathematical model, the dual-sphere model, is proposed to analytically evaluate Dancoff factors of TRISO fuel kernels based on the chord method. The accurate evaluation of fuel kernel Dancoff factors is needed when one analyzes nuclear reactors loaded with TRISO particle fuel. In these reactor designs, fuel kernels are randomly distributed and shield each other, causing a shadowing effect. The Dancoff factor is a quantitative measure of this effect and is determined by the spatial distribution of fuel kernels. A TRISO fuel particle usually consists of four layers that form a coating region outside the fuel kernel. When fuel particles are loaded in the reactor, the spatial distribution of fuel kernels can be affected by the thickness of the coating region. Therefore, the coating region should be taken into account in the calculation of Dancoff factors. However, the previous model, the single-sphere model, assumes no coating regions in the Dancoff factor predictions. To address this model deficiency, the dual-sphere model is proposed by deriving a new chord length distribution function between two fuel kernels that explicitly accounts for coating regions. The new model is employed to derive analytical solutions of infinite medium, intra-fuel pebble and intra-fuel compact/pin Dancoff factors over a wide range of volume packing fractions of TRISO fuel particles, varying from 2% to 60%. Comparisons are made with the predictions from the single-sphere model and reference Monte Carlo simulations. A significant improvement of the accuracy, over the ranges of packing fractions under current investigation, has been observed. The new model provides a reliable numerical evaluation of Dancoff factors for TRISO particle fuel and can be used as a routine method for the neutronic design and analysis of reactors loaded with TRISO particle fuel

[en] Highlights: • Thermal scattering data are fit using linear least squares regression. • Mesh points are optimally selected from phonon frequency distributions. • New meshes give more accurate fits of thermal data than our previous work. • Coefficient data storage is significantly reduced compared to current methods. - Abstract: In a series of papers, we have introduced a new sampling method for Monte Carlo codes for the low-energy secondary scattering parameters that greatly reduces data storage requirements. The method is based on the temperature dependence of the energy transfer (beta) and squared momentum transfer (alpha) between a neutron and a target nuclide. Cumulative distribution functions (CDFs) in beta and alpha are constructed for a range of temperatures on a mesh of incident energies in the thermal range and temperature fits are created for beta and alpha at discrete CDF probability lines. The secondary energy and angle distributions generated from the fit coefficients showed good agreement with the standard Monte Carlo sampling. However, some discrepancies still existed because the CDF probability mesh values were selected uniformly and arbitrarily. In this paper, a physics-based approach for optimally selecting the CDF probability meshes for the on-the-fly sampling method is introduced, using bound carbon in graphite as the example nuclide. This approach is based on the structure of the phonon frequency distribution of thermal excitations. From the study, it was determined that low (<0.1) and high (>0.9) beta CDF probabilities are important to the structure of the beta probability density functions (PDFs) while very low (<1 × 10"−"4) alpha CDF probabilities are important to the structure of the alpha PDFs. The final meshes contain 200 probability values for both beta and alpha. This results in 14.5 MB of total data storage for the on-the-fly coefficients which are used for any temperature realization. This is a significant reduction in data storage from current methods that require around 25 MB per temperature.

[en] The liver X receptors (LXRs) are transcriptional regulators of lipid homeostasis and may be critical for neurodegeneration and neurogenesis in vivo. However, it remains largely unknown about the role of LXRs and its agonists in the in vitro proliferation of neural progenitor cells (NPCs). Here we revealed for the first time that LXRs were markedly expressed in mouse NPCs and were critical for the in vitro proliferation. LXR agonists GW3965 and LXR623 promoted the proliferation of wildtype NPCs, but not NPCs from LXR double-knockout mice. Mechanistically, phosphorylation of MEK1/2 and ERK1/2 in NPCs was enhanced upon LXR agonist treatment, while abrogation of MEK/ERK phosphorylation by the inhibitors PD98059 and U0126 impaired the proliferation of wildtype NPCs in the presence or absence of LXR agonists. Collectively, our findings suggest that LXR agonists GW3965 and LXR623 can stimulate the NPC proliferation in LXR- and MEK/ERK-dependent manner. - Highlights: • LXRs are expressed and functional in the mouse neural progenitor cells (NPCs). • LXRs are essential for the in vitro proliferation of mouse NPCs. • LXR agonists stimulate the proliferation of wildtype NPCs, but not LXRαβ−/− cells. • Phosphorylation of ERK/MEK was enhanced by LXR agonists in wildtype NPCs. • Inhibition of MEK/ERK phosphorylation impaired the proliferation of NPCs.

[en] The sodium salt of hexasulfated β-cyclodextrin has been synthesized and intercalated into a magnesium-aluminum layered double hydroxide by ion exchange. The structure, composition and thermal decomposition behavior of the intercalated material have been studied by variable temperature X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FT-IR), inductively coupled plasma emission spectroscopy (ICP), and thermal analysis (TG-DTA) and a model for the structure has been proposed. The thermal stability of the intercalated sulfated β-cyclodextrin is significantly enhanced compared with the pure form before intercalation

[en] We perform first-principles calculations to study the energetics and electronic properties of silicene and MoSe_2 heterobilayers (Si@MoSe_2 HBLs). It is found that the silicene is bound to MoSe_2 substrate with a binding energy of −0.56 eV per silicon atom, indicating a weak interaction between two layers. The nearly linear band dispersion character of silicene with a sizable band gap is obtained in Si@MoSe_2, due to the variation of on-site energy induced by MoSe_2 substrate. Remarkably, the band gaps and electron effective mass (EEM) of HBLs can effectively be tuned by interlayer spacing, external electric field, and strains. These findings indicate that Si@MoSe_2 HBLs are promising candidates for high-performance silicene-based FET channel operating at room temperature, in which both finite band gap and high carrier mobility are obtained. - Highlights: • The nearly linear band dispersion of silicene is remained. • The band gaps can be modulated by spacing distance, strain and electric field. • The electron effective mass also demonstrates an ideal adjustable range. • The high carrier mobility can also be preserved

[en] A novel method, Flash Spark Plasma Sintering (FSPS), combining flash sintering and electric field assisted sintering was used to densify B_4C ceramics. B_4C powder was densified up to 99.2% in 1 min with limited grain size increase at 1931 °C under an applied pressure of 15.3 MPa. TEM analysis of the grain structure of the resulting ceramics suggests that the Joule heating and sparking initiated by the pulsed current are highly localized to particle interfaces and contribute significantly to the character of densification. We show that plastic deformation under high current and low pressure is the dominant densification mechanism.