Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/ PuO2 fuel designs which have an excellent performance record for normal operation and most transients. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U3Si2). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LW R design (I2S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a U.S. DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S- LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I S-LWR design adopts an integral configuration and a fully passive emergency decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and reactor physics behaviour of the I2S-LWR with U3Si2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I2S-LWR. The results were found to be in a good agreement (e.g. < 200 pcm in reactivity) among the compared codes, giving confidence that the WIMS/PANTHER reactor physics package can be reliably used in modelling LWRs with ATFs.This is the final version of the article. It first appeared from Springer via http://dx.doi.org/10.1051/epjn/201601

Numerical stability of the predictor-corrector method in Monte Carlo burnup calculations of critical reactors

a b s t r a c t Monte Carlo burnup codes use various schemes to solve the coupled criticality and burnup equations. Previous studies have shown that the simplest methods, such as the beginning-of-step and middle-ofstep constant flux approximations, are numerically unstable in fuel cycle calculations of critical reactors. Here we show that even the predictor-corrector methods that are implemented in established Monte Carlo burnup codes can be numerically unstable in cycle calculations of large systems

Review article: Use of ultrasound in the developing world

As portability and durability improve, bedside, clinician-performed ultrasound is seeing increasing use in rural, underdeveloped parts of the world. Physicians, nurses and medical officers have demonstrated the ability to perform and interpret a large variety of ultrasound exams, and a growing body of literature supports the use of point-of-care ultrasound in developing nations. We review, by region, the existing literature in support of ultrasound use in the developing world and training guidelines currently in use, and highlight indications for emergency ultrasound in the developing world. We suggest future directions for bedside ultrasound use and research to improve diagnostic capacity and patient care in the most remote areas of the globe

Full-Core Coupled Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Analysis of Low-Enriched Uranium Nuclear Thermal Propulsion Reactors

Nuclear thermal propulsion is an enabling technology for future space missions, such as crew-operated Mars missions. Nuclear thermal propulsion technology provides a performance benefit over chemical propulsion systems by operating with light propellants (e.g., hydrogen) at elevated engine chamber conditions. Therefore, nuclear thermal propulsion reactor cores exhibit high propellant velocities and elevated propellant and fuel temperatures, subsequently leading to relatively high thermal stresses and geometrical deformation. This paper details the numerical approach to solve the thermo-elastic equations, which was implemented into the recently developed ntpThermo code. In addition, this paper demonstrates the extension of the Basilisk multiphysics framework to perform full-core coupled neutronic, thermal-hydraulic, and thermo-mechanical analysis of nuclear thermal propulsion reactors. The analyses demonstrate and quantify thermo-mechanical feedback, which for the investigated cases, acted to reduce maximum fuel temperatures and pressure drop across the fuel element channels. Thermo-mechanical feedback had a significant impact on the mass flow distribution within the reactor core and, thus, a substantial impact on solid-material temperatures and stresses, but only a minor impact on reactivity and local power distributions. Sensitivity studies revealed that the friction factor correlation applied to perform the analysis has a significant impact on the pressure drop across the fuel element channels. The most important observation of this research is the importance of incorporating the thermo-mechanical feedback within an integrated multiphysics solution sequence to enable the consistent design of future nuclear thermal propulsion systems

APPLICATION OF A CUSTOM DEPLETION FRAMEWORK TO THE PREDICTION OF NEUTRON FLUX DISTRIBUTION THROUGH DEPLETION

Previous works by the authors have introduced the spatial flux variation method (SFV) for predicting the changes in neutron flux due to a change in material compositions. In order to remove a full transport solution at the end-of-step, this work presents a framework responsible for computing macroscopic cross sections after a depletion event. These end of-step cross sections are estimators of changes in neutron loss and production, and enable the prediction of neutron flux using only information obtained from a single beginning of-step transport solution. The framework reads in all relevant data needed to model the depletion system, including one-group cross sections and effective fission yields to reproduce the problem using an external solver. The framework also supports extrapolating microscopic cross sections in order to rebuild the end-of-step macroscopic cross sections needed for the flux prediction. Results indicate that the SFV method is not adversely effected by the external depletion solution, and can be implemented alongside an existing transport-depletion framework

Multiple vortex with different topological charge generated by means of SLM

We present techniques to generate multiple vortex with different topological charge by means of diffractive optical elements. Analytical formulae to describe the Fresnel and Fraunhofer diffraction of the Gaussian beam by a helical axicon (HA) are introduced. The relations are presented as a series of the hypergeometric functions. By setting the axicon parameter equal to zero, the solution for the HA changes to that for the spiral phase plate (SPP). The performance of the aforesaid optical elements is tested both through computer simulation and by experiments using a spatial light modulator, in view of optical miroparticle manipulation

A Modeling and Neutron Diffraction Study of the High Temperature Properties of Sub-Stoichiometric Yttrium Hydride for Novel Moderator Applications

Low-enriched-uranium (LEU) reactor systems utilize moderators to improve neutron economy. Solid yttrium hydride is one of the primary moderator candidates for high-temperature (>700 °C) nuclear reactor applications. This is due to its ability to retain hydrogen at elevated temperatures compared to other metal hydrides. For reactor modeling purposes, both neutronic and thermos-mechanical modeling, several high-temperature properties for sub-stoichiometric yttrium hydride (YH2−x) are needed. In this paper, we present an atomistics and a neutron diffraction study of the high-temperature properties of Y  and  YH2−x. Specifically, we focus on the thermal lattice expansion effects in yttrium metal and yttrium hydride, which also govern bulk thermal expansion. Previously reported physical and mechanical properties for sub-stoichiometric yttrium hydride at ambient conditions are expanded using lattice dynamics to take into account high-temperature effects. Accordingly, an array of newly generated properties is presented that enables high-fidelity neutronics, and thermomechanical modeling. These properties include various elastic moduli, thermal expansion parameters for yttrium and yttrium hydride, and single-phase (YH2−x) and two-phase (Y + YH2−x) density as a function of stoichiometry and density

A Modeling and Neutron Diffraction Study of the High Temperature Properties of Sub-Stoichiometric Yttrium Hydride for Novel Moderator Applications

Low-enriched-uranium (LEU) reactor systems utilize moderators to improve neutron economy. Solid yttrium hydride is one of the primary moderator candidates for high-temperature (&gt;700 &deg;C) nuclear reactor applications. This is due to its ability to retain hydrogen at elevated temperatures compared to other metal hydrides. For reactor modeling purposes, both neutronic and thermos-mechanical modeling, several high-temperature properties for sub-stoichiometric yttrium hydride (YH2&minus;x) are needed. In this paper, we present an atomistics and a neutron diffraction study of the high-temperature properties of Y&nbsp; and &nbsp;YH2&minus;x. Specifically, we focus on the thermal lattice expansion effects in yttrium metal and yttrium hydride, which also govern bulk thermal expansion. Previously reported physical and mechanical properties for sub-stoichiometric yttrium hydride at ambient conditions are expanded using lattice dynamics to take into account high-temperature effects. Accordingly, an array of newly generated properties is presented that enables high-fidelity neutronics, and thermomechanical modeling. These properties include various elastic moduli, thermal expansion parameters for yttrium and yttrium hydride, and single-phase (YH2&minus;x) and two-phase (Y + YH2&minus;x) density as a function of stoichiometry and density