Abderrafi M. Ougouag

Differential displacement kerma cross sections for neutron interactions in Si and GaAs

The cross-section processing code NJOY has been modified to calculate flux-averaged partially integrated differential displacement kerma cross sections or displacement kerma matrix elements for neutron interactions. These, along with total displacement kerma cross sections, have been calculated using ENDF/B-V and ENDL-84 data files for Si, Ga, and As. These displacement kerma matrices for Si and GaAs allow calculations of the distribution of displacement energy between displacement cascades. The tabulation of kerma cross sections for Si constitutes a complete revision of the data used in current standard practices. Another contribution is the tabulated kerma cross sections for GaAs. >

Carrier removal and changes in electrical properties of neutron irradiated GaAs

The changes in electrical properties of n‐GaAs as a result of irradiations with fast neutrons have been studied. Epitaxial layers doped with Si at concentrations in the range 1.35×1015 to 1.599×1016 cm−3 were irradiated with reactor neutron fluences up to 1.31×1015 cm−2. When the changes in carrier concentration, Hall mobility, and resistivity were more than 25% of their initial values, nonlinear dependence on neutron fluence was apparent. New theory is proposed which explains the changes in electrical properties in terms of rates of trapping and release of charges. A theoretical relationship is derived for the change in carrier concentration as a function of neutron fluence and doping level. A linear relationship between neutron fluence and Fermi level shift was found to be consistent with the observed changes in carrier concentration.

ILLICO-HO: A Self-Consistent Higher Order Coarse-Mesh Nodal Method

A self-consistent nodal method has been developed that directly computes the in-node flux shapes. The method renders the use of an approximation for the transverse leakages no longer necessary. The...

ILLICO: A Nodal Neutron Diffusion Method for Modern Computer Architectures

AbstractA nodal multigroup neutron diffusion method for modern computer architectures has been developed and implemented in the ILLICO code. Vectorization and parallelization strategies that are successful in speeding up modern nodal computations on commercially available supercomputers have been identified and applied. Realistic three-dimensional benchmark problems can be solved in the vectorized mode in <0.73 s (33.86 Mflops). Vector-concurrent implementations are shown to yield speedups as high as 9.19 on eight processors. These results demonstrate that modern nodal methods, such as ILLICO, can preserve essentially all of their speed advantages (demonstrated on scalar computers) over finite difference methods. Several ways of treating two-dimensional reactor problems with nonsquare (“jagged”) boundaries as rectangular domain problems are presented and their effectiveness evaluated. They result in nonnegligible performance improvements and can be devised so as to preserve the physics of the initial problem.

Investigation of the Impact of Simple Carbon Interstitial Formations on Thermal Neutron Scattering in Graphite

Abstract In both the prismatic and pebble bed designs of very high temperature reactors, the graphite moderator is expected to reach exposure levels of 1021 to 1022 n/cm2 over the lifetime of the reactor. This exposure results in damage to the graphite structure. Studies of the thermal properties of irradiated graphite show changes in the thermal conductivity and (to a lesser extent) the heat capacity at fluences <1021 n/cm2. In graphite, these properties depend on the behavior of atomic vibrations (phonons) in the solid. Therefore, it can be expected that alterations in the phonon behavior that would produce changes in these properties would have an impact on the thermal neutron scattering behavior of that material. In this work, an atomistic ab initio investigation is performed to explore the potential impact of simple carbon interstitial formations on the inelastic thermal neutron scattering behavior of graphite. Using the VASP/PHONON code system, graphite supercells were modeled with and without either a single carbon interstitial or a di-interstitial (C2) molecule between the graphite planes. This resulted in the production of the phonon frequency spectra for these structures. From the phonon data, the inelastic thermal neutron scattering cross sections were generated, using the NJOY code system, at temperatures of 300 and 1200 K. A comparison of the generated cross sections shows that accounting for the interstitials in the calculations affects the cross sections mainly in the energy range from 0.01 to 0.1 eV.

Pressure vessel steel embrittlement monitoring by magnetic properties measurements

The magnetic properties of specimens of one heat of A533B nuclear pressure vessel grade steel have been examined in the as-received condition and after neutron irradiation to various fluence levels up to 4 [times] 10[sup 18] cm[sup [minus]2] (E > 0.1 MeV) in the University of Illinois Advanced TRIGA reactor core at two temperatures, approximately 120 and 260[degrees]C. The effect of some heat treatments was also investigated. The magnetic properties were measured by an automated hysteresis curve tracing method using a miniature transformer which incorporated the specimens in its core. Changes in magnetic hysteresis energy loss were correlated with neutron fluence in the case of certain irradiated specimens, and with microhardness measurements in the case of heat treated specimens. At the higher irradiation temperatures, no significant changes in either the magnetic hysteresis properties or the microhardness were noted for the present fluences. The relationship between the observed magnetic properties response and irradiation-induced embrittlement is discussed.

Anisotropic Elastic Resonance Scattering Model for the Neutron Transport Equation

Abstract The resonance scattering transfer cross section has been reformulated to account for anisotropic scattering in the center of mass of the neutron-nucleus system. The main innovation over previous implementations is the relaxation of the ubiquitous assumption of isotropic scattering in the center of mass and the actual effective use of scattering angle distributions from evaluated nuclear data files in the computation of the angular moments of the resonant scattering kernels. The formulas for the high-order anisotropic moments in the laboratory system are also derived. A multigroup numerical formulation is derived and implemented into a module incorporated within the NJOY nuclear data processing code. An ultrafine-energy-mesh cross-section library was generated using these new theoretical models and then was used for fuel assembly calculations with the PARAGON lattice physics code. The results obtained indicate that this new model makes a significant difference to predictions of reactivity, multigroup fluxes, and isotopic inventory during depletion.

Microscale Heat Conduction Models and Doppler Feedback

The objective of this project is to establish an approach for providing the fundamental input that is needed to estimate the magnitude and time-dependence of the Doppler feedback mechanism in Very High Temperature reactors. This mechanism is the foremost contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic (TRISO) coated particles. Therefore, its correct prediction is essential to the conduct of safety analyses for these reactors. Since the effect is directly dependent on the actual temperature reached by the fuel during transients, the underlying phenomena of heat deposition, heat transfer and temperature rise must be correctly predicted. To achieve the above objective, this project will explore an approach that accounts for lattice effects as well as local temperature variations and the correct definition of temperature and related local effects.

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

The Cell Analytic-Numerical Method for Solution of the Two-Dimensional Advection-Dispersion Equation

Abstract A new numerical method, called the Cell Analytic-Numerical Method, is developed for solution of the Two-Dimensional Advection-Dispersion Equation by using a transverse integration technique followed by analytical solution of the transverse- integrated local equations. Continuity of the mass flux is then used to obtain a set of coupled tridiagonal equations which can be efficiently solved. This new method is demonstrated to have high accuracy, even when applied on coarse meshes, and to have minimal grid orientation error.