Development of neutronic models for fusion-fission hybrid reactors

Los reactores híbridos fusión-fisión (HFF) constituyen una alternativa interesante para la generación de energía en gran escala sin emisión de gases de efecto invernadero y para la eliminación de residuos de alta actividad producidos por los reactores de fisión. Los HHF potencian las ventajas relativas de la fusión y fisión y minimizan sus desventajas. La fusión produce muchos neutrones pero relativamente poca energía mientras que la fisión produce mucha energía pero necesita un flujo constante de neutrones para sostenerse. El diseño básico de un HFF consiste de un reactor de fusión, que proporciona un flujo constante de neutrones, rodeado por un manto subcrítico de material físil. El presente proyecto apunta a desarrollar la capacidad de calcular la neutrónica de un reactor híbrido fusión-fisión, tomando de partida la fuente de neutrones de un reactor nuclear de fusión tipo Tokamak. El proyecto se desarrolla en colaboración entre la Sección Fusión Nuclear y Física de Plasmas y el Departamento Física de Neutrones de la Comisión Nacional de Energía Atómica.Fusion-fission hybrid (FFH) reactors are an interesting alternative for the generation of electricity without the emission of greenhouse gases, and for the transmutation of spent nuclear fuel produced in fission reactors. FFHs multiply the advantage of separate fusion and fission reactors and minimize their disadvantages. Fusion naturally produces neutrons and a relatively low amount of energy, whereas fission produces more energy per reaction but requires a constant flux of neutrons. The basic design of a FFH consists on a fusion reactor surrounded by a subcritical blanket of fissile material.The present project aims to develop models for the neutronic calculation of fusion-fission hybrid reactors, taking as a start point the neutron source produced from a Tokamak-type fusion reactor. The project is a collaboration between the Neutron Physics Department and the Nuclear Fusion and Plasma Physics Section of the National Atomic Energy Commission

Vibrational spectra of light and heavy water with application to neutron cross section calculations

The design of nuclear reactors and neutron moderators require a good representation of the interaction of low energy (E < 1 eV) neutrons with hydrogen and deuterium containing materials. These models are based on the dynamics of the material, represented by its vibrational spectrum. In this paper, we show calculations of the frequency spectrum for light and heavy water at room temperature using two flexible point charge potentials: SPC-MPG and TIP4P/2005f. The results are compared with experimental measurements, with emphasis on inelastic neutron scattering data. Finally, the resulting spectra are applied to calculation of neutron scattering cross sections for these materials, which were found to be a significant improvement over library data.Fil: Marquez Damian, Jose Ignacio. Comisión Nacional de Energía Atómica. Gerencia del Area de Energía Nuclear. Instituto Balseiro; Argentina. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Malaspina, David Cesar. Northwestern University; Estados Unidos. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Granada, Jose Rolando. Comisión Nacional de Energía Atómica. Gerencia del Area de Energía Nuclear. Instituto Balseiro; Argentina. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; Argentin

Atomic scale Monte-Carlo simulations of neutron diffraction experiments on stoichiometric uranium dioxide up to 1664 K

The neutron transport in nuclear fuels depends on the crystalline structure of materials when neutron energies lie below a few eV. For that purpose, the theoretical formalism that describes the neutron elastic and inelastic scatterings by crystals has been implemented in the CINEL processing tool in order to provide temperature-dependent neutron cross sections usable by the Monte-Carlo code TRIPOLI4®. The performances of the Monte-Carlo calculations are illustrated with the analysis of neutron powder diffraction data on UO2 measured up to 1664 K with the D4 and D20 diffractometers of the Institute Laue–Langevin (Grenoble, France). The comparison of the experimental and simulated pair distribution functions confirms the unusual decrease of the U–O atomic distances with increasing temperature when an ideal fluorite structure (Fm3̄m space group) with harmonic atomic vibrations is assumed over the full temperature range. The flexibility of the CINEL code allowed to explore disorder or anharmonic oxygen vibrations in the Fm3̄m space group by using either a four-site model with a relaxation term or a structure factor equation with a non-zero anharmonic third-cumulant coefficient. As none of these models succeeded to improve the agreement with the experiments, recent works that propose other local crystalline symmetries for UO2 at elevated temperatures were investigated with the CINEL code. The case of the Pa3̄ symmetry is briefly discussed in this paper.Fil: Xu, S.. Commissariat à l'énergie atomique et aux énergies alternatives. Institut de REcherche sur les Systèmes Nucléaires pour la production d’Energie bas carbone; FranciaFil: Noguere, G.. Commissariat à l'énergie atomique et aux énergies alternatives. Institut de REcherche sur les Systèmes Nucléaires pour la production d’Energie bas carbone; FranciaFil: Desgranges, L.. Commissariat à l'énergie atomique et aux énergies alternatives. Institut de REcherche sur les Systèmes Nucléaires pour la production d’Energie bas carbone; FranciaFil: Marquez Damian, Jose Ignacio. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Patagonia Norte; Argentina. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentin

Multilevel transport solution of LWR reactor cores

This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations

Calculation of kinetic parameters βeff and Λ with modified open source Monte Carlo code OpenMC(TD)

This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction βeff and neutron generation time Λ. The modified code, OpenMC(Time-Dependent) or OpenMC(TD), was then used to calculate the effective delayed neutron fraction by using the prompt method, while the neutron generation time was estimated using the pulsed method, fitting Λ to the decay of the neutron population. OpenMC(TD) is intended to serve as an alternative for the estimation of kinetic parameters when licensed codes are not available. The results obtained are compared to experimental data and MCNP calculated values for 18 benchmark configurations.Fil: Romero Barrientos, J.. Comision Chilena de Energia Nuclear; Chile. Universidad de Chile; ChileFil: Marquez Damian, Jose Ignacio. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Patagonia Norte; Argentina. European Spallation Source; SueciaFil: Molina, F.. Comision Chilena de Energia Nuclear; Chile. Universidad Andrés Bello; ChileFil: Zambra, M.. Comision Chilena de Energia Nuclear; Chile. Universidad Diego Portales; ChileFil: Aguilera, P.. Comision Chilena de Energia Nuclear; Chile. Universidad de Chile; ChileFil: López Usquiano, F.. Comision Chilena de Energia Nuclear; Chile. Universidad de Chile; ChileFil: Parra, B.. Instituto de Física Corpuscular; EspañaFil: Ruiz, A.. Comision Chilena de Energia Nuclear; Chile. Universidad de Chile; Chil

The Bariloche Neutron Physics Group Current Activities

Our group has evolved around a small accelerator-based neutron source (ABNS), the 25 million electron Volt (MeV) linear electron accelerator at the Bariloche Atomic Centre. It is dedicated to applications of neutronic methods to tackle problems of basic sciences and to technological applications. Among these, the determination of total cross section of a material as a function of neutron energy by means of transmission experiments for thermal and sub-thermal neutrons is very sensitive to the geometric arrangement and movement of the atoms, over distances ranging from the 'first-neighbour scale' up to the microstructural level or 'grain scale'. This also allowed to test theoretical models of calculated cross sections and scattering kernels. Interest has moved from pulsed neutron diffraction towards deep inelastic neutron scattering (DINS), a powerful tool for the determination of atomic momentum distribution in condensed matter and for non-destructive mass spectroscopy. In recent years non-intrusive techniques aimed at the scanning of large cargo containers have started to be developed with this ABNS, testing the capacity and limitations to detect special nuclear material and dangerous substances in thick cargo arrangements. More recently, the use of the ever-present “bremsstrahlung” radiation has been recognized as a useful complement to instrumental neutron activation, as it permits to detect other nuclear species through high-energy photon activation. The facility is also used for graduate and undergraduate students experimental work within the frame of Instituto Balseiro Physics and Nuclear Engineering courses of study, and also MSc and PhD theses work.Fil: Mayer, Roberto Edmundo. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; ArgentinaFil: D'Amico, N. M. B.. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; ArgentinaFil: Granada, Jose Rolando. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Dawidowski, Javier. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Santisteban, Javier Roberto. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Blostein, Juan Jeronimo. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Tartaglione, Aureliano. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Rodriguez Palomino, Luis Alberto. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Marquez Damian, Jose Ignacio. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Instituto Balseiro; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; ArgentinaFil: Sepúlveda Sosa, C.. Comision Chilena de Energia Nuclear; Chil

Multilevel acceleration of neutron transport calculations

Nuclear reactor design requires the calculation of integral core parameters and power and radiation profiles. These physical parameters are obtained by the solution of the linear neutron transport equation over the geometry of the reactor. In order to represent the fine structure of the nuclear core a very small geometrical mesh size should be used, but the computational capacity available these days is still not enough to solve these transport problems in the time range (hours-days) that would make the method useful as a design tool. This problem is traditionally solved by the solution of simple, smaller problems in specific parts of the core and then use a procedure known as homogenization to create average material properties and solve the full problem with a wider mesh size. The iterative multi-level solution procedure is inspired in this multi-stage approach, solving the problem at fuel-pin (cell) level, fuel assembly and nodal levels. The nested geometrical structure of the finite element representation of a reactor can be used to create a set of restriction/prolongation operators to connect the solution in the different levels. The procedure is to iterate between the levels, solving for the error in the coarse level using as source the restricted residual of the solution in the finer level. This way, the complete problem is only solved in the coarsest level and in the other levels only a pair of restriction/interpolation operations and a relaxation is required. In this work, a multigrid solver is developed for the in-moment equation of the spherical harmonics, finite element formulation of the second order transport equation. This solver is implemented as a subroutine in the code EVENT. Numerical tests are provided as a standalone diffusion solver and as part of a block Jacobi transport solver.M.S.Committee Chair: Stacey, Weston M.; Committee Co-Chair: de Oliveira, Cassiano R.E.; Committee Member: Hertel, Nolan; Committee Member: van Rooijen, Wilfred F.G

Towards a covariance matrix of CAB model parameters for H(H2O)

Preliminary results on the uncertainties of hydrogen into light water thermal scattering law of the CAB model are presented. It was done through a coupling between the nuclear data code CONRAD and the molecular dynamic simulations code GROMACS. The Generalized Least Square method was used to adjust the model parameters on evaluated data and generate covariance matrices between the CAB model parameters