Modeling of hybrid fusion/fission systems for development of advanced nuclear fuel cycles

Abstract

ABSTRACT: Two main critical applications of fast neutrons can be performed over nuclear industry, in one hand the necessity of spent fuel´s burning minor actinides and in the other, the possibility of breeding fissile material of fertile isotopes. In a fusion reaction between deuterium and tritium, among other reactions, nuclei yields a 14.5 MeV energy neutron. These very energetic neutrons can easily convert for instance 238U and 232Th isotopes (fertile material) into 239Pu and 233U isotopes (fissile material), and also the destruction of minor actinides such as Pu, Np, Am and Cm which are present in the spent fuel, reducing significantly its radiotoxicity. Nevertheless, economy of neutrons becomes critical as per each neutron generated in the fusion reaction, one neutron is consumed by tritium, 6Li nucleus and a neutron are required to generate tritium in the nuclear reactions, that is why an effective hybrid reactor model is necessary. MCNP and SCALE codes are widely used around the world and in this work are used in conjunction characterizing the flux of neutrons that are generated from a low aspect ratio machine known as tokamak to the rest of zones of the hybrid reactor. A transport code such as ASTRA is also part of this hybrid reactor modeling tasks which can be used in the determination of densities and temperatures of the volume plasma profiles. MCNP main task is about the evaluation of the flux of neutrons given this external neutron source. Once the neutron flux is known, ORIGEN, part of the SCALE code suite for model and simulation of nuclear safety design and analysis, is used to evaluate the creation and destruction of isotopes given the cross sections, an initial composition and the energy-resolved neutron flux. An internal tool was developed to link the nuclear codes used for this purpose with all nuclear data and to build a hybrid reactor model for the simulation. Fusion-Fission HYBrid (FFHYB) is the tool that helps to build an MCNP input reactor model, geometry, data specification, material specification, neutron source definition, and tally specification. The ORIGEN inputs are also built by the tool. Linking and communicating inputs and outputs between these codes is an integral part of the simulation of fuel assemblies irradiated by fast neutrons coming from a fusion plasma. This fusion and fission concept as a hybrid reactor, with a plethora source of fast neutrons, can be performed as a qualitatively tool opening “new” cycles for the transmutations of nuclear fertile material into fissile fuel and for the spent fuel burning actinides. One of the advantages of this concept is the possibility of having a total reprocessing free (ReFree) conversion fuel cycle. 232Th–233U is selected as the case study for this thesis. Virgin fuel assemblies are exposed to tokamak neutron source in the hybrid rector filled with fertile fuel rods, these same assemblies can be then burned into a traditional thermal water reactor, avoiding violating the fuel rods integrity. This “new” technic with a highly capacity to be a proliferation-resistant mode for the production of nuclear fissile fuel, which is a different way from dominated reprocessing path via the standard UREX/PRUREX processes, this can be of a great acceptance to the nuclear industry for the production of fissile material, helping with scarcity of 235U naturally available, augmenting nuclear fuel reserves without throttling the nuclear energy expansion.