A micro heteregenous reprocessed fuel spiked with thorium in a PWR fuel element considering (TRU-Th) cycle was simulated using three different configurations and different fissile materials that varied from 6.0% to 7.0%. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a burned PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment, which was remained in the cooling pool for five years and then reprocessed using UREX+ technique. The keff and plutonium generation during the burnup were evaluated and compared with the standard fuel. This study was performed using the SCALE 6.0.

One of the most important parts of a hybrid reactor is the cladding because it should withstand high temperatures, neutrons with high energy, high neutron flux, as well as provide the first security contention. Besides, the material should have good mechanical properties to remove the heat. Although, the cladding material choice will have a great influence on criticality calculations. In previous works on ADS there is no cladding used, therefore in this paper it is tested different cladding materials based on SS-316, ODS, T91 and 15-15Ti used in nuclear reactors, to study the variations on the fuel depletion and variations on the neutronic parameters. The results using the cladding are compared with the one obtained without using it. The best material choice is based on the neutronic parameters that presents the closest behavior to the ADS simulated without clad.

Stress Corrosion Cracking (SCC) in a general term describing stressed alloy fracture that occurs by crack propagation in specifically environments, and has the appearance of brittle fracture, yet it can occur in ductile materials like AISI 304L used in internal components of Boiling Water Reactors (BWR). The high levels of oxygen and hydrogen peroxide generated during an operational Normal Water Condition (NWC) promotes an Electrochemical Corrosion Potential (ECP), enough to generate SCC in susceptible materials. Changes in water chemistry have been some of the main solutions for mitigate this degradation mechanism, and one of these changes is reducing the ECP by the injection of Hydrogen in the feed water of the reactor; this addition moves the ECP below a threshold value, under which the SCC is mitigated (-230mV vs SHE). This paper shows the characterization by Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD) and Raman Spectroscopy of the oxide film formed in to a crack propagated during a Rising Displacement Test method (RDT), on Hydrogen Water Chemistry (HWC) conditions: 20 ppb O2, 125 ppb H2, P=8MPa, T=288°C, using a CT specimen of austenitic stainless steel AISI 304L sensitized. The characterization allowed identifying the magnetite formation since an incipient way, until very good formed magnetite crystals.

This paper studies the performance of (U, Pu)C fuel in a hexagonal assembly of a GFR (Gas Fast Reactor). The SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation version 6.0) code was used in the calculation. The goal is to evaluate the behavior of the infinite multiplication factor (kinf) for a heterogeneous assembly model using four nuclear data libraries: V6-238, V7-238, ENDF/B-VI.8 and ENDF/B-VII.0. The burnup of (U, Pu)C was performed by the TRITON-6 module, and the isotopic concentrations were evaluated during the cycle. The present work comprises calculations at Zero Power and Full Power condition. This study intends to achieve more information about different Fast Reactors.

The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of rising the burning and maintaining the safety of nuclear plants, is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and a low absorption cross-section for thermal neutrons. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with burnable poison variations, the impact on multiplication factor and reactivity coefficients to both claddings: zircaloy and silicon carbide. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This code system is widely accepted and used worldwide for safety analysis, and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR.