Claubia Pereira

Neutronic evaluation of the non-proliferating reprocessed nuclear fuels in pressurized water reactors

Abstract Alternative reprocessing methods with low grade decontamination and uranium—plutonium coextraction characteristics are proposed for increasing proliferation resistance and permitting recovered fissile materials utilization in LWR reactors. This work brings out a preliminary neutronic study of fuels reprocessed by Coprocessing and AIROX techniques.

Nuclear fuel loading pattern optimisation using a neural network

Abstract An algorithm to optimise the fuel loading pattern (LP) in nuclear reactors was developed using an artificial neural network (ANN) to generate arrangements for the fuel in the core. The core parameters were calculated with the WIMS-D4 and CITATION-LDI2 codes, and the minimization of the maximum power peaking factor (FP max ) was used to choose the best arrangements. To verify the algorithm a PWR reactor with approximately 1/3 reprocessed fuel loaded was considered. The neutronic performance of the obtained arrangements and the efficiency of the implemented method were analysed. Several configurations were found for the core presenting better characteristics than the reference configuration adopted, so indicating the viability of the developed methodology. The algorithm was applied to a core considering part of the loading with reprocessed fuels, however this technique can be used for standard loadings.

Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

Abstract The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR–R1 reactor are calculated. The idea is to obtain the systematic error for k ∞ for this methodology comparing the calculated and the experimental results.

Flux and dose rate evaluation of iter system using MCNP5

The International Thermonuclear Experimental Reactor (ITER) will perform Deuterium-Tritium (DT) plasma experiments and the neutrons production rate at 14.1 MeV will achieve the level of 1013 n.cm-2.s-1. In this work, the neutron flux and the dose rate during ITER operation has been calculated using the one-dimensional model of the Monte Carlo code MCNP5 and the FENDL/MC-2.1 nuclear data library. The neutron flux and dose rate associated during normal ITER operation were determined along the radial machine direction. Calculations for two different types of concrete compositions were performed to investigate the impact of the bioshield filling materials on the dose rate estimation. The results show that the dose rate level near to the outer wall of the tokamak hall is close to the allowable limit dose. Taking into account the use of large boron concentration in the biological shield concrete (2.9% weight fraction), it was obtained a dose rate reduction by one order of magnitude. The dose rate is dominated by the secondary gamma ray. The magnitude of the dose rate on the outside hall of bioshield during normal ITER operation can not be considered low in accordance with the result found in the simulation performed in this work, i.e., 1 µSv.h-1.

Study of an ADS Loaded with Thorium and Reprocessed Fuel

Accelerator-driven systems (ADSs) are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods was for production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use of without the initial requirement of enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations). The multiplication factor () evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use of and reprocessed fuel allowed production without the initial requirement of enrichment.

Neutron production evaluation from a ADS target utilizing the MCNPX 2.6.0 code

Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by particle accelerator. These particles are injected into a target for the neutrons production by spallation reactions. This target is of utmost importance for an ADS, representing the coupling of the accelerator and the sub-critical core. The determination of optimal materials for these targets is fundamental for the design of an ADS. The main characteristic of an ideal target is the high production of neutrons per incident proton. In this work are shown results for the neutron production of various types of targets using the MCNPX 2.6.0 code. Furthermore, it is performed a comparative study of transport models to describe the spallation reactions available in this code.

Non-proliferating reprocessed nuclear fuels in pressurised water reactors: Fuel cycle options

Abstract Many alternatives have been studied in the attempt to burn Plutonium produced in the last decades. Among these, there are the reprocessing of the Plutonium and the insertion of Thorium in thermal reactors. Nevertheless, the proliferation risk is always associated with conventional reprocessing techniques, such as PUREX. Hence, alternative reprocessing methods with low-grade decontamination and uranium-plutonium co-extraction characteristics have been proposed. Besides that, the insertion of Thorium in fuels coming from PUREX has shown good results. Therefore, based in these data, fuels coming from alternative techniques of reprocessing such as Coprocessing and AIROX, have been analysed, with and without the insertion of Thorium. The goal is study fuel options with good neutronic and environmental performance.

Fast Accelerator Driven Subcritical System for Energy Production Using Burned Fuel

Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by a particle accelerator. These particles are injected into a target for neutrons production by spallation reactions. The neutrons are then used to maintain the fission chain in the subcritical core. The aim of this study is to investigate the nuclear fuel evolution of a lead cooled accelerator driven system used for energy production and high-level waste transmutation. The fuel studied is a mixture based upon 232Th-233U and Pu-MA extracted from PWR spent. The target is a lead spallation target and the core is filled with a hexagonal lattice. In order to reduce the negative reactivity caused by the presence of protactinium, moderator is not used. In this work is used the Monte Carlo code MCNPX 2.6.0, that presents the depletion/burnup capability, combining a ADS source and kcode-mode. The keff evolution, the neutron energy spectra in the core and the nuclear fuel depletion during the burnup are evaluated. Keywords: ADS, thorium, MCNPX.

A neutronic evaluation of the Americium and Neptunium co-insertion in UO2 fuel

In this work the neutronic performance of a core of UO2 fuel with the Americium and the Neptunium co-insertion in a standard PWR (pressurized water reactor) core is evaluated. The homogeneous mode is used and changes in the moderator to fuel volume ratio (Vm/Vf) are evaluated. For the simulation the WIMS-D5 code was utilised.

Analysis of Loss of Flow Events on Brazilian Multipurpose Reactor Using the Relap5 Code

This work presents the thermal hydraulic simulation of the Brazilian multipurpose reactor (RMB) using a RELAP5/MOD3.3 model. Beyond steady state calculations, three transient cases of loss of flow accident (LOFA) in the primary cooling system have been simulated. The RELAP5 simulations demonstrate that after all initiating events, the reactor reaches a safe new steady state keeping the integrity and safety of the core. Moreover, a sensitivity study was performed to verify the nodalization behavior due to the variation of the thermal hydraulic channels in the reactor core. Transient calculations demonstrate that both nodalizations follow approximately the same behavior.