Maria Auxiliadora F. Veloso

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.

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.

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.

Experimental Investigation of the Onset of Subcooled Nucleate Boiling in an Open-Pool Nuclear Research Reactor

The IPR-R1 TRIGA nuclear research reactor at the Nuclear Technology Development Center (CDTN), in Belo Horizonte (Brazil), is a pool type reactor. The maximum core power is 250 kW cooled by natural circulation of light water and an open surface. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel/cladding (gap) and the cladding to coolant interfaces. The objective of the thermal and hydrodynamic projects of the reactors is to remove the heat safely, without producing excessive temperature in the fuel elements. The regions of the reactor core where boiling occurs at many different power levels can be determined from the heat transfer coefficient data. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Experimental results indicated that subcooled pool boiling occurs at the cladding surface in the reactor core central channels at power levels in excess of 60 kW. However, due to the high heat transfer coefficient in subcooled boiling the cladding temperature is quite uniform along most of the active fuel rod region and do not increase very much with the reactor power. An operational computer program and a data acquisition and signal processing system were developed as part of this research project to allow on line monitoring of the operational parameters.

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

Antonella L. Costa1, Patricia A. L. Reis1, Clarysson A. M. Silva1, Claubia Pereira1, Maria Auxiliadora F. Veloso1, Bruno T. Guerra1, Humberto V. Soares1 and Amir Z. Mesquita2 1Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq 2Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear – CDTN/CNEN Brasil

Neutronic Evaluation of a MHR System to Transmutation of Minor Actinides

The goal is to simulate the modular helium reactor (MHR) core to analyze the neutronic parameters behavior due the insertion of Pu isotopes and minor actinides (MAs) using shuffling scheme without compromising the safety parameters. Initially the core is filled with driver fuel (DF). After the burn-up, these fuels are then reprocessed and used to produce the transmutation fuel (TF). Some cycles after, the core is filled with DF and TF fuels. DF fuel is composed of Pu and Np while TF fuel is a mixture of Pu and MAs. The shuffling scheme was evaluated after each cycle. It was verified that neutronic parameters and isotopic composition reach equilibrium when this scheme is used. The WIMS code was used to perform the simulations and the following neutronic parameters were evaluated: infinite multiplication factor, spectrum hardening, and reactivity temperature coefficients.

Transuranics Transmutation Using Neutrons Spectrum from Spallation Reactions

The aim is to analyse the neutron spectrum influence in a hybrid system ADS-fission inducing transuranics (TRUs) transmutation. A simple model consisting of an Accelerator-Driven Subcritical (ADS) system containing spallation target, moderator or coolant, and spheres of actinides, “fuel,” at different locations in the system was modelled. The simulation was performed using the MCNPX 2.6.0 particles transport code evaluating capture and fission reactions, as well as the burnup of actinides. The goal is to examine the behaviour and influences of the hard neutron spectrum from spallation reactions in the transmutation, without the contribution or interference of multiplier subcritical medium, and compare the results with those obtained from the neutron fission spectrum. The results show that the transmutation efficiency is independent of the spallation target material used, and the neutrons spectrum from spallation does not contribute to increased rates of actinides transmutation even in the vicinity of the target.

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.