Study of aluminothermic slag leaching for uranium and thorium recovery

The need for energy by modern society is increasing. On the other hand, it is necessary to reduce costs and environmental impact. In this perspective, the recovery of uranium present in industrial waste from the processing of naturally occurring radioactive material (NORM) appears as a possible complement to the mining stage of the Nuclear Fuel Cycle. NORM's uranium recovery can reduce environmental liabilities and mineral processing costs (especially blasting, crushing, and grinding). The industrial residue of this study, a type of aluminothermic slag, comes from the metallurgical processing of columbite (niobium and tantalum mineral) and has a content, measured by X-ray fluorescence, of 1.78% of U3O8. This content is higher, for example than those found in Lagoa Real-BA (0.2% in rock) and Santa Quitéria-CE (0.1% in rock). Another material that will be studied is ThO2, which is also present in the slag with a content, measured by X-ray fluorescence, around 3.66%. The process parameters analyzed were pH of the solution, time, granulometry and percentage of solids. The metallurgical recovery of U3O8 reached a maximum value of 71,3% with pH = 1, time of 8 hours, 65 % percentage of solids, and 200 µm of granulometry. The metallurgical recovery of ThO2 reached a maximum value of 69,7% with pH = 1, time of 8 hours, 65 % percentage of solids, and 200 µm of granulometry

Evaluation of the acid leaching technique for recovery of U3O8 and ThO2 in niobium/tantalum slag

This study presented the recovery of uranium oxide and thorium oxide from aluminothermic slag from the metallurgical processing of columbite – a niobium/tantalum mineral with the presence of U3O8 and ThO2 – from a mining-industrial facility. The methodology consisted of sampling, comminution, and leaching using sulfuric acid. The head sample showed contents of (1.78 + 0.14) % for U3O8 and (3.66 + 0.04) % for ThO2. The metallurgical recovery reached values above 80% for the uranium oxide and above 70% for the thorium oxide for pH < 1.5 and process time greater than eight hours

Tempo de ebulição da água em uma piscina de combustível usado com condições adiabáticas e nãoadiabáticas

The Spent Fuel Pool (SFP) of Angra II, from Brazil, has received standard spent fuel (SF) assemblies of Uranium dioxide (UO2) discharged from Pressurized Water Reactors (PWR) of the Nuclear Power Plants (NPP) of Angra since the beginning of its operation. However, in case of using Mixed Oxide (MOX) or Thorium-based fuels, it would require further thermal studies of wet storage. It includes the determination of the water boiling time (Tb) of the SFP in case of breakdown of its external cooling system (ECS). This work presents studies of Tb of a simulated SFP storing mixed SF discharged from PWRs. The types of mixed SF studied include MOX plus UO2, oxide of thorium/uranium (U-Th)O2 plus UO2, and oxide of Thorium/transuranic (TRU-Th)O2 plus UO2. The simulations were implemented in CFX Ansys considering the top of the SFP as either adiabatic or non-adiabatic wall. Tb is considerably higher when the non-adiabatic boundary condition is used.O Pool de Combustível Irradiado (SFP) de Angra II, do Brasil, recebeu desde o início de sua operação conjuntos padrão de combustível queimado (SF) de dióxido de urânio (UO2) descarregado de Reatores de Água Pressurizada (PWR) das Usinas Nucleares (PNPP) de Angra. Entretanto, em caso de utilização de combustíveis à base de óxido misto (MOX) ou de tório, seriam necessários mais estudos térmicos de armazenamento úmido. Inclui a determinação do tempo de ebulição da água (Tb) do SFP em caso de falha de seu sistema de resfriamento externo (ECS). Este trabalho apresenta estudos de Tb de um SFP simulado que armazena SF misturado descarregado de PWRs. Os tipos de SF mistos estudados incluem MOX mais UO2, óxido de tório/urânio (U-Th)O2 mais UO2, e óxido de tório/transurânico (TRU-Th)O2 mais UO2. As simulações foram implementadas no CFX Ansys considerando o topo do SFP como adiabático ou parede nãoadiabática. Tb é consideravelmente maior quando a condição de limite não-adiabático é utilizada.Fil: Pereira de Faria, Fernando. Universidade Federal de Minas Gerais; BrasilFil: Godino, Dario Martin. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones en Métodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones en Métodos Computacionales; Argentina. Universidad Nacional de Rosario. Facultad de Ciencias Exactas Ingeniería y Agrimensura. Escuela de Ingeniería Mecánica; ArgentinaFil: Corzo, Santiago Francisco. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones en Métodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones en Métodos Computacionales; Argentina. Universidad Nacional del Litoral. Facultad de Ingeniería y Ciencias Hídricas; ArgentinaFil: Ramajo, Damian Enrique. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones en Métodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones en Métodos Computacionales; Argentina. Universidad Nacional del Litoral. Facultad de Ingeniería y Ciencias Hídricas; ArgentinaFil: Costa, Antonella Lombardi. Universidade Federal de Minas Gerais; BrasilFil: Pereira, Claubia. Universidade Federal de Minas Gerais; Brasi

Numerical simulation of the open-pool reactor coolant system using a multi-domain approach

The computational simulation of large-scale reactors is currently limited by the high computational cost. The system codes allow addressing these problems, although with the well-known loss of local information. The use of coupling domains to reduce the problems looks like a proper alternative to settle this issue. In the present paper, a multi-domain coupling 3-dimensional/0-dimensional method to solve the thermal hydraulics of the TRIGA Mark I IPR-R1 reactor was implemented into a Finite Volume suite. Despite of the broadly literature about coupling methods, even in the nuclear engineering community, most of them manage with different codes in a fully explicit way. In the other hand, the benefit of solve different domain approaches inside the same software is in the use of monolithic algorithms. The proposed method consists on using 3-dimensional full CFD to simulate the reactor pool and 0-dimensional modelling for the external cooling loop. This is made by implementing a set of ad hoc dynamics boundary conditions to model the momentum and energy balances along the pipeline. This strategy was used to perform long-time steady state simulations of the reactor at the design power of 100 kW as well as for the repowering up to 265 kW. The results demonstrated that the core is efficiently cooled at the higher power without need to increase the coolant mass flow rate of the external system. Moreover, two accidental events were simulated: the first case was the Station Black Out at full power of 265 kW. The results indicated that the loss of the external heat sink led to a slow pool heating, but the core remains being cooled by the natural circulation in the pool. In fact, the mass flow rate through the core is only reduced in 15% by the loss of the external loop circulation. Finally, a large-Loss of Coolant Accident for the operational power of 100 kW and keeping the pump running is performed. In this case, the pool is quickly empty if safety systems do not take action and the core is uncovered after 450 s completely losing the core cooling capacity.Fil: Corzo, Santiago Francisco. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones en Métodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones en Métodos Computacionales; ArgentinaFil: Godino, Dario Martin. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones en Métodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones en Métodos Computacionales; ArgentinaFil: Costa, Antonella Lombardi. Universidade Federal de Minas Gerais; Brasil. Conselho Nacional de Desenvolvimiento Científico e Tecnológico; BrasilFil: Reis, Patricia A. L.. Universidade Federal de Minas Gerais; Brasil. Conselho Nacional de Desenvolvimiento Científico e Tecnológico; BrasilFil: Pereira, Claubia. Universidade Federal de Minas Gerais; Brasil. Conselho Nacional de Desenvolvimiento Científico e Tecnológico; BrasilFil: Ramajo, Damian Enrique. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones en Métodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones en Métodos Computacionales; Argentin

THERMAL MODELING OF THE HTR-10 USING THE RELAP5-3D CODE

ABSTRACT Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10

Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes

A very complex type of power instability occurring in boiling water reactor (BWR) consists of out-of-phase regional oscillations, in which normally subcritical neutronic modes are excited by thermal-hydraulic feedback mechanisms. The out-of-phase mode of oscillation is a very challenging type of instability and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, simulations of out-of-phase instabilities in a BWR obtained by assuming an hypothetical continuous control rod bank withdrawal are being presented. The RELAP5/Mod3.3 thermal-hydraulic system code coupled with the PARCS/2.4 3D neutron kinetic code has been used to simulate the instability phenomenon. Data from a real BWR nuclear power plant (NPP) have been used as reference conditions and reactor parameters. Simulated neutronic power signals from local power range monitors (LPRM) have been used to detect and study the local power oscillations. The decay ratio (DR) and the natural frequency (NF) of the power oscillations (typical parameters used to evaluate the instabilities) have been used in the analysis. The results are discussed also making use of two-dimensional plots depicting relative core power distribution during the transient, in order to clearly illustrate the out-of-phase behavior

GB -A PRELIMINARY LINKING CODE BETWEEN MCNP4C AND ORIGEN2.1 -DEN/UFMG VERSION

ABSTRACT Nowadays it is possible to perform burnup simulation in a detailed 3D geometry and a continuous energy description by the Monte Carlo method. This paper describes an initial project to create and verify a connection code to link Origen2.1 (Oak Ridge National Laboratory) and MCNP4C (Los Alamos National Laboratory). Essentially the code includes point depletion capability to the MCNP code. The incorporation of point depletion capability is explicit and can be summarized by three steps: 1-Monte Carlo determines reaction rates, 2-the reaction rates are used to determine microscopic cross sections for depletion equations, 3-solution of depletion equations (given by Origen2.1) determines number densities for next MCNP step. To evaluate the initial version of the program, we focused on comparing the results with one of the major Monte Carlo burnup codes: MCNPX version 2.6.0. The input files for all codes share the same MCNP geometry, nuclear data library and core thermal power. While simulating 75 time steps at 800kw of a Heat Pipe Power System model, we have found that the codes generate very similar results. The neutron flux and criticality value of the core agree, especially in the begin of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB)

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 (n,γ) and fission (n,f) 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