Brazilian Multipurpose Reactor Research Articles

Stochastic modeling of a neutron imaging center at the Brazilian Multipurpose Reactor

Neutron imaging is a non-destructive technique for analyzing a wide class of samples, such as archaeological or industrial material structures. In recent decades, technological advances have had a great impact on the neutron imaging technique, which has meant an evolution from simple radiographs using films (2D) to modern tomography systems with digital processing (3D). The 5 MW research nuclear reactor IEA-R1, which is located at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) in Brazil, possesses a neutron imaging instrument with 1.0×106n/cm2s in the sample position. IEA-R1 is over 60 years old and the future of neutron science in Brazil, including imaging, will be expanded to a new facility called the Brazilian Multipurpose Reactor (RMB, Portuguese acronym), which will be built soon. The new reactor will house a suite of instruments at the Neutron National Laboratory, including the neutron imaging facility, viz., Neinei. Inspired by recent author’s works, we model the Neinei instrument through stochastic Monte Carlo simulations. We investigate the sensitivity of the neutron imaging technique parameter (L/D ratio) with the neutron flux, and the results are compared to data from the Neutra (PSI), Antares (FRM II), BT2 (NIST) and DINGO (OPAL) instruments. The results are promising and provide avenues for future improvements.

Simulating Araponga - The High-Resolution Diffractometer of Brazilian Multipurpose Reactor

The Brazilian Multipurpose Reactor (RMB) is a fundamental project that aims to turn Brazil into a self-sufficient country in the production of radioisotopes and radiopharmaceuticals to supply the Unified Health System (SUS) as much as the private institutions. In addition, the RMB project describes other applications as irradiation and testing of nuclear fuels and structural material analysis, for instance. There are many techniques in the project to study structural aspects of materials, where neutron diffraction represents one of the priorities for implementation. This technique will take place mainly on two diffractometers on Thermal Neutron Guide 1 (TG1), namely Araponga, a high-resolution diffractometer, and Flautim, a high-intensity diffractometer. In this work, we study the performance of the Araponga diffractometer through McStas simulations with input produced by the MCNP code of the RMB core. We investigate the neutron flux values considering a state-of-art high-resolution diffractometer, and the results are promising since some simulated scenarios present values compatible with high-intensity devices.

Critical Velocity Experimental Assessment in Flat Plate Fuel Element for Nuclear Research Reactor

Aluminum-coated plates, containing a uranium silicide (U3Si2) meat dispersed in an aluminum matrix, are commonly used in the fuel elements of Material Testing Reactors (MTRs). These fuel elements are typically comprised of narrow channels formed by parallel flat plates, which allow coolant flow to remove the heat of fission reactions. It is important to mention that the thickness of the plates is much smaller than their width and height. The high flow rates needed to ensure efficient fuel-element cooling may cause fuel-plate mechanical failures due to instability induced by the flow in the channels. In the case of critical velocity, excessive permanent deflections of these plates can cause blockage of the flow channels and lead to overheating. An experimental facility that simulates a plate-like fuel element with three coolant channels was developed for this work. The test-section dimensions were based on the Fuel Element design of the Brazilian Multipurpose Reactor (RMB), project being coordinated by the National Commission of Nuclear Energy (CNEN). Experiments were performed to reach Miller's critical velocity condition. This critical condition was reached at 14.5 m/s leading to consequent plastic deformation of the fuel plates

The Neglected Role of Education in Brazilian Research Regarding Nuclear Theme versus Scientific and Technological Knowledge, Perception and Risk Communication.

Studies of risk perception and risk communication concerning the nuclear area are quite common in scientific literature. However, though numerous studies on this topic point to the importance of scientific and technological knowledge in facilitating the reduction of perceived risks, there are few papers that effectively discuss the role of education in risk communication on the subject, particularly in Brazil. The objective of the present work was to reflect on this problem by analyzing the results obtained from two case studies on the implementation of the Brazilian Multipurpose Reactor, in the city of Iperó, in the state of São Paulo, Brazil. As was verified in the investigation, most participants of the study had high perception of risk with respect to the implementation of the reactor. Nevertheless, although important in the project of the reactor implementation, it was verified that the state, municipal and community schools of the city of Iperó had not participated, in the public hearings nor had they discussed the issue inside the school community, until the moment of the research.

Manufacturing LEU-foil annular target in Brazil

Molybdenum-99 is the most important isotope because its daughter isotope, technetium-99m, has been the most used medical radioisotope. The primary method used to produce Mo-99 derives from the fission of U-235 incorporated in so-called irradiation targets. Two routes are being developed to make Mo-99 by fissioning with low enriched uranium (LEU) fuel. The first adopts UAlx-Al dispersion plate targets. The second uses uranium metal foil annular targets. The significant advantage of uranium foil targets over UAlx-Al dispersion targets is the high density of uranium metal. This work presents the experience obtained in the development of the uranium metal annular target manufacturing steps. An innovative method to improve the procedure for assembling the uranium foil on the tubular target was presented. The experience attained will help the future production of Mo-99 in Brazil through the target irradiation in the Brazilian Multipurpose Reactor (RMB).

Increasing productivity in the manufacture of UAl2–Al dispersion-plate targets for Mo-99 production

Molybdenum-99 is the most important isotope because its daughter isotope, technetium-99 m, has been the most widely used medical radioisotope. The primary method employed to produce Mo-99 derives from the fission of U-235 incorporated in so-called irradiation targets. Pushed by the international Mo-99 crisis that occurred in 2009/2010, Brazil has decided to construct a new research reactor, the Brazilian Multipurpose Reactor (RMB), to produce this vital radioisotope to meet the Brazilian domestic demand. As part of this effort, it has been developed the process for manufacturing the target to be used in the production of Mo-99 by nuclear fission. The low enriched uranium (LEU) aluminide with the predominant phase UAl2 was the starting material. The picture-frame technique was used to clad UAl2–Al briquette with aluminum to obtain plate-type targets. It was developed an innovative method that allows increasing the productivity of this type of target based on multi-core rolling. A thermomechanical treatment was designed to get targets composed basically of a mixture of UAl3/UAl4 that are the required phases for a proper radiochemical dissolution after irradiation. The manufacturing process proved to be suitable for domestic production of targets, fulfilling the specification to produce Mo-99 in the Brazilian Multipurpose Reactor.

Study of new routes for purification of fission 99Mo

99mTc is the most applied medical radioisotope in the world, especially for cancer diagnosis procedures. It is provided by 99Mo radioactive decay, which is one of the fission products from the uranium irradiation in nuclear reactors. At the main production plants, the 99Mo chemical processing may be lined up in several steps to separate it from other fission products according to the features of the used targets or the local requirements as well. In this work, two routes of 99Mo purification, MR1, and MR2, were purposed as an alternative to be set up in the Brazilian Multipurpose Reactor project (BMR). The MR1 route was performed by three chromatographic columns packed with Dowex 1x8 resin, Chelex resin, and alumina, respectively. The route MR2 was carried out also by chromatography applying two columns filled with Dowex 1x8 and alumina respectively, but including a sublimation process performed in a tubular oven under programmed conditions. The final yield for the MR1 route was 84.4 % and the overall time process was about 7 hours, whereas the MR2 route reached 75.3 % in 9 hours.

A diffractometer project for Brazilian Multipurpose Reactor (RMB): McStas simulations and instrument optimization

High-resolution diffractometer is one of the first instruments of the set of 15 priority neutron scattering instruments to be installed at the Brazilian Multipurpose Reactor (RMB). A basic project of this instrument consists of the existence of three guides through which neutrons pass from source to sample to guarantee maximum neutron flux at the sample position. In this study, we investigate guide geometry performance considering fixed diffractometer geometry and spatial arrangement. Comparisons between different guide shapes and supermirrors are performed using software based on the Monte Carlo method, McStas. Our conclusion shows that a better solution is splitting the initial flux into two different guides to obtain the maximum flux at the sample position.

RELAP5 Modeling of a Siphon Break Effect on the Brazilian Multipurpose Reactor

This work presents the thermal-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermal-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.

RMB experimental program on the hydrodynamical behavior of fuel assemblies

The Brazilian Multipurpose Reactor - RMB is a 30 MW pool type research reactor, that uses Materials Testing Reactor - MTR type fuel assemblies. It has a 5x5 square array core with 23 fuel assemblies and two in-core irradiation positions, operating with upward flow and average velocities nearly 10 m/s in the fuel plates channels. The IEA-R1 is a 5 MW pool type research reactor, which also has a 5x5 square array core with 19 standard fuel assemblies, four control fuel assemblies and a central beryllium irradiation device. It operates with downward flow nearly 1.8 m/s in the channels. In order to verify and provide data and information about the dynamical behavior of fuel assemblies under nominal and critical conditions, the experimental circuit ORQUÍDEA is being designed. This circuit will permit upward and downward flow and dynamical behavior of the fuel assemblies and its parts will be tested and verified. Flow rate, temperature, pressure and differential pressure transducers are the instruments of the circuit. Endurance and critical flow velocity tests will be performed. The COLIBRI experimental circuit is being designed to make tests that allow the studies of the fluid-structure phenomenology of fuel plates similar to those of the RMB fuel assemblies when subjected to high flow velocities, which can induce pressure differences between the channels formed by the fuel plates. This work presents a preliminary design for the ORQUÍDEA and COLIBRI experimental circuits to be built at the Instituto de Pesquisas Energéticas e Nucleares - IPEN of the Comissão Nacional de Energia Nuclear - CNEN.

Neutron Guide Building instruments of the Brazilian Multipurpose Reactor (RMB) project

A growing community of scientists has been using neutrons in the most diverse areas of science. In order to meet the researchers demand in the areas of physics, chemistry, materials sciences, engineering, cultural heritage, biology and earth sciences, the Brazilian Multipurpose Reactor (RMB) will provide 3 thermal guides and 3 cold guides, with the installation of several instruments for materials characterization. In this study, we present a standard design requirement of two primordial instruments, namely Sabi'a and Araponga. They are, respectively, cold and thermal neutron instruments and correspond to a Small-Angle Neutron Scattering (SANS) and High-Resolution Powder Neutron Diffractometer (HRPND) to be installed in the Neutron Guide Building (N02) of RMB . To provide adequate flux for both instruments, we propose here an initial investigation of the use of simple and split guides to transport neutron beams to two different instruments on the same guide. For this purpose, we use Monte Carlo simulations utilizing McStas software to check the efficiency of thermal neutron transport for different basic configuration and sources. By considering these results, it is possible to conclude that the split guide configuration is, in most cases, more efficient than cases that use transmitted neutron beams independently of source. We also verify that the employment of different coating indexes for concave and convex surfaces on curved guides is crucial, at least on simulated cases, to optimise neutron flux (intensity and divergence) and diminish facility installation cost.

Transient analysis of the Brazilian Multipurpose Reactor by the coupled neutronics and thermal hydraulics code NTHC1

Dynamical behavior of the Brazilian Multipurpose Reactor (RMB) is investigated by using the coupled neutronics and thermal hydraulics code NTHC1, which consists of the neutron point kinetics (NPK) equations with six groups of delayed neutrons, the heat conduction equations in a fuel plate, and the energy transport equation in a coolant subchannel. The Serpent 2 Monte Carlo code is employed to calculate the RMB kinetic parameters of the NPK equations and steady state power distribution. Improved lumped formulations are adopted for heat conduction in the transversal direction of the fuel and cladding, while the fluid energy equation is discretized along the subchannel by finite difference method. The resulting system of ordinary differential equations is solved by numerical method. The NTHC1 code was verified against the IAEA 10 MW MTR benchmark problems, showing good agreement with literature data. The simulation results demonstrated that RMB operates safely under the postulated initiating events (PIE) of loss of flow and reactivity insertion.

Two-phase CFD simulation of research reactor siphon breakers: A verification, validation and applicability study

During a nuclear reactors design phase many possible accidents are postulated. The main accident that could occur in an open pool reactor, such as the Brazilian Multipurpose Reactor (RMB) is the Loss-Of-Coolant Accident (LOCA). This type of accident occurs when the primary cooling piping breaks and the pressure gradient induces the refrigerant (water) to flow through this pipe by siphon effect, draining the pool. This accident can be avoided with a siphon breaking mechanism, which could consist of an orifice in the coolant pipe at a level above the minimum required water level in the pool. If the pipe breaks, air will enter through this orifice, interrupting the siphon flow. However, tests have shown that the mechanism in most cases does not stop the water flow immediately, causing a reduced water level. The siphon break phenomenon is a complex two-phase transient flow. Experimental studies of the siphon break are very expensive and time-consuming tasks. In this case, it would be appropriate to make a Computational Fluid Dynamics (CFD) analysis of this phenomenon, allowing a detailed evaluation of the flow characteristics in a simpler and cheaper way than an experiment. However, the complexity of this phenomenon imposes difficulties to its computer modeling and its accuracy must be assessed. In this study, a methodology is proposed for the simulation of orifice type siphon breaker for research reactors. In order to properly evaluate the numerical methodology, an experimental model available in the literature was simulated and a Verification and Validation (V&V) procedure was performed based on the Grid Convergence Index (GCI) method. Numerical uncertainties due to mesh were evaluated and estimated for other simulations with different siphon breaker orifice diameters to assess the applicability of the proposed methodology. Results showed good agreement with experimental results with reasonable uncertainty range. The proposed simple numerical uncertainty estimation procedure aims to encourage the discussion towards an uncertainty quantification culture in the CFD community.

Neutronic analysis of the fuel loaded irradiation loop device of the RMB Multipurpose Brazilian Reactor

The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work neutronic calculations have been carried out in order to determine parameters to the Loop basic design. Safe position determined to be 52.7 cm displaced from the core, where linear heat rate (q’) < 1 W/cm fission power is released in the rig loading fissile sample. Since irradiation rig moves during normal operation of the reactor, displacement of the samples must be investigated in terms of perturbation in the core. Results confirmed efficacy of the 12 cm heavy water layer designed to suppress reactivity effects in the core. Linear heat rate in the fissile sample was calculated in order to estimate predefined power ramps of 50–350 W/cm, information necessary to the project of the displacement device. Ramp lengths of 9.59 cm–13.22 cm were found depending on the peak factor considered for the sample rod and the number of beryllium rows arranged in the reflector.

Neutronic analysis of a U–Mo–Al fuel and europium as burnable poison

This work presents the neutronic analysis of the U–Mo–Al dispersion fuel concerning uranium density increase and shows comparisons relatively to the U3Si2–Al fuel. The U3Si2–Al uranium density varied from 3.0 to 5.5gU/cm3 while that of U–Mo–Al fuel varied from 4.0 to 7.5gU/cm3. The molybdenum mass content in the former case varies from 7% to 10% in mass. Here, it is also proposed the utilization of burnable poison nuclides in the U–Mo–Al fuel meat. Since the fuel is metallic, gadolinium and europium were chosen as candidates to cope with this task. A recently developed cell code at IPEN (HRC) composed of the coupling of the codes HAMMER-TECHNION for the cell analysis, ROLAIDS for the actinide self-shielding calculations and CINDER-2 for the actinide and fission transmutation was employed for the neutronic analyses of U–Mo–Al. The simulated reactor core was similar to the one of RMB (Brazilian Multipurpose Reactor) composed of an array of 5×5 positions with 23 fuel elements and 2 aluminum blocks. A second analysis of the europium case employed the SERPENT code in an explicit core modeling. The burnup calculations were performed considering a power of 30MW during three cycles of RMB and 30days. The analyses revealed that the molybdenum content has a great impact in the core reactivity due to its high absorption cross section. A value of 7% was found adequate for the molybdenum mass content. The analyses also reveal that europium is a better burnable poison than gadolinium for the core cycle length and power level under consideration. However, for the U–Mo–Al case, k∞ increases up to a maximum value and decreases afterwards. This is a striking result since the reactivity for the U–Mo–Al fuel does not increase steadily as verified for the U3Si2–Al case. Beyond a certain uranium density, the reactivity will decrease making useless the addition of more uranium.

Rejeitos radioativos gerados na produção de Mo-99 por fissão nuclear

The Brazilian Government intends to construct a facility for molibdenium-99 production by nuclear fission, among other installations, at the Brazilian Multipurpose Reactor enterprise. This study aims to investigate in which extension the activity of the radioactive waste generated in the process is enhanced if different mass of the U-235 and irradiation time of the target is employed to produce the same quantity of Mo-99. The simulation was conducted using the computational code Origen S®, assuming that the target is composed of UAlx alloy with 19.9% enriched U-235 and irradiated in a thermal neutron flux of 1x10 14 n.cm -2 .s -1 , to produce 185 TBq of Mo-99 per week. The results obtained for 3 to 21 days of continuous irradiation of targets with different mass of U-235 show that the amount of radioactive waste generated, in terms of total activity, does not appear to be a dominant aspect to be considered in the choice of Mo-99 production parameters.

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.

Around the World

This week, take a look ahead at science in 2013. Exciting areas to watch include the Curiosity rover, the Large Hadron Collider, the Atacama Large Millimeter/submillimeter Array, the Brazilian Multipurpose Reactor, the Intergovernmental Panel on Climate Change, and more.