WALMIR MAXIMO TORRES

WALMIR MAXIMO TORRES

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Verification and validation of seven turbulence models for a natural circulation loop under transient conditions
2024 - ANGELO, G.; ANGELO, E.; SCURO, N.L.; TORRES, W.M.; ANDRADE, D.A.
A numerical study of the vertical heater, vertical cooler (VHVC) natural circulation loop (NCL) at IPEN/CNEN-SP was conducted using a three-dimensional and transient mathematical model analyzed with the commercial software ANSYS CFX. The study focused on the stable and single-phase flow regime, with a Rayleigh number ranging from zero to 2.8×108. Seven turbulence models have been benchmarked: Zero Equation, Eddy Viscosity Transport Equation (EVTE), k−ω, k−ɛ, Shear Stress Transport (SST), Reynolds Stress (SSG), and Detached Eddy Simulation (DES). The results of these models were compared against each other and against experimental results obtained specifically for this purpose, focusing on the spatial distribution and temporal evolution of temperature at various points in the natural circulation loop. Among all tested models, the k−ɛ model demonstrated superior performance with the lowest average deviation, exhibiting lower initial turbulence production and buoyancy effects than the more complex models. This behavior suggests that the k−ɛ model is more accurate in predicting temperature distribution and is a better choice for transient flow analysis in natural circulation loops with similar geometries to those presented in this study.
Assessment of the IEA-R1 nuclear reactor using a nonstandard fuel assembly with six fuel plates of the Brazilian Multipurpose Reactor
2024 - SOARES, HUMBERTO V.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; BELCHIOR, ANTONIO; ANDRADE, DELVONEI A. de
In order to qualify the fuel plates of the Brazilian Multipurpose Reactor (RMB), a nonstandard Instrumented Fuel Assembly (IFA) was designed and is being constructed to be burned in the IEA-R1 nuclear research reactor. IFA has fuel plates of different uranium densities (10 fixed fuel plates of 3.0 gU/cm3 – IEA-R1 standard; 6 removable fuel plates of 3.7 gU/cm3 – RMB; and a central aluminum plate). This paper is the first step to demonstrate that IEA-R1 can safely operate with this IFA. To verify the IFA thermal behavior inside the IEA-R1 core during reactor operation and certify the no power peaks occurrence, the power distribution was calculated for each fuel plate. LEOPARD and HAMMER-TECHNION codes were utilized to calculate the core thermal neutron cross section and CITATION code to calculate the core power distribution. Calculations were performed for 5 MW reactor power considering the IFA placed in a core peripheral position. The RMB fuel plates average power was 4.73 % higher compared to IEA-R1 fuel plates. This was expected due to the higher density of uranium in these plates. The power of each IFA fuel plate was compared with a fresh IEA-R1 Fuel Assembly (FA) at the same core position. The power in the IFA hottest plate is only 6.79 % higher than the correspondent IEA-R1 fuel plate. The IFA power distribution was also compared to the hottest FA of the core. The power of each IFA fuel plate was below its correspondent hottest FA fuel plate. In addition, the total IFA power is 18.40 % less than the hottest FA in the core. No significant power peaks occur in the IFA during operation. As future works, thermal–hydraulic calculations will be performed considering this calculated power distribution and no hot spots are expected.
Análise de temperaturas em um elemento combustível do reator de pesquisas IEA-R1 durante evento de perda lenta de vazão com RELAP
2023 - CAMPOS, ROGERIO C. de; BELCHIOR JUNIOR, ANTONIO; SOARES, HUMBERTO V.; UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; ANDRADE, DELVONEI A. de
O código RELAP (Reactor Excursion and Leak Analysis Program) é amplamente utilizado para realizar análises de acidentes em reatores nucleares de potência ou de pesquisa. O presente trabalho apresenta uma simulação do transiente de perda lenta de vazão no núcleo do reator a partir de um modelo com RELAP para o reator de pesquisas IEA-R1 contemplando a piscina, o núcleo do reator, toda tubulação e válvulas do circuito primário, o tanque de decaimento, bomba de circulação principal, trocador de calor e tubulação de retorno à piscina. A modelagem proposta conseguiu representar toda a fenomenologia do acidente, ou seja, o comportamento das temperaturas desde o início da perda de vazão, desligamento do reator, seguida da abertura da válvula de circulação natural até a reversão da direção do escoamento no núcleo do reator. A comparação com resultados experimentais mostrou diferenças de temperaturas de 2,3°C para o fluido e de até 4°C para o revestimento.
Computational fluid dynamics analysis of an open-pool nuclear research reactor core for fluid flow optimization using a channel box
2023 - SCURO, N.L.; ANGELO, G.; ANGELO, E.; PIRO, M.H.A.; UMBEHAUN, P.E.; TORRES, W.M.; ANDRADE, D.A.
A channel box installation in the IEA-R1 research reactor core was numerically investigated to increase fluid flow in fuel assemblies (FAs) and side water channels (SWCs) between FAs by minimizing bypasses in specific regions of the reactor core, which is expected to reduce temperatures and oxidation effects in lateral fuel plates (LFPs). To achieve this objective, an isothermal three-dimensional computational fluid dynamics model was created using Ansys CFX to analyze fluid flow distribution in the Brazilian IEA-R1 research reactor core. All regions of the core and realistic boundary conditions were considered, and a detailed mesh convergence study is presented. Results comparing both scenarios are presented in the percentage of use of the primary circuit pump. It is indicated that 21.4% of fluid bypass to unnecessary regions can be avoided with the channel box installation, which leads to the total mass flow from the primary circuit for all FAs increasing from 68.9% (without a channel box) to 77.6% (with a channel box). For the SWCs, responsible for cooling LFPs, an increment from 9.7% to 22.4%, avoiding all nondesired cross three-dimensional effects, was observed, resulting in a more homogeneous fluid flow and vertical velocities. It was concluded that the installation of a channel box numerically indicates an expressive mass flow increase and homogeneous fluid flow distribution for flow dynamics in relevant regions. This gives greater confidence to believe that lower temperatures, and consequently oxidation effects in LFPs, can be expected with a channel box installation.
RANS-based CFD calculation for pressure drop and mass flow rate distribution in an MTR fuel assembly
2021 - SCURO, N.L.; ANGELO, G.; ANGELO, E.; UMBEHAUN, P.E.; TORRES, W.M.; SANTOS, P.H.G.; FREIRE, L.O.; ANDRADE, D.A.
This work presents a Reynolds-averaged Navier Stokes–based computational fluid dynamics methodology for the calculation of pressure drop and mass flow rate distribution in a material test reactor flat-plate-type standard fuel assembly (SFA) of the IEA-R1 Brazilian research reactor to predict future improvements in newer SFA designs. The results improve the understanding of the origin of fuel plate oxidation due to high temperatures, and consequently, due to the internal flow dynamics. All numerical analyses were performed with the ANSYS-CFX® commercial code. The observed results show that the movement pin decreases the central channel mass flow due to the length of the vortex at the inlet region. However, the outlet nozzle showed greater general influence in the flow dynamics. It should have a more gradual cross-section transition being away from the fuel plates or a squarer-shaped design to get a more homogeneous mass flow distribution. Optimizing both regions could lead to a better cooling condition. The validation of the IEA-R1 numerical methodology was made by comparing the McMaster University’s dummy model experiment with a numerical model that uses the same numerical methodology. The experimental data were obtained with laser Doppler velocimetry, and the comparison showed good agreement for both pressure drop and mass flow rate distribution using the Standard k-ω turbulence model.
Total and partial loss of coolant experiments in an instrumented fuel assembly of IEA-R1 research reactor
2020 - MAPRELIAN, EDUARDO; TORRES, WALMIR M.; BELCHIOR JUNIOR, ANTONIO; UMBEHAUN, PEDRO E.; BERRETTA, JOSE R.; SABUNDJIAN, GAIANE
The safety of nuclear facilities has been a growing global concern, mainly after the Fukushima nuclear accident. Studies on nuclear research reactor accidents such as the Loss of Coolant Accident (LOCA), many times considered a design basis accident, are important for ensure the integrity of the plant. A LOCA may lead to the partial or complete uncovering of the fuel assemblies and it is necessary to assure the decay heat removal as a safety condition. This work aimed to perform, in a safe way, partial and complete uncovering experiments for an Instrumented Fuel Assembly (IFA), in order to measure and compare the actual fuel temperatures behavior for LOCA in similar conditions to research reactors. A test section for experimental simulation of Loss of Coolant Accident named STAR was designed and built. The IFA was irradiated in the IEA-R1 core and positioned in the STAR, which was totally immersed in the reactor pool. Thermocouples were installed in the IFA to measure the clad and fluid temperatures in several axial and radial positions. Experiments were carried out for five levels of uncovering of IFA, being one complete uncovering and four partial uncovering, in two different conditions of decay heat. It was observed that the cases of complete uncovering of the IFA were the most critical ones, that is, those cases presented higher clad temperatures when compared with partial uncovering cases, for the specific conditions of heat decay intensity and dissipation analyzed. The maximum temperatures reached in all experiments were quite below the fuel blister temperature, which is around 500 °C. The STAR has proven to be a safe and reliable experimental apparatus for conducting loss of coolant experiments.
Analytical and experimental analysis on safety related aspects of the RMB research reactor
2020 - BELCHIOR JUNIOR, A.; SANTOS, A.A.C. dos; FREITAS, R.L.; SOARES, H.V.; JUNQUEIRA, F.C.; MANTECON, J.G.; MATTAR NETO, M.; MENZEL, S.C.; TORRES, W.M.; UMBEHAUN, P.E.
This paper presents some numerical and experimental safety related activities developed at the Brazilian Multipurpose Reactor (RMB) project by CNEN research institutes. Brief comments on the models and results are presented with emphasis to their relation to the safe design and operation of the reactor. Thermal-hydraulic analysis for Siphon Breaker of the Core Cooling System (CCS); pools hot water layer; core chimney of CCS and spent fuel transport cask are presented, showing results, advantages, difficulties and drawbacks for each analyzed case. All are very distinct cases, involving phenomena that range from two-phase flow and thermal-stratification to lead melting. Beside the one-dimensional thermal hydraulic system Code RELAP5, Computational Fluid Dynamics (CFD) is shown to play an important role in the analysis being performed as it can detail the flow and temperature fields of complex components and phenomena, which are extremely difficult to model analytically or experimentally. Two experimental circuits designed to test RMB fuel elements performance are also presented.
Preliminary numerical analysis of the flow distribution in the core of a research reactor
2019 - SCURO, NIKOLAS L.; ANGELO, GABRIEL; ANGELO, E.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A. de
The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary three-dimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX® commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that a homogeneous flow distribution for all standard fuel assemblies were found, with 2.7% deviation from the average mass flow. What turned out to be negligible and can be assumed that there is a homogeneous distribution in the core. Complex structures were find in the computational domain. Once known the core flow dynamics, it allows future studies to determine whether the heat flux and temperature conditions abbeys thermal-hydraulic safety criteria.
Lower plenum holes for research reactor core flooding
2019 - MAPRELIAN, EDUARDO; BELCHIOR JUNIOR, ANTONIO; TORRES, WALMIR M.
Modern and high power pool type research reactors generally operate with upward flow in the core. They have a chimney above the core, where the heated fluid is suctioned by the pumps. It passes through the decay tank and is sent to the heat exchangers for the cooling and returns to the core. The pipes inside the reactor pool have passive valves (natural circulation valves) that allow the establishment of natural circulation between the core and the pool for the decay heat removal, when the pumps are inoperative. These valves also have the siphon-breaker function in case of Loss of Coolant Accidents (LOCA), avoiding the pool emptying. In some reactors, these valves are located above the core chimney to facilitate the maintenance. When a LOCA causes a water level below these valves, they loose the natural circulation function. If the water level is the same of the chimney top, the available fluid for the core cooling is only that contained in the chimney and core, and a significant quantity of water in the pool is unavailable for core cooling. To bypass this problem during the reactor design phase, the inclusion of small holes of 10 mm of diameter on the lower plenum lateral side is proposed. These holes will allow a flow path between the pool and the core. Theoretical calculations were performed and analyzed for different drilling configurations: 4, 6 8, and 10 holes. A theoretical analysis of the estimated leakage rate during normal operation and evaporation and replacement rates during a hypothetical LOCA were performed. The calculation results showed that the four configurations analyzed are able to supply the water evaporated from chimney. An experiment is being proposed to validate the theoretical calculations and the considered hypotheses.
Status of the development of a fuel assembly decay heat calorimeter for the IEA-R1 nuclear research reactor
2019 - PRADO, ADELK C.; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; BELCHIOR JUNIOR, ANTONIO; PENHA, ROSANI M.L.
The heat release due to decay of fission products following a nuclear reactor shutdown is important matter for determining cooling requirements as well as for predicting postulated accident consequences. Accurate evaluation of decay heat can also potentially provide independent data for the cross examination of fuel burnup calculations, which is useful where few resources are available for examination of spent fuel. The evaluation of decay heat from unloaded fuel assemblies of the IEA R1 research reactor was proposed in order to seize that opportunity. With that purpose a special measuring device is under development at the Nuclear and Energy Research Institute (IPEN). Since average heat flux as low as 0.1W/cm2 is expected and since decay heat release must be accurately evaluated, the device design had to overcome the difficulties of measuring small amounts of heat released over a large boundary surface. The design had also to ensure the safe cooling of the fuel assemblies and proper radiological protection for the personnel. In view of the tight constraints, a novel design was adopted. The device features a submersible measurement chamber, which allows all measurement procedures to be performed without removing the fuel assemblies from the reactor pool, and an array of semiconductor thermoelectric modules, which provides highly accurate decay power measurements. The assemblage of the device is currently in progress, the main parts have already been acquired or manufactured and key components passed partial tests. Commissioning and main experiments will be performed up to the end of 2019.

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