Coolant Temperature Profile Research Articles

Transpiration cooling with phase change by functionally graded porous media

Transpiration cooling using liquid coolants is considered one of the most promising thermal protection techniques for new generation aerospace vehicles. The leading edges are subjected to a very high and non-uniform heat flux. Unfortunately, few coolants can flow through the part near the stagnation point using traditional porous structures, which will cause ablation. Unlike conventional fabrication methods, additive manufacturing is capable of directly tailoring porous structures with well-controlled pore features and functional graded pore topology, thereby optimizing the coolant distribution and improving the cooling performance. Smoothly graded pore structures can be fabricated using additive manufacturing technology. We experimentally assessed the transpiration cooling performance of uniform and graded porosity porous plates to compare the coolant distributions, temperature profiles, and features of the two-phase region. Graded porous media exhibited much better cooling performance than uniform pore topology. Through computations, the graded porosity improved the imbalanced coolant distribution and minimized the hot-spot temperature under a non-uniform heat flux. Importantly, the temperature profile at the outlet should match with the local flow resistance in porous media and the heat flux at the outlet. The temperature and phase change zone were highly dependent on the topology of the porous media, heat flux profile applied to the outlet, and coolant mass flow rate.

A multilayered approach to scale-up forced convection-based freezing of human induced pluripotent stem cells

This work presents a multilayered model-based approach to scale-up slow and forced convection-based freezing of human induced pluripotent stem cells. The approach combines a hybrid single-cell freezing model (covering the cell and cryovial layers) with a computational fluid dynamics model (covering the freezer layer). The proposed approach can calculate the cell survival rate considering spatial and time-dependent temperature profiles within a freezer unit. The approach was first demonstrated on a single cryovial problem where it was found that the inlet coolant velocity affected the cell survival rate when fast cooling rates were applied. The freezing of 235 cryovials was then studied, where 18 operation conditions consisting of six cooling rates (three constant and non-constant cooling rates, respectively) and three inlet coolant velocities were investigated. The multilayered approach captured the spatial heterogeneity in the cell survival rates for a given inlet coolant velocity and temperature profile.

Coupled Analysis of Thorium-based Fuels in the High-Performance Light Water Reactor Fuel Assembly

One of the six selected concepts to be part of Generation IV nuclear reactors is the Supercritical Light Water Cooled Reactor. The High-Performance Light Water Reactor (HPLWR) is the European version and it is a very promising design. In recent years, interest in the study of thorium-based fuel cycles has been renewed and its possibilities for current LWRs have been evaluated. The use of thorium-based fuels will be fundamental in the future sustainability of nuclear energy, since in addition to its abundance in nature, thorium has an important group of advantages. In this paper, performance of thorium-based fuels in the typical fuel assembly of the HPLWR reactor is evaluated, using a computational model based on CFD and Monte Carlo codes for the neutronic/thermal-hydraulic coupled analysis. The volumetric power density profiles, coolant temperature profiles, fuel temperature profiles and others are compared with those obtained for standard UO 2 fuel. When the thorium-based fuels are used, the obtained infinite multiplication coefficients are smaller than the value obtained when UO 2 is used, since the 232 Th isotope has a lower contribution to the multiplicative properties of the medium than 238 U. As a result, a difference of approximately 12 000 pcm was observed. The results verified that the HPLWR is a thermal reactor with a hard spectrum. There are no notable changes in the neutron spectrum if the mass fraction of thorium is slightly varied. With coupled analysis, the potential benefits of the utilization of thorium-based fuels were verified. Moreover, a significant temperature decrease by 136 K on the center line of the fuel elements was observed. When the mass fraction of thorium increases in the oxides mixture, the weighted average temperature on the fuel elements decreases .

VALIDATION OF COOLANT THERMAL RESPONSE IN A TRANSIENT FINITE DIFFERENCE THERMAL TRANSPORT MODEL WITH APPLICATIONS TO FAST SPECTRUM REACTORS

Transient behavior in nuclear reactors is important in accidents and with reactivity control systems that are driven by thermal feedback. Here, we describe a transient finite difference model for a pin cell system. The fidelity of the model is shown by validation against the thermocouple measurements of the CABRI BI1 experiment and the Safety Analysis System-Sodium Fast Reactor model of the experiment. In the BI1 experiment, a sodium-cooled mixed oxide fuel pin was subject to a loss of flow transient to coolant boiling within a sodium test loop positioned in the center of the CABRI research reactor. Comparisons to the initial steady-state coolant temperature profile, coolant temperature profile at twenty seconds into the transient, and at four axial locations within the coolant show agreement of the simple model with the experimental results better than or similar to those of the Safety Analysis System-Sodium Fast Reactor model. The model can be used to determine the thermal response times of coolant in fast reactors currently operating or in the design phase when subject to loss of flow accidents or other transients. Here, we investigate the difference in coolant thermal response for metal fueled and mixed oxide fueled sodium fast reactors when subject to transient overpower and loss of flow events. Additionally, we determine the effect of pin pitch on outlet coolant temperatures during the overpower event. Finally, we return to the CABRI experiment and show the importance of porosity in fuel temperature calculations.

Numerical analysis of thermal stratification dependence on mass flow rate of coolant during LOFA in pools of CLEAR-S

During the start-up or operation of pool type lead cooled fast reactor, it is probable that loss of flow transient appears from forced circulation or mass flow rate less than that during forced circulation, due to any flow blockage, station blackout or pump malfunction. In this type of scenario temperature distribution and movement of thermal stratification layer will be different from that during loss of flow transient from forced circulation. Computational fluid dynamics (CFD) transient analysis of three-dimensional model of CLEAR-S was performed in this research with the use of commercial code FLUENT to analyze the influence of initial coolant mass flow rate on flow and coolant temperature profile in different constituents of engineering validation facility (CLEAR-S) during loss of flow transient. Firstly, steady state was simulated at mass flow rate equal to 100%, 50% and 25% of the rated mass flow rate respectively. Than loss of flow transient was simulated after started from the steady state of 100%, 50% and 25% of the rated mass flow rate as t = 0sec respectively to understand the movement and temperature distribution of thermal stratification within the pools of CLEAR-S. Results showed that after loss of flow transient from 100%, 50% and 25% mass flow rate, LBE in the hot pool was distributed into five to six layers of different temperatures and temperature of LBE on both sides of stratification layer increased with the decrease of initial mass flow rate. As time progressed hot pool stratification layer proceeded upward and width of top most layer of LBE with high temperature reduced which ultimately vanished after giving birth to a new layer at hot pool bottom. Stratification initiated from DHR outlet in the cold pool after loss of flow transient from 100% mass flow rate. Thermal stratification was observed to be started at heat exchanger (HX) outlet during loss of flow transient initiated form 50% and 25% of mass flow rate. Temperature of LBE on both sides of stratification layer increased with the decrease of initial cold pool mass flow rate of CLEAR-S. With the proceeding of time Stratification layer raised upward and top most layer vanished.

URANS simulations of the single-phase mixing in a 4*4 LWR fuel bundle with mixing vanes, validated against a sub-channel experiment

A relatively uniform coolant temperature profile at the outlet of reactor core, resulted from good mixing behavior inside fuel assemblies, is desirable for better safety and economic performance. In order to enhance mixing between adjacent sub-channels, various kinds of mixing vanes have been designed and attached to the edge of spacer grids. Both experimental and computational investigations have been ongoing on this topic for decades. However due to the complexity of the geometrical configuration of the spacer grids and mixing vanes, the flow field in the rod bundle involves complicated mechanism, and is not yet fully understood by researchers.A sub-channel mixing test facility SUBFLOW was designed by PSI jointly with ETH Zurich. The test section is an adiabatic 4*4 rod bundle at atmospheric pressure. Both single-phase and two-phase mixing experiments were carried out and valuable data has been obtained. This paper presents the numerical prediction of the single-phase mixing at a Reynolds number of 28000. Unsteady Reynolds Averaged Navier-Stokes (URANS) simulations were performed on a combination of hexahedral and trimmed meshes. A passive scalar source was introduced as an indicator of the mixing behavior. Various turbulence models have been studied. Lateral mixing of passive scalar, velocity profiles and swirl behavior along the channel have been investigated. Fluctuations induced by the mixing vanes were analyzed. The results showed that among the tested methods only low-Reynolds number model v2f with unsteady solver was capable of obtaining quantitatively correct lateral mixing rate and swirl decay rate along the channel, and capturing the fluctuations of the flow field with qualitative agreement with experimental data.

Introductory measurements on the physical model of the VVER 440 nuclear reactor II. fuel cell assembly

Qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel cell assembly outlet and the mean coolant temperature profile in the thermocouple plane is required for safe and effective loading of nuclear fuel cells. Physical model of the VVER 440 nuclear reactor fuel cell assembly serves to analyze the influence of the coolant mass flow on the coolant velocity and temperature profiles at the plane of the thermocouple position in the fuel cell assembly.

Computational simulation of thermal hydraulic processes in the model LMFBR fuel assembly

The aim of this study was to verify a developed software module on the experimental fuel assembly with partial blockage of the flow section. The developed software module for simulation of thermal hydraulic processes in liquid metal coolant is based on theory of anisotropic porous media with specially developed integral turbulence model for coefficients determination. The finite element method is used for numerical solution. Experimental data for hexahedral assembly with electrically heated smooth cylindrical rods cooled by liquid sodium are considered. The results of calculation obtained with developed software module for a case of corner blockade are presented. The calculated distribution of coolant velocities showed the presence of the vortex flow behind the blockade. Features vortex region are in a good quantitative and qualitative agreement with experimental data. This demonstrates the efficiency of the hydrodynamic unit for developed software module. But obtained radial coolant temperature profiles differ significantly from the experimental in the vortex flow region. The possible reasons for this discrepancy were analyzed.

CFD model of a moderator tank for a pressure vessel PHWR nuclear power plant

The CFD thermal hydraulic simulation of a Pressurized Heavy Water Reactor (PHWR) moderator tank was performed. Heating was achieved by considering two heat sources, conduction/convection through the coolant channels and neutron thermalization. For the first method, a suitable 2D model for coupling the 3D with a 1D code to solve the in-channel coolant was implemented. For thermalization, a volumetric heat source was estimated. Due to the relevance of natural and forced convection in the heat transfer, the Boussinesq approach was assessed for high Rayleigh number flow before it was used to model the moderator tank. The results clarified the complex flow behavior and the temperature field. As expected from design, the thermal power transferred from coolant to moderator is not directly related to the fission power distribution but to the axial coolant temperature profile (nearly the same for all channels) and the non-homogeneous moderator temperature profile. The former was as expected for high-Rayleigh natural convection systems.

Investigations on the Influence of Flow Migration on Flow and Heat Transfer in Oblique Fin Microchannel Array

In order to scrutinize the coolant mass distribution and its effect to the heat transfer in oblique fin microchannel array, extensive numerical studies are performed on planar oblique fin configuration. Full-domain simulations using common-flow down (CFD) approach are employed to provide better insights into the flow distribution, flow stability, and heat transfer performance at a global level. The flow field and temperature profile analysis shows that nonuniform coolant distribution and coolant migration occur in the oblique fin microchannel, and the heat transfer performance for both edges of the heat sink is affected due to changing secondary flow rate. However, the flow migration does not affect the local coolant velocity and temperature profiles significantly in the middle region (0.2 < Z′ < 0.8). Meanwhile, it is also found that Reynolds number affects the coolant migration, the stability of the fluid flow, and heat transfer performance significantly. Higher Reynolds number increases the percentage of secondary flow rate and, hence, enhances the heat transfer for fin surfaces in secondary channels.

Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel

The axial power and coolant-temperature distributions in a fuel channel of the Generation IV pressure-tube super-critical water-cooled reactor (PT-SCWR) are found using coupled neutronics-thermal-hydraulics calculations. The simulations are performed for a channel loaded with a fresh, 78-element Th-Pu fuel assembly. Neutronics calculations are performed using the DONJON diffusion code using two-group homogenized cross sections produced using the lattice code DRAGON. The axial coolant temperature profile corresponding to a certain axial linear heat generation rate is found using a code developed in-house at University of Ontario Institute of Technology (UOIT). The effect of coolant density, coolant temperature, and fuel temperature variation along the channel is accounted for by generating macroscopic cross sections at several axial positions. Fixed-point iterations are performed between neutronics and thermal-hydraulics calculations. Neutronics calculations include the generation of two-group macroscopic cross sections at several axial positions, taking into account local parameters such as coolant temperature and density and average fuel temperature. The coolant flow rate is adjusted so that the outlet temperature of the coolant corresponds to the SCWR technical specifications. The converged axial power distribution is found to be asymmetric, resembling a cosine shape skewed toward the inlet (reactor top).

Subchannel Analysis of Wire Wrapped SCWR Assembly

Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can obtain a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) the pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump effectively; (3) the wire wrap pitch has significant effect on the flow in the assembly. SmallerHwire/Drodwill result in stronger cross flow a more uniform coolant temperature profile, and also a higher pressure drop.

Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO2–PuO2 was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions.

Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maâmora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.

Experimental studies on heat and mass transfer performance of a coiled tube absorber for R134a-DMAC based absorption cooling system

Absorber is an important component in vapor absorption refrigeration system and its performance has greater influence in overall efficiency of absorption machines. Falling film heat and mass transfer in an absorber is greatly influenced by fluid properties, geometry of heat exchanger and its operating parameters. This paper presents on the results of experimental studies on the heat and mass transfer characteristics of a coiled tube falling film absorber, using 1,1,1,2-Tetrafluroethane(R-134a) and N-N Dimethyl Acetamide (DMAC) as working fluids. The effects of film Reynolds number, inlet solution temperature and cooling water temperature on absorber heat load, over all heat transfer coefficient and mass of refrigerant absorbed are presented and discussed. Normalized solution and coolant temperature profiles and refrigerant mass absorbed along the height of absorber are also observed from the experimental results. The optimum over all heat transfer coefficient for R-134a–DMAC solution found to be 726 W/m2K for a film Reynolds number of 350. The R-134a vapour absorption rate is maximum in the normalized coil height of 0.6 to 1.

Transient Behavior of Low Enriched Uranium Silicide Miniplate Fuel under a Triplet Configuration.

This paper describes the results of a transient experiment using low enriched (≤20wt% 235U)uranium silicide research reactor miniplate fuel. Three silicide minlplate fuel specimens assembled in a triplet configuration with a water gap of 2.38 mm were pulsed in the Nuclear Safety Research Reactor (NSRR) with an energy deposition of 78cal/g·fuel. The results obtained in this study are summarized as follows: 1. Under a restricted condition of coolant cross flow, the mid-plate in the triplet configuration (pulsed at 78cal/g·fuel) showed an extremely high peak cladding surface temperature (PCST), 475°C, which was higher than that of the other two plates (173 and 192°C). The temperature of coolant in the water gap increased rapidly to 140°C while that at the non-gap locations increased to about 50°C. Despite the poor cooling conditions, especially at the mid-plate, the three silicide miniplate fuels did not fail mechanically. Owing to the comparatively slow quenching rate from PCST and/or the small magnitude of the temperature drop during quenching, the fuels might not be subjected to the local tensile stress necessary to cause through-plate cracking.2. The bowing of the miniplate fuel caused the 2.38 mm water channel gap to be narrowed to 48% of its maximum. The uneven coolant temperature profiles in the water gaps between the miniplates might be a cause for the bowing.

Dynamic Behaviour of Fixed Bed Catalytic Reactors

Characterizing dynamic behaviour depends on identifying all the significant factors which must be included in models which are used to describe the system. As operation at higher performance is demanded the level of approximation assumes greater importance. At the same time there is a need to examine the way in which reduced models can be developed for use in control. This paper examines these issues with reference to the multitubular fixed bed reactor and in particular looks at how the coolant temperature profiles influence the dynamics. It specifically examines how the temperature profiles affect the inverse response both in terms of the extent and location in the reactor and the way in which disturbances are propagated. Consideration is also given to the effect of changes in reactant flow rate and the way in which it affects the inverse response.

Three-Dimensional Thermal-Hydraulic Analysis of Wire-Wrapped Rods in Liquid-Metal Fast Breeder Reactor Core Assemblies

A study is presented to determine the detailed coolant velocity and temperature profile around the entire rod circumference in liquid-metal fast breeder reactor (LMFBR) core assemblies as well as the detailed radial and circumferential temperature profile in the rod. The digital computer code FATHOM-360 developed to perform the above calculations is described. Fuel, radial blanket, and control assembly rods (both wire-trapped and bare) can be analyzed. Coolant, cladding, and fuel (or absorber) temperature profiles are calculated for uniform and nonuniform heat generation (i.e., accounting for power skew across the pellet) in the rod. Temperature distributions can be calculated for both concentric and eccentric positions of the pellet with respect to the fuel rod cladding. Typical examples of the calculational capabilities of the code are presented. Such capabilities are needed for a reliable design of LMFBR core assemblies and rods to provide detailed cladding temperature profiles and accurately calculate the cladding strain on which the fuel rod lifetime and allowable burnup depend. Overall, a more realistic core thermofluids design is possible by implementing the study presented here.