Coolant Temperature Distribution Research Articles

Design, Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.

Subchannel analysis of a small ultra-long cycle fast reactor core

Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498°C on average and 551°C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498°C on average and 548°C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499°C on average and 538°C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

Investigation of Mixing Flow Process Using PIV and PLIF Techniques

A coolant mixing investigation in a head of a half-size model of VVER-440 fuel assembly (simulator) has been performed at KFKI. The PIV and PLIF measurements have been done under a selected list of power distribution options, flow rates and powers. The experiments were focused on obtaining a data for investigating the trends in temperature difference between the value registered by a thermocouple and that obtained using PLIF technique. The coolant temperature distribution has been measured in many positions along the coolant trajectory and where coolant flow leaves the rod bundle and in the cross section location of thermocouple, thus the dynamics of effect of mixing process is also declared. PIV and LPIF results show their ability to verify the primary results of CFD calculations.

An Analytical and Experimental Study on Coolant Flow and Temperature Distribution within Pem Fuel Cell Stack

In the present study, a mathematical model is developed for investigation of coolant flow and temperature distribution within PEMFC stack. The model consists of two main sub models: (1) a coolant flow distribution model, which is used to determine the coolant mass flow distribution between different cooling flow fields of the fuel cell stack; and (2) a thermal model, which computes the coolant temperature increase along the flow field and also the temperature distribution within each cell of the stack. By using this model, optimized amount of effective parameters on stack coolant flow and temperature distribution could be obtained with respect to lower pressure drop along the manifolds and flow fields and more stack temperature uniformity. Test results of a 65-cells stack are presented to verify the simulation results.

Thermal–hydraulic analysis of wire-wrapped SFR test subassemblies by subchannel analysis method

Thermal–hydraulic analysis of sodium-cooled fast reactors (SFRs) test subassemblies has been performed employing subchannel analysis method. In order to improve the accuracy of the calculation, recent empirical correlations of pressure drop coefficient and inter-channel mixing for evaluation of coolant flow in wire-wrapped fuel subassemblies have been incorporated into the subchannel analysis model. To confirm the prediction capability of the model, calculations and comparisons with available experimental data of 19- and 37-pin subassemblies have been performed. A comparison with previous results of a 127-pin subassembly has also been performed. The results show that empirical correlations which consider three subchannel types separately are capable of predicting the coolant temperature distribution at bundle outlet best among models. Moreover it has been found that the agreement between calculation result and experimental data of outlet temperature distribution improves when bundle hot dimensions are employed. In addition to the conventional flow resistance models, performance of a method of directly varying coolant flow areas to model geometrical disturbance due to the presence of wire spacers has been evaluated. It has been found that the method is capable of simulating crossflows between adjacent subchannels. All the results show reasonably good agreement with the reference data, in which the root mean square (RMS) differences are within 10% for all cases analyzed in this study.

Effect of spacer grid mixing vanes on coolant outlet temperature distribution

Abstract In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper.

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.

Thermo-hydraulic design verification of the neutral beam liner for the ITER vacuum vessel

The ITER vacuum vessel has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping and access inside the vessel for maintenance. A neutral beam (NB) port of equatorial ports provides a path of neutral beam for plasma heating and current drive. An internal duct liner is built in the NB ports, and copper alloy panels are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. Global NB liner models for the upper panel and the lower panel have been developed, and flow field and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux are applied in the analyses for beam steering and misalignment conditions. For the upper panel, it is found that unbalanced flow distribution occurs in each flow path, and this causes poor heat transfer performance as well. In order to improve flow distribution and to reduce pressure losses, hydraulic analyses for modified cooling path schemes have been carried out, and design recommendations are made based on the analysis results. For the lower panel, local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying design coolant flow rate. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired from the analyses. Based on the analysis results, it is concluded that the lower panel for the NB liner is relatively well designed even though the given heat loads are very severe.

Systematic staging in chemical reactor design

Abstract The foundation and implementation of a method for systematic reactor design is described. The reactor path is sectioned into stages where each stage is designed so as to optimize an overall objective. This is a further development of a previously proposed method for designing chemical reactors ( Hillestad, 2004, 2005 ). Reactants pass through a series basic operations or functions to form the desired products. The basic operations are represented by design functions on the path volume. The design functions are fluid mixing (dispersion), distribution of extra feed points, distribution of heat transfer area and coolant temperature, catalyst dilution distribution and more. The conceptual reactor design problem is solved as an optimal control problem. A direct method is applied where both the design functions and the state variables are discretized. The realization of the optimization is a staged process string of multi-functional units. A kinetic model of the gas phase methanol synthesis is used as an example. By applying the method on the model, a staged reactor design with less heat transfer area and higher production is possible.

Experimental and theoretical comparison of fuel temperature and bulk coolant characteristics in the Oregon State TRIGA ® reactor during steady state operation

In September of 2008 Oregon State University (OSU) completed its core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Experimental bulk coolant temperatures were collected in various locations throughout the Oregon State TRIGA ® Reactor (OSTR) core in order to supplement the validity of the numerical thermal hydraulic results produced in RELAP5-3D Version 2.4.2. Axial bulk coolant temperature distributions were collected by acquiring discrete thermocouple measurements in individual subchannel locations during steady state operation at 1.0 MW th. The experimental axial temperature distribution collected was compared to one-channel, two-channel, and eight-channel RELAP5-3D models and found to match within 11.94%, 11.69%, and 8.78%, respectively, on average. Comparisons to similar studies were made based on a dimensional analysis of fluid body forces in the discrete core locations, indicating that the chosen approach produces conservative results for use in the OSTR safety analysis.

DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP IN LIQUID METAL REACTORS

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

Design verification of the heating neutral beam port liner for the ITER vacuum vessel

The ITER vacuum vessel provides a high quality vacuum for the plasma, as well as the first confinement barrier of radioactive materials. It has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping, and access inside the vessel for maintenance. The heating neutral beam (HNB) ports are located side by side at two equatorial ports and provide a path of neutral beam (NB) for plasma heating and current drive. An internal duct liner around the NB is built in the HNB ports, and special panels with tubes armored by copper tiles are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. A global HNB liner model including panels, bulk shielding and tube shielding has been developed, and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux without beam steering and misalignment are applied in the analyses. Particularly, severe NB power conditions such as local increment due to the tile displacement and peak NB power are considered. Local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying three different coolant flow rates. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired. Utilizing temperature gradient data of the panels, thermal stress analyses have been conducted for various analysis conditions. As results of the thermo-hydraulic and the thermal stress analyses, the overall design adequacy of the HNB liner has been evaluated.

Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

Thermal–hydraulic analysis of the HCLL DEMO blanket

The present design of the helium-cooled lithium-lead (HCLL) blanket foresees the successive cooling of three sub-components (the first wall, the stiffening grid, and the breeder cooling plates) by parallel He-cooling channels. The optimisation of the coolant flow is complicated by the fact that the heat recovered by the cooling elements depends on the coolant parameters (temperature, velocity, heat exchange coefficient). Since all these parameters are intrinsically related, the classical approach of obtaining the coolant temperature distribution by enthalpy balance and using this distribution as an input for finite element analyses cannot be used. In this paper we present new finite elements models of the blanket that allow to obtain the temperature distribution in the coolant on the basis of the detailed power density distribution in the surrounding structures. At the same time, the models allow to improve the description of the heat transfer between the solid walls and the coolant. A review of the most used correlations available in literature to estimate the heat transfer coefficient has then been performed and the most appropriate one has been selected. The new models have been used to carry out a thermal–hydraulic analysis of the cooling system of the HCLL breeding blanket. Moreover, an estimation of the pressure drops in the different cooling elements has been made. The obtained results are discussed in terms of differences with those obtained in previous analyses and of the consequences on the blanket design.

Improved Turbulence Modeling for Performance Evaluation of Novel Fuel Designs

An improved quadratic nonlinear eddy viscosity model (NLEVM) is introduced that respects the constraints of realizability for calculation of detailed coolant velocity and temperature distributions inside tight lattice fuel bundles. The model adopts an optimized low-Reynolds formulation based on direct numerical simulation data, combined with an enhanced nonlinear stress-strain relationship to correctly capture the anisotropy of the flow up to the solid wall. The capabilities of the model are first assessed on the prediction of fully developed flow inside triangular lattice bundles; it is shown how the ability to correctly reproduce the turbulent-driven secondary flow enables the model to accurately reproduce wall shear stress and velocity distributions inside the bundle. The model is applied to the evaluation of the thermal-hydraulic performances of novel fuel designs, discussing potential advantages and limitations of the newly proposed solutions.

Subchannel Analysis of Fuel Assemblies of European Experimental ADS

In the frame of the European PDS-XADS (Preliminary Design Study-EXperimental Accelerator-Driven System) project, two concepts of a subcritical reactor core cooled by liquid lead-bismuth eutectic (LBE) were proposed. In this paper, the local thermal-hydraulic behavior of both LBE-cooled reactor-core concepts was analyzed. For this purpose, the MATRA subchannel analysis code (Multichannel Analyzer for Steady States and Transients in Rod Arrays) was selected, and modification was made for its applications to XADS conditions. Compared to the small-core concept, the large-core concept has a much lower hydraulic resistance, lower local coolant velocity, and lower temperatures of coolant and fuel pins. This enables the natural convection approach for removing reactor heat and for short-term realization of the core design using available technologies. The fuel assembly of the small-core concept has a tight configuration that leads to a high flow velocity and high pressure drop. The high power density of the small core results in high local temperatures of coolant, cladding, and fuel. Both coolant velocity and cladding temperature are such that special attention has to be paid to avoid corrosion and erosion damage of cladding materials. A parametric study shows that under the parameters considered, the mixing coefficient has the biggest effect on the coolant temperature distribution, whereas the cladding temperature is strongly affected by the selection of heat transfer correlations.

Thermal-hydraulic assessment of topaz irradiation processes at the ETRR-2 research reactor

Abstract The present work is carried out in order to evaluate the cooling process of irradiated topaz at the Egyptian Second Research Reactor (ETRR-2). A mathematical model is developed to simulate the cooling process of a part of the irradiation grid containing one topaz irradiation box. The coolant flow rate distribution through the model channels is estimated taken into consideration the buoyancy effect in each channel. The coolant temperature distributions along the model channels are predicted as well. The COSMOS/M finite element code is used to estimate the temperature distribution through one irradiated topaz stone. The predicted coolant and surface temperature values show the efficient cooling of topaz stones irradiated at the ETRR-2 research reactor.

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.

Comparative Analysis of the Results of a Benchmark Experiment on the Thermohydraulics of a Model of a Target of an Accelerator-Driven System

A benchmark experiment on heat transfer in a model of a target of an accelerator-driven system was performed as part of a conference of the International Working Group on the Thermohydraulics of Advanced Nuclear Reactors of the Inernational Association for Hydraulics Reasearch. Terms of reference for this experiment were prepared by specialists at the State Science Center of the Russian Federation – Physics and Power-Engineering Institute, using experimental data obtained previously. Specialists from Japan, Spain, and China participated in the experiment. The computational results obtained by specialists from Italy and Russia were taken into account. A considerable discrepancy was observed in the membrane and coolant temperature distributions calculated using different computer codes and the measured distributions. More accurate approaches to simulating the shaping grid, the energy release, and flow turbulization due to the interaction of the streams after the shaping grid are needed to predict the membrane and coolant temperatures.

Development of a two-dimensional model for the thermohydraulic analysis of the hot pool in liquid metal reactors

The SSC-K code has been developed at KAERI based on SSC-L originally developed at BNL to analyze loop-type LMR transients. Because the dynamic response of the primary coolant in a pool-type LMR, particularly the hot pool concept, can be quite different from that in the loop-type LMR, major modifications of SSC-L have been made for the safety analysis of the KALIMER. In particular, it is necessary to predict the hot pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs because the temperature distribution of a hot pool can alter overall system response. In this paper a two-dimensional hot pool model is developed and compared with the experimental data. A preliminary evaluation of unprotected loss of heat sink accidents for the KALIMER design with updated SSC-K code has been performed and analyzed.