Fuel Rod Assemblies Research Articles

Actual problems of modeling thermal-hydraulic processes in fast neutron reactors

The results of studies on hydrodynamics and heat transfer processes in fast neutron reactors are presented. Data on turbulent momentum transfer in rod bundles are analyzed. It is shown that the intensification of turbulent momentum transfer in the rod bundle channels is due to large-scale turbulent momentum transfer (secondary currents). The intensification of interchannel turbulent exchange in close-packed lattices of rods is explained. A dependence is obtained for the dissimilarity coefficients of interchannel convective exchange forced by wire wrapping in bundles of rods. The methods and results of numerical modeling of thermal hydraulics using the Monte Carlo method, thermomechanical analysis of the temperature field in fuel rod assemblies in the lifetime process are presented. The results of modeling based on a water model of temperature fields and the structure of coolant movement in the primary circuit of the reactor in various regimes are presented. A stable temperature stratification of the coolant was revealed in the peripheral zone of the upper chamber of the reactor above the side screens. It is shown that the process of boiling liquid metals in fuel assemblies has a complex structure, characterized by stable and pulsating regimes and a heat transfer crisis. The agreement between the results of experimental and numerical modeling is shown. A cartogram of the flow regimes of a two-phase flow of liquid metals in fuel rod assemblies has been plotted. The influence of the surface roughness of fuel elements on the boiling process and heat transfer during boiling of liquid metals is analyzed. Long-term cooling of a fuel assembly with a “sodium cavity” above the reactor core in accident regimes with boiling of liquid metals is shown. The objectives of further research are formulated.

Euler–Euler modeling of turbulent annular regime

Euler–Euler methods are one of the many approaches available in literature for modeling droplet-laden flows. These models consider both the gas and droplet phases as inter-penetrating continua, thereby separate sets of transport equations are solved for these phases. The current work focuses on modeling of thin liquid films on solid surfaces in the context of the two-fluid approach to multiphase simulation. In majority of the literature, liquid films are modeled using a 2D Laminar approximation with a parabolic profile for the velocity across the film. However, a number of industrial applications demonstrate a turbulent behavior of the film which would affect the effective viscosity and thereby the thickness of the film. In the current work, a turbulent model for liquid films is proposed based on the Universal Velocity Profile. Additional models are employed to capture the deposition of droplets onto solid surface and stripping off of droplets from the film due to surface waves. The liquid film model is validated on experimental data from available literature. Finally, the newly implemented turbulent film model is applied to study the flow over a fuel rod assembly in a Boiling Water Reactor.

DNBR Limit Analysis for the Inner channel of Annular Fuel

The departure from nucleate boiling ratios (DNBR) limit for annular fuel might vary from that of conventional nuclear fuel rod assemblies due to the distinct features of annular fuel assemblies. Precise estimation of DNBR is crucial for ensuring the safe and optimal functioning of nuclear reactors equipped with annular fuel assemblies. This study examines the DNBR limit of inner channel for annular fuel, using the look up table-2006 (LUT-2006).A comprehensive database having data for both normal operation and accident situations in pressurized water reactors (PWRs) using annular fuel was built to analyze the DNBR limit of the inner channel of annular fuel. The critical heat flux (CHF) data in this database were categorized into two groups with the limiting quality as the boundary. The prediction accuracy of LUT-2006 and the DNBR limits were evaluated respectively. Given that the inner diameter of the inner channel of annular fuel in mainstream designs typically ranges from 8-9 mm, and considering the potential occurrence of the limiting quality phenomenon in the inner channel, it would be more suitable to utilize the LUT-2006 based on an 8 mm tube standard to establish the DNBR limit for the inner channel of annular fuel. A conservative DNBR limit of 1.308 is determined for the inner channel of an annular fuel using the LUT-2006. In future thermal-hydraulic optimization designs of annular fuel, it is advisable to utilize LUT-2006 for calculating the DNBR of the inner channel of the fuel.

Numerical Analysis of Effects of Vane Deflection Angle on Flow-Thermal Characteristics in a Fuel Rod Assembly

The utilization of a grid spacer with vane is a significant component within reactor fuel channels. The presence of the vane has a notable impact on the mixing of flow and the enhancement of heat transfer within subchannels. The purpose of this work was to perform a numerical analysis of the effects of the vane deflection angle on the flow-thermal characteristics in a fuel rod assembly. In the current analysis, a square array consisting of a 5 × 5 rod bundle was utilized. The pitch-to-rod diameter ratio was set to 1.33, while the blockage ratio of the grid spacer was determined to be 0.16. A relative study was made for flow-thermal characteristics with four different vane deflection angles, such as 21 deg, 25 deg, 29 deg, and 33 deg. Analyses were made for a fluid pressure of 15.5 MPa, an inlet temperature of 583 K, and a velocity of 4.74 m/s. The present study investigated the Shear Stress Transport (SST) k-ω and Renormalization Group (RNG) k-ε turbulence models to analyze flow phenomena and thermal performance. The numerical results were validated through experimental data and also compared with correlations proposed by researchers. The analysis of the results was carried out using various methods, including the examination of data curves and streamlines, as well as vector and contour plots. The results indicate that a higher deflection angle leads to a greater reduction in temperature at the grid spacer. The swirl ratio was observed to be maximum close downstream to the grid spacer, and the persistence of the swirl ratio in the downstream can enhance the performance of departure of nucleate boiling. The vane on the grid spacer with a higher deflection angle enhances the coefficient of heat transfer remarkably close to the downstream grid spacer.

Modulation of flow meandering in the gap of a model of a peripheral subchannel of a fuel rod assembly

Modulation of flow meandering in the gap of a model of a peripheral subchannel of a fuel rod assembly

Fundamental and applied investigations of the liquid-metal cooled fast reactor thermal hydraulics (achieved results and further investigation issues)

The results of experimental and computational-theoretical investigations in the area of hydrodynamics and heat transfer of fast reactors and accelerator-driven system with liquid metal coolants are presented, and the problems and tasks for further investigations are formulated. Physical phenomena, effects, laws and process characteristics occurring in reactor power plants, including a flow path and a fast reactor core: the velocity and temperature fields, the structure and characteristics of turbulent transfer of momentum and energy, the hydraulic resistance of channels and fuel rod assemblies, the collector system hydrodynamics, the vibroacoustics, the heat transfer in channels and fuel rod assemblies with liquid metal cooling and inter-channel exchange are considered and analyzed. The data of experimental investigations are presented on a single-tube model of a large-module steam generator and on a fragmentary thermohydraulic model of a steam generator of a reactor power plant with twisted steam-generating tubes operating at subcritical and supercritical water pressure. The results of investigations of temperature and velocity fields on a small-scale water model of a fast reactor vessel with an integral layout in nominal, transient and emergency operating regimes are demonstrated. The investigation results on model of a fast reactor vessel with an integral layout are demonstrated that the effect of thermogravitational forces leads to temperature stratification with stagnant and recirculating formations, internal waves appear at the stratified boundaries, temperature pulsations, thermal fatigue of structural materials and a decrease in the service life of the equipment. It is shown that the liquid metal boiling process in fuel rod assemblies is formed under the influence of various factors, has a complex structure, is characterized by both stable (nucleate, annular-dispersed) and pulsation (slug) regimes with significant fluctuations of technological parameters, which can cause a crisis of heat transfer. Heat transfer was studied, a cartogram two-phase flow was obtained for liquid metal boiling in fuel rod assemblies, the effect of the surface roughness of fuel rods on liquid metal heat transfer and boiling regimes in fuel rod assemblies was found. The possibility of long-term stable cooling of the core during sodium boiling was shown by using “sodium cavity” over the reactor core. The kinetic and mechanical characteristics of the degradation process of the simulated fuel assembly in fast reactor core during the thermal interaction of the uranium-containing simulators of fuel with static sodium and their dependence on the parameters and design system were determined.Information about the key problems of thermophysical investigations of a high-temperature fast reactor with a sodium coolant for the production of hydrogen and reactors cooled by water at supercritical pressure is given.

Numerical investigation on transverse flow of helical cruciform fuel rod assembly in a lead-bismuth cooled fast reactor

The helical cruciform fuel (HCF) rod assembly is a new type of fuel for the lead-bismuth (LBE) fast reactor. It can be self-positioned and realize a coolant mixing without wrapped wire or grid spacer. This kind of fuel assembly can not only realize the function of the traditional wire-wrapped fuel assembly, but also omit the wire -wrapped structure, which provides a satisfactory prospect for the development of LBE reactor. In this study, the coolant mixing in the LBE cooled reactor with HCF assembly was analyzed numerically. The influence of the Reynolds number and different coolant mediums were investigated. Advantages and disadvantages of the HCF assembly were analyzed and compared with the wire-wrapped fuel.

Thermal analysis of baffle jetting in fuel rod assembly

Baffle jetting plays a significant role when it comes to safe operation of nuclear power plants. The baffle jetting phenomenon is the generation of horizontal flow impingement on fuel/control rods during the outward flow of the primary coolant into a nuclear reactor. To understand the flow and heat transfer characteristics under the baffle jetting conditions, large eddy simulations (LES) of flow around a 6×6 fuel rod assembly were conducted. Three Reynolds numbers based on jet width and inlet velocity were considered 5010, 10 000, and 20 000. A temperature difference of 5°C between the inlet fluid and the heated rods was considered to analyze the heat transfer characteristics within the assembly under baffle jetting. Various flow parameters were computed such as pressure coefficients along different rods, mean and fluctuating forces, Strouhal number, local and averaged Nusselt numbers. LES results were validated against experimental measurements and other numerical data. It was observed that the effect of the baffle jet was more significant on the first stream-wise row of rods with the stagnation points at the lower part of these rods. Furthermore, the averaged Nusselt number was found to be higher on rods in the stream-wise direction of the jet, rather than at other locations.

Numerical Investigation on Transverse Flow of Helical Cruciform Fuel Rod Assembly in a Lead-Bismuth Cooled Fast Reactor

Numerical Investigation on Transverse Flow of Helical Cruciform Fuel Rod Assembly in a Lead-Bismuth Cooled Fast Reactor

Trip turbine transient and stability analysis of two different fuel assembly designs aimed for the recycling of minor actinides and plutonium in BWRs

Recycling in thermal reactors is an alternative that could help to reduce minor actinide inventories produced by depleted fuel assemblies of nuclear reactors; the current paper assesses two proposed BWR fuel assemblies to recycle MA. These fuel designs aim to no introduce any perturbation in the electricity production from the Boiling Water Reactor, and they must satisfy the reactor core safety constraints. Such constraints must be considered under steady and transient conditions without the need to make any modification to the nuclear reactor safety systems. Previously, these two fuel designs were assessed under steady state conditions. In this study, the two proposed fuel assembly designs are tested for stability analysis and trip turbine transitory conditions. These fuel assembly designs are based on a 10 × 10 standard fuel assembly, and they have four MA-bearing rods that are replacing four fuel rod assemblies in three of the six axial fuel cells that compose the fuel assembly. The first design is a UO2 fuel assembly; the second one is a MOX fuel assembly where also, 8 fuel pin rods have been replaced by water rods in three of the six axial fuel cells that compose the fuel assembly. These two proposed fuel assemblies have been designed to be equivalent by producing the same amount of energy of the BWR reactor. The stability analysis is performed considering a global reactor oscillation; results show that in the MOX fuel assembly, the control rod pattern must be modified to provide an acceptable shutdown margin and to have adequate stability conditions. Both fuel designs provide acceptable conditions under the trip turbine transient condition and provide two fuel assembly designs that could be used for MA recycling in a BWR without any perturbation in energy production.

Development of a flow sweeping mixing model for helical fuel rod bundles

The helical fuel rod is a kind of axially twisted fuel element with a three or four-petaled profile. The helical fuel rod assembly assembled by helical fuel rods against one another in an array is self-supporting and free of spacer grids. This study analyses the flow field in a four-petaled helical fuel rod bundle from a subchannel method perspective, and found that the mixing effects are dominated by the flow sweeping but turbulence mixing. The crossflow is two-way at interior gaps, while it is one-way at side gaps. A basic form of flow sweeping mixing model is proposed and implemented in CTF code. The coefficients are determined based on reported mixing experiments, and the model predicts the twist angle-dependent mixing rates well. Although the proposed model coefficients are geometry dependent, the derivation methodology and basic form of the model could be applied in other helical fuel rod assemblies effectively.

Incorporation of Taguchi approach with CFD simulations on laser welding of spacer grid fuel rod assembly

The high heat density laser welding process minimizes the thermal distortion and residual stresses. Optimal welding process parameters are necessary to ensure the quality of weld. Identification of such welding process parameters is possible through incorporation of Taguchi approach with CFD simulations. This paper deals with CFD simulations on laser beam welding (LBW) of spacer grid fuel rod assembly as per the Taguchi’s L9orthogonal array and identification of optimal LBW parameters with pulsed laser for spot and seam welding of spacer grid. Empirical relations are developed for the weld-bead geometry in terms of the LBW process parameters and validated with numerical simulations.

Analysis of Computer Modelling Results on Fuel Rods Strength and Condition at Reduced or Absent Cooling Caused by Accident

The paper describes the phenomenology of fuel rod behaviour in severe accident. As an example, an experiment is described resulting in severe damage of 19 fuel rod assembly of VVER type; it was carried out in the CORA facility in 1993 (Research Centre, Karlsruhe, Germany). Testing conditions and results of post-test investigations of fuel assembly are given. The fuel rod code RAPTA-SFD is briefly dealt with; the code was a participant in the International Standard Problem ISP-36. The basic results are presented acquired by computer modelling CORA-W2 experiment using RAPTA-SFD code. Among the presented experimentally acquired and calculated results, the scope of the data on stainless steel component behaviour is substantial. The tested CORA-W2 fuel assembly contained a significant quantity of steel components, viz., spacer grids, a guide thimble, and a cladding of an absorber element. It is to be borne in mind that the spacer grids and a guide thimble of the updated and upgraded fuel assembly of VVER-1000 are fabricated from Zr-alloy, hence, the relative quantitative characteristics of chemical interactions between materials and stainless steel (Cr-Ni alloy) will be much lower for the up-to-date upgraded fuel assembly under identical conditions.

Thermohydraulics of Alkaline Liquid-Metal Coolant: A Retrospective-Perspective Look

Sixty years of experience in mastering alkaline liquid metals jointly with industry institutes, the Academy of Sciences, and engineering design agencies developing Nuclear Power Facilities (NPF) has culminated in the development of the scientific foundations for their use in nuclear power and the scientific substantiation of the hydraulic parameters and highly efficient technological processes ensuring successful operation of fundamentally new NPF. The primary directions of research were the hydrodynamics of and heat transfer in channels with complicated shapes and in nominal and deformed fuel-rod assemblies, including the blocking of flow sections, and the mechanisms of turbulent heat and mass transfer, interchannel mixing of the coolant, and the hydrodynamics of collector systems in reactors and heat-exchange equipment. Significant attention was devoted to heat exchange in intermediate heat-exchangers and steam generators of reactor facilities and the boiling and condensation of liquid metals. Questions concerning the generalization and experimental thermophysical databases, development of heat pipes, thermophysics of thermionic converters, high-temperature space NPF, and thermonuclear plants were examined. Future studies resulting from the need to improve the safety, cost-effectiveness, and reliability of developed nuclear power facilities are formulated.

Numerical investigation on the effect of transversal fluid field deformation on heat transfer in a rod bundle with mixing vanes

Spacer grids of fuel rod assemblies are equipped with vanes, which promote flow mixing and turbulence within and across the sub-channels, thereby enhancing the heat transfer. First, a literature study about the various effect of the spacer grid has on the sub-channel thermo-hydrodynamics is provided. It follows, that the multiple effects on the vane angle are insufficiently understood. The effect of the vane angle on design parameters, namely the evolution of the Nusselt number, the pressure drop, the cross and swirl flows, is here further discussed and supplemented by own simulations. The effect of the velocity gradient tensor ∇⊗u, decomposed into a strain and a vorticity contribution, is also looked at downstream of the spacer grid. The RNG k-ε turbulence model was found to provide results best matching the experimental data available in the literature. The use of vanes results in the formation of a downstream vortex. As the flow develops downstream of the spacer grid, the vortex migrates away from the sub-channel center and eventually weakens. In line with the presented literature survey, it is confirmed that a vane angle of about 30° provides optimal swirl and cross flows, resulting in an enhanced heat transfer.

РЕЗУЛЬТАТЫ МОДЕЛИРОВАНИЯ ТЕСТОВОЙ ЗАДАЧИ «BENCHMARK ROSTOV 2» С ИСПОЛЬЗОВАНИЕМ КОМПЛЕКСА ПРОГРАММ САПФИР_95&RC_ВВЭР И КОРСАР

This paper presents the results of “Benchmark Rostov 2” test task simulation using SAPFIR_95&RC_VVER program package; the test task is based on a full-scale experiment performed at the Rostov NPP, where the transition mode with boric acid dilution and a working group insertion to compensate the reactivity were implemented. The purpose of this task is to develop the methods of non-stationary computation of pin-by-pin power energy distribution in calculation programs for neutronic and thermohydraulic characteristics of VVER-type reactors. The SAPFIR_95&RC_VVER program package uses two calculation methods of pin-by-pin power energy distribution: micro- and macro-flux superposition method and combined fine-grid method, where the nodes of radial computational grid in the reactor core coincide with the centers of fuel elements in fuel rod assembly. The combined fine-grid method for calculating pin-by-pin power energy distribution is carried out with the known values of burn-up distributions, fuel temperature, and coolant density obtained from superposition method. Both approaches are used for solving stationary and non-stationary tasks. When simulating the non-stationary processes, the test task can be solved in conjunction with KORSAR thermohydraulic code. This paper describes neutronic models of VVER-1000 reactor core used in “Benchmark Rostov 2” test task for superposition method and combined method. Calculation results for power energy distribution fields are given.

Numerical Simulation of Thermal–Hydraulic Processes in Liquid-Metal Cooled Fuel Assemblies in the Anisotropic Porous Body Approximation

The article presents an anisotropic porous body model in which the transfer anisotropy is taken into account through determining—by means of tensor analysis techniques—the drag force, effective viscosity, and thermal conductivity. The model is intended for describing heat-and-mass transfer in fuel assemblies and tube bundles. For closing the system of anisotropic porous body equations, the integral turbulence model developed by the authors is used. To verify how correctly the hydrodynamics and heat transfer are described, a few hydrodynamic and thermal–hydraulic processes in water- and liquid-metal-cooled fuel rod assemblies are simulated in the anisotropic porous body approximation. The results from simulating the flow patterns of lead–bismuth eutectics in the experimental 19-rod assembly and water in a 61-rod nonheated assembly with its flow cross-section locally blocked in the central and corner parts are presented. The thermal–hydraulic processes in the BREST reactor fuel assembly’s heated 19-rod fragment with its flow cross-section locally blocked in the central part were also simulated using the CONV-3D DNS code in the framework of model cross-verification activities. The numerical analysis was carried out using the developed APMod software module implementing the anisotropic porous body model jointly with the integral turbulence model. It was demonstrated from a comparison of the numerical analysis results with both experimental data and simulation results obtained using the CONV-3D computer code that the APMod software module adequately describes the 3D fields of coolant velocities, pressure, and temperature arising in fuel rod assemblies with a locally blocked part of their flow section. The obtained results testify that the anisotropic porous body model can be used for simulating thermal–hydraulic processes in the cores and heat-transfer equipment of prospective reactors.

Simplified representative models for long-term flow and advective transport in fractured crystalline bedrock

Simplified representative models (SRMs) of the hydrogeological system at radioactive-waste repository sites are presented and demonstrated to give useful predictions of the key hydrogeological factors affecting long-term safety. The SRM is constructed from complex site-descriptive models, which have been developed to be consistent with detailed site information and data from short-term (with duration of days, weeks, up to months) field experiments, by keeping elements that are important for long-term predictions into thousands of years and simplifying features of less importance. The simplified approach relies only on fundamental hydrogeological principles and the mathematics can be kept relatively simple. The purpose of SRM is to provide a means of verifying predictions from complex numerical models, with an approach that is easy to evaluate and allows transparent evaluation of factors influencing long-term results. The approach is applied to evaluations of sites for two repositories in fractured crystalline bedrock in Sweden: one for spent nuclear fuel rod assemblies and one for waste with lower levels of radioactivity. The results indicate that the SRMs are able to yield results similar to those of calculations based on much more complex models. Further, the approach allows an evaluation of additional sources of uncertainty that are difficult or expensive to conduct with the complex models. These capabilities make SRMs a very useful and transparent tool for regulatory review.

Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University (XJTU) in China. A three-dimensional CFD study of the fluid flow and heat transfer at supercritical pressures for the rod-bundle geometries was performed with the key parameter being the fuel rod wall temperature. This investigation used Reynolds-averaged Navier–Stokes turbulence models with wall functions to investigate the behavior of flow through the wire-wrapped fuel rod bundles with water subjected to a supercritical pressure of 25 MPa. Along with the selection of turbulence models, CFD results were found to be dependent on the value of turbulent Prandtl number used in simulating the experimental test conditions for the wire-wrapped fuel rod configuration. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 deg.

Reflooding investigation on light water reactor model fuel assemblies and debris under the loss of coolant accident simulation

The reflooding investigations of large-scale model fuel assemblies and sphere beds under Design Basis Accident and Beyond Design Accident of Light Water Reactors WWER and PWR are presented. The generalizing correlation on the quench front velocity for model fuel rod assemblies and sphere beds was developed on the base of heat balance and Leidenfrost temperature correlation. It is shown that reflooding duration calculation results of model fuel assemblies and sphere beds are in good agreement with the experimental data.