Fuel Element Assembly Research Articles

Direct Numerical Simulation of the Peripheral and Internal Configurations of a Model Assembly of Fuel Elements

Direct Numerical Simulation of the Peripheral and Internal Configurations of a Model Assembly of Fuel Elements

Computation-and-Experiment Study of Behavior of Molten Metal in Fuel Element and Fuel Assembly: Preliminary Experiments and Computational Models

This work investigates the motion of the liquid metal on the surface of a heated cylindrical rod. Experiment data are presented, and computational models modified in accordance with the experimental results are described. Exploration of the motion of molten metal over the surface of fuel column simulator both in gravitational runoff and at gas blowoff involved high-speed recording. Formation of rivulets and droplets has been recorded, as well as their separation from the surface at gas blowoff. Based on the experiments to simulate the melt motion, it is proposed to take into account the capillary interaction forces and to consider—in addition to solid, film draining—the rivulet runoff mode. Analysis of literature made it possible to suggest a criterion for the transition from the film mode to the rivulet mode, as well as a method of calculating the basic parameters for the rivulet mode.

The Effect of Spacer Grids on the Formation of Added Mass under the Bending Vibrations of a Fuel-Element Bundle in a Covered Fuel-Element Assembly of a VVER-440 Reactor

A significant effect of spacer gratings exerted on the dynamic characteristics of the covered fuel-element assemblies is shown. The calculated and experimental values of the added mass in the course of the bending vibrations of a bundle of fuel elements (fuel elements) are in good agreement between each other only when taking into account the flow of liquid in a narrow gap between the hexahedral surface of the spacer gratings and the hexahedral border of the cover of fuel-element assemblies. The small dimensions inherent in the spacer gratings in the axial direction of fuel-element assemblies lead to uneven distribution of the added mass along the length of the fuel-element assemblies, which is a significant factor in the study of bending oscillations of a bundle of fuel elements in liquid.

Experimental Investigations of the Characteristics of Heat-Exchange and Temperature Fields in the Channels and Fuel-Element Assemblies of Lead-Cooled Fast Reactors

The results of experimental investigations of heat exchange and temperature fi elds in channels and fuel assemblies in lead-cooled fast reactors performed at the Leipunskii Institute for Physics and Power Engineering are presented. The main investigations were performed using mercury and lead-bismuth and sodium-potassium alloys. The effect of the contact thermal resistance on heat exchange is analyzed. Data on heat transfer and the azimuthal temperature nonuniformity of the fuel elements are compared for assemblies with a wide lattice, relative spacing s/d = 1.33, smooth fuel elements, and spacing of fuel elements by an ‘edge on edge’ double winding and transverse lattices are compared.

Conjugate heat transfer simulations of advanced research reactor fuel

The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with helical fins, which may be an attractive consideration in future designs to enhance performance and/or safety.

Application of “Leak Before Break” Assessment for Pressure Tube in Delayed Hydride Cracking

The Ignalina NPP uses an RBMK-1500 reactor, which is graphite-moderated with a water-cooled reactor core. The fuel cell assembly in the center of the moderator column consists of a pressure tube containing the fuel element assembly and through which coolant flows. Pressure tubes are made of Zr-2.5Nb zirconium alloys. Hydrogen absorbed by the zirconium alloy during corrosion is one of the factors determining pressure tube lifetime. If the pressure tube hydrogen concentration exceeds solubility limitations, delayed hydride cracking (DHC) may occur. Hydrides forming under certain conditions may reduce resistance to brittle fracture. Here we evaluate hydride influence on pressure tube fracture and the application of leak before break (LBB) for tubes with DHC. Deterministic analysis employing LBB concept used experimental data. Deterministic LBB analysis confirms that pressure tubes comply with LBB requirements.

High resolution gamma ray tomography scanner for flow measurement and non-destructive testing applications

We report on the development of a high resolution gamma ray tomography scanner that is operated with a Cs-137 isotopic source at 662 keV gamma photon energy and achieves a spatial image resolution of 0.2 line pairs/ mm at 10% modulation transfer function for noncollimated detectors. It is primarily intended for the scientific study of flow regimes and phase fraction distributions in fuel element assemblies, chemical reactors, pipelines, and hydrodynamic machines. Furthermore, it is applicable to nondestructive testing of larger radiologically dense objects. The radiation detector is based on advanced avalanche photodiode technology in conjunction with lutetium yttrium orthosilicate scintillation crystals. The detector arc comprises 320 single detector elements which are operated in pulse counting mode. For measurements at fixed vessels or plant components, we built a computed tomography scanner gantry that comprises rotational and translational stages, power supply via slip rings, and data communication to the measurement personal computer via wireless local area network.

Table for Critical Heat Fluxes in Square-Packed Fuel-Element Assemblies

This paper is devoted to the construction of a tabular method for calculating critical heat fluxes in fuel-element assemblies with a square arrangement. The method consists of a basic table and a system of correctional functions, describing the dependence of the critical heat flux on the important independent parameters. Experience in developing such a method for triangular assemblies is used. The method is compared with experimental data and other computational methods. The advantages of the new method are shown: maximum width of the range of parameters, possibility of taking account of the influence of the most important parameters, possibility of using in cell-by-cell (channelwise) calculations, and high accuracy of the description.

Determination of oxidation states of uranium and plutonium in cooling solution of spent fuel element assemblies from NPP A-1

The knowledge of oxidation states of uranium and plutonium is necessary from the point of view of the proposed vitrification technology for the final liquidation of the cooling solution as a high radioactive waste containing alpha activity. The valence states of uranium and plutonium have a major influence on the subsequent leaching rate of these elements from vitrification matrices. Using the various chemical behaviours of these elements, in accordance with their valence states, we made an attempt to establish their oxidation state in the original solution.

Temperature flattening in fuel-element assemblies by using the method of eigenfunctions

This article describes a method of accurately estimating the expected nonuniformity of the temperature distribution in a fuelelement assembly(FEA). It is generally done by a numerical calculation. The approximate analytic solution presented in the article is said to have advantages from the point of view of estimating the situation and analyzing possible measures for smoothing the nonuniformities. Temperature perturbations in a FEA are classified as stochastic and regular and both are described. The representation of the temperature distribution within a FEA as the sum of harmonics corresponding to the eigenfunctions is useful for the analysis of effects of temperature flattening.

Heat removal in fast reactor cores

This paper addresses the problems of a thermal-hydraulic calculation of fast reactor fuel element assemblies (FEA), particularly, that of finding on the basis of the reactor core design parameters the thermal-hydraulic characteristics which determine its operating ability. Research at the FEI is presented here, including studies of the mechanism of interchannel mixing, comples longitudinal-transverse flow in the reactor elements, the characteristics of hydrnDynamics and heat transfer in systems of parallel fuel elements and in atypical channels, and the temperature distributions in symmetrical and distorted fuel element lattices. Methods of mathematical modeling of coolant velocity and temperature distributions in fast reactor FEAs have been developed, and a set of programs has been written for calculating the local thermal-hydraulic characteristics of FEA. They have been tested on a large amount of experimental material and offer an effective method for performing thermal-hydraulic calculations.

Enhancement of heat transfer in the RBMK and RBMKP

The increase of unit power of nuclear reactors and the related increase in heat removal is one of the pressing problems of the development of nuclear power. An effective way of solving this problem is the enhancement of heat transfer in fuel-element assemblies (FEA). Heat-removal enhancement is a complicated problem, related not only to an increase in heat removal while maintaining the accepted criteria of the working capacity of the fuel elements, but also to the solution of such concomitant problems as the decrease of the energy expended to pump through the coolant, the guarantee of the necessary technological effectiveness of the proposed enhancement devices, the account of the possible effect of these devices on the physical parameters of the core, etc. This paper discusses the different approaches to the increase of heat removal in FEA with two- and single-phase coolants.

Material behavior investigations of the VK-50 reactor fuel element assemblies

Material behavior investigations of the VK-50 reactor fuel element assemblies

Operating experience with the VV�R-1000 reactor fuel-element assemblies of the fifth unit of the Novovoronezh nuclear power station

The initial operation of the fifth unit was characterized by an increased number of transitional regimes, the number of which did not exceed the design values. No effect of the considerable number of power variations on the working efficiency of the fuel elements was observed. An analysis of the primary circuit coolant activity with respect to fission products shows that operation of the fifth unit is characterized by a low degree of depressurization of the fuel element cans. The number of fuel elements with microdefects is significantly lower than the limiting value, which confirms the high reliability of the fuel elements. After attaining nominal power level, no marked depressurization of the fuel elements develops.

Mass- and heat-exchange processes in fuel-element assembly models

Mass- and heat-exchange processes in fuel-element assembly models

Soviet-Canadian Seminar on the Vibration of Steam-Generator Tubes and Fuel-Element Assemblies in Nuclear Power Stations

Soviet-Canadian Seminar on the Vibration of Steam-Generator Tubes and Fuel-Element Assemblies in Nuclear Power Stations

A 6 MeV gamma photon facility

The 16N radiation facility provides a source of 6 MeV gamma photons for shielding and dosimetry studies. It utilises the 16N produced in the cooling water of the 300 kW Universities Research Reactor, Risley, as it flows through the fuel element assemblies. The facility incorporates a specially designed disc radiator providing an isotropic uniform disc source suitable for benchmark experiments. A shielded cell houses a lead collimator weighing 1.6 Mg which is capable of being moved in three directions by a traversing gear with external control and position indicator.

The measurement of mechanical integrity in a reactor fuel element by the analysis of external vibration signals

The mechanical integrity of a HIFAR (=HIgh Flux Australian Reactor) fuel element assembly, specifically the clearances associated with an internal, concentric specimen-irradiation tube (“liner”) and the mechanical rigidity of the annular fuel element plates, can be monitored by analysing signals obtained from vibration sensors mounted in an easily accessable location on the fuel element shield plug. The approach employed could be adapted for use as a diagnostic tool on other types of fuel element assemblies.

The temperature distribution in liquid metal cooled fuel elements

This article presents the results of experimental and analytical studies of temperature distributions and heat transfer in reactor cores with hexagonal fuel element assemblies. The temperatured is tributions in fuel elements in the central region and at the plane walls are analyzed. Some information i s given on the overheating of fuel element cladding under the spacing fins. Effects related to the redistribution of coolant flow rates over the crosssection of the assembly in the transition from turbulent to laminar flow are discussed.

An Automatic Flow Balancing Device for Boiling Water Reactors

A prototype device to control the exit steam quality in the individual fuel element assemblies of a boiling water reactor was designed, constructed and tested at simulated reactor operating conditions. The device consists of two venturis and a mechanism for comparing the pressure signals from the venturis and controlling the flow by means of a hydraulically operated valve. The device, which is completely self-contained within one channel of a boiling water reactor, held the discharge volume quality to ± 6% of the design value during transients in the heat transfer rate of up to 25%.The addition of the device to the high heat flux channels of a boiling water reactor would permit designing reactors at a higher heat flux for the same departure from nucleate boiling. The pressure across the core (header to header) is increased by ≍ 20% for a reactor of the Dresden type. Overall economic effects on reactor operations, including reactivity effects, were not evaluated.