On the fluctuating wall pressure field in tube banks

This paper presents the experimental analysis of pressure and velocity fluctuations of the cross flow in tube banks, with triangular and square arrangements, and four different aspect ratios. Air is the working fluid, driven by a centrifugal blower, passed by a settling chamber and a set of honeycombs and screens, before reaching the tube bank at an incidence angle of 90°. Both triangular and square arrangements have pitch-to-diameter ratios P/D=1.60, 1.26, 1.16 and 1.05, with Reynolds numbers, calculated with the tube diameter and the velocity of the flow in the narrow gap between the tubes, from Re=4×10⁴ to 7×10⁴. Velocity and velocity fluctuations were measured by a constant-temperature hot-wire anemometer, while pressure fluctuations were measured by a piezo-resistive pressure transducer mounted inside one of the tubes in the bank. Behavior of fluctuating quantities is described by means of dimensionless root mean square values and autospectral density functions, while their interdependence is discussed based on cross-correlation functions.     

Analysis of flow distribution a PWR fuel rod bundle model containng A 90% blockage

An experimental investigation, covering a Reynolds number range from 2 × 10³ to 3.5 × 10⁴, was conducted to study the velocity and turbulence intensity distributions due to the presence of a blockage in an unheated 7 × 7 rod bundle. The blockage configuration, consisting of a 4 × 4 rod array, created a maximum flow area reduction of 90% in the central nine subchannels. The blockage sleeve length was 38.3 × rod diameter and the 90% blockage zone length extended for 16.4 × rod diameter. The results showed that upstream of the blockage, the flow was not influenced by the blockage until it reached the location where the inlet taper section of the swelling started. At the downstream end, the flow disturbance was extensive and persisted over a distance of about 83 rod diameters. Compared to the downstream velocity profiles, the turbulence intensity measurements however showed a faster recovery from the blockage influence. At the higher Reynolds number, velocity profiles calculated using the COBRA subchannel computer code compared consistently with the experimental data. The general flow behaviour of the various subchannels was reasonably well predicted. However, at low Reynolds number, due mainly to the frictional form loss calculation scheme in COBRA and uncertainty in the flow transition, the flow diversion due to the blockage to the surrounding unblocked subchannels was overestimated. The influence of the degree of recovery from the rod swelling on the flow was also studied using COBRA.     

Wavelet time–frequency analysis of accelerating and decelerating flows in a tube bank

In the present work, the steady approximation for accelerating and decelerating flows through tube banks is discussed. With this purpose, the experimental study of velocity and pressure fluctuations of transient turbulent cross-flow in a tube bank with square arrangement and a pitch-to-diameter ratio of 1.26 is performed. The Reynolds number at steady-state flow, computed with the tube diameter and the flow velocity in the narrow gap between the tubes, is 8 × 10⁴. Air is the working fluid. The accelerating and decelerating transients are obtained by means of start and stop of the centrifugal blower. Wavelet and wavelet packet multiresolution analysis were applied to decompose the signal in frequency intervals, using Daubechies 20 wavelet and scale functions, thus allowing the analysis of phenomena in a time–frequency domain. The continuous wavelet transform was also applied, using the Morlet function. The signals in the steady state, which presented a bistable behavior, were separated in two modes and analyzed with usual statistic tools. The results were compared with the steady-state assumption, demonstrating the ability of wavelets for analyzing time varying signals.     

Experimental study on the Reynolds number dependence of turbulent mixing in a rod bundle

An experimental study for Reynolds number dependence of the turbulent mixing between fuel-bundle subchannels, was performed. The measurements were done on a triangular array bundle with a 1.20 pitch to diameter relation and 10 mm rod diameter, in a low-pressure water loop, at Reynolds numbers between 1.4 × 10³ and 1.3 × 10⁵.The high accuracy of the results was obtained by improving a thermal tracing technique recently developed. The Reynolds exponent on the mixing rate correlation was obtained with two-digit accuracy for Reynolds numbers greater than 3 × 10³. It was also found a marked increase in the mixing rate for lower Reynolds numbers.The weak theoretical base of the accepted Reynolds dependence was pointed out in light of the later findings, as well as its ambiguous supporting experimental data.The present results also provide indirect information about dominant large scale flow pulsations at different flow regimes.     

Wall temperature measurement and heat transfer correlation of turbulent supercritical carbon dioxide flow in vertical circular/non-circular tubes

Experimental data are presented which describe heat transfer characteristics of turbulent supercritical carbon dioxide flow in vertical tubes with circular, triangular, and square cross-sections. Experiments are conducted at a constant pressure of 8 MPa under various conditions such as inlet bulk temperatures ranging from 15 to 32 °C, imposed heat fluxes from 3 to 180 kW/m², and mass velocities from 209 to 1230 kg/m² s. The corresponding Reynolds and Grashof numbers are in the range of 3 × 10⁴ to 1.4 × 10⁵, and 5 × 10⁹ to 4 × 10¹¹, respectively. The test section is composed of an entrance region of 0.6 m long and a heating region of 1.2 m long. Wall temperatures are measured by thermocouples installed at the outer surface of the heating region. In order to identify the effect of the cross-sectional shape on the supercritical heat transfer, wall temperature distributions in the streamwise direction are compared at the same heat flux and mass velocity conditions. Based on the wall temperature data, an improved heat transfer correlation, which can be applicable to both forced convection and mixed convection regimes, is proposed, and compared with previous ones.     

Vapour void fraction in subcooled flow boiling

Experimental data on steam void fraction and axial temperature distribution in an annular boiling channel for low mass-flux forced and natural circulation flow of water with inlet subcooling have been obtained. The ranges of variables covered are: mass flux = 1.4 × 10⁴−1.0 × 10⁵ kg/hr m²; heat flux = 4.5 × 10³−7.5 × 10⁴ kcal/hr m²; and inlet subcooling = 10–70°C. The present and literature data match well with the theoretical void predictions using a four-step method similar to that suggested by Zuber and co-workers.     

Effect of irradiation by 140 Mev Ag ions on the optical and electrical properties of polypropylene/TiO2 composite

Changes in the optical, structural, dielectric properties and surface morphology of a polypropylene/TiO₂ composite due to swift heavy ion irradiation were studied by means of UV–visible spectroscopy, X-ray diffraction, impedance gain phase analyzer and atomic force microscopy. Samples were irradiated with 140 MeV Ag¹¹⁺ ions at fluences of 1 × 10¹¹ and 5 × 10¹² ions/cm². UV–visible absorption analysis reveals a decrease in optical direct band gap from 2.62 to 2.42 eV after a fluence of 5 × 10¹² ions/cm². X-ray diffractograms show an increase in crystallinity of the composite due to irradiation. The dielectric constants obey the Universal law given by ε α f ⁿ⁻¹, where n varies from 0.38 to 0.91. The dielectric constant and loss are observed to change significantly due to irradiation. Cole–cole diagrams have shown the frequency dependence of the complex impedance at different fluences. The average surface roughness of the composite decreases upon irradiation.     

Evaluation of heat-transfer coefficient at direct-contact condensation of cold water and steam

The estimation of the heat transfer coefficient at the direct-contact condensation of cold water and steam is a very hard task since the phenoma are essentially undsteady and the interface motion is so complicated that an exact estimation of its area is almost impossible. The present study shows the heat transfer coefficient evaluated experimentally by assuming simple interface shapes for complicated surfaces and estimated those through comparison of the numerical analyses to the data of experiments related to the loss of coolant accidents of light water reactors.At chugging, the heat transfer coefficient reached up to 2 × 10⁶W/(m^{2 K}). At condensation oscillation, it ranged between 10⁵–10⁶ W/(m² K). At a jet region of cold water injected into the steam flow in a pipe or the stationary steam in a vessel, the value was around 2 × 10⁵W/(m²K), and at the surface of stratified flow, it was between 3 × 10³–3 × 10⁴W/(m²K).     

Experimental study of heat transfer enhancement in narrow rectangular channel with longitudinal vortex generators

In order to enhance heat transfer in cooling channels of plate-type fuel elements in reactor cores, the experimental research is conducted on the heat transfer and pressure drop in horizontal narrow rectangular channels with mounted longitudinal vortex generators (LVGs) for water flow with Prandtl number Pr = 4–5. The parameters examined were: flow velocity from 0.5 to 3.4 m/s, Reynolds number from 3000 to 20,000, heat flux 43.6 kW/m², maximum system pressure 1.3 atm, and viscosity ratio from 1.05 to 1.2. It is found that the LVGs could greatly improve the heat transfer rate by 10–45%. Thermal performance is compared under three constraints, i.e., identical mass flow rate (IMF), identical pressure drop (IPD) and identical pumping power (IPP). It is found that the heat transfer performance of channel with LVGs on two sides are better than those on one side. Application of LVGs to plate-type fuel element is a potential technique for next generation advanced nuclear reactors concepts.     

Simulating test for thermal mixing in the hot gas chamber of the HTR-10

The helium coolant at the outlet of the pebble bed core of the 10 MW High Temperature Gas-cooled Reactor-Test Module exhibits a severe radial temperature deviation. In order to avoid damages at the downstream components due to alternating thermal loads such as the steam generator, a hot gas chamber is especially designed to solve the problem. Thermal mixing performance of the coolant in the hot gas chamber is experimentally investigated on a 1:1.5 scale model by air. The experimental result shows that within the Reynolds number range of 1.4×10⁵–5.8×10⁵, the hot gas chamber with a radial mixer reaches excellent thermal mixing of the coolant of about 94%. The flow resistance coefficient for the hot gas chamber is also presented.     

Experimental studies on rewetting of hot vertical annular channel

Studies on the rewetting behaviour of hot vertical annular channels are of interest in the context of emergency core cooling in nuclear reactors following LOCA. Experimental studies were carried out to study the rewetting behaviour of a hot vertical annular channel, with hot inner tube, for bottom flooding and top flow rewetting conditions. The length of the inner tube of the test section was 3030 mm for bottom flooding rewetting experiments and 2630 mm for top flow rewetting experiments. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 lpm (11.7×10⁻⁵ m³ s⁻¹). The initial surface temperature of the inner tube was varied from 200 to 500°C. The experimental studies show that for a given initial surface temperature of the tube, the rewetting velocity increases with an increase in flow rate of water and it decreases with an increase in the initial surface temperature for a given water flow rate. For a given water flow rate and initial surface temperature, the rewetting velocity is higher in the case of rewetting under bottom flooding conditions as compared to that in the case of rewetting under top flow conditions. These conclusions agree with the conclusions reported in the earlier literature. Using the experimental data of the present work, correlations for bottom flooding and top flow rewetting velocities are developed.     

Ductile–brittle transition of polycrystalline iron and iron–chromium alloys

Fracture toughness of polycrystalline Fe, Fe–3%Cr and Fe–9%Cr was measured by four-point bending of pre-cracked specimens at temperatures between 77 K and 150 K and strain rates between 4.46 × 10⁻⁴ and 2.23 × 10⁻² s⁻¹. For all materials, fracture behaviour changed with increasing temperature from brittle to ductile at a distinct brittle–ductile transition temperature (T_c), which increased with increasing strain rate. At low strain rates, an Arrhenius relation was found between T_c and strain rate in each material. At high strain rates, T_c was at slightly higher values than those expected from extrapolation of the Arrhenius relation from lower strain rates. This shift of T_c was associated with twinning near the crack tip. For each material, use of an Arrhenius relation for tests at strain rates at which specimens showed twinning gave the same activation energy as for the low strain rate tests. The values of activation energy for the brittle–ductile transition of polycrystalline Fe, Fe–3%Cr and Fe–9%Cr were found to be 0.21, 0.15 and 0.10 eV, respectively, indicating that the activation energy for dislocation glide decreases with increasing chromium concentration in iron.     

Experimental investigation and CFD simulation of horizontal stratified two-phase flow phenomena

For the investigation of stratified two-phase flow, two horizontal channels with rectangular cross-section were built at Forschungszentrum Dresden-Rossendorf (FZD). The channels allow the investigation of air/water co-current flows, especially the slug behaviour, at atmospheric pressure and room temperature. The test-sections are made of acrylic glass, so that optical techniques, like high-speed video observation or particle image velocimetry (PIV), can be applied for measurements. The rectangular cross-section was chosen to provide better observation possibilities. Moreover, dynamic pressure measurements were performed and synchronised with the high-speed camera system.CFD post-test simulations of stratified flows were performed using the code ANSYS CFX. The Euler–Euler two fluid model with the free surface option was applied on grids of minimum 4 × 10⁵ control volumes. The turbulence was modelled separately for each phase using the k–ω-based shear stress transport (SST) turbulence model. The results compare very well in terms of slug formation, velocity, and breaking. The qualitative agreement between calculation and experiment is encouraging and shows that CFD can be a useful tool in studying horizontal two-phase flow.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Turbulent momentum transport in rod bundles

An experimental investigation was performed with air to obtain detailed information on the velocity and turbulence distribution for parallel turbulent flows through subchannels of rod bundles. Experimental results were obtained for wall and corner subchannels of rod bundles of four parallel rods. The pitch-to-diameter ratios were varied between 1.07 and 1.4. The Reynolds numbers ranged from 6 × 10⁴ to 2 × 10⁵, depending on the rod bundle arrangement.On the basis of the data measured, the eddy viscosities in the directions normal and parallel to the wall were calculated. The experimental results of the velocity and wall shear stress distributions are compared with the predictions by the VELASCO code. There are considerable differences between computed and experimental results especially for low pitch-to-diameter ratios. The reasons for the discrepancies are discussed together with the results of attempts to adjust the VELASCO code against the experimental data.     

Dissolution of uranium metal without hydride formation or hydrogen gas generation

This study shows that metallic uranium will cleanly dissolve in carbonate-peroxide solution without generation of hydrogen gas or uranium hydride. Metallic uranium shot, 0.5–1 mm diameter, was reacted with ammonium carbonate–hydrogen peroxide solutions ranging in concentration from 0.13 M to 1.0 M carbonate and 0.50 M to 2.0 M peroxide. The dissolution rate was calculated from the reduction in bead mass, and independently by uranium analysis of the solution. The calculated dissolution rate ranged from about 4 × 10⁻³ to 8 × 10⁻³ mm/h, dependent primarily on the peroxide concentration. Hydrogen analysis of the etched beads showed that no detectable hydrogen was introduced into the uranium metal by the etching process.     

Fluid-to-fluid modeling of critical heat flux in 4 × 4 rod bundles

Fluid-to-fluid modeling of critical heat flux (CHF) is to simulate the CHF behaviors for water by employing low cost modeling fluid, and the flow scaling factor is the key to apply the technique to fuel bundles. The CHF experiments in 4×4 rod bundles have been carried out in Freon-12 loop in equivalent nuclear reactor water conditions (P=10.0–16.0 MPa, G=488.0–2100.0 kg/m² s, X_cr=−0.20–0.30). The models in fluid-to-fluid modeling of CHF is verified by the CHF data for Freon-12 obtained in the experiment and the CHF correlation for water obtained by Nuclear Power Institute of China (NPIC) in the same 4×4 rod bundles. It has been found that the S.Y. Ahmad Compensation Distortion model, the Lu Zhongqi model, the Groeneveld model and Stevens–Kirby model overpredict the bundles CHF values for water. Then an empirical correlation of flow scaling factor is proposed. Comparison of the CHF data in two kinds of test sections for Freon-12, in which the distance of the last grid away the end of heated length is different, shows that the spacer grid, which is located at 20 mm away from the end of the heated length, has evidently influenced on the CHF value in the 4×4 rod bundles for Freon-12. This is different from that for water, and the need for further work is required.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Lateral turbulent diffusion for longitudinal flow in a rectangular channel

The turbulent diffusivity for mass transfer in the wide direction of a 0.008 × 0.1 m rectangular duct was determined by continuous tracer injection. A hybrid computation method was used to analyse the axial development of time average lateral concentration distributions. For the investigated Reynolds number range, 30 000–86 000, the turbulent diffusivity of mass can be correlated by = 0.178y₀(τ ) From the diffusivity results one may infer the magnitude of the Lagrangian integral spatial scale of turbulence.     

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.