Search

The lead-bismuth eutectic(LBE)-cooled nuclear systems have significant natural circulation due to its large thermal expansion. Therefore, the passive safety of the reactor in postulated accidents is greatly improved. Moreover, the flow resistance and heat transfer characteristics are important to natural circulation ability of LBE-cooled system. In this paper, the natural circulation capability, the flow resistance and heat transfer characteristics were experimentally studied in the N atural C irculation C apability L oop(NCCL) facility. The research mainly includes four aspects: 1) start-up process of natural circulation; 2) capability of natural circulation; 3) flow resistance of LBE in an annular channel and 4) heat transfer in an annular channel. The results showed that a significant natural convection flow was generated in the experiment. Due to high thermal expansion of LBE, the natural circulation is quickly established and stabilized. The flow rate of natural circulation depends on the loop resistance and the temperature between heating section and cooling section. In our experiment, the maximum natural circulation flow rate was measured to be ∼0.81 kg/s. A new friction coefficient correlation was fitted and the results is slightly larger than the value calculated by the Blasius formula. For the convection heat transfer of liquid LBE in an annular channel, the heat conduction term is dominant, and the Nusselt number increases with the increase of Peclet number. Fitted experimental correlation of convection heat transfer of liquid LBE in annular channel is given. • The flow resistance and heat transfer experiments were conducted in a lead bismuth eutectic loop. • The start-up process of natural circulation and capability of natural circulation were studied in this experiment. • Fitted experimental correlations of friction coefficient and convection heat transfer of LBE in annular channel were given. [ABSTRACT FROM AUTHOR]

The Steam-Water Co-Current Flow (SWCOCF) in the AP1000 Pressurizer Surge Line (PZR SL) are important phenomena during the depressurization stage of the Small Break Loss-of-Coolant Accidents (SB-LOCAs), which affect the predictions of SWCOCF pressure drop in PZR SL and safety analyses. To investigate the pressure drop characteristics of SWCOCF in the PZR SL of AP1000, a PZR SL test model made of stainless steel is designed and a test loop is also established. The test section is mainly composed of a long inclined tube with a short acrylic visual section, a whole surge line tube and a vertical section. Two experiments were conducted: the single phase water experiment and the SWCOCF experiment. The total friction factor and the full-liquid two-phase friction multiplier as well as the liquid-only two-phase friction multiplier of the PZR SL tube were obtained, separately. The research results show that: the whole surge line two-phase friction multiplier increases with the increasing of mass quality, and decreases with the increasing of pressure. Two prediction models of two-phase friction multiplier for the surge line were developed based on the Martinelli-Nelson and Lockhart-Martinelli method, respectively. The fitted full-liquid two-phase friction multiplier model and the liquid-only two-phase friction multiplier model can predict the two-phase friction pressure drop in PZR SL very well. This paper may provide contributions to the design and safety analysis of AP1000 surge line. [ABSTRACT FROM AUTHOR]

This work mainly focuses on the verification process of the thermal-hydraulic and safety analysis code (TSACO) applied to the Chinese helium-cooled ceramic breeder test blanket module cooling system. The HE-FUS3 facility, which is an integral test facility for helium cooling systems, has been chosen to assess the capability of the TSACO code. The steady state and transient behavior of two loss-of-flow accidents (LOFAs) were simulated by TSACO. The steady state temperature distributions of the two LOFA tests were well reproduced by TSACO. The transient results of mass flowrates and outlet temperatures predicted by TSACO at the compressor and test section compared well with the experimental data. The good agreement between the calculated results and experimental data verifies the capability of the TSACO code to simulate helium cooling systems. In addition, an uncertainty analysis was conducted for the test section outlet temperatures of LOFA30 and LOFA50. The power, inlet temperature, mass flowrate, physical material properties, and heat transfer models were chosen as the input uncertainty parameters. The results show that the experimental data are easily covered by the uncertainty bands calculated by TSACO during transient analysis. This demonstrates the accuracy and fidelity of the TSACO code. [ABSTRACT FROM AUTHOR]

In this study, the saturated flow boiling heat transfer of γ-Al 2 O 3 /H 2 O nanofluids with 20 nm diameter and 0.1%, 0.5% volume concentration in a vertical tube is experimentally carried out. An ultrasonic oscillation was used to prepare nanofluid. The influences of such important parameters as surface heat flux (50–300 kW m −2 ), pressure (0.2–0.8 MPa) and mass flux (350–1100 kg m −2 s −1 ) on boiling characteristics are taken into consideration. It is confirmed that the most enhancement is about 86% for γ-Al 2 O 3 /H 2 O nanofluids saturated flow boiling heat transfer compared with deionized water, and the average value of Nusselt number was enhanced 23% and 45% respectively for 0.1 vol.% and 0.5 vol.% in this study. And the Nusselt number of nanofluid flow boiling increases with increasing the surface heat flux, the volume concentration of nanoparticle and pressure. It is confirmed that nanoparticles deposited on the heating surface by SEM observation and nanoparticles do not change obviously after boiling by TEM observations, they are attribute to the continuous operation of ultrasonic oscillation. In addition, the influence of mass flux on the enhancement rate of nanofluid saturated flow boiling heat transfer is negligible. Furthermore, a dimensionless parameter was proposed for nanofluid saturated flow boiling heat transfer data processing. [ABSTRACT FROM AUTHOR]

The flow distribution instability in parallel thin rectangular multi-channel system has been researched in the present study. The research model of parallel channel system is established by using RELAP5/MOD3.4 codes. The transient process of flow distribution instability is studied at imposed inlet mass flow rate and imposed pressure drop conditions. The influence of heating power, mass flow rate, system pressure and channel number on flow distribution instability are analyzed. Furthermore, the flow distribution instability of parallel two-channel system under asymmetric inlet throttling and heating power is studied. The results show that, if multi-channel system operates at the negative slope region of channel Δ P – G curve, small disturbance in pressure drop will lead to flow redistribution between parallel channels. Flow excursion may bring the operating point of heating channel into the density-wave oscillations region, this will result in out-phase or in-phase flow oscillations. Flow distribution instability is more likely to happen at low power/flow ratio conditions, the stability of parallel channel system increases with system pressure, the channel number has a little effect on system stability, but the asymmetry inlet throttling or heating power will make the system more unstable. [ABSTRACT FROM AUTHOR]

The present work discusses the natural circulation operational characteristics of integrated pressurized water reactor (IPWR). A thermal hydraulic model of IPWR was established using RELAP5 code. Under the ideal steady-state control scheme, IPWR can realize the reactor power change with steam flow. The conversion process from forced circulation to natural circulation has been studied. The results show that the reactor power has large fluctuations during the conversion process based on the traditional control scheme. Two optimized control schemes were applied to achieve a smooth conversion process. From the perspective of reducing power oscillation, the result obtained by canceling the reactor power control at initial conversion process is better than the results obtained by shutting down each pump one by one. Under natural circulation conditions, the stable operation boundary of IPWR system was determined. The steady state operation characteristics and rapid load change process under natural circulation show that primary coolant can achieve stable natural circulation operation at low load conditions. The obtained analysis results are significant for improving operation characteristics of the IPWR. [ABSTRACT FROM AUTHOR]

The flow and power distribution characteristics of plate-type fuel reactor core have been investigated in the present study. A reactor core thermal-hydraulic and neutronics coupled model was established using RELAP5-3D codes. The coupled modeling method was used to establish the thermo-hydraulic model, and the phenomenon of two-phase flow instability in a parallel two-channel system was used to verify the reliability of the method. Besides, the influence on power and flow distributions of uneven core inlet temperature was also studied by the application of coupled three-dimensional neutron-kinetics and thermal-hydraulics. The results show that, the coupled modeling method is reliable and can be used to establish the thermo-hydraulic model of a reactor core. By this method, both the effects of heating power and inlet temperature on flow distribution have been studied. The flow and power distributions of reactor core are inseparable and have a great effect on each other under the strong coupling of thermal-hydraulic and neutronics, but the non-uniformity of power distribution is greater than flow distribution. When reactor core inlet temperature distribution is uneven, reactor power peaking factor shifts to the low inlet temperature region, and flow peaking factor shifts to the high inlet temperature region. The effect of uneven core inlet temperature on flow distribution has a far greater impact than reactor power. [ABSTRACT FROM AUTHOR]

Abstract: Using the source-based SIMPLE algorithm based on a fixed grid method, a two-dimensional numerical model for a convection-diffusion controlled mushy region phase-change problem was developed to investigate the heat transfer characteristics of LIVE L4 melt pool subjected to a partial solidification process in a Pressurized Water Reactor (PWR) lower plenum during a hypothetical severe accident. For the binary non-eutectic mixtures of L4 melt, a linear liquid fraction temperature relationship was implemented on the calculations of the velocity and enthalpy in the mushy zone. The effect of fluid flow in the melt pool was analyzed, and numerical results for the cases with and without phase change model were calculated to investigate the effects of solidification on the heat transfer characteristics of L4 melt pool. Numerical results indicated that the phase-change model could well predict the main parameters of melt pool, e.g. the melt pool temperatures, heat flux through the melt pool, and the crust thickness. Results also indicated that the predicted Nu number without solidification was overestimated by about 12%, compared with that with solidification. [Copyright &y& Elsevier]

Highlights: [•] Modeling and nodalization for AP1000 were conducted using RELAP5/MOD3.0. [•] A typical 2-inch cold leg small break LOCA was analyzed. [•] RELAP5 was modified with entrainment correlations from the ATLATS facility. [•] This article is useful for research of liquid entrainment through ADS-4 in AP1000. [Copyright &y& Elsevier]

Highlights: [•] A reasonable model for the primary loop of an IPWR is built and then coded with Fortran 90. [•] The model has good prediction ability for full pressure start-up mode of an IPWR. [•] The effects of the start-up mode on the thermal hydraulic characteristics are also discussed. [•] The proposed model has clear physics meaning, good expansibility and accuracy. [Copyright &y& Elsevier]

Highlights: [•] A RELAP5 model for RCS and passive safety systems in AP1000 was developed. [•] A spectrum of cold leg small break LOCAs was analyzed. [•] The PCTs are far below the limit value of 1478K and meet the safety criterion. [•] This article is useful for design and operation of AP1000 and other plants. [Copyright &y& Elsevier]

Abstract: Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from the direct contact of feed-water and lead bismuth eutectic (LBE) in the chimney of 3 m height, which eliminates the bulky and flimsy steam generators. Moreover, as the coolant LBE is driven by the buoyancy of steam bubbles, the primary pump is not necessary in the reactor. The conceptual design makes the reactor simple, compact and economical. Owing to the large thermal expansion coefficient of LBE and good performance of steam lift pump, the reactor is expected to have good passive safety. A new computer code is developed to investigate the thermal–hydraulic behaviors and safety performance of PBWFR in the present work. Unprotected rod run-out transient over power (UTOP) and unprotected loss of flow (ULOF)/unprotected loss of heat sink (ULOHS) are simulated to test and verify its safety. The results show that PBWFR has very good inherent safety due to the satisfactory neutron and thermal–physical properties of LBE. Cladding materials turn to be the key factor to restrict its safety performance and UTOP is more dangerous for PBWFR. It's suggested that it should appropriately reduce the maximum value of the control rods to mitigate the consequence of UTOP due to good reactivity feedbacks in the core. [Copyright &y& Elsevier]

Abstract: The LIVE-L4 test was conducted to investigate the transient and steady state behavior of the molten pool and the crust influenced by different heat generation rates. The main purpose of this work is to develop a simple novel model of the LIVE code to calculate the entire process of the LIVE-L4 test after the melt of KNO3–NaNO3 poured into the test vessel. The LIVE code is a transient code and can be used as a fast computational program to calculate the LIVE tests. Natural convection heat transfer in the melt pool, crust behavior, heat conduction in the vessel wall, and radiative heat transfer were all considered in the model of the LIVE code. In the LIVE code, Asfia–Dhir correlations were used to calculate average and local heat transfer coefficients in the melt pool. With the assumption of no considering the composition change of local melt at melt/crust interface, many important parameters, including the melt pool temperature, heat flux distribution along the vessel wall, the thickness of the crust in steady state, and crust growth rate during the test, were calculated and compared with the LIVE-L4 experimental data. The melt pool Nu calculated by the LIVE code is larger than experimental data due to the use of Asfia–Dhir correlation in the LIVE code, which caused the average heat flux through the vessel wall larger than experiment data except the heating phase of 5 kW. It is attributed that the temperature difference between the melt pool temperature and the interface temperature at melt/crust measured in the test is larger than that calculated by the LIVE code due to the constant interface temperature at melt/crust of 284 °C used in the LIVE code. Crust growth rate calculated by the LIVE code was consistent well with the experiment data. Calculation results indicated that the LIVE code could generally predict the main parameters of the melt and crust well during the LIVE-L4 test. [Copyright &y& Elsevier]

Abstract: A series of reasonable mathematical and physical models for the thermal–hydraulic characteristic analysis of a passive residual heat removal system (PRHRS) in an integral pressurized water reactor were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, and flow and heat transfer models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code with Visual Fortran 6.5 has been developed to analyze the thermal–hydraulic characteristics of the PRHRS. Visual input, real-time processing and dynamic visualization output were achieved with Microsoft Visual Studio.NET 2008, which greatly facilitate applications in the engineering. The software was applied to analyze the effects of the heat transfer area of the PRHRS, the height difference between the residual heat exchanger center and the steam generator center, and the main steam valve turn-off time on the transient thermal–hydraulic characteristics of PRHRS. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the integral pressurized water reactor. [Copyright &y& Elsevier]

Abstract: The transient critical heat flux (CHF) experiments with forced sinusoidal inlet flow oscillation (oscillation period in 1–11s, normalized amplitude of inlet flow oscillation in 0–3.0) were conducted in a vertical tube under low pressure condition. To analyze the triggering mechanism and aftermath of periodic dryout, the wall temperature fluctuation characteristics at the onset of periodic dryout and during post-periodic dryout were investigated. Under inlet flow oscillation condition, periodic dryout would be triggered at the wave trough of liquid film oscillation as wall heat flux far below the stable-flow CHF. The transient periodic dryout would give rise to temperature fluctuations on the tube wall, the amplitude of which increased with oscillation period and heat flux. The large wall temperature fluctuation during long-playing periodic dryout could significantly pre-trigger continuous dryout. The changing trends of the periodic dryout heat flux show a reasonable agreement with Okawa’s theoretical model, in which the liquid film oscillation was supposed be weakened by the axial mixing of liquid film. Moreover, the droplet entrainment at the oscillatory interface also has noticeable influence on the oscillation characteristics of liquid film. Based on the analysis of parameter effects on periodic dryout, a semi-empirical correlation was proposed to predict the periodic dryout heat flux under inlet flow oscillation condition. [Copyright &y& Elsevier]

Abstract: Since the TMI accident in 1979, a lot of attention in the nuclear engineering field has been drawn to the small break LOCA issue, around which plenty of work has been done both experimentally and theoretically. Subsequent reactor designs have also been greatly influenced. As a Generation III + reactor that received Final Design Approval by U.S. NRC, AP1000 employs a series of passive safety systems to improve its safety. However, the thermal hydraulic phenomena related to small break LOCAs in AP1000 have not been fully understood and further studies are still required. This paper investigated the available literature and information on thermal hydraulic phenomena that occur during small break LOCAs in AP1000, which included the critical flow, natural circulation, counter-current flow limiting, entrainment, reactor vessel level swell, direct contact condensation and thermal stratification. In particular, the physical phenomena, theoretical and experimental research conducted in the past few decades, and prediction models as well as their comparison and evaluation for the thermal hydraulic phenomena related to the small break LOCAs in AP1000 were concluded. [Copyright &y& Elsevier]

Abstract: Critical heat flux (CHF) is experimentally studied on a relatively large downward-facing surface with a heated stainless steel disk diameter of D =300mm in confined space at atmospheric pressure using water as the working fluid. The bulk working fluid is subcooled. The gap size s can be adjusted to 0.9, 2.2, 2.6, 3.0, 3.2, 5.0, 7.0, 10.0, 13.0, 15.6, 19.5, 25.0, 36.0, 51.0 and 77mm. We found that the average CHF under the present condition is approximately 0.25MW/m2 which is only about 23% of which occurs on an upward-facing surface without confined space in pool boiling. The CHF increases with the increase of the gap size when the gap size is smaller than 7mm and it is a function of Bond and Jakob numbers when the gap size is larger than 7mm. [ABSTRACT FROM AUTHOR]

Abstract: In this study, we have developed a thermo-hydraulic and safety analysis code named TSAC1.0 with Visual Fortran 6.5 to analyze the thermal-hydraulic characteristics of the China advanced research reactor (CARR) under reactivity insertion accident (RIA) which was induced by unexpected control rod withdrawal in full power condition. The neutron kinetic model depended on the point kinetics with six groups of delayed neutrons including reactivity feedback effects and it was adopted for the solution of reactor power. Furthermore, a new simple and convenient model was adopted for the solution of the transient behaviors of main pump instead of the complicated four-quadrant model. Visual input, real-time processing and dynamic visualization output were achieved using Microsoft Visual Studio.NET 2003 to make the application of TSAC1.0 much more convenient in the engineering. The simulated results of TSAC1.0 were found to be in reasonable agreement with those of RELAP5/MOD3 and showed that the parameters, including the peak coolant temperature, the peak heat structure temperature, and MDNBR, were in the acceptable range of design safety limit under RIA. [ABSTRACT FROM AUTHOR]

Abstract: In this paper, a local radial point interpolation method (LRPIM) is presented to obtain the numerical solutions of the coupled equations in velocity and magnetic field for the fully developed magnetohydrodynamic (MHD) flow through a straight duct of rectangular section with arbitrary wall conductivity and orientation of applied magnetic field. Local weak forms are developed using weighted residual method locally for the governing equations of fully developed MHD flow. The shape functions from LRPIM possess the delta function property. Therefore, essential boundary conditions can be applied as easily as that in the finite-element method. The implementation procedure of LRPIM method is node based, and it doesn’t need any “mesh” or “element”. Computations have been carried out for different Hartmann numbers, wall conductivities and orientations of applied magnetic field. [ABSTRACT FROM AUTHOR]

Abstract: In this study, the 3D flow and heat transfer characteristics in rod bundle channels of the super critical water-cooled reactor were numerically investigated using CFX codes. Different turbulent models were evaluated and the flow and heat transfer characteristics in different typical channels were obtained. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) was analysed. For typical quadrilateral channel, it was found that HTC increases with P/D first and then decreases significantly when P/D is <1.4. There exists a “flat region” at the maximum value when P/D is 1.4. If P/D is larger than 1.4, heat transfer deterioration (HTD) occurs as main stream enthalpy is quite small. Furthermore, the HTD under low mass flow rate and the non-uniformity of circumferential temperature were also discussed. [ABSTRACT FROM AUTHOR]

Abstract: A study on ONB is performed by GNN in vertical narrow annuli with bilateral heating. Based on GA and ANN, a GNN model predicting qONB has been developed. The qONB prediction is investigated by GNN with distilled water flowing upward through narrow annuli with 0.95mm, 1.5mm and 2.0mm gaps, respectively. The GNN prediction results have a good agreement with experimental data. A parametric study has been conducted to investigate the effects of system pressure, mass flux, wall superheat and geometrical structure on qONB by applying GNN. Simulation and analysis results show that GNN can effectually predict qONB. [Copyright &y& Elsevier]

Abstract: Onset of nucleate boiling (ONB) experiments using deionized water as working fluid have been conducted in a range of pressure from 1 to 4 MPa, mass flow velocity from 56 to 145 kg/m2 s and wall heat flux from 9 to 58 kW/m2 for vertical narrow annuli with annular gap sizes of 0.95, 1.5 and 2 mm. We found that the ONB sometimes occurs only on outer annulus surface, sometimes occurs only on inner annulus surface and sometimes occurs on both annulus surfaces. The heat flux of the other side has great influence on the heat flux of the ONB and the latter will decrease with the increase of the heat flux of the other side. It is also found that the heat flux of the ONB increases with the increase of the pressure, the mass flux and wall superheat. However, the heat flux of the ONB will decrease as the gap size increases in narrow annuli. The heat flux of the ONB in narrow annuli is much lower than that calculated by correlations for conventional channels and a new correlation, which has good agreement with the experimental data, has been developed for predicting the heat flux of the ONB in narrow annuli. [Copyright &y& Elsevier]

Abstract: A point reactor neutron kinetics model, a drift-flow U-tube steam generator model, a non-equilibrium three-region pressurizer model and other models were established and a transient analysis code with Visual Fortran 6.5 has been developed to analyze the thermal-hydraulic characteristics of the Chinese advanced pressurized water reactor (AC-600). Visual input, real-time processing and dynamic visualization output were achieved with Microsoft Visual Studio.NET 2003, which greatly facilitate applications in the engineering. The software were applied to analyze the transient thermal-hydraulic characteristics of the loss of feed-water accident, the double loops loss-of-flow accident, the reactivity insertion accident, the sudden increase of feed-water temperature accident and the loss of offsite power accident for the Qinshan nuclear power plant in China. The obtained analysis results are significant to the improvement of design and safety operation of the plant. [Copyright &y& Elsevier]

Abstract: In this study, the data set of the 2006 CHF look-up table is partitioned into five subsets by using Fuzzy c-means (FCM) clustering algorithm. The elements of the same subset are ‘similar’ to each other in some sense while those assign to different subsets are ‘dissimilar’. At the same time, a Genetic Neural Network (GNN) model for predicting critical heat flux (CHF) is set up. It has some advantages of its globe optimal searching, quick convergence speed and solving non-linear problem. The methods of establishing the model and training of GNN are discussed particularly in the article. Local condition type CHF is predicted by GNN on the basis of 6930 CHF data from the 2006 CHF look-up table. The prediction results are consistent with database very well. Next, the mainly parametric trends of the CHF are analyzed by applying GNN. At last, prediction of dryout point is investigated by GNN with distilled water flowing upward through narrow annular channel with 0.95mm and 1.5mm gaps, respectively. The prediction results by GNN have a good agreement with experimental data. Simulation and analysis results show that the network model can effectively predict CHF. [Copyright &y& Elsevier]

Abstract: An experimental study of horizontal narrow gap heat transfer of porous media under a round downward-facing heated plate has been carried out, using water as the working fluid. The boiling curves are obtained with different gap size, plate diameter and solid spherules. The results show that the heat transfer increases significantly with porous media in the gap especially under boiling condition, and the occurrence of pool boiling crisis would be brought forward when the gap size is very small. The results also indicate that the heat transfer in horizontal narrow gap can be enhanced by increasing the ratio of gap size to plate diameter and using porous media with high thermal conductivity. Based on the mechanism of heat transfer of porous media, the correlations for natural convection and nucleate boiling are proposed to predict the heat flux. [Copyright &y& Elsevier]

Abstract: Local singularity of a signal includes a lot of important information. Wavelet transform can overcome the shortages of Fourier analysis, i.e., the weak localization in the local time- and frequency-domains. It has the capacity to detect the characteristic points of boiling curves. Based on the wavelet analysis theory of signal singularity detection, Critical Heat Flux (CHF) and Minimum Film Boiling Starting Point (q min) of boiling curves can be detected by using the wavelet modulus maxima detection. Moreover, a genetic neural network (GNN) model for predicting CHF is set up in this paper. The database used in the analysis is from the 1960s, including 2365 data points which cover a range of pressure (P), from 100 to 1000kPa, mass flow rate (G) from 40 to 500kgm−2 s−1, inlet sub-cooling (ΔT sub) from 0 to 35K, wall superheat (ΔT sat) from 10 to 500K and heat flux (Q) from 20 to 8000kWm−2. GNN mode has some advantages of its global optimal searching, quick convergence speed and solving non-linear problem. The methods of establishing the model and training of GNN are discussed particularly. The characteristic point predictions of boiling curve are investigated in detail by GNN. The results predicted by GNN have a good agreement with experimental data. At last, the main parametric trends of the CHF are analyzed by applying GNN. Simulation and analysis results show that the network model can effectively predict CHF. [Copyright &y& Elsevier]

Abstract: Critical heat flux (CHF) experiments using deionized water as working fluid have been conducted in a range of pressure from 0.6 to 4.2MPa, mass flow velocity from 60 to 130kg/m2 s and wall heat flux from 10 to 90kW/m2 for vertical narrow annuli with annular gap sizes of 0.95 and 1.5mm. We found that the CHF, occurring only on the inside tube, or on the outside tube or on both tubes of the annular channel, depends on the heat flux ratio between surfaces of the outside and inside tubes. The CHF, occurring on the surface of the inside tube, reaches the maximum value under the pressure of 2.3MPa while it occurring on the surface of the outside tube keeps increasing with the increase of the pressure. The CHF, occurring on the inside or outside tubes, increases with the increase of the mass flow velocity and the annular gap size; and decreases with the increase of critical quality and the other tube wall heat flux. Empirical correlations, which agree quite well with the experimental data, have been developed to predict the CHF occurring on surfaces of the inside or outside tubes of the narrow annular channel on the conditions of low pressure and low flow. [Copyright &y& Elsevier]

Abstract: Experimental studies on heat transfer characteristics for single-phase natural circulation flow under a rolling motion condition are performed. Experiments with and without rolling motions are conducted so that the effects of rolling motion on natural circulation heat transfer are obtained. The experimental results show rolling motion enhances the heat transfer. The heat transfer coefficient of natural circulation flow increases with the rolling amplitude and frequency. A modified Reynolds number that considers the influence of the acceleration is employed to express the effect of heat capacity. Using experimental data, an empirical equation for the heat transfer coefficient under a rolling motion condition is obtained. The calculated results agree with the experimental data. [Copyright &y& Elsevier]

Abstract: Two-phase flow instability of two-channel system has been theoretically studied in the present study. Based on the homogeneous flow model, the parallel channels model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The marginal stability boundary (MSB) at different system pressure conditions is obtained. The typical MSB shape is usually a classical inclination “L” at some operation condition (i.e. the system pressure is low and the inlet resistance coefficient is small). The three-dimensional instability spaces (or instability reefs) with different inlet resistance coefficients are obtained. The three coordinates consists of phase change number (N pch ), subcooling number (N sub ) and nondimensional pressure (P +). The lower part of the instability space is larger than the upper one. Increasing the system pressure or inlet resistance coefficient can strengthen the system stability. However, increasing the heating power destabilize the system stability. The influence of inlet subcooling on the system stability is multi-valued. [Copyright &y& Elsevier]

Abstract: Two-phase flow instability of natural circulation under a rolling motion condition is experimentally studied. The experimental results show the rolling motion induces a fluid flow fluctuation. At the trough point of the flow fluctuation, rolling motion can cause the early occurrence of natural circulation two-phase flow instability, and this case is defined as trough-type flow oscillation. The system stability decreases with increasing rolling amplitude and effect of rolling frequency is nonlinear. The complex overlap effect of trough-type flow oscillation and density wave oscillation can enhance the system coolant fluctuation; this case is defined as complex flow oscillation. Complex flow oscillation may be divided into two types: regular and irregular complex flow oscillations. Irregular complex flow oscillation is a transition type from trough-type flow oscillation to regular complex flow oscillation. Under the same thermal hydraulic conditions, the marginal stability boundary (MSB) of regular complex flow oscillation is similar to that of density wave oscillation without rolling motion, and the influences of rolling parameters on the MSB are slight. [Copyright &y& Elsevier]

Abstract: In this study, the method of characteristic line (MOC) was adopted to evaluate the valve-induced water hammer phenomena in a parallel pumps feedwater system (PPFS) during the alternate startup process of parallel pumps. Based on closed physical and mathematical equations supplied with reasonable boundary conditions, a code was developed to compute the transient phenomena including the pressure wave vibration, local flow velocity and slamming of the check valve disc, etc. Some interesting results were obtained and it was shown that severe slamming between the valve disc and valve seat occurred during the alternate startup of parallel pumps. The induced maximum pressure vibration amplitude is up to 5.0MPa, which occurs under the high–high speed startup condition. The scheme of appending a damping torque with the check valve disc was also numerically performed to eliminate the water hammer for the optimum design purpose. The adoption of damping torque slows down the closing speed of the check valve and has been approved to be an effective approach. This work is expected to be instructive for the optimum design of the PPFS in NPPs so as to mitigate the potential damage caused by valve-induced water hammer. [Copyright &y& Elsevier]

Abstract: Pool boiling is experimentally studied on a relatively large downward-facing surface with heated stainless steel disk diameters of D =100 and 300mm in confined space at atmospheric pressure using water as the working fluid. The bulk working fluid is subcooled. The gap size s can be adjusted to 10, 15 and 20mm for D =100mm and 0.9–77mm for D =300mm. We found that pool boiling under the present condition is far weaker than that occurring on an upward-facing surface. Furthermore, we found that the larger the diameter of the stainless steel plate, the weaker the pool boiling heat transfer. The heat transfer rate may be predicted by the Kutateladze correlation for s >20mm and by its modified form for s <20mm for both D =100 and D =300mm. Two different typical bubble circulation motions are found. Type I motion occurs with a probability of 90.9% and type II occurs with a probability of 9.1% according to the statistical calculations. Most coalesced bubble diameters are from 90 to 100mm for D =100mm and from 100 to 200mm for D =300mm. [Copyright &y& Elsevier]

Abstract: Experimental study has been conducted to examine the pool boiling occurs on a relative large downward-facing round surface with a diameter of 300mm in confined water pool at atmospheric pressure. An artificial neural network (ANN) has been trained successfully based on the experimental data for predicting Nusselt number of transition boiling in the present study. The input parameters of the ANN are wall superheat, , the ratio of the gap size to the diameter of the heated surface, δ/D, Prandtl number and Rayleigh number. The output is Nusselt number, Nu. The results show that: Nu decreases with increasing , and increases generally with an increase of δ/D. Nu increases with increasing Pr when gap size is smaller than 4.0mm. And Nu decreases initially and then increases with increasing Pr as gap size bigger than 5.0mm. The results also indicate that the influence of Grashof number, Gr, could be negligible. Finally, a new correlation was proposed to predict the transition boiling heat transfer under the present condition. The comparisons between the prediction of the new correlation and experimental data show a reasonable agreement. [Copyright &y& Elsevier]

Abstract: An experimental study on natural convection heat transfer on a horizontal downward facing heated surface in a water gap has been carried out under atmospheric pressure conditions. A total of 7204 experimental data points are correlated using Rayleigh versus Nusselt number correlations in various forms, based on different independent variables. The effects of different characteristic lengths and film temperatures are discussed. The buoyancy force acts as a resistance force for natural convection heat transfer on a downward facing horizontal heated surface in a confined space. For the estimation of the natural convection heat transfer under the present conditions, empirical correlations in which Nusselt number is expressed as a function of Rayleigh number, or Rayleigh and Prandtl numbers both, may be used. However, the best accuracy is provided by an empirical correlation which expresses the Nusselt number as a function of the Rayleigh and Prandtl numbers, as well as the gap width-to-heated surface diameter ratio; and uses the temperature difference between the heated surface and the ambient fluid in the definition of Rayleigh number. The characteristic length is the gap size and the film temperature is the average fluid temperature. [Copyright &y& Elsevier]

Abstract: In this paper, two-phase flow instability in natural circulation loops of China Advanced Research Reactor (CARR) has been investigated. CARR is a low pressure and low power density research reactor. A natural circulation instability analysis model is developed for the natural circulation loop of CARR. The homogeneous flow model is used to establish the system control equations. The non-uniform heating and subcooled boiling heat transfer is included. The accumulation heat of the wall is also included. Numerical method of Gear is employed to solve the system equations documented in terms of ordinary differential equations. According to the calculation results, stability maps of the natural circulation loop, which confirm the presence of an instability region under the conditions of low equilibrium quality in the outlet and low pressure, are obtained. It is a special kind of density wave oscillation (DWO) that occurs in very low equilibrium quality region with the characteristics of geysering and ‘Type-I’ DWO at the same time. The calculation results show such oscillation course clearly. The variations of the mass flow rate, the pressure drop and the boiling boundary are analyzed separately. Especially, the phase-space trajectory of the boiling boundary and the mass flow rate is discussed. Finally the oscillation frequency is discussed. The calculated results have important significance for the safety operation and accidental analysis of CARR. [Copyright &y& Elsevier]

This paper presents a non-discrete approach for vibration and buckling analysis of rectangular Kirchhoff plates with mixed edge support conditions. A plate is decomposed into multiple rectangular subdomains along the change of the discontinuous support conditions. A dominant subdomain is established, whose Ritz coefficients in the displacement trial function are incorporated into the neighbouring subdomains, by imposing compatibility conditions for transverse displacement and slope along the non-discretized interconnecting boundaries. The automated Ritz method is employed to derive governing eigenvalue equations for the vibration and buckling analysis of the plate problem. The validity and accuracy of the proposed method are verified with convergence and comparison studies. Numerical results are presented for vibration and buckling of several selected rectangular plates with various combinations of mixed edge support conditions. [Copyright &y& Elsevier]

In order to deal with the environmental pollution problem caused by excessive emission of greenhouse gases, hydrogen energy, a green low-carbon secondary energy, acts as an important role in the process of energy transformation and the realization of global carbon neutrality goals. With the development of nuclear reactors of Generation IV, there is a new way using the high-temperature residual heat of nuclear reactors to couple with advanced chemical reaction in producing hydrogen. This research based on a small Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), gives a hydrogen production method derived by biomass. Exergy analysis of this method by Aspen HYSYS shows that the material exergy consumption (including physical exergy consumption and chemical consumption) is the main part affecting exergy losses. The energy utilization of a closed-loop heating system is better than that of an open system that directly heats gasification agents. Specifically, When the temperature reaches 700 °C, the mass flow rate of CO 2 and Biomass is 1:15 and the reaction finally reaches equilibrium, the total exergy consumption of the system is 1.47 × 109 kJ/h and the mole fraction of H 2 is 49.9 % in theory. In addition, the change of temperature mainly affects physical exergy consumption while the change of mass flow rate mainly affects chemical exergy consumption. Furthermore, when the system is coupled with Brayton cycle power generation system, in which a small amount of SCO 2 is drawn out as a high-temperature gasification agent for biomass hydrogen production, the exergy consumption is barely increased. • The concept of a biomass and nuclear hydrogen production system is presented. • Supercritical carbon dioxide acts as a gasification agent and heat-transfer medium. • Detailed exergy analysis on thermal-hydraulic parameters of FuSTAR. • Exergy consumption barely increases when coupled with the Brayton power cycle. • A reference for the optimal design of comprehensive utilization of nuclear energy. [ABSTRACT FROM AUTHOR]

An enhanced moving particle semi-implicit method for simulation of incompressible fluid flow and FSI (Fluid-Structure Interaction) is developed in the present paper. Based on the recent studies on LSMPS (Least Square Moving Particle Semi-implicit) method [1,2] , the fluid models for FSI are improved, and the elastic structure models and fluid–structure coupling schemes are developed. Many strategies are developed to reduce the influence of mechanical vibration of elastomers on fluid calculation. The models are validated step by step. The improved fluid models with different Dirichlet and Neumann boundary are validated through hydrostatic pool problem, connecting tube problem and dam break problem. The elastic structure models under the velocity boundary and the loading boundary are validated through free oscillation cantilever problem and pressurized ring problem respectively. The fluid-structure coupling schemes are validated by the violent sloshing problem without considering the elastic deformation and the problem of dam break with an elastic gate. The results show a good performance of the developed method on the simulation of incompressible fluid flow and fluid-structure interaction. [ABSTRACT FROM AUTHOR]

Pebble beds are used in many tritium breeder designs of fusion blanket, which is purged by a flow of low velocity helium. In this paper, experimental study of the pressure drop of helium through packed sphere beds was carried out. The pebbled beds were sealed in a rectangular container. The spheres of five different diameters varied from 0.5 mm to 2.0 mm. The Darcy regime and Forchheimer regime were identified within the range 2.47 < Re < 27.59. The transition region was identified and discussed in detail. The pressure drop was characterized by Kozeny-Carman constants or Ergun's constants for the experimental range. Some commonly used correlations were evaluated by the experimental data. [ABSTRACT FROM AUTHOR]

Accelerator Driven System (ADS) has been developed to provide external neutron source for transmutation and multiplication to handle issues of short nuclear fuel supply and increased nuclear waste. With heavy nuclides bombarded by high-energy proton beam, neutron yields in spallation material which is accompanied by formidable energy deposition. Due to little survival of solid target from high power density and many unresolved issues in liquid target, granular target is motivated with higher sustainable power density. Although many investigations into cooling capability in spallation targets have been conducted, little literature is available on granular target which merits further study. In the present study, investigation is focused on cooling characteristics of granular target compared with liquid target based on the same target model. Due to combination of granules and fluid, coupled CFD-DEM method is employed to simulate movement and heat exchange for granular target, but only CFD technique for liquid target. Directed against same studied domain, comparison and analysis are carried out for two types of targets in the present study. Our research indicates that the granular target owns higher safety margin to withstand severe conditions especially under high energy deposition. The present findings may help to clarify and confirm the safe and reliable spallation target for some specific applications. [ABSTRACT FROM AUTHOR]

TEXAS-VI is a transient, three-field, one-dimensional mechanistic model for the steam explosion phenomena. A fuel solidification model and associated fragmentation criteria of the solidifying particle for both the mixing phase and explosion phase were developed and incorporated into TEXAS-VI to account for solidification. In the present study, TEXAS-VI was used to analyze the KS-2 and KS-4 steam explosion experiments, which were performed in the KROTOS facility as part of the OECD-SERENA-2 program. In the simulation, the KROTOS experimental facility was modeled as Eulerian control volumes based on the facility geometry. The molten corium jet was divided up into a series of LaGrangian master particles equal to the initial jet diameter. Both the mixing phase and the explosion phase of the experiments were simulated by TEXAS-VI. Comparison to test data indicates that the fuel jet kinematics and the vapor volume during the mixing phase were well predicted by TEXAS-VI. The TEXAS-VI prediction of the dynamic explosion pressure at different axial locations in the test was also in good agreement with the experimental results. The maximum pressure of KS-2 and KS-4 predicted by TEXAS-VI were 16.7 MPa and 41.9 MPa, respectively. The KS-4 maximum steam explosion pressure predicted by TEXAS-VI was higher than that of KS-2, which was consistent with experiment observation. The observed differences of the dynamic explosion pressure between the KS-2 and KS-4 experiments were also successfully simulated by TEXAS-VI. This suggests that TEXAS-VI is able to analyze the effect of prototypic melt compositions on the steam explosion phenomena. Additional benchmarking and evaluations are ongoing. [ABSTRACT FROM AUTHOR]

During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes. [ABSTRACT FROM AUTHOR]

• This study utilizes the increasing flow regime experimental data to extend the prediction range and improve the accuracy. • Integrated flow regime prediction models considering vertical and horizontal flow direction were developed and evaluated. • The MLP, RF, and SVM algorithms were utilized to develop the flow regime prediction model respectively. • The flow regime prediction model shows better performance than existing models in three directions respectively. The nuclear reactor core is the pivotal component in a nuclear power plant to generate and transfer heat, so accurate prediction of reactor core flow and heat transfer characteristics is a crucial problem for the reactor system design. Flow regime is closely related to the thermal–hydraulic characteristics of two-phase fluid. Still, flow regime models used in thermal–hydraulic calculation codes rely heavily on earlier experimental data, featuring a narrow application range and unsatisfactory accuracy. To extend the prediction range and improve the accuracy of flow regime prediction by making full use of the increasing experimental data, integrated flow regime prediction models considering vertical flow direction for light water reactor (LWR) and horizontal flow direction for Canadian Natural Deuterium Uranium (CANDU) were developed and evaluated based on three standard machine learning algorithms in this study. Firstly, the experimental data of horizontal and vertical flow collected from the literature was modified and preprocessed to be training data. Then, the multi-layer perceptron (MLP) algorithm, random forest (RF) algorithm, and support vector machine (SVM) algorithm were utilized to develop the flow regime prediction model respectively. The performance of the three models was evaluated and compared and the results showed that the flow regime prediction model based on RF was the optimal model with higher prediction and more efficiency than the other two models. Finally, the flow regime prediction model was compared with existing models in three directions respectively, which indicated that the range and accuracy of the model based on RF were superior to the existing models. This study provides an implantable and scalable approach for flow regime prediction, and the application range and accuracy of flow regime prediction can be continuously improved with updating training data. [ABSTRACT FROM AUTHOR]

Spray cooling is an efficient cooling technology with high heat dissipation, and has been extensively utilized in various industry areas. The interaction mechanism between the spray droplet and environmental space with high pressure and temperature is extremely complex, especially with non-condensable gas. In this paper, a pseudopotential multicomponent multiphase lattice Boltzmann model for droplet migration with non-condensable gas in high temperature and pressure was proposed. The mechanism of transfer characteristics for droplet in steam were studied. It is found that the mass transfer of droplet is controlled by density and condensation, droplet growth is mainly impacted by density at initial stage, and in later period, the growth is greatly weakened by the condensation suppression of non-condensable gas. The calculated results show that when Fo > 0.4, mass transfer is significantly decreased. Heat transfer is also affected by non-condensable gas, and 2% gas declines the droplet heat transfer rate by about 10%. It is also found that the mass transfer rate is approximately linearly associated with the ratio of non-condensable gas content inside and outside the droplet. The results can be used for development of spray cooling model, and have positive guiding significance for optimizing spray cooling system. • A heat and mass transfer model of droplet and steam with non-condensable is established. • The mechanism of transfer characteristics for droplet in steam were studied. • The mass transfer of droplet is controlled by density and condensation in different stage. • The mass transfer rate is approximately linearly associated with the ratio of non-condensable gas content. [ABSTRACT FROM AUTHOR]

• Empirical correlations are fitted in the helical cruciform seven-rod bundle. • Mixed convection in the laminar flow region is analyzed. • Transverse mixing in inner, side, and corner subchannels is investigated. A novel type of nuclear fuel named helical cruciform fuel may realize self-supporting for adjacent elements and improve heat transfer, both of which contribute to increasing the power density of the reactor. The flow and heat transfer characteristics of a hexagonal helical cruciform seven-rod bundle were experimentally investigated in the present work. 200 sets of isothermal frictional resistance experiments and 199 sets of heat transfer experiments were finished. The laminar flow region (200< Re <1000), the transitional flow region (1000< Re <3000), and the turbulent flow region (3000< Re <16,000) were determined based on the experimental data. Several correlations are compared with the present experimental data. The helical cruciform single-rod correlation underestimates the Nusselt number in the seven-rod bundle for Re <7000, while it has an average relative error of 7.96 % with the resistance experimental data. The empirical correlations to calculate the frictional resistance coefficient and the Nusselt number were fitted. When the Richardson number is greater than 0.1 in the laminar flow region, the Reynolds number is less than 600, and flow is considered to be mixed convection. The temperature distributions of the inner, side, and corner subchannel are analyzed. The temperature of the inner subchannel is higher than the side and corner subchannel. The increase of Reynolds number is beneficial to the fluid transverse mixing between different subchannels. This work gives validation data to support the numerical model in the helical cruciform rod bundle. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

• Two coupled parallel frameworks for shell and tube heat exchangers are proposed. • Fully three-dimensional parallel computation on both sides of the heat exchanger is realized. • One-to-one method achieves 85% parallel efficiency in 14 cores computation. • One-to-many method improves computational efficiency by about 20 times over serial in 28 cores computation. Shell and tube heat exchanger is widely used as a common type of heat exchanger. When the size and the number of heat exchanger tubes are enormous, it is difficult to use the direct modeling method for numerical simulation of heat exchangers. The simplified modeling method based on porous media utilizing user-defined function (UDF) has been widely applied. However, when the three-dimensional (3D) modeling is required on the tube side, the parallelization problem still restricts the computational efficiency of numerical simulation. In this paper, a one-to-one parallel method was first developed to solve the problem of coupled parameter transfer between the primary and secondary sides in the parallel framework, which was tested to achieve a parallel efficiency of 85% under 14 cores. However, the method acquires complicated data transfer between nodes, large memory occupation, and the modeling of the tube side is different from the actual structure under certain circumstances. To solve the problems, a one-to-many parallel method is proposed, and the computational efficiency is tested to be enhanced by about 20 times under 28 cores compared with single core, with a parallel efficiency of 70%. The two methods proposed in this paper are of great significance to the parallel transformation of the future 3D full-size code for shell and tube heat exchangers development and the calculation efficiency improvement. [ABSTRACT FROM AUTHOR]

• A Hi-fi space thermionic nuclear reactor system model is established, and its dynamic performance is obtained. • A model predictive controller is designed for the space reactor system using discrete state-space model and Kalman filter. • Several transient scenarios were designed to evaluate the control performance of the model predictive controller applied to space reactors from various perspectives. • Stable and rapid control of the space reactor has been achieved. Space nuclear reactors possess the inherent benefits of independence from solar energy and the capability to withstand intricate external circumstances. Distinguishing themselves from terrestrial reactors, space nuclear reactors function within the realm of outer space, operating autonomously throughout their entire operational cycles. Consequently, the establishment of a dependable automatic control system assumes utmost significance. In this paper, a high-fidelity nonlinear space thermionic nuclear reactor model is established, encompassing neutron physics, thermal–hydraulic, and thermoelectric conversion system models. An automatic control system with a model predictive controller (MPC) is designed based on the system. A simplified state-space model is established based on system identification as the internal model for the MPC controller, demonstrating good agreement with the dynamic response of the nonlinear system model. The deployment of the MPC controller for system control simulations exhibits excellent control performance under various operating conditions. In the case of a nuclear power step response, the MPC controller achieves an overshoot of 0.07 % and a settling time of 0.65 s. For an electric power step response, the MPC controller yields an overshoot of 0.28 % and a settling time of 36.6 s. These control performances are significantly better than those achieved by the PID controller. In power tracking control, the MPC controller demonstrates virtually error-free tracking performance. When external reactivity disturbances are introduced, the MPC controller outperforms the PID controller in swiftly compensating for the disturbances, maintaining the system's operating state at the desired level. [ABSTRACT FROM AUTHOR]

In this study, we investigate the steam oxidation behaviors of molten zirconium alloy at exceptionally high temperatures of 1870 ℃ and 1970 ℃. A specialized high-temperature steam oxidation apparatus was established to conduct the steam oxidation experiment. The experiments yielded crucial data including oxidation weight gain results, macroscopic appearance of the oxide surface, and section morphology. Several new phenomena: α-Zr(O) precipitates and two-layer oxide structure, were observed in this study. Based on these observations, a complete mechanism of molten zirconium oxidation in steam was developed, providing an interpretive framework for the oxidation observed in this experiment. This newly proposed mechanism reevaluates the formation process of nonceramic precipitates in the molten bulk and its consequential impact on oxidation kinetics at temperatures below 2000 ℃. • High-temperature steam oxidation behaviors of molten zirconium alloy at 1870℃ and 1970℃ were studied. • The oxidation mechanism of molten zirconium was discussed. • The enhanced effect of the melt on the oxidation kinetics was analyzed. [ABSTRACT FROM AUTHOR]