The Multiphysics Object Oriented Simulation Environment (MOOSE) is a code package that couples a variety of physics modules, allowing for highly accessible multiphysics simulations. The physics modules include a finite-element Navier-Stokes (N-S) module that is designed to solve laminar fluid dynamics problems. The usage of this module in multiple recent studies coupled with the growing interest in MOOSE for usage in non-LWR safety studies by the Nuclear Regulatory Commission (NRC) prompted the authors to investigate the computational fluid dynamics (CFD) capabilities of MOOSE. A 2D flow past a round cylinder scenario is simulated in the MOOSE environment to investigate the effectiveness of the N-S module. The simulation assumed an unsteady laminar flow with a Reynolds number of 200, where the unsteadiness in the vortex street is captured using a Euler-implicit time-integration scheme. To verify the results from MOOSE, a similar simulation is conducted using the well-utilized OpenFOAM finite volume code, and the velocity profiles of both transient simulations are compared. Grid and input sensitivity studies are conducted in MOOSE to quantify the numerical and input errors, respectively. The numerical and input errors are also visualized with contour plots in order to qualitatively understand the evolution of the errors across time and space. Results show that MOOSE struggles with capturing the instabilities in the flow, as it takes more than 20 seconds before a vortex street is observed. Excellent agreement in the velocities is seen between MOOSE and OpenFOAM upstream of the cylinder, but huge discrepancies can be seen where instabilities are expected around the cylinder and upstream of it. The discrepancies become more severe as the transient simulations progress in time. The numerical error is most prominent in the wake of the cylinder and around the boundaries of the channel. The input errors are also prominent in the wake but are most prominent around the cylinder walls. The results demonstrate the difficulties faced by the N-S module in MOOSE when it comes to simulating unsteady flow and prompts questions about the effectiveness of finite element CFD solvers.

In order to more accurately simulate the strong neutronics physical and thermal-hydraulic coupling phenomenon in a typical PWR, ARMcc (a software for the physical thermal coupling calculation of PWR core) is developed. In the ARMcc, the physical calculation module is based on the fourth-order nodal expansion method (NEM) and nodal Green’s function method (NGFM), the thermal-hydraulic calculation module is based on one-dimensional single-phase single channel heat transfer model and one-dimensional cylinder heat conduction calculation model, and the finite volume method (FVM) and finite difference method (FDM) are used to solve one-dimensional cylinder heat conduction model in the program. Based on the typical PWR benchmark NEACRP-L-335, the steady-state coupling calculation ability of the program is verified. The key parameters of the core calculated by the program, such as critical boron concentration and core Doppler temperature, are in good agreement with the reference results. The deviation between the critical boron concentration and the reference results is less than 5 ‰. In addition, the influence of four calculation modes on the simulation of core physical thermal coupling process is studied, and the results of PARCS program are selected as reference. It is found that NGFM + FDM calculation mode can more accurately simulate the core fuel Doppler temperature, core outlet coolant temperature and core power distribution; NEM + FVM can more accurately simulate the core fuel maximum temperature.

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

Hydraulic control rod drive technology (HCRDT) is a newly invented patent owned by Institute of Nuclear and New Energy Technology of Tsinghua University with independent intellectual property rights. Hydraulic cylinder is the core part of this technology, so the seal of hydraulic cylinder directly affects the performance of the hydraulic cylinder and HCRDT. A new experiment table was designed and used to study the leakage and internal cylinder friction at various types of friction coupling between internal cylinder and seal ring. The result shows that Motion-Resistance Force is the smallest when the internal cylinder film is TiN. Furthermore, when seal ring film is DLC (Diamond-like carbon) the leakage is the smallest. At last, the best friction coupling scheme was obtained. The experimental results provide the basis for the establishment of the motion model of hydraulic cylinder, which provides a foundation for further design and optimization of hydraulic drive technology of control rod.

The description of the fluid flow and the coolant velocity field within a Pressurized Water Reactor (PWR), for the safety analysis is always desirable. In this work, a validation of CFD code ANSYS Fluent performed in the frame of experimental investigations to study the velocity field of coolant flow in a single phase at a reactor vessel down-comer. Under steady-state conditions. Several Experiments using Particle Image Velocimetry, have done at small hemispherical vessel, which designed by the College of Nuclear science and Technology at Harbin Engineering University (HEU) to obtain experimental data for velocity field. The measurements of the fluid flow velocities at two different locations as investigation area in the down-comer under the inlet part, for two different inlets velocities have experimentally calculated and comprehensively simulated using FLUENT. A 3D model of hemispherical cylinder, created by commercial package SolidWorks 2016 and star –CCM code has used as tool for geometry meshing and the simulations process performed with ANSYS-FLUENT code. RANS was first calculated with boundary conditions such as the velocity used at the inlet and the outlet. Two different locations were been selected as in the experimental model and the velocity has measured. The results of CFD analyses demonstrated very good agreement with the experimental data. The results that obtained by the experiment and Fluent help to understand the velocity distribution in the down-comer.

Knowledge of the heat transfer phenomenon under flow decay transient condition is important for the safety assessment of a very high temperature reactor (VHTR) during a loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally investigated. The experiment was conducted by using a forced convection heat transfer experimental apparatus. A flow control value with its control system was used to realize a flow decay condition. Helium gas was used as a coolant, and a platinum cylinder with a diameter of 1 mm was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Subsequently, the flow rate of the helium gas began to exponentially decrease with different time constants ranging from 3 s to 15 s. The initial flow velocity ranged from 7 m/s to 10 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under a wide range of experimental conditions such as heat generation rates and flow decay time constants. The results indicated that the temperature of the test heater exhibits a rapid increase during this process, and the increasing rate of the temperature is higher for a lower time constant. An increase in the heat generation rate leads to a higher increase in the surface temperature. Therefore, the heat generation rates of the fuel rods are high when a VHTR operates at high power, and it is more challenging to implement passive safety design to ensure the temperature limitation of the fuel rods during a loss-of-coolant accident. Moreover, the heat transfer coefficient relative to time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was examined based on the experimental data.

The flow distribution at core inlet plays a vital part on hydraulic design of pressurized water reactor. Nonuniform coolant flow distribution at core inlet is caused by many factors, among which the behavior of flow in the lower plenum is the most direct cause. Therefore, a further research on the flow behavior of coolant in the lower plenum is very important and necessary. However, the flow behavior is dominated by the variation of the flow environment related to the complex structure and the condition of upstream uniformity. Using CFD methods, the flow field in the pressure vessel under the uniform flux operation condition of three-loops is simulated in this paper. The standard k–ε turbulence model and upwind solver scheme are selected. Pressure and velocity along with the flow direction are investigated. The variation trend of flow characteristics is discussed by analysis on streamlines at different locations ranging from inlet pipe to lower core plate, which provides evidence for the formation of swirling eddies in the lower plenum and uneven flow flux at lower core plate. In order to understand the formation process of eddies of different sizes in the lower plenum, the velocity fields in the downcomer and lower plenum under the conditions of different inlet velocity are analyzed. Furthermore, the effects of key structures on the formation of swirl are also presented. The results show that the value of velocity flowing into the lower plenum is an important factor on the range of vertical reflux effect, also slowing down because of the resistance of the vortex suppression plate. And the value of angular momentum of swirls in the lower plenum is mainly determined by the inlet angle of converging streams flowing from the downcomer, which is caused by dispersed flow along the cylinder wall flowing from two cold legs, related with the inlet conditions.

A square array of cylinders subjected to axial flow is commonly encountered in nuclear reactors and other heat exchangers. Large-scale vortices have been observed in the gaps between the cylinders, both experimentally and numerically. These periodic flow instabilities occur in tightly-spaced cylinder arrays and originate from the velocity difference between the gap and the subchannel regions. The pressure fluctuations caused by the coherent vortex structures are possibly a source of fretting and fatigue in the aforementioned applications. In order to quantify and comprehend this phenomenon, Large-Eddy Simulations are performed on an incompressible, Newtonian fluid flowing adiabatically through a numerical domain containing a single rigid cylinder with periodic boundary conditions, representative for a cylinder in an infinite square array. Subsequently, the temporal frequency spectrum of the wall pressure profile is calculated. The spatial autocorrelation function of this Fourier spectrum, the so-called Cross Spectral Density function, contains information regarding the amplitude and convection speed of the pressure fluctuations. It is shown that the flow instability is strongest for a pitch-over-diameter ratio of 1.03. Also, the simulations indicate that the convection speed is monotonously increasing with the pitch-over-diameter ratio. An updated model for this convection speed is proposed. Finally, it is shown that the single-cylinder approximation has its limitations, but provides valuable information with minimal computational cost.

As one of the significant equipment in CEFR operation system, cold trap plays an important role in purifying the coolant sodium and ensuring the safe and steady operation of the reactor. After several years’ of operation, the cold trap of CEFR can’t work very well as it has been designed previously, which results in the low impurity aggradation efficiency, being easily blocked and frequent replacement. Analyzing the structure of the current cold trap, there are two possible reasons causing the problems mentioned above. The first one is that the sodium flow path is improper in the cold trap resulting in the blockage in specific area. The second one is that the density and layout of wire mesh didn’t take the zoning scheme, which would make the precipitation of impurities become easier in exterior area and the flow path is thus blocked. Based on the two reasons, a modified scheme is proposed from the view of structure and hydraulics to improve the impurities precipitation mode so that the lifetime of the cold trap can be prolonged. The modified scheme is to block about 1/3 orifices on interior cylinder wall in wire mesh zone and to open two rows of diversion holes on the upper support plate of wire mesh so that the velocity distribution can be improved. Meanwhile, in order to make impurities precipitate from the bottom of the cold trap, the wire mesh is arranged to be in upper, middle and lower areas using different wire mesh density. By 3-D modeling and numerical analysis and then comparing the results to that of current cold trap of CEFR, it can be concluded that the velocity distribution in modified scheme is more reasonable and the new layout of the wire mesh can meet the impurities precipitation requirements better. Further comparative analysis with the results under the actual operational condition will be carried on after the completion of the manufacture for modified cold trap.

The heat transfer tubes in steam generator of High Temperature Reactor-Pebble bed Modules (HTR-PM) suffer a high-speed and high temperature Helium flow environment. To avoid high thermal stress and decrease the fluid-induced vibration, tubes are packed in the form of rectangular tubes bundles. The empirical formulas of flow-induced vibration in ASME code are applicable to circular cylinder and incompressible media, which is not available for the proposed problem. In the present paper, rectangular tubes bundles are simplified as square cylinder with specific stiffness. The vibration of six square cylinders with horizontal and vertical arrangements are analyzed in displacement and frequency. The results show that the six square cylinders reveal same vibration amplitude and frequency, but different balance positions. For the horizontal arrangement and the vertical arrangement, according to the different balance positions, the balance position boundary line can be drawn. The cylinders at two sides of the balance position boundary line are closer to each other, where collisions or damages are easy to occur.

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

The dry-out heat flux (DHF) for the copper sphere beds was measured using the hydrogen generated at a high potential in the copper electroplating system of mass transfer. The uniform self-heating condition was simulated by the experimental methodology. The experimental apparatus consisted of a power supply, a data acquisition system, a polycarbonate cylinder containing the copper cathode beds and copper anode submerged in an aqueous solution of H 2 SO 4 . This study reported the experimental results for the bottom flooding condition with the 6 mm copper beds. The DHF data was obtained for the bed height of 20, 40 and 60 mm. The tendency of DHF on the bed height is similar to that of existing heat transfer studies. However, the absolute DHF value was quite different. We inspected the potential of the non-heating experimental method and reviewed further studies for reducing the discrepancy.

Due to the inherent safety characteristics of passive core injecting system, it is one of the most important mitigation strategies under loss of coolant accident (LOCA). However, flow instability during passive gravity-driven reflooding may occur, which can obstruct core reflooding process and reduce the heat removal rate from the reactor core. Therefore, exploring the characteristic and influencing mechanisms of flow instability during gravity-driven reflooding is necessary to analyze the effectiveness of passive core injecting system and mitigate the consequences of LOCA. In this work, small scale experiments are performed to investigate the flow instability during gravity-driven reflooding when the open ratio of steam outlet are in the ranges of 3.33–50%, gravity driving head are 16–20kPa and system pressure are 0.1–0.3MPa. The test facility consists of a cooling water storage tank, a test section and a condensate tank. The results show that flow instability could occur under specific conditions and have a strong regularity, which is divided into three stages: cooling water initial injection, cooling water expulsion and cooling water re-injection. In addition, the effects of open ratio of steam outlet, gravity driving head and system pressure on flow instability phenomenon are investigated. Increasing the open ratio of steam outlet accelerates the discharge rate of steam and prevents the accumulation of steam at the inner cylinder, which can contribute to cooling the heated rods and restrict the flow instability. And under low gravity driving head of 16kPa and low system pressure of 0.1MPa the temperature and pressure acutely change and flow instability is more likely to occur.

This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of Japan sodium cooled fast reactor (JSFR), there may be dissimilar welded joints between ferritic and austenitic steels especially in intermediate heat exchangers (IHX) and steam generators (SG). Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and type 304 austenitic stainless steel (304SS) was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the heat affected zone (HAZ) in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in base metal (BM) of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.

A new type of Steam separator for CAP1400 Steam Generator is introduced. The result of characteristic experiment, the variables of which contain steam load, saturated water flow and water level, shows its separation characteristics are suitable for CAP1400 Steam Generator. Steam generator is a major equipment of pressurized water reactor power plant, one function of which is to transfer heat from reactor to feedwater and to generate qualified steam. Steam separator, located in upper shell of Steam generator, is used to separate steam from saturated water/steam mixture which exits from the tube bundle region. Steam separating process usually consists of three stages. Centrifugal separation first happens in Primary separator, where most of water is separated. Gravity separation is achieved in the space formed between Primary separator and the dryer. Finally, as the relatively low moisture content mixture flows through the chevron-type dryer, the qualified steam is obtained. The new developed primary separator, named SP3 type, consist of raise cylinder, swirl impeller, downcomer and vane. As the steam/water mixture enters the primary separator, a centrifugal force is induced by the swirl. Thus the greater dense water in the steam-water mixture is forced toward the outside diameter of the separator, while the less dense steam stays in the central portion of the separator. The separated water leaved the separator through the downcomer. Relatively high quality steam exits the separator, passing upward through the center of raise cylinder. The paper introduced the SP3 primary separator developed by hot test, which includes test device, device design, test method, and result analysis. Separation characteristics of SP3 primary separator are obtained by various tests. The result of varying steam load test shows the outlet moisture is 13.3% under the condition of 100% steam load and 4.3 recirculation ration. The result of varying saturated water flow test shows the outlet moisture increases as the saturated mixture flow rate increases. The outlet moisture increased slowly as saturated water flow rate increases when steam load is at 50% and 100%, while the outlet moisture increases fast when steam load reaches at 120%. The result of varying water level test shows the water level is not a major factor for the SP3 primary separator performance. By analyzing experimental data, a fitting formula of the primary separator performance is obtained. the steam flow rate is the most sensitive parameter affecting primary separator performance in that formula. The water flow is the subordinate parameter for the primary separator performance. the resistance characteristics of separator is also obtained, which supports the research and development of the CAP1400 steam generator..

The fatigue damage and lift force caused by vortex induced vibration occur very often in the core of the Pressurized Water Reactor (PWR) [1] It is extremely complex to illustrate the mechanism of vibration which induced by Cross-flow. With the spacer grids and wings, the flow direction which in axial direction at the inlet will change and create swirls, so there are many flow directions in the nuclear fuel component. Assumed the tube endure cross-flow only in this article to simplify the fluid model. Most researchers in this field often ignore the displacement of structure induced by the cross flow because the value is so small that not enough to change the fluid region. In truth conditions, the motion of the cylinder caused the wake oscillation and strengthen the vortex shedding, in turn, the vortex shedding will aggravate the vibration amplitude. According that, one way FSI (Fluid Solid Interaction) can’t capture the influence from the cylinder vibration. In this article, Two-way FSI method was executed to get the vibration in time history in order to get the random vibration induced by the cross flow more close to the actual project. Using Finite Volume Method to discrete the fluid control equation and finite element method to discrete structure control equation combined with moving mesh technology. An interface between the fluid region and the structure region was created to transfer the fluid force and the structure displacement. Coupling CFD code and CSD (Computational Solid Dynamics) code to solve the differential equation and obtain the displacement of the cylinder in time history. A Fast Fourier Transfer (FFT) has been done to get the vibration frequency. An Analysis of the vortex shedding frequency and vibration frequency to find the correlation between the vortex shedding and the vibration frequency has been done. A modal analysis for the cylinder without water has been done to get the natural frequency. Results shows the cylinder has different response to the vortex shedding at different position of the cylinder in the same condition. There are more works need to be done aim to get the vibration mechanism in tandem tube and parallel tube to get clearly mechanism of vortex induced vibration in nuclear fuel assembly. The research of the vortex induced vibration in this article is a key to get on the follow research in more tubes array in different methods.

A computational fluid dynamics based simulation is performed to optimize the design of the flow distribution device in the lower plenum of the intermediate heat exchanger (IHX) of a pool-type sodium-cooled fast reactor (SFR) in this work. As a typical shell and tube heat exchanger, hot primary sodium flows in the IHX from the top and flows over the tube bundles, called shell-side. The secondary sodium (tube-side) runs through heat transfer tubes and its inlet plenum is specified at the bottom. The flow distribution device is arranged in the lower plenum of IHX, to change the flow distribution of the secondary sodium before into the heat transfer tubes. The CFD tool used in the work is ANSYS Fluent code. Two separated flow distribution devices have been simulated and compared. First, the orifice plates, three flow distribution orifice plates with different positions in the cylinder of lower plenum are respectively set as the model 1, 2 and 3. Secondly, the conical disk model, which is arranged at the bottom of the lower plenum, is established as model 4. And changing the size of the conical disk, the model 5 is established to predict the influence of the size of the conical disk on flow distribution. The results show that all of these models have similar velocity distributions at the outlet of lower plenum, which can be divided into three separate regions, where the flow velocity is higher at the inner and outer, and the velocity in the middle is lowest. When the orifice plate is set at the higher position, the overall velocity distribution is more uniform at the outlet. And the larger conical disk could make a more uniform velocity distribution as well.

Natural convection heat transfer from vertical 7×7 rod bundle in liquid sodium was numerically analyzed to optimize the thermal-hydraulic design for the bundle geometry with equilateral square array, ESA. The unsteady laminar three dimensional basic equations for natural convection heat transfer caused by a step heat flux were numerically solved until the solution reaches a steady-state. The PHOENICS code was used for the calculation considering the temperature dependence of thermo-physical properties concerned. The 7×7 test rods for diameter ( D = 7.6 mm), heated length ( L = 200 mm) and L/d (= 26.32) were used in this work. The surface heat fluxes for each cylinder were equally given for a modified Rayleigh number, ( R f,L ) ij and (R f,L ) Nx×Ny,S/D , ranging from 3.08×10 4 to 4.28×10 7 ( q = 1×10 4 ∼7×10 6 W/m 2 ) in liquid temperature ( T L = 673.15 K). The values of S/D , which are ratios of the diameter of flow channel for bundle geometry to the rod diameter, for vertical 7×7 rod bundle were ranged from 1.8 to 6 on the bundle geometry with equilateral square array. The spatial distribution of average Nusselt numbers for a vertical single cylinder of a rod bundle, (Nu av ) ij , and average Nusselt numbers for a vertical rod bundle, (Nu av,B ) Nx×Ny,S/D , were clarified. The average value of Nusselt number, (Nu av ) ij and (Nu av,B ) Nx×Ny,S/D , for the bundle geometry with various values of S/D were calculated to examine the effect of array size, bundle geometry, S/D , (R f,L ) ij and (R f,L ) Nx×Ny,S/D on heat transfer. The bundle geometry for the higher (Nu av,B ) Nx×Ny,S/D value under the condition of S/D = constant was examined. The general correlations for natural convection heat transfer from a vertical N x ×N y rod bundle with the equilateral square and triangle arrays including the effects of array size, (R f,L ) Nx×Ny,S/D and S/D were derived. The correlations for vertical N x ×N y rod bundles can describe the theoretical values of (Nu av,B ) Nx×Ny,S/D for each bundle geometry in the wide analytical range of S/D (= 1.8 to 6) and the modified Rayleigh number ( (R f,L ) Nx×Ny,S/D = 3.08×10 4 to 4.28×10 7 ) within −9.49 to 10.6 % differences.

Knowledge of the heat transfer phenomenon during flow decay transient condition is important for the safety assessment of very high temperature reactor (VHTR) during the loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally studied. The experiment was performed by using a forced convection heat transfer test loop. A flow control value with its control system was used to realize the flow decay condition. Helium gas was used as coolant and platinum cylinder with 1 mm in diameter was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Then, the mass flow rate of the helium gas starts to decrease exponentially with different time constants ranged from 4.3 s to 15.4 s. The initial flow velocity ranged from 10 m/s to 4 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under wide experimental conditions such as initial flow rate, flow decay time constant. It was found that the temperature of the test heater shows rapid increase during this process, the increasing rate of the temperature is higher for a shorter time constant. The heat transfer coefficient versus time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was clarified based on the experimental data.

Control rod hydraulic drive system (CRHDS), which is invented by INET, Tsinghua University, is a new type of internal control rod drive technology. Control rod hydraulic deceleration device (CRHDD), which consists of the plug, the hydraulic deceleration cylinder, the hydraulic buffer, etc., is one of the main components of the CRHDS. The plug is connected with the top of the control rod driving shaft and moves along with the control rod inside the hydraulic deceleration cylinder. The CRHDD performs the rod dropping deceleration function through the interworking of the plug and the deceleration cylinder which is filled with water, and reduces the rod dropping peak acceleration and the impact force acting upon the control rod to prevent the control rod cruciform blade from being deformed or damaged. The working mechanism of the CRHDD is presented and analyzed. The rod dropping performance of the CRHDS was tested experimentally under room temperature. The theoretical model of the control rod dropping process, which is composed of the three dimensional flow field analysis and flow resistance calculation of the hydraulic deceleration cylinder, the kinematics and dynamics model of the control rod, is built whose results are compared and validated by the CRHDS scram test results under room temperature. Then the model takes into account of the influence of the fuel assembly box on the control rod scram process under high temperature working conditions, and is used to analyze the influence of the key parameters, including the helical spring stiffness inside the deceleration cylinder and the working temperature on the CRHDD working performance. The research results can give guidance for the design and optimization of the CRHDD.