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• An equivalent mechanical model for liquid sloshing applied to annular cylindrical containers with evenly distributed cylinders is presented. • The influence of internal cylinders on the parameters of the equivalent mechanical model is evaluated by the finite element method. • The formulas for parameters of the equivalent mechanical model are presented and validated by experimental results. The dynamic response of the liquid sloshing in the reactor vessel is an important issue for seismic designs of liquid metal reactors (LMR). In engineering design, equivalent mechanical models are generally used to evaluate the response of liquid containers. However, several equipment support cylinders are arranged in the reactor vessel forming an annular cylindrical container with evenly distributed cylinders. The equivalent model of such containers is absent. Therefore, we developed an equivalent mechanical model for liquid sloshing in the irregular annular region of LMR. Using the finite element method, parameters of the equivalent mechanical model were evaluated. Moreover, calculation formulas for parameters of the equivalent model were obtained. The calculations were compared with the experiments in the reference, achieving a good agreement. This study can provide parameters for a simplified model of liquid sloshing in similar containers and provide a reference for the seismic design of LMR. [ABSTRACT FROM AUTHOR]

The performance of printed circuit heat exchangers (PCHEs) has a significant impact on the heat transfer efficiency and compactness of the supercritical carbon dioxide (SCO 2) Brayton cycle system suitable for the fourth-generation nuclear energy system. Compared to PCHEs with other types of channels, the airfoil fin PCHEs have superior flow heat transfer performance. The installation of vortex generators (VGs) in the heat transfer channels can enhance heat transfer efficiently, but the application of them to PCHEs is very limited. In order to improve the performance of PCHE with airfoil fin channels, a novel channel structure combining airfoil fin channels with rectangular VGs is proposed, and the effect of geometric parameters of VGs on the thermal-hydraulic performance of the channels is also investigated. The results show that the VG improves the synergistic effect of velocity field and temperature gradient field, leading to the increase of turbulent kinetic energy and decrease of heat transfer entropy production in the channel, which finally promotes the enhancement of heat transfer. The comprehensive performance of the airfoil fin channel with VGs is 5.95%–19.47% higher than that of the conventional airfoil fin channel. The increase in the height and length of the VGs improves the heat transfer performance of the channel, and the increase in the width of the VGs weakens the heat transfer performance of the channel, and both increase the flow resistance of the channel. By comparing the comprehensive performance, it is found that the VGs have the best height of 0.2 mm, the best length of 0.8 mm and the best width of 0.04 mm. In addition, according to the noise performance statistics (NPS), the height of the VG contributes more to the Nu and Fanning friction coefficients of the airfoil fin channels. • The effects of vortex generators on the thermal-hydraulic performance are investigated. • The improved channels increased the overall performance of PCHE by 5.95%–19.47%. • The mechanism of heat transfer enhancement caused by the VGs is analyzed. • The optimum geometric parameters of VGs are detected. • The contribution ratio of each geometric parameter is investigated. [ABSTRACT FROM AUTHOR]

• Correlation between the temperature field of HEFA and original reactor was built. • Original reactor temperature was obtained by numerical simulation and extrapolation. • The results of extrapolation and numerical simulation were mutually verified. Liquid sodium serves as the coolant in the pool-type sodium-cooled fast reactor. The primary penetration equipment on the primary side is immersed in a sodium pool. Argon gas is employed as the covering gas above liquid sodium. To determine the temperature distribution of the roof slab of the main vessel for strength analysis, it is necessary to analyze the heat transfer characteristics. Owing to the large volume of the original reactor, it is difficult to establish a full-scale experimental model. Consequently, a scaled-down experiment is required to obtain the temperature distribution of the roof slab and argon space. In addition, the complex heat transfer characteristics within the roof slab and argon space make it difficult to maintain similarities between the experimental and original facilities. Numerical simulations are a useful approach for extrapolating an experiment to the original reactor. In this study, a scaled-down heat transfer experimental facility for an argon space (HEFA) was established based on the similarity principle, and a numerical model was set up according to the HEFA. The validity of the numerical methods was confirmed by comparing the experimental and numerical results for the HEFA. Based on similarity theory, the temperature distribution in the HEFA was extrapolated to that in the original reactor. A numerical model was established based on the original reactor. By comparing the extrapolated temperature distribution and numerical distribution of the original reactor, it was found that they were in good agreement. Therefore, the validity of the extrapolated temperature distribution was confirmed. [ABSTRACT FROM AUTHOR]

• The effects of airfoil structural parameters on heat transfer and flow resistance are investigated. • The optimum structural parameters for the airfoil fin channels are detected. • The improved channels increased the overall performance of PCHE by 0.13–6.3%. • The field synergy principle explains the enhanced heat transfer mechanism well. • The influence of entropy production on heat transfer and flow resistance is studied. As an important component of the system, the performance of the printed circuit heat exchanger (PCHE) has a significant impact on the efficiency and compactness of the supercritical carbon dioxide (SCO 2) Brayton cycle system. Optimization of the structural parameters of airfoil fins can improve the performance of PCHE flow channels, but it has not received enough attention. Therefore, it is of great significance to investigate the effects of structural parameters such as the maximum thickness (W f), the distance from the position where the maximum thickness is located to the leading-edge (D f), and the radius of the leading-edge (R f) of the airfoil fins on the flow heat transfer performance of SCO 2 in the channel of PCHE. A three-dimensional numerical model is developed and validated for performance analysis of the PCHE, meanwhile the thermal-hydraulic performance of channels with different structural parameters of the airfoil fins is compared. The results show that with the increase of the W f of the airfoil fins, the heat transfer characteristics of the PCHE channel are gradually enhanced, and the flow characteristics are gradually weakened. When the maximum thickness of the airfoil fins is in the front, the heat transfer and flow characteristics of the PCHE channel are better. As the R f of the airfoil fins increases, the heat transfer characteristics and flow characteristics of the PCHE channel fluctuate. The change of the turbulent kinetic energy, the boundary layer thickness, the synergistic effect of the velocity field and the temperature gradient field, and the local entropy production in the channel can well explain the mechanism of the flow and heat transfer of the SCO 2 in the airfoil fin channel. In addition, among the channels with different structural parameters of the airfoil fins, the Fin-5 airfoil fin channel with W f of 1.5 mm, the Fin-6 airfoil fin channel with D f of 0.6 mm and the Fin-13 airfoil fin channel with R f of 1.46 mm are recommended. Their comprehensive performance is 2.0–6.3%, 0.9–5.6% and 0.13–1.81% higher than that of the Fin-3 airfoil fin channel, respectively. [ABSTRACT FROM AUTHOR]

The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model. [ABSTRACT FROM AUTHOR]

The complex structure of a pool-type sodium-cooled fast reactor may lead to uncertainty and asymmetry of flow and temperature field distributions under a pump stuck accident. This phenomenon has obvious three-dimensional (3-D) thermal-hydraulic characteristics and cannot be analyzed by one-dimensional or two-dimensional models. Previous research has been limited and lacking of 3-D numerical data. Therefore, the commercial computational fluid dynamics software FLUENT is used to simulate a full-scale 3-D integrated model of the China Experimental Fast Reactor (CEFR) in order to obtain 3-D thermal-hydraulic characteristics of key structures and components in the pool-type sodium-cooled fast reactor under a pump stuck accident in the primary loop. The results show that a special asymmetrical backflow phenomenon may occur in the pressure tube and the intermediate heat exchangers (IHXs) of the failure loop under the accident, further leading to complicated flow and thermal characteristics in both the hot and the cold pools. There is obvious thermal stratification and asymmetric temperature distribution, within a temperature difference of more than 90°C between the different loops' IHX outlet. The temperature difference between the upper and lower areas of the baffles is 20°C to 105°C. This research provides a detailed reference for engineering design and operation. [ABSTRACT FROM AUTHOR]

• A design procedure for the cladding of central measuring shroud based on elastoplastic model was proposed. • Considering actual operating situation, design of cladding against cyclic thermal shock loads was performed. • Compared to the elastic model, the design results of cladding thickness and gap distance is greatly reduced. • The increase of cladding thickness is beneficial for central measuring shroud, but harmful for cladding. The thermal shock caused by scram conditions greatly affects the service life of central measuring shroud. Cladding, as a protective layer, can be adopted to protect the central measuring shroud. In this study, a cladding design procedure based on elastoplastic model was proposed and verified by experiment. Then, the cladding thickness and gap distance were designed. And the temperature and strain distribution were analyzed and discussed. It was found that under thermal shock condition, the cladding thickness and gap distance should be at least 2 mm and 2.66 mm to guarantee safety. Compared to elastic model, the design results of elastoplastic model are reduced by 50 % for cladding thickness and 33.5 % for gap distance. The increase of cladding thickness is beneficial for central measuring shroud, but harmful for cladding. The alternating tensile and compressive strain caused by temperature changes is the main reason for total strain fluctuation. [ABSTRACT FROM AUTHOR]

• We study the modes of short coaxial multi-layer cylinders subjecting to sine wave with different prominent frequencies. • We compare the added mass of flexible double-layer shells by experimental and FEM methods, as well as shells with rigid outer shell by Au-Yang's theory. • We obtain the pressure components of the circumferential modes subjecting to seismic waves. As the most important component of pool-type fast reactors, the main vessel contains most of the equipment for the primary circuit and thousands of tons of liquid sodium. The fluid–structure interaction effects of the coaxial multi-layer thin-walled cylindrical shells formed by the main vessel and its thermal shield cannot be neglected. Though the research on the fluid–structure interaction of the liquid in the pool is sufficient, the research on the short coaxial multi-layer cylindrical shells is not enough, especially the axial and circumferential mode shapes and the ratio of each mode under seismic conditions. So, it is necessary to study the fluid–structure interaction of short coaxial multilayer shells. This paper presents shaking table tests performed on double- and triple-layered cylindrical shells to investigate fluid–structure interaction and seismic response, respectively. The experiments reveal that when the aspect ratio is less than 2, Mode (1,6) exhibits the most significant response under sine wave excitation conditions, while Mode (1,1) demonstrates the highest response under seismic conditions. The first three modes are characterized as out-of-phase modes. Additionally, the pressure component of the first circumferential mode contributes 20% to the overall response of the first 12 modes under seismic conditions. [ABSTRACT FROM AUTHOR]

As an important component of the supercritical carbon dioxide (SCO 2) Brayton cycle system, the performance of the printed circuit heat exchanger (PCHE) has a significant impact on the efficiency and compactness of the whole system. It is important to investigate the effect of airfoil fin angle of attack on the flow and heat transfer performance of SCO 2 in the channels of airfoil fin PCHEs. A three-dimensional numerical model is developed and the thermal-hydraulic performance of channels with different airfoil fin angles of attack is compared. The results show that channels with consistent distribution of the airfoil fins with angles of attack of 15°–25° and channels with inconsistent distribution of the airfoil fins with angles of attack of 20°–30° have enhanced heat transfer capacity. This improvement can be attributed to the changes in the flow behavior of the SCO 2 , increases of the turbulent kinetic energy, changes of the thermal properties of the SCO 2 and the enhanced synergy of velocity and temperature gradient fields. And the flow performance of the PCHE channel gradually reduces with increasing fin angle of attack. And the channel with the airfoil fins whose angle of attack is 15°–25° has the best overall performance. • Effect of the angle of attack of fins on the channel performance is investigated. • The channels with consistent and opposite distribution of airfoil fins are proposed. • The field synergy principle explains the enhanced heat transfer mechanism well. • The optimum angles of attack for the airfoil fin channels are detected. • The improved channels increased the overall performance of PCHE by 7.6%–15.9%. [ABSTRACT FROM AUTHOR]

Spent fuel from nuclear power plants is usually stored in specific equipment in spent fuel pools. To improve the efficiency of spent fuel transportation, a new free-standing spent fuel storage basket has been developed. In the event of an earthquake, the spent fuel storage baskets may collide with each other. If the spent fuel baskets break, this could lead to a leak of radiative materials. As the spent fuel is stored in high density in the spent fuel pool, the spent fuel leak may also lead to re-criticality. In addition, the fluid has a strong fluid-structure interaction with the baskets and the fuel assemblies. Therefore, it is very important to make the dynamic response of the baskets within the safe range under earthquake. To ensure the safety of the spent fuel storage baskets, the baskets need to be tested by a shaker in advance. During the scale analysis of the shaking table tests, there are two similarity laws, namely, a velocity-similarity law (C v = 1) and an acceleration-similarity law (C a = 1). However, there are some debates on which similarity law to obey in the complex phenomenon of baskets' seismic response. To solve the above problem, a series of 1/2 scale-down tests were carried out to investigate the response of the baskets. By the sine beat-wave tests, we obtain the friction coefficient between the spent fuel storage baskets and the pool, and the sliding and rocking patterns. Moreover, we carry out seismic shaking table tests on freestanding spent fuel storage baskets with different similarity laws. By observing the movement of the basket at different water levels of the working conditions, it is found that the fluid decreases the new basket's rocking amplitudes but increases the sliding amplitudes. The experiment can be the benchmark for the seismic analysis of spent fuel storage baskets. [ABSTRACT FROM AUTHOR]

• The turbulent flow in a rod bundle was studied using large eddy simulation. • The first- and second-order turbulence statistics were provided. • The cross-sectional mixing and vortexes were captured and analyzed. • Turbulent energy budget was studied by decomposing the Reynolds stress. The turbulent flow and interchannel mixing in rod bundles is a concerned issue in the thermal-hydraulic design of fuel assemblies. In this paper, a large eddy simulation was performed to study the turbulent flow of water in a 2 × 2 rod bundle. The diameter of each rod and the pitch-to-diameter ratio were set to 20 mm and 1.25, respectively. The grid resolutions were carefully studied by examining the two-point correlation of velocity components along the streamwise direction, and by comparing the power spectral density with Kolmogorov's −5/3 law. The numerical results showed that the root-mean-square (RMS) velocity is highly anisotropic in the gap area. The secondary flow on the cross-section of the rod bundle was nonuniform, and the intensity is strong close to the wall where small eddies were generated. The distributions of Reynolds normal and shear stresses and the budget of primary normal stress were presented. [ABSTRACT FROM AUTHOR]

A large quantity of argon is typically provided at the top of the main vessel of a sodium-cooled fast reactor (SFR) to serve as covering gas. The sodium coolant in an SFR evaporates and mixes with the argon covering gas during reactor operation. This mixed gas has a critical influence on the heat transfer at the top of the main vessel, particularly the temperature of the roof slab. The temperature distribution in the roof slab significantly affects reactor safety as it serves as the pressure boundary of the main vessel. Research on the behavior of sodium vapor has typically employed numerical simulations, and any experimental verifications have generally used simplified experimental models. To study the heat transfer characteristics of sodium vapor and eliminate the influence of simplifications on the results, the Heat transfer Experimental Facility for Argon space at the top of the main vessel (HEFA) was established based on a scaled-down top of the main vessel. The HEFA was employed to analyze influence of sodium vapor on the heat transfer characteristics of the reactor space by evaluating the effects of various heating temperatures with and without liquid sodium on the temperature distributions in the argon and roof slab. Furthermore, the emissivity of the liquid sodium surface was measured. The results show that the radiative absorption effect of sodium vapor primarily occurs near the liquid sodium surface. When the heating temperature was less than 400 °C, the presence of sodium vapor increased the temperature of the main vessel. When the heating temperature was 540 °C, the radiative absorption effect of the sodium vapor was enhanced, reducing the roof slab temperature. The observed phenomena can be considered to ensure reactor safety during the design of SFRs. [ABSTRACT FROM AUTHOR]

• Experimental investigation was carried out on convectional heat transfer characteristics of argon space in the vessel of pool-type fast reactor with a geometrically similar facility. • Convection of the argon increases the overall temperature of the roof slab and reduces its temperature gradient. • The circumferential temperature difference of the annular gap is observed, and it is judged that there is natural convection with one side rising and the other falling. A large amount of sodium is used as coolant in the pool-type fast reactor, and argon is used as covering gas in the top part of main vessel. At the top of the main vessel, an annular cavity is formed between the rotating plug and its support, where natural convection of argon is formed. This kind of structure makes the heat transfer complicated at the top of main vessel. Excessive temperature or large temperature gradient at the top of main vessel may occur. Therefore, it is necessary to carry out analysis of the temperature distribution at the top of the main vessel. Many researches based on CFD have been carried out on the analysis of the top domain, but few of them are verified by experiment. In addition, a few experiments have been performed using simplified facility, which is not so similar to the real geometry. Therefore, a scaled-down Heat-transfer Experimental Facility for Argon space at the top of main vessel (HEFA) is established, which is much similar to the real geometry. In HEFA, the annular gap is kept, in addition, a heated stainless bottom plate is used to simulate the liquid sodium surface. Two conditions were set up, one for vacuum, the other for argon. The temperature of the facility was measured in the two conditions. By comparing the results of two condition, the results show that convection of the argon increases the overall temperature of the roof slab and reduces its temperature gradient of roof slab. The circumferential temperature difference at annular gap around rotating plug was observed in the argon condition. It is judged that the circumferential temperature difference is caused by natural convection in the annular gap with one side rising and the other falling. This phenomenon must be paid much attention in the design of SFR. [ABSTRACT FROM AUTHOR]

Highlights • Experimental investigation on pool boiling of a vertical heated bundle under flooded condition. • Heat transfer coefficients of the bundle under different heated fluxes were obtained. • Flow patterns changing with time were recorded. Abstract Heat removal of the spent fuel in the spent fuel pool normally depends on the forced-circulation system in the pool. Once the forced-circulation cooling system fails, decay heat of the spent fuel can only be removed by the natural convection of the coolant. If this situation lasts for a long time, temperature of the water in the pool may continue increasing till the pool-type boiling occurs. It is important to research the boiling heat transfer characteristics for the design and safety assessment of the spent fuel pool. Researches on the pool boiling have already been extensively performed, but few is on the spent fuel assemblies which are vertical, long-span and tight bundles. This paper presents the pool boiling behavior of the heated vertical bundles flooded in the water which models the spent fuel stored in the pool. A series of experiments were carried out with different heated powers. Temperatures of the bundles were measured and heat transfer coefficients were obtained. High-speed camera was used to capture the two-phase flow patterns in the experiments. The highest temperature of the spent fuel cladding is about 115 °C during the saturate boiling state. This datum confirms the safety of the spent fuel under flooded condition. [ABSTRACT FROM AUTHOR]

Highlights • The theoretical and experimental investigations on two-phase pressure gradient of steam-water mixture through the curved flow channel in case of IVR-ERVC have done in view of exit quality within 0 < x < 0.030 under bubble flow regime. • The sub pressure gradient terms viz. friction, acceleration and gravitation have considered for the sake of housing every aspect in the evaluation of overall pressure gradient. • The frictional pressure gradient was predicted twice by considering fluid as homogeneous and separatedby incorporating HFM and SFM respectively. The experimental measurements showed same trends as that of the predicted values. • The RMSE of HFM and SFM were calculated as 0.389 and 1.822 respectively. Abstract In-Vessel Retention (IVR) of corium through External Reactor Vessel Cooling (ERVC) is an ongoing technique for the extraction of heat out of the reactor pressure vessel (RPV). The efficiency of heat transfer and critical heat flux (CHF) are dependent on the optimal design of the external flow channel. Insertion of flow field modulators i.e. fins, grooves etc. within the channel are considered as promising factors for maximizing heat dissipation out of the vessel via flowing fluid, but on the other hand, their insertion may impose excessive back pressure on the channel which may lead it to be ruptured. So, it is necessary to analyze pressure effect while designing flow channel in order to overcome associated safety issues. By keeping this in view, the present study has been executed in order to analyze the two-phase pressure gradient through the curved channel. Two-phase (steam-water) pressure gradient calculation through the 90° square, even cross-sectional, curved channel occupying 40 mm of hydraulic diameter and 476 mm of curvature radius has been deliberated by conferring Homogeneous Flow Model (HFM) as well as Separated Flow Model (SFM) for the computation of combined effect of frictional, acceleration and gravitational pressure gradients. The calculated pressure gradient with respect to exit steam quality (x) was validated by the experimental results conducted at ' Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy '. The experimental and calculated results were in good agreement within 0 < x < 0.030 under 'bubble flow regime' having RMSE of 0.389 and 1.822 with HFM and SFM respectively. [ABSTRACT FROM AUTHOR]

Highlights • Sloshing inside annular region separated by several cylinders was studied. • The frequency and wave height were relied on the number of internal cylinders. • The results using VOF approach were in good agreement with experimental results. Abstract The liquid sloshing behavior analysis of reactor vessel is an important topic of sodium-cooled fast reactor design. It is known that there are two pumps and four heat exchangers immersed in sodium liquid. Although lots of research work has been performed on the liquid sloshing inside simple cylindrical tank and coaxial circular cylinders, few about sloshing inside annular region separated by several cylinders (ARSSC), this paper deals with the analysis of the liquid sloshing inside the ARSSC geometry. A simplified cylindrical tank of fast reactor was designed, in which there are six internal cylinders used to model pumps and heat exchangers. By shaking table experiments, the number of internal cylinders was gradually increased, and the natural frequency and wave height of liquid sloshing were measured under each case. In numerical investigation, the VOF models were built to record wave height using the CFD simulation approach, the conditions were consistent with experiments. Much different with previous research, the characteristics of sloshing inside ARSSC geometry were strongly dependent on the number of internal cylinders. Experiment results showed that the natural frequency and wave height all decrease with the increase of internal cylinders gradually. The wave height response obtained by numerical simulation was in good agreement with the experimental results. In addition, as the increase of internal cylinders gradually, the maximum sloshing response occurred when the forcing frequency matched high order natural frequency of sloshing. Moreover, a correction factor was obtained by experiments for the natural frequency theoretical calculation formula. The conclusions can provide favorable reference for the seismic design of the sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]

Highlights • This study provides a numerical method for researching the dynamic characteristic of the fluid-structure interaction of the spent fuel storage rack. • This paper studied the fluid-structure interaction effect of the rack which densely placed in spent fuel pool. • This study can offer the assistance for the design and manufacture of spent fuel pools and high-density spent fuel storage racks, ultimately. Abstract Spent fuel storage rack is the most important equipment for storing spent fuel after refueling in nuclear power station. Free standing storage rack has an advantage to protect the structural integrity of the rack, because the fluid-structure interaction of the free standing rack can dissipate the earthquake energy and reduce the movement of rack, when earthquake occurs. The fluid added mass is a main parameter for studying the fluid-structure interaction effect, which provides an important parameter for the transient structural analysis of the storage rack. However, the calculation methods of fluid added mass, which in engineering, are only for the simple geometric structure such as circular cylinder or square cylinder. It cannot apply to the spent fuel storage rack, which the geometric structure is extraordinarily complex. In this paper, in order to investigate a more accurate fluid added mass, a 2-dimensional dynamic model was constructed. Furthermore, the transient analysis of the 2-dimensional model was conducted by Computational Fluid Dynamics (CFD) software, Fluent. Then, the added mass of the model was obtained according to CFD transient fluid force. Moreover, a model which corresponding to the previous vibration experiment of rack was conducted to verifies the correctness of the model. From numerical results, some important features for the design of rack were found. This study can offer the assistance for the design and manufacture of spent fuel pools and high-density spent fuel storage racks, ultimately. [ABSTRACT FROM AUTHOR]

Under normal conditions, the decay heat of the spent fuel can be removed by the forced-circulation cooling system of the spent fuel pool. In the case of a system failure, the water level in the pool will be gradually decreased owing to an evaporation of the water, and the spent fuel bundles may be exposed to air, i.e., a dry out. Since the Fukushima accident, researchers have been paying significant attention to the safety of the spent fuel under such dry-out conditions. To study the boiling behaviour of the spent fuel bundle under a partial dry-out condition, an experimental investigation has been carried out. A newly constructed 9 × 9 heated tube bundle with the same geometry as the prototype spent fuel assembly in an AP1000 was used as the test section. Two-phase flow patterns were recorded. The heat transfer coefficient of the rods in different locations was obtained. A temperature fluctuation was observed at the beginning of the dry-out process mainly owing to the reduced heat transfer effectiveness of the steam. Although the maximum wall temperature (about 570 °C) exceeded the safety criterion when 50% of the tube bundle at the top end was exposed to air, the bundle maintained its integrity. [ABSTRACT FROM AUTHOR]

Supercritical carbon dioxide (SCO 2 ) has been extensively adopted in heat-transfer experiments due to its low critical point compared to supercritical water (SCW). While it helps to largely reduce the technical difficulty and cost, the experimental results need to be extrapolated to SCW using a set of reliable and well-validated fluid-to-fluid scaling criteria. In the present paper, a criterion for scaling the mass flux in normal and enhanced heat transfer regimes was achieved by using Computational Fluid Dynamic (CFD) approaches. This criterion as well as the available scaling laws were assessed based on a wide range of the experimental data. It was found that Prandtl number shows a significant and nonmonotone effect on the scaling of mass flux (and consequently Nusselt number). Similar exponents of the Prandtl number were obtained using different approaches, but a constant exponent failed to capture the nonmonotone effect, especially when the bulk temperature surpasses the pseudo-critical temperature where large variations in the thermophysical properties appear. A modified scaling criterion, which is applicable to the heat transfer deterioration occurred at low mass fluxes, was proposed for the Nusselt number. The deviation from the experimental data was significantly reduced after Grashof number was incorporated into the criterion. [ABSTRACT FROM AUTHOR]

The fluid has great influence on the motion and deformation of the structure under vibration condition, which makes the structure produce a strongly nonlinear vibration. The spent fuel storage rack, which submerged in liquid, is an important equipment in nuclear power plant. Thus, it is extremely important to study fluid-structure interaction problem of the spent fuel storage rack among the equipment in nuclear power plant. Moreover, the fluid-structure interaction between the water around the spent fuel storage rack and the structure may compromise the integrity of the storage rack. Therefore, it is highly necessary to research the characteristic of the fluid-structure interaction among the racks. In this paper, the free vibration experiments, based on the 3:10 model of the spent fuel storage rack in CAP1400 (a new reactor type designed by China), were conducted to study the fluid-structure interaction effect of adjacent racks. The parameters of fluid-structure interaction among the racks was measured under different clearances through free vibration experiment. Ultimately, the formula for calculating the added mass of the rack under different gap sizes was put forward in this paper. This study can provide fluid-structure interaction parameters for structural dynamic analysis of the spent fuel storage rack, and offer the assistance for the design and manufacture of spent fuel pools and high-density spent fuel storage racks, ultimately. [ABSTRACT FROM AUTHOR]

Supercritical carbon dioxide (SCO 2) Brayton cycle is considered as one of power conversion systems for small modular reactors (SMRs) due to its compact and simple layout. SCO 2 Brayton cycle direct cooled reactor further reduces the number of system components and simplifies the system layout, which makes it suitable for the design of offshore floating nuclear power plant. In this paper, a conceptual design is proposed for 1MWe offshore floating nuclear power plant (SCFNP-1), applying SCO 2 direct cooled reactor system. Considering that the design of mass flowrate for low-power system is small, pre-pressure safety containment is applied to ensure system safety under LOCA condition. Based on the safety system for SCFNP-1, its transient performance is analyzed through simulation according to the protection logic. The pre-pressure safety containment effectively limits the drop of system pressure and core mass flowrate, and it also reduces the leakage mass of SCO 2. This work provides helpful references for the safety of low-power SCO 2 direct cooled reactor systems with pre-pressure safety containment under LOCA. • A conceptual design is provided for 1MWe offshore floating nuclear power plant (SCFNP-1), applying SCO 2 direct cooled reactor system and pre-pressurized safety containment. • Transient simulation codes are developed for LOCA simulation, which is based on Modelica codes for SCO 2 cooled reactor system. And benchmark calculation is carried out to verify the developed codes. • The influences of pre-pressurized safety containment with different volume and initial pressure are analyzed for system safety. • Transient performances of the system with safety control system are simulated under LOCA under different accident conditions, considering the influences of different parameters of pre-pressure safety containment. [ABSTRACT FROM AUTHOR]

Under seismic condition, for CAP1400 (a new nuclear reactor type designed by China) spent fuel storage rack; water can flow through the channels between the rack storage cells. Unfortunately, the fluid structure interaction (FSI) effect for such perforate structure has not been studied in depth. In engineering approach of spent fuel rack’s structural analysis, the perforations are ignored, leading to an over-conservative added mass. To study the friction coefficient and FSI effect for CAP1400 rack, a series of shaking table experiments had been carried out using a reduced-scale spent fuel rack model in a water tank (spent fuel pool model). To derive the friction coefficient between rack and tank wall, sweep frequency tests were carried out in shallow-water-level condition and deep-water-level condition. Large uncertainty was found in the measurements of friction coefficient; consequently, the data were organized by statistical methods. Moreover, the relationship between gap size and fluid pressure were studied. Finally, the response to seismic loading was studied in different gap size conditions. The research can provide the basic input parameter for the dynamic analysis for spent fuel storage rack; moreover, the experimental seismic response can be the benchmark for dynamic analysis of CAP1400 spent fuel rack. [ABSTRACT FROM AUTHOR]

Steam Generator Tube Rupture (SGTR) accident was reclassified as Class III Design Basis Accident (DBA) in 1991 due to its higher frequency of occurrence. Under Class III condition, radioactive release safety criteria are even stricter than thermal hydraulic criteria. In this paper, HPR1000 SGTR accident was simulated with two kinds of mitigation measures after steady-state initialization. Key system responses (such as pressure, mass flow rate in primary and secondary system, SG level, short term phase terminating time and radioactive release) were analyzed and compared for the two strategies. No SG overfilled nor other thermal hydraulic limit occurred for both strategies, while lower SG overfill risk and higher radioactive release risk existed for strategy 1 and converse situation existed for strategy 2. Specially, it is recommended VDA to be actuated by SG level, and instant main feedwater and auxiliary feedwater isolations were strongly required for strategy 2. When designing SGTR mitigation measure, there should be a compromised balance between SG overfill risk and radioactive release. Optimized mitigation strategy can be obtained through more accurate activity signal and higher auxiliary feedwater mass flow rate. [ABSTRACT FROM AUTHOR]

All advanced NPPs are required to improve safety strategy to cope with design extension condition (DEC) after Fukushima accident. One design improvements for HPR1000 is adoption of secondary Passive Residual Heat Removal System (PRS) to cope with station blackout accident together with loss of all feedwater, which can help to strengthen defense in depth capability and enrich DEC mitigation measures. To make the best use of PRS decay heat removal capability, investigation of PRS startup performance and decay heat removal capability after MFLB accident are investigated based on HPR1000 RELAP5 model. PRS startup response featured by pressure, mass flow rate and void fraction are investigated under three startup strategies. Strategy 3(backflow valve open strategy) is recommended due to minimum fluctuation during initiating stage, in which strategy steam isolation valve is normally open and backflow isolation valve is used to initiate PRS. Then Main Feedwater Line Break (MFLB) accident was simulated with the assumption of loss of auxiliary feedwater. Response characteristics in primary loop and PRS are investigated. After PRS being initiated after scram with 60 s delay, natural circulation in PRS will establish rapidly, and decay heat will be removed under natural circulation operation mode. No overall boiling occurred in primary circuit before isolation of affected SG. Secondary PRS is adequate to remove decay heat after DEC. [ABSTRACT FROM AUTHOR]

Under seismic conditions, the motion of spent fuel storage racks will be influenced greatly by fluid-structure interaction (FSI) between water, racks and walls. In the design configurations of CAP1400 (a new reactor type designed by China), known as end tube connection construction, the water between the rack storage cells can flow out of the rack, forming a perforate structure. However, the FSI effect for such structure has not been studied in depth. In engineering, the perforations are ignored, leading to an over-conservative added mass. The FSI effect of racks can be categorized into rack-wall FSI (also known as the FSI effect of neighboring wall) and rack-rack FSI. Free vibration experiments for a rack model were carried out to study the FSI effect of neighboring wall of CAP1400 rack. By dimension analysis, it is determined that KC number, oscillatory Re number, and nondimensional gap size are key factors for added mass. The experiments obtained the fluid added mass with different combinations of key parameters. An added mass expression for CAP1400 rack near a wall was obtained and the error was evaluated. The results can be utilized as a reference for the FSI input of the dynamic analysis of CAP1400 spent fuel rack. [ABSTRACT FROM AUTHOR]

• The modal tests and FEM simulations are carried out on single-end-fixed coaxial cylindrical shells. • The rigidity of the outer shell has a great impact on the fluid-structure interaction effect of the single-end-fixed model. • There is a 50% difference in added mass for two types of boundary conditions when the circumferential wave number is 3. There are many coaxial cylindrical thin-wall structures with narrow fluid gaps in the reactor vessel of the pool-type fast reactor. The fluid–structure interaction characteristics of such structures are important for the reactor's safety in an earthquake. However, the fluid–structure interaction characteristics of coaxial cylindrical shells with different boundary conditions and the inner-outer-shell coupling remains to be explored. In this paper, we study the added mass for the shell-type vibration of single-end-fixed coaxial cylindrical shells by vibration experiments and the finite element method (FEM). Then, we study the influence of the outer shell's rigidity and different boundary conditions (single-end-fixed and double-end-fixed cylinder) on the modal frequency and the added mass by the FEM. It shows that different boundary conditions cause a 50% difference in the added mass with modes with small circumferential wave numbers. This work provides a benchmark for strong fluid–structure interaction for flexible double shells with small fluid gaps. [ABSTRACT FROM AUTHOR]

Decay heat removal system (DHRS) is an essential system for pool-type sodium-cooled fast reactor (SFR) to ensure natural circulation and remove decay heat during station blackout (SBO) accident. Decay heat exchanger (DHX) is the key equipment in DHRS. In China Experimental Fast Reactor (CEFR), DHX layout scheme consisting of two DHXs in hot pool (DHX Layout 1) is adopted. Different DHX layout scheme is designed in the advanced pool-type SFR, for example DHXs are installed in both cold and hot pool (DHX Layout 2). It is necessary to analyze the influence of different DHX layout schemes. In this paper, two DHX layout schemes were proposed. Numerical simulation was carried out, of which method was preliminarily validated. The numerical results show that DHX Layout 2 can reach long-term cooling stage of natural convection earlier than DHX Layout 1 with strengthened flow rate, resulting in more uniform temperature distribution in the hot pool. There is a local asymmetrical temperature distribution in the cold pool for DHX Layout 2. These findings by numerical simulations provide important numerical reference for the subsequent DHRS design of the advanced pool-type SFR. [ABSTRACT FROM AUTHOR]

Whether the control rod can drop down in time is one of the important guarantees for the safe operation of the nuclear power plant. The drop-down process of the control rod is very complicated. For a long time, the researchers have done a lot of work on that, but it is hard to consider all the nonlinear factors. This paper considers the main factors together. Based on the theoretical analysis, we developed the nonlinear dynamics response analysis software for the nuclear power plant, which can be used to calculate the rod’s drop-down time. Compared with the results of the experiments, the software we developed proves to be applicable and reliable. [ABSTRACT FROM AUTHOR]

To cope with SBLOCA in absence of High-Head Safety Injection (HHSI) from design of HPR1000, atmospheric relief system (originally named as VDA in French) is uniquely designed to help to trigger Middle Head Safety Injection (MHSI) or Low Head Safety Injection (LHSI) earlier through cooling primary system quickly after SBLOCA. To make the best use of VDA decay heat removal capability, primary and secondary system of HPR1000 was modeled with RELAP5/SCDAP computer code. After steady-state initialization, a cold leg 30 mm break SBLOCA was simulated with six simulation conditions and five additional cases including availability of ACCU, different VDA discharge locations and area. Response characteristics of primary loop under different VDA working conditions are investigated. Pressurizer pressure decreases rapidly to lower level to trigger the reactor scram, VDA activation and accumulator safety injection sequently. Peak cladding temperature is 899.45 K occurring at 222 s, which is far below the safety limit. Activation of VDA can trigger ACCU SI earlier with a critical point, while positive reactivity will be introduced due to negative moderator temperature effect and Doppler effect. Larger VDA discharge capability will introduce larger reactivity feedback, as well as induce lower core level and SG level. It's suggested that VDA discharge condition should be chosen before the critical point, with the compromise with reactivity feedback introduced due to the negative moderator temperature effect. [ABSTRACT FROM AUTHOR]

A nodal neutron kinetics model with a higher-accuracy algorithm for three-dimensional multi-group neutron diffusion calculation is developed for transient analysis of reactor core in hexagonal-z geometry. The spatial discretization technique is based on a high order nodal expansion method for hexagonal-z geometry. In this method, one dimensional radial and axial spatially flux of each node and energy group are defined as quadratic polynomial expansion and four order polynomial expansion respectively. The partially-integrated radial and axial leakages are both approximated by the quadratic polynomial. The time discretization technique is based on a fully implicit scheme combined with a Runge-Kutta method for solving the neutron flux and delayed neutron precursor equations. A flux weighting method is adopted to calculate homogenized multi-group cross sections for transient analysis of control rod ejection accident. The described numerical methods are implemented into the computer code HNHEXK. This code is tested with one static benchmark problem and several kinetic benchmark problems. The numerical results show that the current method has a good agreement with the reference. [ABSTRACT FROM AUTHOR]

The heat transfer performance of Passive Residual Heat Removal Heat Exchanger (PRHR HX) immerged in the In-containment Refueling Water Storage Tank (IRWST) is of great importance for the efficient and safe removal of the residual heat in the AP1000 plant. The C-shape heating rods bundle is being used in the PRHR HX, yet the heat transfer characteristics of this special C-shape heat exchanger are not completely clear. The Westinghouse Electric Corporation has performed the PRHR HX experiment using three vertical tubes to validate the design of PRHR HX. However, the horizontal sections of the C-shaped bundle were neglected, and attentions were mainly focused on the local heat transfer effect of the vertical tubes. In the present work, an overall scaled-down IRWST and PRHR HX models are built to simulate the thermal hydraulic process in the residual heat removal accident. The natural convection heat transfer characteristics of both the overall IRWST model and the local C-shaped PRHR HX bundle model are researched in the first stage. For the overall IRWST model, the natural convection flow vectors are qualitatively measured by PIV technique. The velocity and temperature distributions are influenced mutually, then thermal stratification phenomenon develops in the overall IRWST. The existence of the “thermal interface” has great influences on the thermal stratification. For the local PRHR HX heat transfer, traditional natural convection correlations for both horizontal and vertical bundles are compared to the experimental data. For the vertical bundle of the PRHR HX model, a revised Rayleigh number is proposed as a criterion for the slender vertical rod bundles, when the vertical rod bundle outside flow cannot be treated as flat plate. Then revised correlations are proposed to predict the PRHR HX average heat transfer coefficient. Moreover, the analysis shows that the forced convection also plays an important role in enhancing the heat transfer effect of the PRHR HX model upper horizontal bundle, so the modified Numix is recommended for the PRHR HX upper horizontal bundle secondary heat transfer calculation. [ABSTRACT FROM AUTHOR]

A “single-hole plate + tube bundles” (SHTB) hybrid porous media model was developed to investigate the hydraulic characteristics of sodium cooled fast reactor (SFR) steam generator (SG). In this hybrid model, single-hole plate porous media (SHPPM) model was used to discover subtle features induced by irregularly-broached tube support plates (TSP). While the tube bundles porous media was aimed at simplifying the heat transfer tube region, the combination of this two sub-models seized the balance between computational accuracy and calculating scale. In the SHPPM model, the commercial CFD software FLUENT was used to simulate the pressure drop of sodium when flowing through the single-hole plate under different inlet velocities. Two key factors-viscous resistance coefficient and the inertia resistance coefficient-were calculated by the V– Δ P formula. Comparing the results with the lateral flow velocity caused by the original sized single-hole plate model, this SHPPM model verified the validity and accuracy of the simulation. The hybrid porous media model reasonably simulated the maximum lateral flow velocity on the sodium side of SG, particularly in regions surrounding the TSPs. As a result, the maximal lateral flow velocity of the tube side entrance is 2.834 m/s and in the region around the TSPs is 1.3 m/s. Therefore, this work not only provided some useful flow field data to help forecast flow-induced vibration (FIV) phenomenon, but also proposed an innovative approach for simulating the SG of SFR. [ABSTRACT FROM AUTHOR]

A three-dimensional, multigroup, diffusion code based on a high order nodal expansion method for hexagonal-z geometry (HNHEX) was developed to perform the neutronic analysis of hexagonal-z geometry. In this method, one-dimensional radial and axial spatially flux of each node and energy group are defined as quadratic polynomial expansion and four-order polynomial expansion, respectively. The approximations for one-dimensional radial and axial spatially flux both have second-order accuracy. Moment weighting is used to obtain high order expansion coefficients of the polynomials of one-dimensional radial and axial spatially flux. The partially integrated radial and axial leakages are both approximated by the quadratic polynomial. The coarse-mesh rebalance method with the asymptotic source extrapolation is applied to accelerate the calculation. This code is used for calculation of effective multiplication factor, neutron flux distribution, and power distribution. The numerical calculation in this paper for three-dimensional SNR and VVER 440 benchmark problems demonstrates the accuracy of the code. In addition, the results show that the accuracy of the code is improved by applying quadratic approximation for partially integrated axial leakage and four-order approximation for one-dimensional axial spatially flux in comparison to flat approximation for partially integrated axial leakage and quadratic approximation for one-dimensional axial spatially flux. [ABSTRACT FROM AUTHOR]

The effective and reliable operation of the Passive Residual Heat Removal System (PRHRS) is very significant for the safety of the reactor. However, the passive heat sink tank IRWST in PRHRS is vulnerable to get stratified, which has deleterious impacts on the heat transfer effect of the Passive Residual Heat Removal Heat Exchanger (PRHR HX). In the present work, both the numerical simulations and experiments are conducted to investigate the passive baffles designs of the PRHR HX&IRWST. An overall scaled-down separate-effects PRHR HX&IRWST experimental test bench has been built, and the experimental data are used to validate the numerical methods in the prototype PRHR HX model. Different passive baffles Designs A–D are tested employing the commercial CFD software CFX, and then validated by the baffles experiment. The influencing factors of baffles number, thermal conductivity, and baffles shape are investigated. Both the numerical and experimental temperature distributions, natural convection velocity, thermal stratification extent, as well as the Heat Transfer Coefficient are analyzed. The complicated factors combine to determine the overall HTC in baffle design. The numerical and experimental results provide possible improvement ways for the optimization of the PRHR HX baffles design, which have practical engineering application significance in the optimum design of the passive safe system in AP1000 plant. [ABSTRACT FROM AUTHOR]

• A coupled calculation method is validated and applied to SFR safety analysis. • The transient process of the SBO accident in CEFR is simulated by coupling the SAC-DRACS and FLUENT. • The hysteresis in heat transfer among DHRS circuits is simulated. • The residual heat removal capacity of DHRS is evaluated. In station blackout(SBO) accident, the establishment and maintenance of natural circulation in the China Experimental Fast Reactor(CEFR) mainly relies on the decay heat removal system(DHRS). However, due to the complex structure of the reactor, it's very difficult to model the full-scale CEFR. In the present work, DHRS and the primary circuit system of the CEFR are modeled by the System Analysis Code for Direct Reactor Auxiliary Cooling System (SAC-DRACS) code and three-dimensional CFD software FLUENT, respectively. Then the long-term transient simulation of 4000 s is calculated by coupling SAC-DRACS and CFD codes. The overall temperature stratification and local core temperature variations in the pool caused by DHRS are analyzed. Meanwhile, the residual heat removal capacity of DHRS is evaluated, the average core outlet temperature does not exceed 500.3 °C under the action of DHRS. This work provides an extension of the research method for the pool-type fast reactor safety analysis. [ABSTRACT FROM AUTHOR]

• The influence of the connecting pieces between the rack storage units on the FSI effect was studied. • The influence of the distance between the ends of the rack and the pool wall on the FSI effect was studied. • An added mass coefficient formula with four impact factors is proposed for spent fuel storage rack of CAP1400. Spent fuel in nuclear power plants is usually stored temporarily in spent fuel pools in the form of racks. Seismic analysis of the racks is required for the structural safety of the spent fuel. Since the racks are arranged in water, the seismic analysis of the racks should consider the fluid–structure interaction effect (i.e. the added mass of fluid imposed on spent fuel storage racks). The traditional computational method for the added mass is to simplify the racks to a concentric square column with a two-dimensional flow. This simplification improves computational efficiency but results in excessive margins. In addition, since the rack of CAP1400 (a Chinese AP-type PWR) uses a porous structure, which is different from the conventional rack, the traditional simplified method cannot be used to calculate the added mass of the rack of CAP1400. Therefore, this paper proposes a computational method for the added mass, in which the three-dimensional flow, as well as the internal pores of the rack, are considered so that the excessive margin can be reduced. Using this new method, a formula with four factors was given to calculate the added mass coefficient. Among four factors, the dimensionless gaps in X and Y directions represent two-dimensional flow effect, while the dimensionless height of the rack represent three-dimensional flow effect. The new method is verified by a scaled down experiment for rack of CAP1400. [ABSTRACT FROM AUTHOR]

Safe shutdown behavior of China Experimental Fast Reactor (CEFR) in emergency status is one of the key points in its safety assessment. The CEFR design incorporates a pool-type structure. However, the flow and heat transfer characteristics in the pool are complex, which results in obvious three-dimensional (3D) effects. Therefore, obtaining the 3D flow paths and accurate temperature distribution characteristics using a one-dimensional system code is difficult. In the present work, 3D computational fluid dynamics simulations were performed on the pool-type sodium-cooled fast reactor CEFR in full-scale. The "decoupling modeling and coupling calculation" method was used for performing numerical simulation to comprehensively investigate the 3D thermal and flow characteristics under emergency shutdown conditions. The convection and the corresponding heat transfer effects were evaluated by investigating both inter-flow and intra-flow paths in the reactor core. FLUENT 18.0 results were compared and verified with system code SELTAC simulation results. The results revealed that the residual heat was mainly dissipated by the intra-subassembly flow under reactor emergency shutdown conditions. The maximum core outlet temperature did not exceed 543.4 °C under emergency shutdown event, which is lower than the safety limit. This phenomenon indicates that if the heat-removal capability is maintained in the main heat exchange system, CEFR shutdown protection is effective. The three-dimensional results of the present work provide important numerical references for CEFR safety analysis. [ABSTRACT FROM AUTHOR]

Spent fuel rack is the key equipment for the storage of spent fuel after refueling. In order to investigate the performance of the spent fuel rack under the earthquake, the phenomena including sliding, collision, and overturning of the spent fuel rack were studied. An FEM model of spent fuel rack is built to simulate the transient response under seismic loading regarding fluid-structure interaction by ANSYS. Based on D’Alambert’s principle, the equilibriums of force and momentum were established to obtain the critical sliding and overturning accelerations. Then 5 characteristic transient loadings which were designed based on the critical sliding and overturning accelerations were applied to the rack FEM model. Finally, the transient displacement and impact force response of rack with different gap sizes and the supporting leg friction coefficients were analyzed. The result proves the FEM model is applicable for seismic response of spent fuel rack. This paper can guide the design of the future’s fluid-structure interaction experiment for spent fuel rack. [ABSTRACT FROM AUTHOR]

There is a large water storage tank installed at the top of containment of AP1000, which can supply the passive cooling. In the extreme condition, sloshing of the free surface in the tank may impact on the roof under long-period earthquake. For the safety assessment of structure, it is necessary to calculate the impact pressure caused by water sloshing. Since the behavior of sloshing impacted on the roof is involved into a strong nonlinear phenomenon, it is a little difficult to calculate such pressure by theoretical or numerical method currently. But it is applicable to calculate the height of sloshing in a tank without roof. In the present paper, a simplified method was proposed to calculate the impact pressure using the sloshing wave height, in which we first marked the position of the height of roof, then produced sloshing in the tank without roof and recorded the maximum wave height, and finally regarded approximately the difference between maximum wave height and roof height as the impact pressure head. We also designed an experiment to verify this method. The experimental result showed that this method overpredicted the impact pressure with a certain error of no more than 35%. By the experiment, we conclude that this method is conservative and applicable for the engineering design. [ABSTRACT FROM AUTHOR]

This study is motivated by a lack of information regarding the seismic analysis and design of AP1000 passive containment cooling system water storage tank (PCCWST). The major challenge faced in this specific type of fluid–structure interaction (FSI) problem is the estimation of the forces associated with the hydrodynamic pressure. In this paper, an equivalent mechanical model for the water tank of PCCWST was developed regarding the hydrodynamic pressure. Parameters of this newly-developed mechanical model, which was evaluated by finite element method (FEM), were given in chart form as functions of geometrical parameters including ratio of inner and outer tank radii, tank bottom’s inclination angle, and tank height. The values of impulsive mass were compared between the proposed model and the cylindrical tanks reported in the literature when their inclination angle approaches zero. The agreement shown by this comparison confirmed the validity of the proposed mechanical model. Shaking table experiments of a scaled PCCWST were also carried out to verify the feasibility of the proposed model. Finally, the transient response of PCCWST subjected to horizontal ground excitation was provided to illustrate the application of the proposed model. Therefore, the study shows that the equivalent mechanical model can provide an effective method for the design of PCCWST. [ABSTRACT FROM AUTHOR]

The heat transfer effect of Passive Residual Heat Removal Heat Exchanger (PRHR HX) and buoyancy-induced flow in the In-containment Refueling Water Storage Tank (IRWST) are of great importance for the efficient and safe removal of the residual heat in the AP1000 reactor. Although some numerical studies have been conducted, only the standard k – ɛ model has been applied. Experimental validation of the simulation results was also not sufficient because of the lack of appropriate experimental data. In the present work, the applicability of different Reynolds Average Navier–Stokes (RANS) turbulence models and Large Eddy Simulation (LES) were examined, utilizing the commercial CFD software CFX 14.5. Further, two types of grids were built for the high/low-Reynolds turbulence models, and the y + values as well as grids sensitivity were carefully analyzed. Meanwhile, overall scaled IRWST and PRHR HX models were built to simulate the thermal–hydraulic process in the residual heat removal accident, which was a new overall scaled separate effect IRWST&PRHR HX experiment. More than 150 thermocouples were utilized to measure the temperature in the key regions, and Particle Image Velocimetry (PIV) was utilized for the measurement of the flow velocity. Based on the validation of turbulence models in simulating the overall variations of temperature and velocity field in the IRWST model, the transient heat transfer capacity of PRHR HX was then analyzed. The results indicated that the low-Reynolds Shear Stress Transport (SST) model with multi-sublayer grid was appropriate for the simulation of buoyancy-induced flow. Nusselt numbers obtained from numerical simulations, experimental data, and empirical correlations were further compared to analyze the heat transfer mechanism. Combination factors including the special C-shape, flow resistance, and turbulent mixing effects imposed important influences on the heat transfer effects of the PRHR HX model. It was confirmed by numerical results, experimental data, as well as empirical correlations that the heat transfer capability of the vertical section was better than the horizontal section. [ABSTRACT FROM AUTHOR]

Component cooling water system (CCS) pump building is an important building connecting nuclear island and conventional island in pressurized water reactor (PWR). When main feed water pipe which crosses through CCS pump building breaks, large amount of fluid leakage may cause flooding hazard to the important equipment in CCS pump building. It is necessary to carry out waterproofing design for the safety of CCS pump building. Usually, important equipment in CCS pump building is designed with a certain height of waterproofing base, which needs to be further verified. It is almost impossible to conduct experimental analysis due to the large size and complex structures of CCS pump building. In the present work, numerical simulation was used to analyze the waterproofing ability of CCS pump building under main feed water pipe broken accident. Calculation method was verified by experiment. It shows that large amount of fluid accumulates in CCS pump building. For more serious flooding condition, water level around CCS pumps reaches 861 mm which exceeds the height of existing CCS pump waterproofing base. A waterproofing improvement design was proposed for CCS pump building of AP1000-type PWR, which feasibility was also analyzed by numerical simulation. [ABSTRACT FROM AUTHOR]

Lead cooled Fast Reactor (LFR) is attracting attention due to its advantages in safety. The SVBR 75/100 (Lead–Bismuth Fast Reactor) is a 75–100 MWe concept utilizing Lead–Bismuth eutectic alloy (LBE) as coolant in which the power level depends upon the specific design parameters chosen. A safety analysis code for Lead cooled Fast Reactor (SAC-LFR), developed to conduct analysis of the thermal–hydraulic behaviors and safety performance of SVBR 75/100, is discussed in this paper. A plant control model which is incorporated into SAC-LFR is developed to simulate the plant control system (PCS) of the fast reactor. In the present work, normal transients such as load following are modeled by SAC-LFR to show that the PCS model is available for simulating plant control in this code. Unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) are simulated by SAC-LFR to test and verify the rector’s safety. The results show that SVBR 75/100 has very good inherent safety characteristics due to the neutron and thermal-physical properties of LBE. The key factor that limits its safety performance is the cladding materials. UTOP is more dangerous than ULOF for SVBR 75/100. The safety performance in UTOP could be considerably improved by decreasing the control rod reactivity of SVBR, which can be done thanks to the favorable characteristics of the coolant. [ABSTRACT FROM AUTHOR]

AP1000 passive containment cooling system water storage tank (PCCWST) is set at the top of a shield building. Its function is to hold the water which will flow by gravity to help cool the containment vessel in the highly unlikely event of an emergency – one of the AP1000 plant’s innovative passive safety systems. The huge amount of water may pose a threat to the integrity of the shield building during an earthquake. In this paper, both modal characteristics and transient response of a scaled elevated water tank model were studied by a shaking table experiment. The acceleration of the supporting structure and displacement of water sloshing were recorded under transient loading. Four numerical models were built, including two-mass equivalent model, single lumped-mass model, and finite element models using Fluid30 and Fluid80 elements. The advantages and disadvantages of different numerical methods were discussed. Moreover, finite element (FE) models of AP1000 shield building were established considering the fluid–structure interaction (FSI) effect. The sloshing phenomenon in PCCWST under resonant three-cycle sine (RTCS) wave and seismic loading was studied. The response of shield building under seismic loading was also studied. The results showed that the maximum von Mises stress occurred at the corner of the air inlet. The shaking table experiment can provide a benchmark for seismic response of an elevated water tank. The FE model of shield building may help improve the design of the AP1000 and CAP1400 in the future. [ABSTRACT FROM AUTHOR]

The deposition of particle-like corrosion products on the inner surface of SG (steam generator) tube is likely to have negative effects on the safe operation of a PWR (pressurized water reactor). In this research, a model considering particles as elastic spheres is proposed to predict the deposition distribution of corrosion products. In the model, the collisions between particles and wall are modeled based on a spring–mass system. Based on the model, numerical calculations were performed to analyze the influence of various Reynolds numbers and elbow curvatures on particle deposition. The computation results of the particles’ motion and deposition distribution in SG tubes indicated good agreements with the experimental results. [ABSTRACT FROM AUTHOR]

Huge water storage tank on the top of many buildings may affect the safety of the structure caused by fluid-structure interaction (FSI) under the earthquake. AP1000 passive containment cooling system water storage tank (PCCWST) placed at the top of shield building is a key component to ensure the safety of nuclear facilities. Under seismic loading, water will impact the wall of PCCWST, which may pose a threat to the integrity of the shield building. In the present study, an FE model of AP1000 shield building is built for the modal and transient seismic analysis considering the FSI. Six different water levels in PCCWST were discussed by comparing the modal frequency, seismic acceleration response, and von Mises stress distribution. The results show the maximum von Mises stress emerges at the joint of shield building roof and water around the air inlet. However, the maximum von Mises stress is below the yield strength of reinforced concrete. The results may provide a reference for design of the AP1000 and CAP1400 in the future. [ABSTRACT FROM AUTHOR]

Flow instability investigation is one of the key issues in respect of the design of a nuclear reactor. Considering the possible occurrence of flow instabilities in the supercritical water-cooled reactor (SCWR), numerical analysis has been carried out on the supercritical flow instabilities applying APROS, which is a simulation software product for full-scale modeling and dynamic simulation of industrial processes. Based on the experience in BWR, three cases of density wave oscillations (DWOs) that might occur in the supercritical water-cooled reactor are modeled by APROS, including those in a single channel, two-parallel channels and natural circulation loop. The modeling and analysis of each case of DWO are presented and discussed in detail. Stability boundaries are obtained by APROS, parts of which are compared with previous experimental or numerical studies, showing good coincidence for each DWO case. Several key issues such as the effect of inlet temperature on stability boundary, oscillating period and effect of time step on numerical results are discussed. The discrepancy in oscillating period between the numerical and experimental results is discussed by means of theoretical analysis. Conjecture is raised that some important boundary conditions in experiments are ignored. The work presents an endeavor to establish a common methodology for the stability analysis in SCWR. [ABSTRACT FROM AUTHOR]

Highlights: [•] SAC-CFR is self-developed system analysis code for fast breeder reactor. [•] The three-dimensional hot pool model code is coupled in SAC-CFR. [•] SAC-CFR is used to analyze LOFWS tests of the EBR-II. [•] SAC-CFR can analyze flow patterns of the flow filled in sodium pool. [Copyright &y& Elsevier]

• The stress of central measuring shroud under thermal shock accident is calculated. • The fluid–solid thermal coupling method is used to obtain more accurate temperature and stress fields. • The position of maximum stress and the law of stress change with time are obtained. Thermal shock threatens the safe operation of the fast reactor and damages the structural integrity of the fast reactor. The stress of central measuring shroud under thermal shock accident in the "Monju" reactor is studied in this paper. The method of fluid–solid thermal coupling is adopted to solve the temperature field by fully coupling, while the stress field by one-way coupling. By comparing the temperature field with the experimental data, it is found that the temperature field is in good agreement with the error range within 3.44%. Under thermal shock condition, the core outlet temperature decreases with time, and the stress field shows a law consistent with the temperature field. The maximum stress position is located at the junction of the lower head and the sleeve. Based on the 1/3 model, thermal shock can be divided into three stages. At the initial stage of thermal shock, in the first 5 s after the thermal shock begins, the temperature gradient of the inner and outer walls increases with time, and the stress increases with time in the first 7 s after the thermal shock begins, reaching a maximum value of 345.67 MPa in 7 s. In the middle stage of thermal shock, between 5 and 15 s, the temperature gradient of the inner and outer walls decreases with time. Between 15 and 23 s, the temperature gradient of the inner and outer walls is small. The stress decreases with time between 7 and 24 s, reaching a minimum of 144.15 MPa in 24 s. At the end of thermal shock, after 23 s, the temperature gradient of the inner and outer walls increases with time. The stress increased with time after 24 s and reached a maximum of 339.62 MPa at 60 s. The maximum deformation is 23.38 mm in 8 s. Considering that the stress value is significant, further low-cycle fatigue analysis is recommended throughout the life cycle, which is very important. [ABSTRACT FROM AUTHOR]

• The three-dimensional flood discharge simulation analysis is performed under three accidents in pressure compartment of the CCS pump building. • A new drainage grid is added to discharge the overflow water and the drainage effect is verified. • Four new flood discharging schemes are put forward to ensure the safety of CCS pumps. • The calculation results provide design references and verification of waterproof flooding. The feed water pipes of steam generators (SG) stretch across the pump building of the component cooling water system (CCS), thereby connecting the reactor containment and the steam turbine building in the pressurized water reactor of a nuclear power plant. When main feed water pipes rupture, large quantities of water and steam may discharge into the CCS pump building, which causes the CCS pumps to be submerged and the CCS system to shut down, thereby threatening the safety of nuclear power plants. Therefore, it is necessary to analyze the overflow characteristics after such an accident. However, it is difficult to perform experiments in the CCS pump building, owing to its huge area and complex structure. In this study, numerical simulation was performed via computational fluid dynamics by using the volume of fluid model, in which three phases (water, air, and steam) were considered. Validation & Verification (V&V) is performed by comparing the calculated results with experimental verification of the small model, so that the reliability of the calculation method and results are proved. Overflow characteristics in three different accidental conditions were also analyzed: the pump trip accident triggered by double-end fracture (TDEF), pump trip accident triggered by conservative maximum flow rate discharge (TmaxD), and pump trip accident not triggered by a fracture (TnotF) in the pressure compartment. It has been revealed that under TDEF and TmaxD, the maximum water level around the two CCS pumps is lower than the waterproof foundation bed of the CCS pumps. However, under TnotF, the highest water level around the two CCS pumps is much higher than the waterproof foundation bed of the CCS pumps, which may threaten their safety. To address this problem, certain new optimization design schemes have also been proposed. [ABSTRACT FROM AUTHOR]