Thermal-hydraulic Modeling and Simulation of Piston Pump

This paper presents a kind of modeling approach to the study of the thermal-hydraulic piston pump which is used in the airplane comprehensively. A set of lumped parameter mathematical models are developed which are based on conservation of energy. Heat transfer analysis for the piston pump is also given in the paper in which the heat flow inside the piston pump is described precisely. The theoretical basis and modeling stratagy are applied in a typical thermal-hydraulic circuit containing the piston pump. Simulation results are presented which show a comparison of model/rig performance and the agreement obtained demonstrates the validity of the modeling approach.

Core and blanket thermal-hydraulic analysis of a molten salt fast reactor based on coupling of OpenMC and OpenFOAM

In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released from the fuel salt and transferred to the second loop by fuel salt circulation.Therefore,the MSFR is characterized by strong interaction between the neutronics and the thermal hydraulics.Moreover,recirculation flow occurs,and nuclear heat is accumulated near the fertile blanket,which significantly affects both the flow and the temperature fields in the core.In this work,to further optimize the conceptual geometric design of the MSFR,three geometries of the core and fertile blanket are proposed,and the thermal-hydraulic characteristics,including the three-dimensional flow and temperature fields of the fuel and fertile salts,are simulated and analyzed using a coupling scheme between the open source codes OpenMC and OpenFOAM.The numerical results indicate that a flatter core temperature distribution can be obtained and the hot spot and flow stagnation zones that appear in the upper and lower parts of the core center near the reflector can be eliminated by curving both the top and bottom walls of the core.Moreover,eight cooling loops with a total flow rate of0.0555 m3 s-1 ensur an acceptable temperature distribusure an acceptable temperature distribution in the fertile blanket.

Rapid Thermal-Hydraulic Analysis and Design Optimization of ITER Upper ELM Coils

ITER edge localized mode （ELM） coils are important components of the in-vessel coils （IVCs） and they are designed for mitigating or suppressing ELMs. The coils located on the vacuum vessel （VV） and behind the blanket are subjected to high temperature due to the nuclear heat from the plasma, the Ohmic heat induced by the working current and the thermal radiation from the environment. The water serves as coolant to remove the heat deposited into the coils. Based on the results of nuclear analysis, the thermal-hydraulic analysis is performed for the preliminary design of upper ELM coils using a rapid evaluation method based on 1D treatment. The thermal-hydraulic design and operating parameters including the water flow velocity are optimized. It is found that the rapid evaluation method based on 1D treatment is feasible and reliable. According to the rapid analysis method, the thermal hydraulic parameters of two water flow schemes are computed and proved similar to each other, providing an effective basis for the coil design. Finally, considering jointly the pressure drop requirement and the cooling capacity, the flow velocity is optimized to 5 m/s.

A Thermal-Hydraulic Coolant Channel Module (CCM) for Single- and Two-Phase Flow

A theoretical “drift-flux based thermal-hydraulic mixture-fluid coolant channel model” is presented. It is the basis to a corresponding digital “Coolant Channel Module (CCM)”. This purpose derived “Separate-Region Mixture Fluid Approach” should yield an alternative platform to the currently dominant “Separate-Phase Models” where each phase is treated separately. Contrary to it, a direct procedure could be established with the objective to simulate in an as general as possible way the steady state and transient behaviour of characteristic parameters of single- and/or (now non-separated) two-phase fluids flowing within any type of heated or non-heated coolant channels. Their validity could be confirmed by a wide range of verification and validation runs, showing very satisfactory results. The resulting universally applicable code package CCM should provide a fundamental element for the simulation of thermal-hydraulic situations over a wide range of complex systems (such as different types of heat exchangers and steam generators as being applied in both conventional but also nuclear power stations, 1D and 3D nuclear reactor cores etc). Thereby the derived set of equations for different coolant channels (distinguished by their key numbers) as appearing in these systems can be combined with other ODE-s and non-linear algebraic relations from additional parts of such an overall model. And these can then to be solved by applying an appropriate integration routine. Within the solution procedure, however, mathematical discontinuities can arise. This due to the fact that along such a coolant channel transitions from single- to two-phase flow regimes and vice versa could take place. To circumvent these difficulties it will in the presented approach be proposed that the basic coolant channel (BC) is subdivided into a number of sub-channels (SC-s), each of them being occupied exclusively by only a single or a two-phase flow regime. After an appropriate nodalization of the BC (and thus its SC-s) and after applying a “modified finite volume method” together with other special activities the fundamental set of non-linear thermal-hydraulic partial differential equations together with corresponding constitutive relations can be solved for each SC separately. As a result of such a spatial discretization for each SC type (and thus the entire BC) the wanted set of non-linear ordinary differential equations of 1st order could be established. Obviously, special attention had to be given to the varying SC entrance or outlet positions, describing the movement of boiling boundaries or mixture levels along the channel. Including even the possibility of SC-s to disappear or be created anew during a transient.

Integration of CATHENA Thermal-Hydraulic Model with CANDU 6 Analytical Simulator Controller

This paper introduces a powerful design and analysis tool named SIMCAT, that is developed to support applications to license a CANDU nuclear reactor, refurbish projects, and support the existing CANDU stations. It consists of the CATHENA （Canadian Algorithm for Thermo-Hydraulic Network Analysis）, the control logics from C6SIM （CANDU 6 Analytical Simulator）, and a communication protocol, PVM （parallel virtual machine）. This is the first time that CATHENA has been successfully coupled directly with a program written in another language. The independence of CATHENA and the C6SIM controllers allows the development of both CATHENA and C6SIM controller to proceed independently.

Numerical Simulations of Upper Plenum Thermal-Hydraulics of Monju Reactor Vessel Using High Resolution Mesh Models

In order to evaluate the effects of the FHs （flow holes） on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.

Effects of Rectangular Wing Vortex Generators on the Thermal-Hydraulic Performance of Louvered Fin and Flat Tube Heat Exchanger

In this paper,a novel composite heat transfer enhancement technique comprised of louvered fins(LFs)and rectangular wing vortex generators(RWVGs)is proposed to improve the LF side thermal-hydraulic performance of louvered fin and flat tube heat exchangers(LFHEs).After validation of the LF side pressure dropΔP and heat transfer coefficient hLFof the baseline by experiments,the numerical method is applied to investigate the influential mechanisms of the RWVG parameters(the number N(7 to 15),attack angleβ(30°to 90°),height H_(VG)(0.8 mm to 2 mm)and width W_(VG)(0.8 mm to 1.2 mm))on the performance of the LFHE in the velocity range of 3 m/s to 10 m/s.Results show that thermal-hydraulic performance of the LFHE is significantly impacted by the RWVGs,and according to the performance evaluation criteria(PEC),the LFHE achieves its optimal thermal-hydraulic performance when N=7,β=45°,H_(VG)=1.8 mm and W_(VG)=1 mm.Compared to the baseline,the maximum,minimum and average increments of PEC for the optimal case are 13.85%,4.67%and 8.39%,respectively.

THREE-DIMENSIONAL NUMERICAL SIMULATION OF THERMAL-HYDRAULIC PERFORMANCE OF A CIRCULAR TUBE WITH EDGEFOLD-TWISTED-TAPE INSERTS

Three-dimensional numerical simulations and experiments were carried out to study the heat transfer characteristics and the pressure drop of air flow in a circular tube with Edgefold-Twisted Tape （ETT） inserts and with classic Spiral-Twisted-Tape （STT） inserts of the same twist ratio. The RNG turbulence model for mildly swirling flows, the enhanced wall treatment for low Reynolds numbers, and the SIMPLE pressure-velocity method were adopted to simulate the flow and heat transfer characteristics. Within the range of Reynolds number from 2 500 to 9 500 and the twist ratio y from 5.4 to 11.4, the Nusselt number of the tube with ETT inserts is found to be 3.9% - 9.2% higher than that with STT inserts, and the friction factor of the tube with ETT inserts is 8.7% - 74% higher than that of STT inserts. The heat enhancement is due to higher tangential velocity and asymmetrical velocity profile with the increase and decrease of the periodic velocity within an edgefold length. It is found that main factors affecting the heat transfer of ETT inserts are the twist angle and the gap width between the tube and inserts. A larger twist angle leads to a higher tangential velocity, and larger Nusselt number and friction factor. The thermal-hydraulic performance slowly decreases as the twist angle increases. The gap width between tube and inserts has a significant influence on the heat transfer, while little influence on pressure drops. The thermal-hydraulic performance increases in average by 124% and 140% when the gap width reduces from 1.5 mm to 1.0 mm and 0.5 mm. The larger the gap width, the higher velocity through the gap will be, which would reduce the main flow velocity and tangential velocity. So a small gap is desirable. Comparing experimental and numerical results at variable air flow and tube wall temperature, the numerical results are found to be in a reasonable agreement with the experiment results, with difference of the Nusselt number in a range of 1.6% - 3.6%, and that of the friction factor in a range of 8.2% - 13.6%.

Thermal-Hydraulic Calculation and Analysis on Evaporator System of a 1000 MW Ultra-Supercritical Pulverized Combustion Boiler with Double Reheat

A double reheat ultra-supercritical boiler is an important development direction for high-parameter and large-capacity coal-fired power plants due to its high thermal efficiency and environmental value.China has developed a 1000 MW double reheat ultra-supercritical boiler with steam parameters of 35 MPa at 605℃/613℃/613℃.Reasonable water wall design is one of the keys to safe and reliable operation of the boiler.In order to examine the thermal-hydraulic characteristics of the double reheat ultra-supercritical boiler,the water wall system was equivalent to a flow network comprising series-parallel circuits,linking circuits and pressure nodes,and a calculation model was built on account of the conservation equations of energy,momentum and mass.Through the iterative solving of nonlinear equations,the prediction parameters of the water wall at boiler maximum continue rate(BMCR),75%turbine heat-acceptance rate(THA)and 30%THA loads,including total pressure drops,flow distribution,outlet steam temperatures,fluid and metal temperatures were gotten.The results of calculation exhibit excellent thermal-hydraulic characteristics and substantiate the feasibility of the water wall design of the double reheat ultra-supercritical boiler.

Thermal-hydraulic Stability Analysis of Nuclear Heating Reactors

The two-phase flow instability that can occur in a natural circulation system is of importance in the design of nuclear heating reactors. The time domain code RETRAN-02 and the frequency domain code NUFREQ were applied to estimate the instability boundary and the effects of such parameters as pressure, inlet resistance and riser height in NHR-5 and an experimental loop. The results of the calculations and the experiments are in good agreement. Nonlinear density wave oscillations were analyzed using the RETRAN-02 code. The theory of nonequilibrium thermodynamics was used to find an explicit criterion to estimate the threshold of the stability. Experimental simulation of the nuclear feedback density wave instability was also carried out in a test loop using. computer controlled electric power.

Core designing of a new type of TVS-2M FAs:neutronics and thermal-hydraulics design basis limits

One of the most important aims of this study is to improve the core of the current VVER reactors to achieve more burn-up(or more cycle length)and more intrinsic safety.It is an independent study on the Russian new proposed FAs,called TVS-2M,which would be applied for the future advanced VVERs.Some important aspects of neutronics as well as thermal hydraulics investigations(and analysis)of the new type of Fas are conducted,and results are compared with the standards PWR CDBL.The TVS-2M FA contains gadolinium-oxide which is mixed with UO_(2)(for different Gd densities and U-235 enrichments which are given herein),but the core does not contain BARs.The new type TVS-2M Fas are modeled by the SARCS software package to find the PMAXS format for three states of CZP and HZP as well as HFP,and then the whole core is simulated by the PARCS code to investigate transient conditions.In addition,the WIMS-D5 code is suggested for steady core modeling including TVS-2M FAs and/or TVS FAs.Many neutronics aspects such as the first cycle length(first cycle burn up in terms of MWthd/kgU),the critical concentration of boric acid at the BOC as well as the cycle length,the axial,and radial power peaking factors,differential and integral worthy of the most reactive CPS-CRs,reactivity coefficients of the fuel,moderator,boric acid,and the under-moderation estimation of the core are conducted and benchmarked with the PWR CDBL.Specifically,the burn-up calculations indicate that the 45.6 d increase of the first cycle length(which corresponds to 1.18 MWthd/kgU increase of burnup)is the best improving aim of the new FA type called TVS-2M.Moreover,thermal-hydraulics core design criteria such as MDNBR(based on W3 correlation)and the maximum of fuel and clad temperatures(radially and axially),are investigated,and discussed based on the CDBL.

Preliminary safety analysis for heavy water-moderated molten salt reactor

The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel.Issues arising from graphite in traditional molten salt reactors,including the positive temperature coefficient and management of highly radio-active spent graphite waste,can be addressed using the HWMSR.Until now,research on the HWMSR has been centered on the core design and nuclear fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization.However,the core safety of the HWMSR has not been extensively studied.Therefore,we evaluate typical accidents in a small modular HWMSR,including fuel salt inlet temperature overcooling and overheating accidents,fuel salt inlet flow rate decrease,heavy water inlet temperature overcooling accidents,and heavy water inlet mass flow rate decrease accidents,based on a neutronics and thermal-hydraulics coupled code.The results demonstrated that the core maintained safety during the investigated accidents.

Probabilistic analysis of embankment slope stability in frozen ground regions based on random finite element method

Prediction on the coupled thermal-hydraulic fields of embankment and cutting slopes is essential to the assessment on evolution of melting zone and natural permafrost table, which is usually a key factor for permafrost embankment design in frozen ground regions. The prediction may be further complicated due to the inherent uncertainties of material properties. Hence, stochastic analyses should be conducted. Firstly, Karhunen-Loeve expansion is applied to attain the random fields for hydraulic and thermal conductions. Next, the mixed-form modified Richards equation for mass transfer （i.e., mass equation） and the heat transport equation for heat transient flow in a variably saturated frozen soil are combined into one equation with temperature unknown. Furthermore, the finite element formulation for the coupled thermal-hydraulic fields is derived. Based on the random fields, the stochastic finite element analyses on stability of embankment are carried out. Numerical results show that stochastic analyses of embankment stability may provide a more rational picture for the distribution of factors of safety （FOS）, which is definitely useful for embankment design in frozen ground regions.

Development of a dynamics model for graphite-moderated channel-type molten salt reactor

A molten salt reactor(MSR) is one of the six advanced reactor concepts selected by the generation Ⅳ international forum because of its advantages of inherent safety, and the promising capabilities of Th-U breeding and transuranics transmutation. A dynamics model for the channel-type MSR is developed in this work based on a three-dimensional thermal–hydraulic model(3DTH) and a point reactor model. The 3DTH couples a three-dimensional heat conduction model and a one-dimensional single-phase flow model that can accurately consider the heat conduction between different assemblies. The 3DTH is validated by the RELAP5 code in terms of the temperature and mass flow distribution calculation. A point reactor model considering the drift of delayed neutron precursors is adopted in the dynamics model. To verify the dynamics model, three experiments from the molten salt reactor experiment are simulated. The agreement of the experimental data and simulation results was excellent.With the aid of this model, the unprotected step reactivity addition and unprotected loss of flow of the 2 MWt experimental MSR are modeled, and the reactor power and temperature evolution are analyzed.

Flow field effect of delayed neutron precursors in liquid-fueled molten salt reactors

In molten salt reactors(MSRs),the liquid fuel salt circulates through the primary loop and a part of the delayed neutron precursors(DNPs)decays outside the reactor core.To model and analyze the flow field effect of DNPs in channel-type liquid-fueled MSRs,a three-dimensional space-time dynamics code,named ThorCORE3D,that couples neutronics,core thermalhydraulics,and a molten salt loop system was developed and validated with the Molten Salt Reactor Experiment(MSRE)benchmarks.The effects of external loop recirculation time,fuel flow rate,and core flow field distribution on the delayed neutron fraction loss of MSRE at steadystate were modeled and simulated using the ThorCORE3D code.Then,the flow field effect of the DNPs on the system responses of the MSRE in the reactivity insertion transient under different initial conditions was analyzed systematically for the channel-type liquid-fueled MSRs.The results indicate that the flow field condition has a significant effect on the steady-state delayed neutron fractions and will further affect the transient power and temperature responses of the reactor system.The analysis results for the effect of the DNP flow field can provide important references for the design optimization and safety analysis of liquid-fueled MSRs.

Feasibility of an innovative long-life molten chloride-cooled reactor

Molten salt-cooled reactor is one of the six GenIV reactors with promising characteristics including safety,reliability,proliferation resistance,physical protection,economics,and sustainability.In this paper,a small innovative molten chloride-cooled fast reactor(MCCFR)with 30-year core and a target 120-MWt thermal power was presented.For its feasible study,neutronics,thermal-hydraulics,and radiation damage analysis were performed.The key design properties including kinetics parameters,reactivity swing,reactivity feedback coefficients,maximum accumulated displacement per atom(DPA)of reactor pressure vessel(RPV)and fuel cladding,and maximum coolant,cladding,and fuel temperatures were evaluated.The results showed the MCCFR could operate without refueling for 30 years with overall negative reactivity feedback coefficients up the end of its life.During its 30-year life,the excess reactivity was well managed by constantly pulling out the control rods.The maximum accumulated DPA on RPV and fuel cladding were 8.92 dpa and 197.03 dpa,respectively,which are both below the design limits.Similarly,the maximum coolant,cladding and fuel center temperatures were all below the design limits during its entire lifetime.According to these results,the MCCFR core design with long life is feasible.

Implementation of high-fidelity neutronics and thermal–hydraulic coupling calculations in HNET

To perform nuclear reactor simulations in a more realistic manner,the coupling scheme between neutronics and thermal-hydraulics was implemented in the HNET program for both steady-state and transient conditions.For simplicity,efficiency,and robustness,the matrixfree Newton/Krylov(MFNK)method was applied to the steady-state coupling calculation.In addition,the optimal perturbation size was adopted to further improve the convergence behavior of the MFNK.For the transient coupling simulation,the operator splitting method with a staggered time mesh was utilized to balance the computational cost and accuracy.Finally,VERA Problem 6 with power and boron perturbation and the NEACRP transient benchmark were simulated for analysis.The numerical results show that the MFNK method can outperform Picard iteration in terms of both efficiency and robustness for a wide range of problems.Furthermore,the reasonable agreement between the simulation results and the reference results for the NEACRP transient benchmark verifies the capability of predicting the behavior of the nuclear reactor.

Simulation of a molten salt fast reactor using the COMSOL Multiphysics software

In this study,COMSOL v5.2 Multiphysics software was utilized to perform coupled neutronics and thermal–hydraulics simulations of a molten salt fast reactor,and the SCALE v6.1 code package was utilized to generate the homogenized cross-section data library.The library’s 238 cross-section groups were categorized into nine groups for the simulations in this study.The results of the COMSOL model under no fuel flow conditions were verified using the SCALE v6.1 code results,and the results of the neutronics and thermal–hydraulics simulations were compared to the results of previously published studies.The results indicated that the COMSOL model that includes the cross-section library generated by the SCALE v6.1 code package is suitable for the steady-state analysis and design assessment of molten salt fast reactors.Subsequently,this model was utilized to investigate the neutronics and thermal–hydraulics behaviors of the reactor.Multiple designs were simulated and analyzed in this model,and the results indicated that even if the wall of the core is curved,hot spots occur in the upper and lower portions of the core’s center near the reflectors.A new design was proposed that utilizes a flow rate distribution system,and the simulation results of this design showed that the maximum temperature in the core was approximately 1032 K and no hot spots occurred.

Fusion-Driven Sub-Critical Dual-Cooled Waste Transmutation Blanket:Design and Analysis

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.

A Review of the Scaling Study of the CANDU-6 Moderator Circulation Test Facility

Following the previous relevant works [1]-[3], a scaling analysis is performed to derive a set of scaling criteria which were thought to be suitable for reproducing the major thermal-hydraulic phenomena in a scaled-down CANDU moderator tank similar to that in a prototype power plant during a full power steady state condition. The objective of building a scaled-down moderator tank is to generate the experimental data necessary to validate the computer codes which are used to analyze the accident analysis of CANDU-6 plants. The major variables of interests in this paper are moderator flow velocity and temperature of the moderator which is D2O inside the moderator tank during a steady state and transient conditions. The reason is that the local subcooling of the moderator is found to be a critical parameter determining whether the stable film boiling can sustain on the outer surface of the calandria tube if the contact of overheated pressure tube and cold calandria tube should occur due to pressure tube ballooning during LBLOCA with ECC injection failure [4]. The key phenomena involved include the inlet jet development and impingement, buoyancy force driven by the moderator temperature gradient caused by non-uniform direct heating of the moderator, and the pressure drop due to viscous friction of the flow across the calandria tube array. In this paper, the previous researches are reviewed, some concerns or potential problems associated with them implied by comparing CFD analyses results between the CANDU-6 moderator tank and 1/4 scaled-down test facility are described, and as a way to examine the assumption of the scaling analysis is true an order-of-magnitude analyses are performed. Based on the results of these analyses the assumption of neglecting ?and ?terms cannot be justified for the power of 0.5 MW and 1.566 MW for the 1/4 scaled-down facility. Further investigation is thought to be necessary to confirm this result, i.e. if the scaling of the previous work1 is justifiable by some other independent analyses.