HTGR Research Articles

The deterministic neutron transport calculation for the HTR-PM by the three-dimensional method of characteristics code ARCHER

Due to the strong geometric adaptability, the Three-Dimensional (3-D) Method Of Characteristics (MOC) is a promising candidate for 3-D whole-core high-fidelity neutron transport calculations. However, the 3-D MOC can hardly be applied to large-scale full core problems on account of the huge computational costs in memory and time. In the past, the 3-D MOC was mainly used in traditional Light Water Reactor (LWR) calculations, but rarely used for more complex core calculations. For pebble-bed High Temperature gas-cooled Reactors (HTRs), most 3-D MOC codes are out of ability to construct complex pebble beds, let alone efficient acceleration methods. The 3-D MOC code ARCHER can efficiently simulate the large-scale pebble-bed HTRs through using the Linear Source Approximation (LSA), the Coarse Mesh Finite Difference (CMFD) and the hybrid MPI-OpenMP parallel. In this work, the two-level CMFD acceleration and efficient preconditioners on Krylov subspace linear solvers are implemented in the ARCHER code in order to achieve better performance. In addition, the practical HTR-PM criticality problem is calculated by ARCHER, which is the world’s first high-fidelity neutron transport solution of large-scale pebble-bed HTRs by deterministic numerical method. It also shows that the 3-D MOC has great application potential in other complex reactors.

Numerical study of the flow and heat transfer characteristics in a scale model of the vessel cooling system for the HTTR

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models’ execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system’s size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier–Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

Optimization of preparation process and study on the properties and microstructures of nuclear natural graphite powder for HTR pebble fuel elements

Nuclear natural graphite powder (NNGP) is the key and main component in the matrix graphite (MG) of pebble fuel elements for high temperature gas-cooled reactors (HTRs). In this study, a modified lab-scale powder preparation process of NNGP proposed for the first time would contribute to a more efficient and lower energy-consumption preparation system. The optimized NNGP was successfully prepared by the modified process, whose properties and microstructures were comprehensively investigated. The results indicated that the optimized NNGP possessed a lower specific surface area, a higher graphitization degree and fewer structural defects. MG pebbles manufactured by the optimized NNGP exhibited higher crush strength and thermal conductivity, a lower corrosion rate, as well as better thermal stability during heat treatment. This indicated that the modified NNGP was conducive to improving the integrity and safety of MG pebbles.

Spatial evolution mechanism of vortex structure in the highly-loaded helium compressor cascade

ABSTRACT The highly loaded design method of helium compressors can effectively solve the difficulty in compressing helium in High Temperature Gas-cooled Reactors (HTGR). But it also causes obviously different attack angle characteristics of blade surface loads in a highly loaded helium compressor compared to air compressors. This difference inevitably affects separation characteristics and flow loss within the compressor. In the current study, the effects of highly loaded design methods and changes in attack angle on the separation characteristics of the compressor cascade are analyzed by applying a numerical simulation method first. Then, the influence of Mach number on the loss characteristics of the cascade for a highly loaded helium compressor is systematically analyzed. Finally, the effect of differences in the material properties of working fluid on the separation characteristics is discussed. The results indicate that the proportion of secondary flow loss to the total loss in highly loaded compressor cascades is 2.46 times larger than that in conventionally loaded ones. While the properties of working fluid have an effect on the performance of the compressor cascade, their effects on the weight factor of vortex loss are highly limited.

T production using a high-temperature gas-cooled reactor for the DEMO fusion reactor: Li rod structure for the initial irradiation test

This work proposes tritium (T) production using a high-temperature gas-cooled reactor (HTGR) and studies it for the initial T inventory in a demonstration fusion reactor and a prior engineering test with T handling. At this stage, the aim is to investigate the compatibility between electricity and T production, which makes the stable confinement of T in the Li-loading rods during the HTGR operation period a crucial issue. The total T outflow into the helium gas during the reactor operation period was attempted to be reduced using Zr spheres with Ni coating as a T absorption material to avoid an increase in the inner T pressure. The basic hydrogen absorption properties of the Zr spheres with Ni coating, which coexist in an environment with Al2O3 and/or LiAlO2 oxides, were measured. The structures of the Li-loading rod for a typical commercial HTGR and irradiation test were studied using the obtained data. In addition, an outline of the irradiation test is reported.

Kinetic modeling of IG-110 oxidation in inert atmosphere with low oxygen concentration for innovative high-temperature gas-cooled reactor applications

ABSTRACT The oxidation behavior of IG-110, a graphite core component, was investigated at temperatures ranging from 400 to 1000°C in a 10 ppm Ar/O2 flow to simulate the oxidation process between the graphite core component and helium coolant with low O2 concentrations employed in advanced High-Temperature Gas-cooled Reactors (HTGRs). The results reveal that IG-110 undergoes significant mass loss at temperatures above 700°C, resulting in total mass changes of −1.5%, −5.3%, and −9.0% at 700, 800, and 1000°C, respectively, during a 10-hour oxidation period. No significant mass loss is observed below 600°C. To understand the oxidation mechanism of IG-110 under low O2 concentrations, we propose a kinetic model as the current chemical kinetic-controlled model does not fully explain the oxidation behavior observed in our research. Our analysis shows lower estimated reaction rates compared to studies at higher O2 concentrations; the activation energy values exhibit good agreement. The proposed kinetic model sheds light on the oxidation mechanism of IG-110 under 10 ppm Ar/O2 flow. This study provides new insights into the oxidation behavior of graphite core components in HTGRs and highlights the importance of controlling the O2 concentration in the helium coolant to prevent severe degradation of SiC-matrix fuel compacts.

Assessing potential of high temperature reactor facilities of Russian design for near-term deployment of hydrogen production

In frames of the work, a pilot comparative multi-criteria evaluation of several options for the production of hydrogen using the energy of nuclear reactors was carried out. Technical and economic models of hydrogen production by methods of steam methane reforming, high-temperature and low-temperature electrolysis using nuclear reactors as a source of thermal and electrical energy have been developed. Russian projects of high-temperature gas-cooled reactors and existing VVER-type NPPs are considered. An approach to a comparative multi-criteria evaluation of options for hydrogen production using the energy of nuclear reactors is proposed. The proposed approach has been tested by comparing several variants of hydrogen production. When evaluating the values of the criteria, data obtained from open sources, with the help of developed economic models and expert assessments, were used. As a result, the considered variants of hydrogen production were ranked using several metrics and sets of criteria weights. The sensitivity analysis of metric values and option ratings to changes in the main criteria showed a sufficient sustainability of the multi-criteria analysis carried out.

Evaluation of the spatial self-shielding impact for TRISO-based nuclear fuel depletion

Reactor physics analyses of nuclear cores with nuclear fuel concepts containing tristructural isotropic (TRISO) particles, such as pebbles or compact fuel elements, rely on various degrees of simplification to keep these highly heterogeneous problems computationally tractable. One such limitation regards the level of spatial discretization employed during burnup calculations, where traditionally only a limited number of spatial zones are modeled at the full core level and assume that the spectrum is constant within the fuel elements and TRISO particles in this depletion zone. This type of assumption neglects the impact of spatial self-shielding effect within the kernels (microscale level) as well as within the compact or pebbles (mesoscale level). The Monte Carlo code Serpent 2 contains many relevant features for efficiently modeling this type of geometry, including a collision-based domain decomposition intended for very large burnup calculations, which we leveraged for this work to quantify the impact of capturing neutron flux variations occurring at the micro- and mesoscale level on a series of high-temperature gas-cooled reactor fuel element depletion problems. While spatial self-shielding is observed at both scales, with differences from a volume-averaged burnup of ±7% within the kernels and ±2% between TRISO particles within the fuel element, the conjugated effect on nuclide inventories and multiplication factor are negligible, hence confirming that assuming a single average spectrum value may be sufficient for most applications.

Shutdown dose rate calculations in high-temperature gas-cooled reactors using the MCNP-ORIGEN activation automation tool

Background High-temperature gas-cooled reactors (HTGRs) have many distinct features from the current light water reactor (LWR) fleet, and potential decommissioning strategies should take them into consideration. The proper characterization of the shutdown dose rates can establish a suitable strategy. Methods This article introduces a new shutdown dose rate calculation capability that relies on the MCNP-ORIGEN activation automation tool and the MCNP repeated structures to explicitly model the TRistructural ISOtropic (TRISO) particles as decay radiation sources. Results Three exercises are conducted with this capability, and their results discussed. The first two exercises verify and demonstrate the workflow. The third exercise proposes a decommissioning strategy for a high-temperature gas-cooled microreactor and studies its feasibility from the shutdown dose rate standpoint. The calculations yield a dose rate map above at the reactor citadel after a three month cool-down, showing values above 40 mSv/h. Conclusions The findings imply that the decommissioning strategy based on placing the reactor in a safe shutdown state and allowing it to cool down for three months before removing various components and extracting the fuel assemblies, may not be sufficient to minimize unnecessary exposure of personnel.

Cross section generation for a conceptual horizontal, compact high temperature gas reactor

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

A Whole-Core Transient Thermal Hydraulic Model for Fluoride Salt-Cooled Reactors

Abstract A thermal hydraulic model is developed for a solid pin-fueled fluoride-salt-cooled small modular advanced high temperature reactor (SmAHTR). This preconceptual SmAHTR was developed by the Oak Ridge National Laboratory (ORNL). For the fuel assembly configuration investigated in this study, the fuel and non-fuel pins are arranged in a hexagonal layout. The molten FLiBe salt coolant flows parallel to the bank of pins. A finite volume model is developed and used to compute temperatures in the solid regions (fuel and non-fuel pins, and the graphite reflectors) in the core. The temperature, flow, and pressure profiles for the coolant flowing through the pin bundles in the core are calculated using the conventional subchannel methodology. Pertinent closure relations are used to compute the hydraulic losses, momentum, and energy exchange between adjacent subchannels, and heat transfer between the solid and fluid regions. The resulting model can perform both steady-state and transient computations across the entire core. This fully implicit model also includes an adaptive time-stepping algorithm for automatic time-step adjustment. A preliminary code-to-code comparison demonstrates good agreement between the present subchannel-based model and a computational fluid dynamics (CFD)-based model for a transient case in which the core inlet flowrate varies with time. Following the code-to-code comparison, the thermal hydraulic model is used to analyze the protected loss of heat sink (P-LOHS) accident scenario. Key results such as the transient evolution of peak fuel, non-fuel, and coolant temperatures, and 3-D core temperature distribution are presented and discussed.

Utilization of organic Rankine cycles in a cogeneration system with a high-temperature gas-cooled nuclear reactor – thermodynamic analysis

Utilization of organic Rankine cycles in a cogeneration system with a high-temperature gas-cooled nuclear reactor – thermodynamic analysis

Development of a hybridized small modular reactor and solar-based energy system for useful commodities required for sustainable cities

The current study develops a hybridized small modular nuclear reactor and solar-based system designed specifically for sustainable communities in metropolitan areas to meet their power, heat, clean fuel, fresh water and food requirements. Both floating-type and bifacial-type photovoltaic arrays are integrated with a high-temperature gas-cooled small modular reactor. In the integrated system, a Rankine cycle subsystem for power generation, a solid oxide electrolyzer for hydrogen production, a pressure swing adsorption unit with ammonia reactor for nitrogen and then ammonia production, a multi-effect desalination unit for freshwater generation, and heat recovery units in order to provide heating to residential area, food drying facility, greenhouse, and fish farm facilities. For both floating and bifacial PV options, the 120 MWp PV modules are considered and integrated with two small modular reactor units at 250 MWth capacities each, 500 MWth in total. A thermodynamic analysis, which is based on energy and exergy approaches, is carried out. The time-dependent analyses are carried out for floating-type PV and bifacial-type PV options for the cities of Istanbul in Turkey; Barcelona in Spain; Los Angeles (LA) in the United States; Tokyo in Japan; and Toronto in Canada by using their hourly meteorological data, source and load capacities in a typical meteorological year. Among ten cases, between 112835 and 146180 tonnes of ammonia are generated in a typical year while meeting the heating and power requirements of sustainable communities in metropolitan areas. Among ten cases, the average overall energy and exergy efficiencies are found as 41.04% and 46.88%, respectively. The hybridization of such nuclear and solar systems achieves unique advantages by preventing intermittencies compared to the renewable-alone systems, reducing carbon emissions compared to fossil-based systems, and lowering the levelized costs of energy compared to the nuclear energy-alone systems.

Nuclear hydrogen production through carbonaceous-matter gasification. A physicochemical optimization

To reach the worldwide decarbonisation and the corresponding climate change mitigation, low-carbon energy sources must be used in the near-term and the role of the nuclear energy in pathways to produce net zero emissions is outstanding. This is not only because nuclear energy produces carbon-free energy, but also because it can supply high-temperature process heat to industrial applications such as the hydrogen production. Hydrogen is presently considered as the energy vector of the future to attain these environmental goals.In this paper, the possible use of a nuclear-assisted steam coal gasification process for hydrogen production is investigated in deep. As a modern High Temperature Gas Cooled Reactor (HTGR) is able to produce a gas outlet temperature of 950 °C, the nuclear heat can be transferred to a steam coal gasification plant via an intermediate helium circuit for providing the high-temperature process heat required by the endothermic gasification reactions between the steam and the carbonaceous matter.The nuclear-assisted gasification technology is considered to be attractive for the processing of an Argentine orthoasphaltite that is extracted from a minefield located in Neuquén province. The present research comprises both theoretical and experimental studies, which were addressed to get the necessary information about the optimal conditions for the hydrogen production though the steam gasification process using the Argentine orthoasphaltite as feedstock.A conceptual design of an energy complex for nuclear hydrogen, electricity and liquid fuels production is also proposed. In this conceptual design, the steam coal gasification reactor consists of two concentric vessels specially designed to separate the pyrolysis step occurring at low temperature and the char gasification step that takes place at higher temperatures. In this way, it is possible to give a full use to the volatile compounds that are present in the feedstock and are identified and characterized in this work.

Assessment of Reactor Physics Characteristics of Prismatic Fuel Concepts With a Hydrogen-Based Moderator for Use in a Fluoride Salt-Cooled Small Modular Reactor

Abstract Fluoride-salt-cooled high temperature reactor (FHR) effectively combines the solid fuel and moderator design of high-temperature gas-cooled reactor (HTGR) technology with the fluoride salt coolant (LiF-BeF2, FLiBe) of molten salt reactor (MSR) technology, enabling low-pressure (∼1 atm, 101.325 kPa), and high-temperature (∼700 °C) operations. The design and operational features of the FHR make it a potentially attractive option for a small modular reactor (SMR), provided that it can be modified and made physically small and operate at a low-enough power level (<350 MWth/<150 MWel). Most FHR-SMR designs use high-assay low enriched uranium (HALEU) fuel in the form of tri-structural isotropic (TRISO) fuel particles, combined with the use of a graphite moderator. However, there are alternative design concepts for an FHR-SMR that may offer superior performance characteristics, while utilizing an alternative fissile fuel supply option. In this exploratory study, lattice physics calculations were performed with Serpent to evaluate an alternative FHR-SMR prismatic fuel block design concept using coated annular fuel pellets instead of TRISO-particle fuel compacts, along with the use of hydrogen-based solid moderator rods made of 7LiH. In initial studies, it was found that fuel blocks with 120 moderator rods made of 7LiH tended to have large positive temperature reactivity coefficients (TRCs), which is undesirable for safety reasons. However, reducing the number of moderator rods to 90 or 54, while increasing the number of fuel rods and coolant holes led to low or negative temperature coefficients. For a prismatic fuel block design with 54-7LiH moderator rods, the isothermal temperature coefficient of reactivity (Isothermal TRC), with simultaneous changes in the fuel (F), graphite (G), hydrogen (H), and coolant (C) temperatures, ranges between −0.159 mk/K and −0.054 mk/K, depending on the operating temperature and fuel burnup. Such alternative FHR-SMR fuels could achieve a single-batch core life of ∼10 years with low enriched uranium (LEU) fuel, and ∼45 years with HALEU, in a 350-MWth reactor core.

Two-phase flow instability characteristics of HTGR Once Through Steam Generators

High Temperature Gas-cooled Reactors (HTGR) have the advantage of inherent safety compared with most Gen III reactors. They also have very high temperature, which make the plant have high electricity generation efficiency and have has the potential for hydrogen production. Once Through Steam Generator (OTSG) is responsible for transferring high temperature heat from primary loop helium to secondary loop composed of water and steam. The evaporation tubes of HTGR OTSG are very long due to the low heat transfer coefficient of helium convection over tube bundles. A Two-phase flow instability may occur in HTGR OTSG, which can cause pulsation of the mass flow rate, system pressure and temperature. This will not only interfere with the control system, but also cause thermal fatigue damage in the structures. Thus, it is important to predict and avoid two-phase flow instability in HTGR OTSG. Three types of flow instabilities are investigated for HTGR OTSG using both a linear analytical model and a nonlinear numerical model. They are flow excursion (or Ledinegg instability), density wave oscillation and pressure drop oscillation. For the nonlinear numerical model, RELAP5 software is adopted. Flow excursion and pressure drop oscillation are discussed based on the predicted hydraulic characteristic curve of HTGR OTSG (pressure drop vs. mass flow rate curve). Density wave oscillation is investigated using the transient variation of the outlet mass flow rate. For the linear analytical model, a set of ordinary differential equations are established using integration and small perturbation methods. The variables of differential equations are the length of the subcooling water region, the length of both the subcooling water and saturated two-phase regions, the average density of the saturated two-phase region and the inlet mass flux. The stability of OTSG is investigated using the eigenvalues of the coefficient matrix of the ordinary differential equations. Based on the results calculated using the linear analytical model and nonlinear numerical model, there is no flow excursion, pressure drop oscillation or density wave oscillation at the designed partial and full power levels (30%, 50%, 75% and 100%). Thus, HTGR OTSG can operate stably and have a sufficient safety margin at the designed power levels. Density wave oscillation will happen when decreasing system pressure. However, increasing inlet resistance coefficient can avoid density wave oscillation effectively. The tube lengths’ influences on the stability boundary of OTSG are also investigated and it shows no strong dependency.

Shaking Table Testing of a Scaled Nuclear Power Plant Structure with Base Isolation

To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records and three artificial earthquake motions were used to input the ground motion in the tests. The results demonstrated that the seismic performance of the plant was better and that the structure was in an elastic state, under a safe shutdown earthquake event. Isolation bearings were found to effectively reduce the dominate frequency of the structure. The acceleration amplification factor of the superstructure was found to be less than 1. The isolation test results showed that the peak of the floor response spectrum at the pressure vessel support was less than 0.1 g. In the nonisolation test, the peak of the floor response spectrum was greater than 1 g. In the isolation test, the relative displacement of the structure was less than 1.1 mm, which was relatively small. The structure maintained a good isolation performance and exhibited improved safety under extreme ground motion.

Optimum power determination and comparison of multi-pass and OTTO fuel management schemes of HTGR-based PeLUIt reactor

A co-generation reactor based on pebble-bed High-Temperature Gas-Cooled Reactor (HTGR) has been designed. This study focuses on designing a modular nuclear reactor based on pebble bed high-temperature reactor (Electric Power and Industrial Hot Steam Plant/Pembangkit Listrik dan Uap Panas Industri, abbreviated as PeLUIt), an upgrade from Indonesia’s 10 MW Experimental Power Reactor (Reaktor Daya Eksperimen, RDE). PeLUIt is proposed as a solution for electricity and co-generation application. The 5-pass and Once-Through-Then-Out (OTTO) refueling scheme for the PeLUIt has been analyzed. PEBBED was used for equilibrium core calculations, whereas COMBINE7 was used for spectrum calculations utilizing ENDF/B-VII.0 as nuclear data. Based on the results, it can be inferred that the PeLUIt co-generation reactor analysis with an optimal power of 40 MWt is produced by increasing fuel intake into the core while retaining core geometry and fuel design. In this study, burn-up levels for each refueling scheme are also analyzed, and the value for OTTO is lower than 5-pass refueling schemes. In addition, the Depressurized Loss of Forced Cooling (DLOFC) accident scenario for both types of refueling schemes was also analyzed. The result showed that the OTTO refueling scheme has a higher safety level than the 5-pass refueling scheme.

Theoretical analysis about model simplifications and boundary conditions on two-phase flow instability of parallel heated tube systems

Flow instability may happen in both subcritical two-phase flow boiling systems (such as U tube natural circulation steam generator of Pressurized Water Reactor, once through steam generator of High-Temperature Gas-Cooled Reactor, etc) and supercritical evaporation systems. In previous theoretical researches, the model of a single heated tube (SHT) with constant pressure drop boundary condition (ConPD) was always used to investigate the system stability. However, in real applications, there are usually multiple parallel heated tubes (MHT) and unheated connection regions before the branch heated tubes. The system is usually driven by a pump, and there is always a long unheated main pipe and throttling between the heated tubes (boiler or steam generator) and the pump. However, there lacks rigorous analysis about the effects of these model simplifications and boundary conditions. Take superheated two-phase flow boiling systems as examples, the theoretical models with different flow-driven boundary conditions (ConPD, constant mass flow rate (ConMF), and pump driven (PumpD)), different numbers of heat tubes (SHT, MHT), with or ignoring the unheated region (UHR), with or ignoring the bypass of a pump, with identical or un-identical geometry parameters and operation parameters are established. Linear ordinary differential equations describing the instability are obtained. Eigenvalues of the coefficient matrix of linear ordinary differential equations, and the property of the block matrix and the similarity matrix are used to analyze the effects of the above boundary conditions and model simplifications on the system stability. Two new dimensionless numbers, dimensionless pump number (Npump) and dimensionless bypass number (Nbypass), are proposed to describe the effect of pump and its bypass on stability quantitatively. The effect of the pump bypass on the stability can be attributed to the decrease of Npump. No matter ConPD or PumpD is adopted, SHT considering UHR is proved to be more stable than SHT ignoring UHR. No matter UHR is considered or ignored, SHT with PumpD is proved to be more stable than SHT with ConPD. No matter UHR of MHT at the main pipe is considered or ignored, the stability of MHT with ConPD or ConMF or PumpD is proved to be the same as that of SHT with ConPD ignoring UHR. When ConPD ignoring UHR is adopted, the stability of MHT with un-identical geometry and operation parameters is proved to be the same as that of its most unstable heated tube. The above conclusions about the effect of model simplification and boundary condition are obtained from superheated two-phase flow boiling systems, they also hold for saturated two-phase flow boiling systems and supercritical evaporation systems.

Improved measurement of local thermal conductivities of coatings by the steady-state Raman spectroscopy method: A case study in TRISO particles

Tristructural-isotropic particle (TRISO) is a fully ceramic microencapsulated fuel of high-temperature gas-cooled reactors with complex microstructural morphology and heat transfer process, hindering the accurate measurement of its thermal conductivity. Here, a dual test wavelength steady-state Raman spectroscopy method (SSRS) for determining the thermal conductivities is proposed, resolving the issue of the conventional SSRS failing to measure the thermal conductivity of 3C-SiC. By demonstrating prerequisites for using the SSRS method and modeling the conductive-radiative coupled heat transfer, the thermal conductivities of all layers in TRISO have been accurately measured. It is also found that the thermal conductivity of an underlying material can be estimated by measuring the material on its top if the outer material has low absorption under the applied laser wavelength, offering the possibility of measuring the buried layers in localized structures. The work extends the usable range of the SSRS method for obtaining thermal conductivities in wide applications.