HTGR Research Articles

Printed Circuit Heat Exchangers (PCHEs): A Brief Review

Heat exchangers and other heat transfer devices/systems play vital roles of heat transfer in thermal fluid flow systems for industrial application. Sodium cooled fast reactors are normally designed to have two loops of sodium coolants and one loop of water coolant which generates steam for power production. The two loops of sodium coolants consist of primary cooling system of sodium which cools the fuel rods of the reactor core and secondary cooling system of sodium transferring heat from the sodium primary cooling system. The water-cooling system transfers heat from the secondary cooling system of sodium for steam generation. Lead cooled fast reactors on the other hand are designed to have primary cooling system of lead cooling the fuel rods in the reactor core and secondary cooling system of water transferring heat from the lead cooling primary system for steam generation. Water cooled Nuclear Power Plants used water to cool the reactor core in the primary system and the heat removed from the core is used for steam generation directly as in BWRs and SCWRs or in the secondary system of heat exchanger as in PWRs. Other reactor systems such as Gas-cooled fast reactor (GFR), Molten-salt reactor (MSR), High-temperature gas-cooled reactor (HTGR), and Small Modular Reactors (SMRs) also have various types of heat exchangers in their designs to support power/electricity generation. Appropriate heat exchangers are therefore needed for various stages of heat transfer in power generation systems. Thus, Heat exchangers and other heat transfer devices/systems play vital roles of heat transfer in thermal fluid flow systems for industrial applications. This study presents brief review of PCHEs which have comparable advantages over other types of heat exchangers. Recent studies on PCHEs and other heat exchanger types have been reviewed. Design and optimization of PCHEs, optimization of Brayton and Rankine circles, and fluid flow and heat transfer devices/systems have been discussed briefly. The review findings show that the design and optimization of PCHEs depends on the intended industrial application of the heat exchanger. The various channel types and channel cross-section types available for design and optimisation as well as the design and optimised system being able to withstand high pressure and temperature conditions in addition to its compact size for the intended industrial application make PCHEs unique among other types of heat exchangers.

Fusion-fission hybrid reactor facility: power profiling

The current state of research in the field of nuclear and thermonuclear power aimed at creating power generation plants makes it possible to predict the further development of modern power industry in the direction hybrid reactor power plants. Such hybrid systems include a tokamak with reactor technologies, worked out in detail in Russia, and systems with an additional source of neutrons. Power generation plants using tokamaks and accelerators with the required level of proton energy will be of exceptionally large size and power, which will postpone their construction on an industrial scale to the distant future. The ongoing research is aimed at the development of small generation and has the prospect of entering the field of energy use in a shorter period. The hybrid reactor facility under study consists of an axisymmetric assembly of fuel blocks of a high-temperature gas-cooled reactor and a linear plasma source of additional neutrons. The paper demonstrates the results of optimization plasma-physical, thermophysical and gas-dynamic studies, the purpose of which is to level the distortions of the power density field, which are formed in the volume of the multiplicating part of the facility due to the pulsed operation of the plasma source of D-T-neutrons. The studies on increasing the “brightness” of the source and modeling its operating modes were carried out using the DOL and PRIZMA programs. The thermophysical optimization and gas-dynamic calculations were performed using the verified SERPENT and FloEFD software codes. The calculations were made on a high-performance cluster of the Tomsk Polytechnic University.

A GPU-based DEM model for the pebble flow study in packed bed: Simulation scheme and validation

This study developed a GPU-based DEM (GPU-DEM) model for the pebble flow in the packed bed. The details of the GPU-DEM scheme have been illustrated. The uniform mesh neighbor-search method, unified memory addressing (UMA), and special boundary conditions were incorporated in the GPU-DEM model. Cases of the direct impact of two identical particles, the angle of repose of particles drained from a lifted hopper, and the circulating hopper to mimic a pebble flow in High-Temperature Gas-cooled Reactor (HTGR) reactors were carried out to systematically verify the applicability, computational efficiency, and accuracy of the GPU-DEM model. Related results, like the unloading speed, porosity, and velocity distribution, obtained by the GPU-DEM model have been validated too. In addition, the effects of single-precision floating-point number computation on GPU were discussed. The results demonstrated that the GPU-DEM can achieve 18– 20 times acceleration while maintaining accuracy, even when single-precision floating-point numbers were used for calculation. Therefore, it is a powerful tool to explore the particle flows in chemical engineering applications.

Microstructural analysis of tristructural isotropic particles in high-temperature steam mixed gas atmospheres

High-temperature gas-cooled reactors (HTGRs) use tristructural isotropic (TRISO) particles embedded in a graphitic matrix material to form the integral fuel element. Potential off-normal reactor conditions for HTGRs include steam ingress with temperatures above 1,000 °C. Fuel element exposure to steam can cause the graphitic matrix material to evolve, forming an atmosphere composed of oxidants and oxidation products and potentially exposing the TRISO particles to these conditions. Investigating the oxidation response of TRISO particles exposed to a mixed gas atmosphere will provide insight into the stability under off-normal conditions. In this study, surrogate TRISO particles were exposed to high temperatures (T = 1,200 °C) in flowing steam (3% < pH2O < 21%) and CO (pCO < 1%) to determine the oxidation behavior of the SiC layer when exposed to various mixed gas atmospheres. Scanning electron microscopy, x-ray diffraction, and focused ion beam milling was used to determine the impact of CO and steam on the oxidation behavior of the SiC layer. The data presented demonstrates how the SiC layer showed strong oxidation resistance due to limited SiO2 growth and maintained its structural integrity under these off-normal conditions.

Improved Methods for Calculating the Neutron Diffusion Coefficient of General Pebble Bed

There exists a neutron streaming effect in the high-temperature gas-cooled pebble-bed reactor (HTGR) pebble-bed core caused by the spatial heterogeneity of the neutron’s free path, which has a remarkable impact on neutron leakage. It is necessary to take into consideration the streaming effect in evaluating the homogenized diffusion coefficient of the pebble bed, prior to the whole-core diffusion calculation. In this paper, two methods are proposed for calculating the homogenized multigroup diffusion coefficient of the pebble bed based on migration area conservation theory and Benoist’s theory, respectively. Compared with existing methods, the newly proposed methods are adaptable to a general pebble bed consisting of multitype pebbles and nonvacuum atmosphere. Numerical results demonstrate the proposed methods’ effectiveness and consistency in evaluation of the pebble-bed homogenized diffusion coefficient.

Design of a laser welding plugging system for the steam generators of high-temperature gas-cooled reactor

As fourth-generation nuclear power equipment, high-temperature gas-cooled reactor (HTGR) is attracting more and more attention both in industry and academia. The core outlet temperature of the steam generator (SG) in HTGR is much higher than that of pressurized water reactor steam generators, which is more likely to cause damage. Till now, welding plugging is still one of the important means to prevent the leakage of damaged heat transfer tubes. However, how to improve the automation of welding plugging to reduce the short-range operation time in a radioactive environment is still difficult. This work proposes the design of a laser welding plugging system for the HTGR. Firstly, the design requirements are analyzed and the overall design scheme is given. The whole system consists of a laser generator, a multi-axis robot, and supporting devices. Secondly, the key supporting devices including a laser welding head, a pneumatic clamping jaw, and a single-axis motion platform are explained in detail. The structure simulation based on SolidWorks is adopted, which provides the basis for the design of the single-axis motion platform. Thirdly, the control system which takes the PLC control system as the main control unit and the industrial computer as the monitoring data processing unit is described. Finally, an experimental system is built and the laser welding experiment is carried out to test the feasibility of the designed laser welding plugging system.

Inventories of Short-Lived Fission Gas Nuclides in Nuclear Reactors

Taking inventories in reactor cores is critical for understanding their radioactive source terms and establishing the relationship between the activity concentration in the primary loop and the status of the reactor core’s fuel. However, there is a niche in which a simple but accurate relationship between reactor conditions and nuclide inventories can reliably predict the fission gas nuclide activities of the reactor core in the primary loop. In this study, a simple and efficient model called “Inventories of a Point Reactor for Fission Gas Nuclides” (IPRFGN) was proposed to calculate and interpret such inventories, in which a 10 MW high-temperature gas-cooled experimental reactor (HTR-10) was used as the test case. The present study findings were consistent with those of a general point–depletion burnup code such as the KORIGEN code. Here, the relative error was &lt;1%. Based on the application of the IPRFGN model in HTR-10, the results indicate that the proposed IPRFGN model has provided the relationship between the inventories of fission gas nuclides in the core and the reactor conditions in all types of nuclear fission reactors. In the future, the IPRFGN model will be used for calculating fission gas nuclide inventories in various reactors.

Surface and in-depth structural changes in nuclear graphite irradiated with noble gases described with Raman imaging

4th Generation high-temperature gas-cooled nuclear reactors (HTGR) are regarded as possible sources of industrial heat in Poland and Europe, allowing for a substantial reduction of the dependency on gas and coal import. It is mainly due to their safety of use, reliability and economy in a current energetic crisis. In this work, graphite, as a primary construction material and neutron moderator in HTGR, was evaluated before and after ion irradiation since its properties depend on the material’s structure and purity. Commercial graphite materials (IG-110, NBG-17) and the laboratory’s in-home material were chosen for the exemplary samples. The structural damage in HTGR was simulated with energetic Ar+ and He+ ions with fluencies from 1E12 to 2E17 ion/cm2. Raman imaging was chosen to assess radiation damage build-up: the crystallites’ evolution, occurrence and types of defects. The recorded evolution showed stronger disordering of the material with heavier Ar+ ions than with He+.

Research on non-LWRs operational intervention level based on HTR

Nuclear emergency is the final line of defense for humanity and the environment, and operational intervention level (OIL) becomes a key technical element pre-specified in emergency preparedness and guides emergency response actions. The International Atomic Energy Agency (IAEA) issued a new version of OIL system based on light water reactors (LWRs) in 2017. To explore its applicability for non-LWRs, this paper presents a set of specified OIL determination procedures for high temperature gas-cooled reactors (HTRs). Considering a representative OIL based on ground monitoring (OIL1) as the case study, it concludes that the default OIL1 is 500 μSv/h for HTR-PM, half of the IAEA recommended value. In addition, through sensitivity analysis, we find that the slow release of radionuclides does not affect the determination of the default OIL1 value. From the perspective of OIL, this research further confirms that HTR reactor types can technically simplify off-site emergency requirements.

Development of a hybrid model for fuel performance analysis of spherical fuel elements under normal conditions

The spherical fuel element (SFE) dispersed with many tri-isotropic fuel (TRISO) particles has been primarily employed in the high-temperature gas-cooled reactor. Owing to the complex structure of SFE, behaviors will be evolved for both TRISO particles and the fuel matrix with the increasing burnup. To evaluate the fuel performance of SFE, the homogenization method has been conventionally adopted for SFE scale modeling, and dispersed TRISO particles are analyzed subsequently with fixed temperature boundary conditions. However, owing to the varying properties of TRISO particles under reactor operation, the homogenization models dedicated to certain circumstances can hardly reflect the properties evolution of TRISO with high fidelity, which may lead to uncertainties in analysis results. To this end, a hybrid model, namely the conventional homogenization model informed by the lower scale model online, has been developed in this paper, which may increase modeling fidelity for both SFE and particles. Specifically, property models for the TRISO particle and graphite matrix have been implemented in COMSOL. Subsequently, based on the linking of COMSOL and MATLAB, the hybrid model for SFE analysis was established. Through the hybrid model, the temperature field and stress distribution of TRISO, the evolution of effective properties, and the thermal and mechanical state of the SFE were obtained and analyzed. The results show that within the scope of this paper, the central temperature of TRISO in the innermost region rises with the increasing burnup for the decreasing thermal conductivity of materials, and the peak value is about 1324 K, which is about 60 K higher than that in the outermost region. The difference in the maximum plenum pressure is more than 5 MPa among TRISO particles, which is mainly caused by the variations in CO production. Consequently, the hoop stress of SiC in the outermost region is slightly elevated. The distribution of effective thermal conductivity (ETC) for TRISO is nearly 4.3 W/(m·K) at the beginning of life (BOL), while it was reduced to 3.6 W/(m·K) at the end of life (EOL). There are marginal variations along the radial direction for the ETC of TRISO. Meanwhile, the ETC of SFE decreases from about 21 W/(m·K) for fresh fuel to 13 W/(m·K) for spent fuel. The predicted effective elastic modulus and coefficient of thermal expansion of SFE increase gradually versus time. The former is caused by the densification of pyrolytic carbon, and the latter is mainly determined by the high volume fraction of matrix and the changes of CTE for pyrolytic carbon. In addition, the comparison in the temperature field between the proposed model and the Differential Effective Material Theory (D-EMT) model may further demonstrate the feasibility and effectiveness of the hybrid model.

Numerical study on heat transfer characteris tics of supercritical water in straight and helical tubes

The subcritical steam generators in high-temperature gas-cooled reactor pebble-bed modules can be replaced by supercritical steam generators to improve thermal efficiency. The axial and circumferential nonuniform flow and heat transfer characteristics of supercritical water in straight and helical tubes with varying inclinations (0–90°) are simulated to investigate the combined effects of tube structure, flow direction, thermophysical properties, property variations, buoyancy, and centrifugal forces. With heat fluxes of 233–930 kW/m2 and inlet mass fluxes of 225 or 1260 kg/(m2 s) under a working pressure of 24.5 MPa, the friction coefficients, turbulent kinetic energy, secondary flow, temperature, and heat transfer coefficients are studied in the cross section of variable structure tubes. The wall temperature and heat transfer coefficients of four regions are compared in straight and helical tubes; the maximum heat transfer coefficient appeared before the pseudo-critical point. The heat transfer coefficients are the greatest at the bottom of the straight tubes, where fluidity was high, and at the outer of the helical tubes, where the fluid velocity was high under centrifugal force. The secondary flow in the helical tubes causes a nonuniformity in the wall temperature, causing the circumferential maximum wall temperature to move.

Heat transfer of supercritical pressure water in helical tubes

Subcritical steam generators in the current high-temperature gas-cooled reactor pebble-bed module in China can be replaced by supercritical steam generators to better coordinate the reactors and supercritical steam turbine unit to improve the thermal efficiency with no fluid phase change at supercritical pressures. The heat transfer of supercritical-pressure water flow in helical tubes with various helix diameters was simulated using the re-normalization group k–ε model to investigate the effects of finite-thickness walls, inlet mass flux, and helix diameter. The finite-thickness wall barely affected local heat transfer in the straight tube with inlet mass flux of 1,260 kg/(m2 s) whereas the local heat transfer in the helical tube was very different with relative difference in q/qav between the inside and bottom, caused by the centrifugal force, is 36%. In helical tubes, the heat transfer coefficients in the inner bar were the lowest, with a low fluid velocity under the effect of the centrifugal force. The buoyancy and centrifugal force in the helical tubes caused non-uniformity in the wall temperature; however, when the fluid temperature increased below the pseudo-critical temperature, the temperature distributions showed an increasing trend. The non-uniformities in the wall temperature distributions decreased as the centrifugal force increased, and they were suppressed by the finite-thickness wall.

Simultaneous Solution of Helical Coiled Once-Through Steam Generator with High-Speed Water Property Library

Efficient simulation of the helical coiled once-through steam generator (H-OTSG) is crucial in the design and safety analysis of the high-temperature gas-cooled reactor (HTGR). The physical property and phase transformation of water in the steam generator brings great challenges during simulation. The water properties calculation routine occupies a large part of the computational time in the steam generator solution process. Thus, a thermohydraulic property library is developed based on the IAPWS-IF97 formulation in this work to reduce the computational cost. Here the formulation adopts the backward equation method to avoid iterations in thermodynamic property calculation. Moreover, two Newton-method-based simultaneous solutions are implemented as implicitly nonlinear solvers, including Jacobian-Free Newton–Krylov (JFNK) and Newton–Krylov (NK) methods, due to its excellent computational performance. These simultaneous solution algorithms are combined with the developed water property library to simulate the H-OTSG efficiently. The numerical analysis is performed based on the transient and steady-state cases of the HTR-10 steam generator. Successful simulations of HTR-10 steam generator cases demonstrate the capability of the newly developed method.

Indonesia's experimental power reactor program (RDE)

The experimental power reactor program (RDE) was created to introduce Indonesia to nuclear power plants (NPP). This was done by building a prototype of a non-commercial 10 MWth power reactor. The RDE program aims to show the public that Indonesian engineers can design reactors that can supply enough electricity and steam to support industrial needs. The technology used is a high-temperature gas-cooled reactor. The RDE program is a strategic and intermediate target for the country’s energy security and national sovereignty. In the long term, the development of this RDE-based nuclear power plant is expected to have implications for the use of clean energy and the reduction of the country’s dependence on fossil fuels. In addition, the implementation and development of the RDE program will increase national industrial capacity and global competitiveness. Moreover, the RDE program can be used as a reference for commercial nuclear power plants called PeLUIt (Power and Steam Generators for Industry) to meet the demand for industrial electricity and heat in various regions in Indonesia. However, many challenges arise in the implementation of this program, including the estimated significant development costs and the possibility of a long development period. To date, the RDE program has had a positive impact on strengthening the capacity of human resources in Indonesia in the field of power reactor design by being able to map the capabilities of the national industry to support nuclear power plant development programs. This is in addition to the formation of stakeholders from the Government, academics, and entrepreneurs.

Deep Learning for Multigroup Cross-Section Representation in Two-Step Core Calculations

We investigate using deep learning, a type of machine-learning algorithm employing multiple layers of artificial neurons, for the mathematical representation of multigroup cross sections for use in the Griffin reactor multiphysics code for two-step deterministic neutronics calculations. A three-dimensional fuel element typical of a high-temperature gas reactor as well as a two-dimensional sodium-cooled fast reactor lattice are modeled using the Serpent Monte Carlo code, and multigroup macroscopic cross sections are generated for various state parameters to produce a training data set and a separate validation data set. A fully connected, feedforward neural network is trained using the open-source PyTorch machine-learning framework, and its accuracy is compared against the standard piecewise linear interpolation model. Additionally, we provide in this work a generic technique for propagating the cross-section model errors up to the keff using sensitivity coefficients with the first-order uncertainty propagation rule. Quantifying the eigenvalue error due to the cross-section regression errors is especially practical for appropriately selecting the mathematical representation of the cross sections. We demonstrate that the artificial neural network model produces lower errors and therefore enables better accuracy relative to the piecewise linear model when the cross sections exhibit nonlinear dependencies; especially when a coarse grid is employed, where the errors can be halved by the artificial neural network. However, for linearly dependent multigroup cross sections as found for the sodium-cooled fast reactor case, a simpler linear regression outperforms deeper networks.

High Temperature Test Facility sensitivity and calibration studies to inform OECD-NEA benchmark calculations

The High Temperature Test Facility (HTTF) is a thermal hydraulics test facility for block-type high-temperature gas-cooled reactors (HTGRs) built to provide data to validate HTGR modeling tools. Several experiments were conducted at HTTF in the spring of 2019 that serve as the basis for an upcoming Organization for Economic Cooperation and Development – Nuclear Energy Agency HTGR thermal hydraulics benchmark. We present sensitivity studies of steady state and both the pressurized and depressurized conduction cooldown (PCC and DCC) transients in HTTF in this paper. These studies identified that uncertainties in the thermal conductivity of the ceramic blocks making up the HTTF core are significant drivers of peak temperatures in steady state, PCC, and DCC. The emissivity of the heater rods and core blocks also provide significant control over heater rod temperatures. We also performed calibration studies based on HTTF Experiment PG-26, which was intended to serve as the basis for the DCC benchmark problem. We found that no calibration was able to reproduce PG-26 peak block temperatures within an average of 30%, but by taking the best-estimate PG-26 calibration and applying it to a similar experiment—PG-29—we were able to reproduce the peak block temperatures within an average of 6.2% error. Based on this nearly fivefold improvement in peak block temperature predictions from PG-26 to PG-29 and confounding factors in PG-26 that contributed to previous difficulties modeling the experiment in RELAP5-3D and SAM we conclude that PG-26 data are insufficient for developing a DCC benchmark and recommend using PG-29 as the benchmark experiment instead.

High temperature corrosion of two superalloys in the impure helium environment

ABSTRACT The helium coolant of a very high temperature gas-cooled reactor contains traces of impurities, which have an adverse effect on the structural metallic materials at elevated temperatures. The chromium-containing superalloys Inconel 617 and Incoloy 800 H have been the representative materials with outstanding performance, and, in this study, their corrosion behaviors were investigated under the impure helium, which was the typical HTGR operating environment. A continuous chromia-based surface scale can be developed in the impure helium, but after being above a critical temperature this oxide layer may be destroyed. This destructive phenomenon may be general for the high-temperature alloys such as Inconel 617 and Haynes 230, however, Incoloy 800 H can spontaneously resist this damage by the silica layer. This study was focused on the critical temperature calculations and measurements for Alloy 617 and 800 H, and the different corrosion behavior has been discussed.

Horizontal Split-Table Conceptual Design to Support Advanced Reactor Validation Needs

To improve the nuclear data testing and validation of advanced reactors, such as pebble-bed high-temperature gas-cooled reactors, molten salt reactors, and heat pipe microreactors, a conceptual design for a novel critical assembly using a horizontal split table (HST) was developed jointly by Oak Ridge National Laboratory and Lawrence Livermore National Laboratory. The mechanical design is led by Lawrence Livermore National Laboratory, whereas the neutronics considerations are led by Oak Ridge National Laboratory. The characteristics of the designed HST and the benefits of performing such an experiment to the community are included. As a proof of concept, a proposed critical experiment using tristructural isotropic fuel particles and a graphite moderator/reflector is described, mimicking a pebble-bed-type advanced reactor based on the HTR-10. A critical configuration corresponding to a footprint of about 4.5 m2 was determined with SCALE/KENO-VI to fit the planned dimensions of the HST. The similarity of the pebble-bed design and the HTR-10 reactor application was assessed using SCALE/TSUNAMI, and a similarity coefficient, ck, of 0.9982 was obtained, proving that the concept will be useful for cold-critical validation and for nuclear data validation and assimilation of pebble-bed-type advanced reactors. In the proposed design, the materials with the highest keff sensitivity are graphite and uranium, which demonstrates that particular care must be given to carbon-related cross-section data. A cross-section library study was performed to test the influence of the different recent releases of the ENDF/B cross-section library on the concept’s keff. The effect of mechanical uncertainties between the fixed and moving tables was also assessed by calculating the reactivity change caused by vertical and horizontal gaps, as well as angular and torsion offsets between the two sides of the HST concept. As a last analysis step, the performed nuclear data assimilation of the hypothetical experiment showed that uncertainties can be reduced by several hundred pcm. The same analysis process is currently being used to create a molten salt advanced reactor–type HST concept based on the Molten Salt Reactor Experiment.

Influence of Quantum Oscillations in the Thermal Scattering Law of Zirconium Carbide on Neutron Thermalization and Criticality

Zirconium carbide (ZrC) is a candidate material for advanced high temperature reactors, including space nuclear thermal propulsion applications. Thermal scattering laws (TSLs) are generated in the incoherent approximation for carbon bound in ZrC [C(ZrC)] and zirconium bound in ZrC [Zr(ZrC)], using ab initio lattice dynamics methods. Disordered alloy theory is introduced to improve treatment of isotopic composition within the elastic scattering cross section. Localized higher-energy vibrational modes and the presence of a phonon band gap in C(ZrC) cause quantized oscillation in the TSL atypical of nonhydrogenous solids. These oscillations yield a significant likelihood of large energy downscattering and upscattering interactions such that the quanta of energy transfer affecting neutron thermalization is substantially greater than classically expected. MC21 critical mass calculations of ZrC mixtures with high-enriched uranium demonstrate an impact of TSLs when compared to a free-gas treatment for thermal neutron–driven 235U loadings. The critical mass of homogenous mixed moderator systems of ZrC and reactor-grade graphite are also sensitive to the ZrC TSL. Moreover, the effect of quantized energy exchange on the neutron spectra is found to influence the temperature feedback coefficient.

Developing PCC and DCC integral effects test experiments at the High Temperature Test Facility

Among the Next-Generation Nuclear Plant (NGNP) designs, the High-Temperature Gas-Cooled Reactors (HTGRs) are very attractive, due to their inherent safety features, high power conversion efficiency, and potential of providing high-temperature process heat. To perform a thorough safety study and to license these types of reactors, sufficient information needs to be provided about the phenomena that occur during accident scenarios. While several experimental research efforts have been dedicated in the past to investigate accident scenarios, knowledge gaps still exist in the phenomena characteristic of pressurized and depressurized conduction cooldown (PCC/DCC) transients as well as for normal operation scenarios. This paper summarizes the Oregon State University High Temperature Test Facility (HTTF) test matrix, experimental campaign, and selected tests results. High Temperature Test Facility is a scaled Integral Test Facility (IET) that is capable of mimicking scaled dimensions and operational conditions of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The goal of the High Temperature Test Facility is to provide experimental data on the DCC, PCC and normal operating scenarios of the reference Modular High-Temperature Gas Cooled Reactor design. The DCC, PCC, mixing, heat up and cooldown tests described in this paper were performed at prototypical Modular High-Temperature Gas Cooled Reactor temperatures, scaled initial pressure conditions (∼200 kPa), and thermal power input of less than 70 kW. Presented test data show temperature distributions in the High Temperature Test Facility core, upper plenum, cross duct, or lower plenum. Based on these temperature profiles attempts to investigate stratified flow, natural convection flow, heat up, cooldown and mixing phenomena are made. Furthermore, this paper evaluates the performed test campaign in the light of the Very High Temperature Gas-cooled Reactor Phenomena Identification and Ranking Table (PIRT) and proposes experiments to complement the existing PCC/DCC testing database for the validation of the thermal-hydraulic codes.