HTGR Research Articles

Effect of tip injection on performance of highly loaded helium compressor in high-temperature gas-cooled reactor

As one of the representatives of the fourth-generation advanced nuclear reactor, the continuous development and application of high-temperature gas-cooled reactor (HTGR) technology has promoted the technical progress of the whole nuclear energy field. Helium compressor is one of the core components of HTGR. The highly loaded helium compressor effectively solves the disadvantages of the low single-stage pressure ratio and numerous stages of the traditional helium compressor. But it also brings a more complex tip-leakage flow. Tip injection, which is an active control method, can effectively control the tip clearance leakage and inhibit the development of the leakage vortex. This paper analyses the effects of axial deflection angle and injection pitch angle on the rotor performance of a highly loaded helium compressor via numerical simulation and validates them via experiment. Results show that the leakage vortex can be blown to the pressure surface of the adjacent blade by proper axial deflection angle to reduce the vorticity of the leakage vortex. The injection pitch angle directly affects the intensity of the leakage vortex during its initiation and development. When the axial deflection angle is 60° and the injection pitch angle is 20°, the adiabatic compression efficiency and total pressure ratio increase by 0.554 % and 0.160 % respectively under the design condition, and by 0.822 % and 0.162 % respectively under near-stall conditions.

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

Matching Analysis of Technical Parameters and Safety Standards for Nuclear Replacement of Coal-Fired Units

In the context of the growing share of renewable energy and the impending decommission of a large number of coal-fired units, nuclear energy is the only green energy that can replace coal power for a stable, clean, efficient, and large-scale power supply. This article compares the differences between coal power and nuclear power in terms of thermal system, thermal cycle, turbine parameters, and safety. It discusses the possibility of replacing the boiler of a coal power plant with nuclear power, that is, replacing the boiler of a coal power plant with a nuclear reactor for generation/heating/cogeneration. For coal-fired units with similar capacity that do not use a reheat cycle (at or below high pressure) and nuclear power units (such as a high-temperature gas-cooled reactor), as well as coal-fired units with a reheat cycle (ultra-high pressure and above) and nuclear power units (such as pressurized water reactors), there are great differences in steam parameters. In terms of steam turbines, the size of nuclear power units is relatively larger, requiring additional dehumidification measures. In addition, the safety factors and management methods considered in the site selection, construction, and operation of nuclear power plants are more stringent and complex, and comprehensive analysis and evaluation are needed in aspects such as waste treatment and accident emergency response. Except for the relevant provisions of the American Society of Mechanical Engineers code for pressure vessels, nuclear power units are not compatible with coal-fired units in terms of safety standards. Therefore, considering comprehensively, it is believed that the scheme of nuclear power replacing coal-fired units for power generation/heating/cogeneration program is not feasible at present.

Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs

Abstract The purpose of the present work is to determine the neutronic characteristics of annular cylindrical prismatic Very High Temperature Gas-cooled Reactor (VHTR) with different advanced fuel designs. The study includes detailed neutronic investigations of the eight types of TRISO (containing SiC layer) and TRIZO (containing ZrC layer) fuel particles for annular VHTR core. All calculations were performed using continuous energy Monte Carlo code MVP2.0 and MVPBURN with JENDL4.0 nuclear data library. The obtained results indicated that SiC + PyC and PyC layers in TRISO fuel appeared to be not excellent choice for the annular VHTR core due to its shorter operation year and, lower fuel burnup. The annular cylindrical prismatic core fueled with TRIZO fuel particle shows good neutronic characteristics compared to those designed with other TRISO fuel particles.

Thermal Fluid Modeling Approaches in SAM for High Temperature Gas-Cooled Reactor Applications

This paper provides an overview of the unique modeling and simulation challenges and needs in the system and safety analysis of two common types of High Temperature Gas-cooled Reactor (HTGR) designs. The challenges are associated with the complex core geometry configurations and the change of the dominant heat removal mechanisms between normal operating conditions and decay heat removal transients. This requires that simulation tools utilize models with reasonable length scales and capture both the localized heat transfer and the core-scale heat transfer at the same time. To meet such analysis needs, different methodologies were developed. This includes adopting existing methods and further developing/improving them, as well as proposing innovative methods that leverage advanced computational frameworks such as the Multiphysics Object Oriented Simulation Environment (MOOSE).

An Innovative Fuel Design for HTGRs: Evaluating a 10-Hour High-Temperature Oxidation of the SiC Fuel Matrix During Air Ingress Accident Conditions

Preventing severe corrosion incidents caused by air ingress accidents in high-temperature gas-cooled reactors (HTGRs) while improving heat removal efficiency from the core is of paramount importance. To enhance both safety and efficiency, a sleeveless silicon carbide (SiC)-matrix fuel compact has been proposed. This study evaluates the 10-hour oxidation of reaction-sintered SiC (RS-SiC)-matrix fuel compact under the conditions of an air ingress accident within the temperature range of 1000 to 1400 °C. The oxidation tests were conducted in a stagnant air environment without flow. As a result, it is demonstrated that RS-SiC exhibits exceptional resistance to air oxidation up to 1400 °C, as shown by the thermogravimetric analysis (TGA), with minimal mass loss due to the oxidation of free carbon. Scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM–EDX) analysis reveals that the morphology and thickness of the SiO2 layer formed on the RS-SiC surface vary with temperature. At 1400 °C, uniform oxide layer thickness ranging from 1.59 to 4.10 μm and localized nodule-like oxide formations of approximately 10 μm are observed. In contrast, at 1000–1200 °C, thinner oxide layers are identified, indicating that oxide growth accelerates at higher temperatures. The oxidation rates measured provide insights into the mechanisms of oxide growth.

Neutronics Analysis on High-Temperature Gas-Cooled Pebble Bed Reactors by Coupling Monte Carlo Method and Discrete Element Method

The High-Temperature Gas-Cooled Pebble Bed Reactor (HTG-PBR) is notable in the advanced reactor realm for its online refueling capabilities and inherent safety features. However, the multiphysics coupling nature of HTG-PBR, involving neutronic analysis, pebble flow movement, and thermo-fluid dynamics, creates significant challenges for its development, optimization, and safety analysis. This study focuses on the high-fidelity neutronic modelling and analysis of HTG-PBR with an emphasis on achieving an equilibrium state of the reactor for long-term operations. Computational approaches are developed to perform high-fidelity neutronics analysis by coupling the superior modelling capacities of the Monte Carlo Method (MCM) and Discrete Element Method (DEM). The MCM-based code OpenMC and the DEM-based code LIGGGHTS are employed to simulate the neutron transport and pebble movement phenomena in the reactor, respectively. To improve the computational efficiency to expedite the equilibrium core search process, the reactor core is discretized by grouping pebbles in axial and radial directions with the incorporation of the pebble position information from DEM simulations. The OpenMC model is modified to integrate fuel circulation and fresh fuel loading. All of these measures ultimately contribute to a successful generation of an equilibrium core for HTG-PBR. For demonstration, X-energy’s Xe-100 reactor—a 165 MW thermal power HTG-PBR—is used as the model reactor in this study. Starting with a reactor core loaded with all fresh pebbles, the equilibrium core search process indicates the continuous loading of fresh fuel is required to sustain the reactor operation after 1000 days of fuel depletion with depleted fuel circulation. Additionally, the model predicts 213 fresh pebbles are needed to add to the top layer of the reactor to ensure the keff does not reduce below the assumed reactivity limit of 1.01.

Thermal hydraulic analysis of a high-temperature gas reactor steam generator after loss of feed water accidents

The helical coil steam generator is widely employed in the energy conversion process of high-temperature gas-cooled reactors for its high thermal efficiency in producing superheated steam within the helical tubes. These tubes necessitate materials specifically designed to endure high temperatures, prevent ruptures, and ensure durability under various operational conditions. This study aims to identify the maximum temperatures arising from potential transients in the secondary system and to propose immediate safety measures for minimizing associated risks. The RELAP5/MOD3.4 code was used to simulate the behavior of the next generation nuclear plant steam generator's conceptual design using steady-state conditions. Following the validation of the model, the established steady-state conditions were utilized to simulate accident scenarios, including the feed water line break and the loss of feed water flow to the steam generator. The transient scenarios assumptions include closed flap valves and retained hot helium in the shell side. The simulation tasks were conducted both with and without activating the depressurization system, in order to assess its effect on the temperature of tubes on the wall side. The analysis results indicate that tube wall temperatures can reach 720 °C at the steam generator outlet region and 620 °C at the weld region. The immediate activation of the depressurization system with a controlled relief valve can effectively reduce these temperatures by 239 °C in the steam generator outlet and by 133 °C in the weld region. This also lowers the steam generator's secondary side pressure. As a result, an upgrade and enhancement of operational procedures is straightforward.

Design of experimental facility for helium gas component testing

The current state-of-the-art helium gas test loops for component testing are constrained by the maximum operating temperatures, pressures, and flow rates of such facilities. Idaho National Laboratory is designing a new experimental facility known as HECTOR (HElium Component Testing Out-of-pile Research), which is a helium gas test loop for evaluating the performance of components to helium gas pressures, temperatures, and flow rates up to 8 MPa, 800 °C, and 0.15 kg/s, respectively. The construction of this facility will enable researchers to perform experiments across a wide range of Reynolds and Nusselt numbers, which could result in the maturation of technology for high-temperature gas-cooled reactor (HTGR) components.

Benchmark Analysis of a Prismatic Type-high Temperature Gas-cooled Reactor with the ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 Nuclear Data Libraries

In this work we performed a benchmark analysis of the High Temperature Engineering Test Reactor (HTTR) fully-loaded start-up critical core with a 30-bundle loading configuration using the Monte Carlo code Serpent 2 with the recently released nuclear data libraries ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0. The purpose of the work is to reveal the impacts of using different ENDF nuclear data libraries on neutronics calculations of a prismatic high temperature gas-cooled reactor (HTGR). The benchmark results obtained with Serpent 2 were compared against the available experimental values and those attained by different computer codes (MCNP5 and SCALE) using the nuclear data library ENDF/B-VII.0. The comparative results showed good agreement between Serpent 2, the experimental values, MCNP5 and SCALE. The results also exhibited a notable difference between using ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 in predicting the neutronics parameters of the HTTR that suggests further investigation in future work.

Thermo-economic and environmental analyses of supercritical carbon dioxide Brayton cycle for high temperature gas-cooled reactor

The supercritical carbon dioxide (sCO2) Brayton cycle demonstrates special advantages for the high temperature gas-cooled reactor (HTGR) delivered into commercial operation recently in China, with its high efficiency, compactness, flexibility, and safety compared to the conventional steam Rankine cycle. However, the large temperature rise of 500 °C for the HTGR brings new challenges for the design of sCO2 cycle. Here, we present the first study on the thermodynamic, economic, and environmental performance of the HTGR-sCO2 system using the energy, exergy, economic, and environmental (4E) evaluation method. The cascaded sCO2 cycle made up of two independent sCO2 cycles is proposed, which are arranged in series on the cold side of reactor heat exchanger. We show that the cascaded sCO2 cycle can utilize the heat absorption from HTGR effectively by optimizing the cycle configurations of top and bottom sub-cycles. The improved cascaded sCO2 cycle minimizes the exergy loss, and increases the thermal efficiency to 43.2% when compared to the steam Rankine cycle of HTGR demonstration power plant and the single recompression cycle. By balancing the fixed-capital investment cost with the net power, the levelized cost of electricity can be reduced to 0.0283$/kWh. The life-cycle GHG emission intensity of HTGR-sCO2 systems is about 6.5gCO2,eq/kWh, which is much smaller than that of coal-fired power plants, suggesting a great potential for decarbonization of the HTGR-sCO2 system. Our study may find implications for the advancement of the sCO2 Brayton cycle in next-generation nuclear power plant.

Finite element analysis of single steel-plate concrete composite containment of nuclear power plant under commercial aircraft impact

Single steel-plate concrete composite (SSC) structures, with good impact resistance and high construction efficiency, have emerged as new solutions for nuclear power plant (NPP) containment. This paper presents a commercial aircraft impact analysis of a new SSC containment structure. Before obtaining more reasonable results from large-scale numerical analysis, small-scale models of the SSC local structure, SSC panel elements, and commercial aircraft were established and experimentally or theoretically verified. Based on these small-scale models, a complete model for SSC containment against aircraft impact was established. Parametric analysis was conducted to investigate the influences of various factors on the dynamic characteristics of the containment. The results indicated that while the influence of soil–structure interaction (SSI) effects is relatively minimal, lateral constraints from external plants should be considered. Considering the balance among steel consumption, economic benefits, construction quality, and efficiency, SSC containment is recommended for high-temperature gas-cooled reactors.

A Demonstration of the Risk-Informed NEI 18-04 Design Evaluation Model for the Modular High-Temperature Gas Reactor

Several advanced reactor designs are now under active development in the United States and elsewhere, promising sustainable solutions to the growing world energy needs. The designs currently being considered are quite diverse and different from the more established light water reactor technology that has dominated the operating commercial nuclear landscape. While advanced reactor concepts were first explored in the dawn of the nuclear age, they are now being reconsidered under the light of modern needs, and specifically, for their flexible operating conditions and inherent safety characteristics. In response, the U.S. Nuclear Regulatory Commission staff is moving forward with development of 10 CFR Part 53 rulemaking, which is a more risk-informed, technology-agnostic framework for licensing and regulating such new designs. The nuclear industry response to this regulatory initiative resulted in the technical report by the Nuclear Energy Institute, NEI 18-04 Revision 1, which provides an implementation roadmap of the risk-informed approach when defining the safety case for a new plant design. The implementation of this safety case may be a nontrivial exercise for an actual reactor design. This paper provides a demonstration of performing such an analysis for a representative advanced reactor. Public information from the General Atomics high-temperature gas reactor design was considered in this demonstration. The analysis workflow was facilitated with the FPoliSolutions’ proprietary Risk-Informed System Engineering (RISE) digital platform, a product that was presented in previous publications. RISE is one application of FPoli’s enterprise digital platform, which was created to facilitate orchestration of complex workflows leveraging recent technologies developed at national laboratories, such as Idaho National Laboratory’s RAVEN and EMRALD frameworks. The analysis described in the paper includes the selection and classifications of events, the integration of probabilistic risk analysis artifacts, and event modeling simulations for consequence evaluations. The results are then used for system, structures, and components safety classification and a synthesis of the safety case for the design in line with the frequency-consequence targets presented in NEI 18-04. The purpose of the analysis, as framed in RISE, is to readily produce outputs and views that can aid users and regulators in making risk-informed decisions to demonstrate their plant safety case.

Comparative analysis of irradiation-stimulated hardening in the austenite and ferrite phases of F321 stainless steel

F321 stainless steel is renowned for its outstanding mechanical properties and corrosion resistance, making it a crucial structural material in light water reactors and high-temperature gas reactors. With irradiation (IR), F321 stainless steel undergoes microstructural alterations, such as the formation of IR-stimulated defects (e.g., stacking fault tetrahedrons (SFTs), dislocation loops (DLs), and clusters), resulting in IR hardening. This study compares the IR-stimulated hardening in the austenitic and ferritic phases of F321 stainless steel through both experimental methods and numerical simulations. In the austenitic phase, the size and number density of IR defects (DLs and SFTs) were determined, which were proportional to the IR dose. In the ferritic phase, IR-induced DL parameters were quantified, and the effects of IR on spinodal decomposition were analyzed. The interaction processes and mechanisms between dislocations and IR defects in the austenitic phase were evaluated using molecular dynamics (MD) simulations, leading to the development of a model for IR defect-induced hardening strength. Additionally, an MD data-driven crystal-plasticity-finite-element-model (CPFEM) was established. The nano-indentation hardness of the austenitic and ferritic phases was measured and calculated through both experimental methods and CPFEM simulations. The results show that the nano-indentation hardness of the ferritic phase is reduced compared to the austenitic phase, due to a lower density of IR defects. Understanding these effects provides deeper insights into how IR influences the mechanical and microstructural properties of F321 steel, which is essential for predicting the material's long-term performance and reliability.

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Tristructural isotropic (TRISO)-coated fuel particles are designed for use in high-temperature gas-cooled nuclear reactors, featuring a structural SiC layer that may be exposed to oxygen-rich environments over 1000 °C. Surrogate TRISO particles were tested in 0.2–20 kPa O2 atmospheres to observe the differences in oxidation behavior. Oxide growth mechanisms remained consistent from 1200–1600 °C for each PO2, with activation energies of 228 ± 7 kJ/mol for 20 kPa O2 and 188 ± 8 kJ/mol for 0.2 kPa O2. At 1600 °C, kinetic analysis revealed a change in oxide growth mechanisms between 0.2 and 6 kPa O2. In 0.2 kPa O2, oxidation produced raised oxide nodules on pockets with nanocrystalline SiC. Oxidation mechanisms were determined using Atom probe tomography. Active SiC oxidation occurred in C-rich grain boundaries with low PO2, leading to SiO2 buildup in porous nodules. This phenomenon was not observed at any temperature in 20 kPa O2 environments.

Technology for reprocessing mother liquor and washing solution from crystalization purification of HTGR SNF

The extractants diphenyl-N,N-dioctylcarbamoylmethylphosphine oxide and diphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide for the processing of spent fuel were tested. The conditions for extraction and separation of uranium and TPE–REE were determined. A technological scheme was proposed for processing spent nuclear fuel from high-temperature gas-cooled reactors. During dynamic testing, at least 99.9% of uranium and plutonium and at least 99.5% of americium and rare earth elements were extracted. The TPE + REE and U+Pu fractions were isolated. The U+Pu fraction contained approximately 5% Am, while the TPE + REE fraction had less than 0.1% U and Pu.

Development and assessment of an integrated multigenerational energy system with cobalt-chlorine hydrogen generation cycle

This study aims to develop and assess a new multigeneration system where nuclear heat is utilized as the energy source. The multigeneration system is further designed to generate power, cooling, freshwater and hydrogen. In the present multigeneration system, five main subsystems, including high-temperature gas-cooled pebble bed nuclear reactors, a Rankine cycle, a cobalt-chlorine thermochemical cycle, a multi-effect desalination system and an ammonia-water absorption refrigeration system are integrated for synchronized operation. The analyses of the present system are carried out with the approaches of energy and exergy. This integrated system uses a total of 1000 MW thermal energy that is obtained from four units of high-temperature gas-cooled pebble bed nuclear reactor. Using all the thermal energy that comes from the nuclear reactors, a total of 346.99 MW of electricity, a total of 1.59 MW of cooling, a total of 384.67 kg/s of freshwater, and a total of 0.25 kg/s of hydrogen are produced. The energetic and exergetic performance coefficients of the ammonia-water absorption refrigeration system are 0.74 and 0.83, respectively. While the energy efficiency for the overall system is calculated as 37.83%, the exergy efficiency is found to be higher as 46.32% with the multiple useful outputs which help exergetically improve the overall system.

Biomimetic structure-driven high strength and toughness in continuous SiC skeleton-reinforced graphite composites

Graphite-matrix structural composites with excellent mechanical properties and long-term reliability are ungently required in aerospace and nuclear industries. However, the enhancement of graphite-matrix composites by mainstream surface coating techniques or second-phase particle reinforcement is rather limited. Inspired by the structural-functional feature of plant cells, silicon carbide (SiC, cytoderm) coated mesocarbon microbeads (MCMB, cytoplasm) powders were prepared by combustion synthesis (CS), which were then densified via spark plasma sintering (SPS) to obtain the biomimetic cellular-structured SiC-reinforced MCMB (M@SiC) composites. The in-situ formed SiC ‘cytoderm’ ensures good interfacial bonding and contributed to the densification of the composites. Additionally, the continuous SiC skeleton and cellular-structured configuration improved the mechanical properties of M@SiC composites. Owing to the complex crack deflection, branching and bridging effects at the MCMB/SiC interfaces and graphite nanoflakes inside MCMB, the biomimetic M@SiC composite with SiC content of 47 vol% exhibited strengthening and toughening, with flexural strength and fracture toughness of 245 MPa and 3.82 MPa m1/2, respectively. The developed biomimetic graphite-matrix composites with excellent mechanical properties are expected to be applied in reusable spacecrafts and high-temperature gas-cooled reactors.

Transcritical CO2 mixture power for nuclear plant application: Concept and thermodynamic optimization

The conventional steam power cycle suffers from lower thermal efficiency, larger size, and inadequate cold source matching, failing to meet the demands for high efficiency, compactness, and adaptability required by new-generation underwater power plants. Consequently, the T-CO2 mixture power cycle has been proposed for technological innovation in underwater nuclear plants. Thermodynamic model of the recuperated, precompression, recompression and partial cooling cycles is developed based on the first law of thermodynamics. And the thermodynamic performance of five CO2 mixtures, including CO2/SO2, CO2/H2S, CO2/butane, CO2/propane, and CO2/cyclohexane in the four cycles coupled with a pressurized water reactor (PWR) and an advanced high-temperature reactor (AHTR) is investigated. The findings reveal that when optimized for thermal efficiency, the CO2/cyclohexane is better suited for recuperated and precompression cycles, while CO2/H2S and CO2/propane are more suitable for recompression and partial cooling cycle. Furthermore, regarding the T-CO2 mixture power cycle incorporating PWR and AHTR, the recompression cycle of CO2/propane and CO2/H2S should be opted to achieve maximum thermal efficiencies of 33.07 % and 46.07 %, respectively. Conversely, selecting the precompression cycle of CO2/H2S can yield maximum specific work of 98.61 kJ/kg and 161.62 kJ/kg, respectively. Therefore, investing in reforming nuclear power with new technologies is a feasible policy choice. Additionally, policymakers need to closely monitor the reform requirements of nuclear power plants.

Understanding Ag liquid migration in SiC through ex-situ and in-situ Ag-Pd/SiC interaction studies

The effective containment of fission products (FPs) within tri-structural isotropic (TRISO) fuel particles is crucial for the safety and efficiency of High Temperature Gas-cooled Reactors (HTGRs). Combining multi-scale (µm to nm) and multi-dimensional (2D and 3D) analysis, this study focuses on the migration mechanisms of silver (Ag), for which the Ag-110 m FP isotope is radiotoxic, facilitated by reactions with palladium (Pd) alloys and the silicon carbide (SiC) layer at elevated temperatures (1300–1500 °C). Three primary pathways for Ag migration are identified: (1) diffusion in solid palladium silicide to form (Pd,Ag)2Si, (2) infiltration through cracks and pores in the carbon phase, and (3) liquid phase migration along SiC grain boundaries. Especially at temperatures above the melting point of Pd2Si (1404 °C), a ‘dissolution-recrystallisation’ mechanism is proposed of the Ag-Pd-Si liquid phase migrating along SiC grain boundaries. The study employs state-of-the-art in-situ heating transmission electron microscopy (TEM) techniques to directly observe the dynamic migration processes of Pd silicide within SiC, providing unprecedented insights into the liquid behaviour in TRISO fuel. These findings highlight the significant role of liquid phases in determining the transport of FPs through the TRISO-SiC which is vital for developing strategies that enhance the safety and efficacy of HTGRs.