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High-temperature gas-cooled reactor
High-temperature gas-cooled reactor
High-temperature gas-cooled reactor
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Refueling floor at Fort Saint Vrain HTGR, 1972

A high-temperature gas-cooled reactor (HTGR) is a type of gas-cooled nuclear reactor which uses uranium fuel and graphite moderation to produce very high reactor core output temperatures.[1] All existing HTGR reactors use helium coolant. The reactor core can be either a "prismatic block" (reminiscent of a conventional reactor core) or a "pebble-bed" core. China Huaneng Group currently operates HTR-PM, a 250 MW HTGR power plant with two pebble-bed HTGRs, in Shandong province, China.

The high operating temperatures of HTGR reactors potentially enable applications such as process heat or hydrogen production via the thermochemical sulfur–iodine cycle. A proposed development of the HTGR is the Generation IV very-high-temperature reactor (VHTR) which would initially work with temperatures of 750 to 950 °C.

History

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The use of a high-temperature, gas-cooled reactor for power production was proposed by in 1944 by Farrington Daniels, then associate director of the chemistry division at the University of Chicago's Metallurgical Laboratory. Initially, Daniels envisaged a reactor using beryllium moderator. Development of this high temperature design proposal continued at the Power Pile Division of the Clinton Laboratories (known now as Oak Ridge National Laboratory) until 1947.[2] Professor Rudolf Schulten in Germany also played a role in development during the 1950s. Peter Fortescue, whilst at General Atomics, was leader of the team responsible for the initial development of the High temperature gas-cooled reactor (HTGR), as well as the Gas-cooled fast reactor (GCFR) system.[3]

The Peach Bottom unit 1 reactor in the United States was the first HTGR to produce electricity, and did so very successfully, with operation from 1966 through 1974 as a technology demonstrator. Fort St. Vrain Generating Station was one example of this design that operated as an HTGR from 1979 to 1989. Though the reactor was beset by some problems which led to its decommissioning due to economic factors, it served as proof of the HTGR concept in the United States (though no new commercial HTGRs have been developed there since).[4][failed verification]

Experimental HTGRs have also existed in the United Kingdom (the Dragon reactor) and Germany (AVR reactor and THTR-300), and currently exist in Japan (the High-temperature engineering test reactor using prismatic fuel with 30 MWth of capacity) and China (the HTR-10, a pebble-bed design with 10 MWe of generation). Two full-scale pebble-bed HTGRs, the HTR-PM reactors, each with 100 MW of electrical production capacity, have gone operational in China as of 2021.[5]

Reactor design

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A simplified flow diagram of a 1,100 MWe HTGR.

Neutron moderator

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The neutron moderator is graphite, although whether the reactor core is configured in graphite prismatic blocks or in graphite pebbles depends on the HTGR design.

Nuclear fuel

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The fuel used in HTGRs is coated fuel particles, such as TRISO[6] fuel particles. Coated fuel particles have fuel kernels, usually made of uranium dioxide, however, uranium carbide or uranium oxycarbide are also possibilities. Uranium oxycarbide combines uranium carbide with the uranium dioxide to reduce the oxygen stoichiometry. Less oxygen may lower the internal pressure in the TRISO particles caused by the formation of carbon monoxide, due to the oxidization of the porous carbon layer in the particle.[7] The TRISO particles are either dispersed in a pebble for the pebble bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks. The QUADRISO fuel[8] concept conceived at Argonne National Laboratory has been used to better manage the excess of reactivity.

Coolant

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Helium has been the coolant used in all HTGRs to date. Helium is an inert gas, so it will generally not chemically react with any material.[9] Additionally, exposing helium to neutron radiation does not make it radioactive,[10] unlike most other possible coolants.

Control

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In the prismatic designs, control rods are inserted in holes cut in the graphite blocks that make up the core. The VHTR will be controlled like current PBMR designs if it utilizes a pebble bed core, the control rods will be inserted in the surrounding graphite reflector. Control can also be attained by adding pebbles containing neutron absorbers.

Safety features and other benefits

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The design takes advantage of the inherent safety characteristics of a helium-cooled, graphite-moderated core with specific design optimizations. The graphite has large thermal inertia and the helium coolant is single phase, inert, and has no reactivity effects. The core is composed of graphite, has a high heat capacity and structural stability even at high temperatures. The fuel is coated uranium-oxycarbide which permits high burn-up (approaching 200 GWd/t) and retains fission products. The high average core-exit temperature of the VHTR (1,000 °C) permits emissions-free production of high grade process heat. Reactors are designed for 60 years of service.[11]

List of HTGR reactors

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Constructed reactors

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As of 2011, a total of seven HTGR reactors have been constructed and operated.[12] A further two HTGR reactors were brought on-line at China's HTR-PM site, in 2021/22.

Facility
name 		Country Commissioned Shutdown No. of
reactors 		Fuel type Outlet
temperature (°C) 		Thermal
power (MW)
Dragon reactor[12] United Kingdom 1965 1976 1 Prismatic 750 21.5
Peach Bottom[12] United States 1967 1974 1 Prismatic 700–726 115
AVR[12] Germany 1967 1988 1 Pebble bed 950 46
Fort Saint Vrain[12] United States 1979 1989 1 Prismatic 777 842
THTR-300[12] Germany 1985 1988 1 Pebble bed 750 750
HTTR[12] Japan 1999 Operational 1 Prismatic 850–950 30
HTR-10[12] China 2000 Operational 1 Pebble bed 700 10
HTR-PM[13] China 2021 Operational 2 Pebble bed 750 250

Additionally, from 1969 to 1971, the 3 MW Ultra-High Temperature Reactor Experiment (UHTREX) was operated by Los Alamos National Laboratory to develop the technology of high-temperature gas-cooled reactors.[14] In UHTREX, unlike HTGR reactors, helium coolant contacted nuclear fuel directly, reaching temperatures in excess of 1300 °C.

Proposed designs

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References

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Revisions and contributorsEdit on WikipediaRead on Wikipedia

High-temperature gas-cooled reactor

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from Grokipedia
A high-temperature is a design that employs gas as a and as a moderator, utilizing TRISO-coated particles to achieve core outlet temperatures of 700–950°C while ensuring through passive heat removal mechanisms. HTGRs feature modular construction in either prismatic block or pebble-bed core configurations, with power outputs ranging from small-scale reactors (e.g., 30 MWth) to larger demonstration units (up to 600 MWth per module), enabling high exceeding 45% for and of process heat. The TRISO fuel, consisting of or carbide kernels encased in multiple layers, confines fission products even at temperatures up to 1600–1800°C, preventing release during normal operation or severe accidents. This design's low and graphite's high allow for natural convection cooling, eliminating the risk of meltdown and providing extended grace periods—often days—without active intervention. Development of HTGRs dates back to the , with early experimental plants like the U.S.'s Peach Bottom (115 MWth, 40 MWe, operational 1967–1974) and Fort St. Vrain (330 MWe, 1976–1989) demonstrating the technology's feasibility, followed by international efforts in , , and . As a Generation IV system, HTGRs support decarbonization by providing high-temperature steam or gas for industrial applications, including via thermochemical processes, , and manufacturing, with economic advantages such as lower power generation costs compared to light-water reactors. As of November 2025, operational HTGRs include Japan's High-Temperature Test Reactor (HTTR, 30 MWth, critical since 1998 and restarted in 2021 for ongoing research) and China's demonstration plant (210 MWe, which entered commercial operation in December 2023), marking the first grid-connected pebble-bed HTGR. Ongoing projects, such as the U.S.-based Xe-100 (targeting deployment by 2027) and international collaborations under the Generation IV International Forum, focus on licensing, fuel supply chains for high-assay low-enriched , and integration with to enhance global .

Fundamentals

Core Principles

A high-temperature gas-cooled reactor (HTGR) is a type of nuclear reactor that employs graphite as a moderator and helium as a coolant to achieve core outlet temperatures typically ranging from 750°C to 950°C, facilitating high-efficiency electricity generation and the provision of process heat for industrial applications. This design leverages the inherent properties of its materials to operate at elevated temperatures without the need for active cooling systems under normal conditions, distinguishing it from light-water reactors. Within international frameworks, HTGRs are classified as a variant of the Very High Temperature Reactor (VHTR) under the Generation IV International Forum (GIF), an initiative coordinated by multiple nations to advance sustainable nuclear technologies, and are recognized by the International Atomic Energy Agency (IAEA) as an advanced gas-cooled reactor system emphasizing fuel efficiency and high-temperature performance. HTGRs maintain a predominantly thermal neutron spectrum, where fast neutrons from fission are slowed down primarily through elastic scattering interactions with carbon atoms in the graphite moderator, enabling efficient fission in low-enriched uranium fuel. This spectrum includes contributions from epithermal neutrons, particularly in the resonance energy range, which influence self-shielding effects in fuel particles and are accounted for in core calculations using multi-group neutron transport methods. The core configuration can be either pebble-bed, featuring spherical fuel elements continuously recirculated for online refueling, or prismatic, using stacked hexagonal graphite blocks with embedded fuel compacts for batch refueling, both of which support the thermal spectrum while optimizing neutron leakage and power distribution. The economy in HTGRs benefits from 's low absorption cross-section, approximately 100 times lower than that of , which minimizes parasitic losses and allows for higher fuel utilization compared to water-moderated reactors. By using to moderate without introducing , HTGRs eliminate risks associated with aqueous coolants and structural materials, enhancing long-term core integrity and operational reliability. This solid moderator also provides structural support and heat conduction paths, contributing to a favorable balance of reactivity control and potential exceeding 100 GWd/t in thermal spectra. The thermal power generated in an HTGR core, denoted as PP, arises from the energy released by nuclear fissions and can be derived from fundamental reactor physics principles tailored to graphite-moderated systems. The rate of fissions per unit volume is given by the product of the average neutron flux ϕ\phi (in neutrons/cm²·s) and the macroscopic fission cross-section Σf\Sigma_f (in cm⁻¹), which for thermal neutrons in HTGRs is dominated by the 235^{235}U fission cross-section of approximately 580 barns at 0.025 eV, adjusted for the graphite's moderating ratio of about 200 (mean logarithmic energy loss per collision). Integrating over the core volume VV yields the total fission rate Nf=ϕΣfVN_f = \phi \Sigma_f V fissions per second. Each fission releases an average recoverable energy EfE_f of about 200 MeV (or 3.2×10−113.2 \times 10^{-11} J), primarily as kinetic energy of fission products and prompt neutrons, with approximately 7% additional contribution from delayed beta and gamma decays of fission products. Thus, the total power is P=NfEf=ϕΣfEfVP = N_f E_f = \phi \Sigma_f E_f V. In HTGR-specific analyses, ϕ\phi is typically on the order of 101310^{13} to 101410^{14} n/cm²·s in the core center due to the low power density (around 5-10 MW/m³) enabled by graphite's thermal properties, ensuring the equation aligns with measured outputs in prototypes like the HTR-10 (10 MWth). P=ϕΣfEfVP = \phi \Sigma_f E_f V

Thermodynamic Operation

In high-temperature gas-cooled reactors (HTGRs), heat generated by in the TRISO-coated fuel particles is transferred to the primarily through conduction within the matrix of the fuel elements and via the forced flow of through the core channels or pebble bed. Fission energy deposits directly in the fuel kernels, conducting outward through the pyrocarbon and layers into the surrounding , which acts as both moderator and structural material, before being convected to the stream. During normal operation, enters the core at 250–350°C and exits at 750–900°C, with forced dominating the , enhanced by 's high thermal conductivity of approximately 0.3 W/m·K at operating temperatures. Radiation contributes in pebble bed configurations, particularly across voids and gaps, but remains secondary to under steady-state conditions. HTGRs employ a closed for efficient , where the hot coolant from the core outlet directly or indirectly drives a gas turbine, bypassing traditional steam cycles to achieve compact power conversion. The cycle typically includes a , recuperator, precooler, , and multi-stage , with recirculating at pressures of 6–7 MPa. Turbine inlet temperatures reach up to 850°C in designs like the GT-MHR and PBMR, limited by metallic component materials, enabling direct cycle operation that simplifies the plant layout and reduces capital costs compared to indirect cycles. This integration leverages 's chemical inertness and low neutron absorption, allowing high-temperature operation without issues common in water-cooled systems. The thermodynamic efficiency of the Brayton cycle in HTGRs benefits from elevated temperatures and helium's favorable properties, outperforming light-water reactors (LWRs). The ideal Carnot efficiency provides a theoretical bound, given by η=1−TcoldThot\eta = 1 - \frac{T_\text{cold}}{T_\text{hot}} where temperatures are in Kelvin; for typical HTGR conditions with Thot≈1123T_\text{hot} \approx 1123 K (850°C) and Tcold≈400T_\text{cold} \approx 400 K (compressor inlet after intercooling), η≈64%\eta \approx 64\%. Actual Brayton efficiencies account for irreversibilities and reach 45–48% in HTGRs, compared to 33% in LWRs, due to helium's low molecular weight (4 g/mol), which minimizes compressor work, and its high thermal conductivity, which improves recuperator effectiveness up to 95%. This efficiency advantage enhances overall plant economics and fuel utilization. In pebble bed HTGRs, the across the core, critical for determining power requirements, is governed by the , derived for flow through porous packed beds and adapted for multiphase helium-graphite interactions. The equation combines viscous (laminar) and inertial (turbulent) contributions: ΔPL=150(1−ϵ)2ϵ3μvdp2+1.75(1−ϵ)ϵ3ρv2dp\frac{\Delta P}{L} = 150 \frac{(1-\epsilon)^2}{\epsilon^3} \frac{\mu v}{d_p^2} + 1.75 \frac{(1-\epsilon)}{\epsilon^3} \frac{\rho v^2}{d_p} The viscous term originates from the Blake-Kozeny model, treating the bed as a network of capillaries where loss scales with μ\mu, superficial velocity vv, bed height LL, ϵ\epsilon (typically 0.39–0.42 in HTGR pebbles), and particle diameter dpd_p (60 mm for standard fuel pebbles); it dominates at low Reynolds numbers (Re<1000Re < 1000) common in HTGR flows. The inertial term, from the Burke-Plummer extension, captures form drag in turbulent regimes (Re>1000Re > 1000), proportional to ρ\rho and v2v^2. For HTGR pebble flow, with v≈5–20v \approx 5–20 m/min and μ≈4×10−5\mu \approx 4 \times 10^{-5} Pa·s, total ΔP\Delta P is 0.2–0.5 bar across a 10 m core, ensuring efficient circulation; the KTA correlation refines this for anisotropic beds by incorporating experimental pebble bed data. Core outlet temperature is regulated in HTGRs by modulating helium mass flow via variable-speed circulators and adjusting reactor power through control rods, maintaining exit temperatures at 700–900°C to optimize cycle performance while avoiding fuel limits. In the HTR-10 test reactor, for instance, flow rates of 4.3 kg/s achieve 900°C outlet during high-temperature phases, with bypass valves fine-tuning distribution to prevent hot spots. Helium purity is strictly controlled at 99.99% (total impurities <100 ppm, including H2_2O <0.1 ppm and CO/CO2_2 <10 ppm) using purification loops with getters and catalysts to remove oxidizing species, thereby preventing chronic graphite corrosion that could compromise core integrity over decades of operation.

Design Components

Moderator and Reflector

High-purity graphite serves as the primary moderator in high-temperature gas-cooled reactors (HTGRs) due to its low neutron absorption cross-section, typically Σa<0.0035 cm−1\Sigma_a < 0.0035 \, \text{cm}^{-1}, which minimizes parasitic neutron capture while efficiently slowing fast neutrons to thermal energies through elastic scattering. This material also exhibits exceptional thermal stability, maintaining structural integrity up to 1600°C, well beyond typical HTGR core outlet temperatures of 750–950°C, owing to its high sublimation point and resistance to oxidation in inert helium environments. The reflector in HTGR designs consists of radial and axial layers, often constructed from graphite or beryllium, surrounding the core to minimize neutron leakage and enhance criticality by reflecting escaping neutrons back into the fissile region. Beryllium offers superior reflection efficiency due to its low atomic mass and high scattering cross-section, while graphite provides cost-effective, compatible structural support; these configurations typically achieve an effective multiplication factor keff>1.05k_\text{eff} > 1.05, ensuring sufficient excess reactivity for operational margins. Graphite's performance under operational conditions is influenced by and irradiation-induced dimensional changes, which affect core geometry and neutronics over the reactor's lifetime. The coefficient of thermal expansion (CTE) for nuclear-grade is approximately 4×10−6 K−14 \times 10^{-6} \, \text{K}^{-1} between 20–120°C, leading to manageable expansion at elevated temperatures, but fast neutron irradiation initially causes densification with a change Δρ/ρ=−0.1%\Delta \rho / \rho = -0.1\% per 1021 n/cm210^{21} \, \text{n/cm}^2 (E > 0.1 MeV), followed by swelling at higher doses due to crystal lattice disruption and void formation. These effects are mitigated through allowances for anisotropic growth and periodic monitoring to prevent excessive or cracking in moderator blocks. Manufacturing standards for HTGR graphite emphasize isostatic pressing to achieve isotropic properties and purity levels below 5 ppm for impurities like to preserve low absorption. A representative grade, IG-110, developed for Japanese HTGR applications, features a of 1.78 g/cm³ and in-plane thermal conductivity of 116 W/m·K at (decreasing to approximately 70 W/m·K at 600°C), enabling efficient heat dissipation from the core while supporting structural loads under . In prismatic HTGR cores, the moderator consists of dense hexagonal graphite blocks with integrated fuel compacts and coolant channels, achieving near-solid packing with low void fractions for optimal neutron economy. In contrast, pebble-bed designs employ graphite-moderated fuel spheres in a loosely , resulting in moderator packing densities with 60–70% void fraction to facilitate helium flow and online refueling, though this increases neutron leakage compared to prismatic arrangements.

Fuel and Cladding

The fuel in high-temperature gas-cooled reactors (HTGRs) primarily consists of tri-structural isotropic (TRISO) coated particles, which serve as micro-encapsulated elements designed for exceptional durability under high temperatures and irradiation. Each TRISO particle features a central kernel of (UO₂) or uranium oxycarbide (UCO), typically 350–600 μm in , coated with multiple layers: a porous buffer layer of (PyC) approximately 95 μm thick to accommodate fission gas swelling and ; an inner dense isotropic PyC (IPyC) layer about 40 μm thick for structural support; a (SiC) layer around 35 μm thick acting as the primary fission product barrier; and an outer dense isotropic PyC (OPyC) layer of similar thickness to protect the SiC and facilitate particle handling. The SiC layer, with its high and , withstands temperatures up to 1800°C while maintaining integrity, enabling the fuel to operate in HTGR environments exceeding 1000°C without significant degradation. Enrichment levels for HTGR TRISO fuel are tailored to achieve efficient economy in graphite-moderated cores, typically ranging from 8–20% ²³⁵U for low-enriched uranium (LEU) kernels to support standard operations, though mixed oxide (MOX) variants incorporate up to 19.75% ²³⁹Pu blended with for advanced cycles like plutonium disposition or utilization. These enrichment choices balance criticality, potential, and proliferation resistance, with LEU dominating modern designs to align with non-proliferation goals. The protective multilayer cladding of TRISO particles eliminates the need for traditional metallic sheaths used in light-water reactors, as the coatings provide inherent containment against corrosion and mechanical stress in the helium environment. HTGR fuels achieve high burnup rates, up to 120 GWd/t, owing to their deep-burn capability, which maximizes fuel utilization through extended without compromising particle integrity. (BU) is the total thermal energy extracted per metric ton of initial heavy metal, quantified in GWd/t. This performance stems from the TRISO design's resistance to kernel migration and coating failure, allowing sustained operation at high fluxes. Deep-burn variants can exceed 150 GWd/t in optimized prismatic or pebble-bed configurations. HTGRs employ two principal fuel assembly types: pebble-bed and prismatic. In pebble-bed designs, fuel is embedded in graphite spheres (pebbles) of 60 mm diameter, each containing approximately 15,000 TRISO particles distributed within a 50 mm fueled zone, enabling online refueling and continuous core circulation for uniform burnup. Prismatic assemblies, by contrast, use hexagonal graphite blocks housing cylindrical compacts with around 10,000 TRISO particles per compact, offering higher power density but requiring batch refueling. The pebble-bed approach enhances thermal-hydraulic stability, while prismatic designs suit modular reactors with fixed geometries. Fission product retention in TRISO fuel relies on diffusion-limited transport models, such as Fick's laws and the Booth model, which predict negligible release through intact coatings at operational temperatures. Below 1600°C, these models indicate a release fraction less than 10⁻⁶ for key products like cesium (Cs) and strontium (Sr), with noble gases like krypton (Kr) showing fractions around 10⁻⁹ due to the low diffusivity in PyC and SiC layers (e.g., Cs diffusion coefficient in UO₂: D=0.90×10−18exp⁡(−209 kJ/mol/RT) m2/sD = 0.90 \times 10^{-18} \exp(-209 \, \text{kJ/mol}/RT) \, \text{m}^2/\text{s}). This containment ensures radiological safety even during transients, as metallic fission products remain trapped unless coatings fail, which is rare below design limits.

Coolant and Circulation

Helium serves as the primary coolant in high-temperature gas-cooled reactors (HTGRs) due to its chemical inertness, which prevents reactions with core materials such as and fuel cladding, ensuring long-term structural integrity. Its low thermal neutron absorption cross-section of approximately 0.0005 minimizes parasitic losses, supporting efficient fission chain reactions without significant interference. Additionally, exhibits a high of 5.19 J/g·K at 900°C, facilitating effective removal from the core while maintaining single-phase flow across operational temperatures up to 950°C. The primary coolant loop in HTGRs employs recirculating blowers to maintain circulation, typically operating at pressures of 5–7 MPa to balance efficiency and system compactness. For a representative 350 MWth core, mass flow rates range from 100–300 kg/s, directed through core channels to achieve temperatures around 300–400°C and outlet temperatures of 700–900°C, depending on design. This closed-loop configuration isolates the high-temperature primary from secondary systems, with blowers providing the necessary head to overcome pressure drops in the core, , and . Heat transfer from the primary occurs via intermediate heat exchangers (IHXs), which employ an indirect loop to isolate the reactor from power conversion or process fluids, enhancing by preventing cross-contamination. In some designs, such as those coupled to advanced processes, the intermediate loop uses sodium as a secondary for its high thermal conductivity, while others incorporate to avoid reactivity issues in high-temperature applications. These IHXs, often helical-coil or plate-fin types, operate at helium-side pressures matching the primary loop, transferring heat at efficiencies exceeding 90% while maintaining differential pressures to ensure . Impurity management is critical in HTGRs to prevent corrosion of graphite components, with hydrogen and methane levels controlled using getters such as titanium sponges or palladium-based absorbers integrated into the purification system. These impurities, arising from minor leaks or , can otherwise promote graphite oxidation; however, with effective control maintaining concentrations below 1–10 ppm, corrosion rates are limited to less than 0.1 mm/year at 900°C, preserving core integrity over decades of operation. The helium purification system operates on a continuous bypass cycle, diverting 1–15% of the primary flow through a series of components to remove impurities and restore coolant purity. Incoming helium passes first through cartridge filters to capture particulates, followed by copper oxide beds at 250–400°C to oxidize hydrogen to water and carbon monoxide to dioxide. Subsequent molecular sieve traps adsorb water vapor and carbon dioxide, while low-temperature charcoal beds (around -196°C using liquid nitrogen) capture nitrogen, methane, and residual gases. The purified helium is then recombined with the main flow, with system sizing ensuring a purification constant of at least 2.9 × 10⁻⁵ s⁻¹ to achieve full coolant cleanup within 24 hours post-shutdown.

Control Mechanisms

Control mechanisms in high-temperature gas-cooled reactors (HTGRs) ensure stable reactivity during operation, load following, and shutdown. Primary reactivity control is achieved through control rods, which consist of neutron absorbers such as (B₄C) or encased in sleeves to compatibly interface with the -moderated core. These rods are positioned in channels within the side reflector or core periphery, allowing precise adjustment of absorption to maintain criticality. For emergency shutdown (), the rods are gravity-driven or electromagnetically released, achieving full insertion in less than 2 seconds, typically around 0.6–0.7 seconds to 80% of effective length, ensuring rapid negative reactivity insertion of several percent Δk/k. Burnable poisons are integrated into the fuel elements to manage initial excess reactivity and prevent excessive peaking early in the core life. Common materials include or compounds, which have high absorption cross-sections that diminish over time as the isotopes burn up, providing gradual reactivity hold-down without compromising long-term fuel utilization. These poisons are dispersed within TRISO fuel particles or fuel compacts, typically at concentrations optimized for equilibrium , such as in prismatic or pebble-bed designs. Reactivity feedback coefficients contribute to inherent stability in HTGRs. The Doppler coefficient, arising primarily from fuel temperature broadening of neutron resonances (mainly in ²³⁸U), is negative at approximately α_D = -0.5 pcm/K, enhancing self-regulation during power transients. The , which could be positive due to reduced coolant density increasing leakage in gas-cooled systems, is mitigated by design features like dense moderation and core geometry, ensuring the overall remains negative (typically -1 to -7 pcm/K total). Reactivity balance is quantified by ρ = (k_eff - 1)/k_eff, where k_eff is the effective multiplication factor; in HTGRs, operational designs maintain |ρ| < 1% (1000 pcm) during load following by balancing positions against fuel depletion, buildup, and temperature feedbacks, enabling flexible power adjustment without instability. Reserve shutdown systems provide diversity beyond primary s, often employing soluble injection as a mechanism to flood core channels or reflectors with neutron-absorbing solution, achieving shutdown margins exceeding 1% Δk/k even if rods fail. This is supplemented in some designs by absorber ball systems using pellets for gravity insertion.

Historical Development

Origins and Early Experiments

The conceptual origins of high-temperature gas-cooled reactors (HTGRs) emerged in the United States during the 1940s amid efforts to develop nuclear propulsion for military aircraft, driven by the need for compact, high-temperature heat sources capable of powering air-breathing engines without conventional fuel. The Aircraft Nuclear Propulsion (ANP) program, launched in 1946 under the joint oversight of the U.S. Air Force and the Atomic Energy Commission, explored gas-cooled reactor designs to achieve outlet temperatures exceeding 800°C for efficient jet propulsion. A key milestone was the Gas Cooled Reactor Experiment (GCRE), conducted from 1957 to 1959 at the National Reactor Testing Station in Idaho, which tested a helium- and nitrogen-cooled, graphite-moderated core at up to 32 MWth and 871°C, validating the thermal and neutronic performance of high-temperature gas cooling systems. These military-focused experiments provided foundational data on materials and heat transfer that influenced subsequent civilian HTGR development, shifting emphasis from propulsion to stationary power generation. The first dedicated HTGR prototype was the Dragon reactor in the , an international collaboration under the () and , constructed at and achieving criticality in 1963 with operations continuing until 1976. Rated at 20 MWth, Dragon featured a helium-cooled, -moderated core using prismatic fuel elements, primarily serving as a test platform for coated-particle fuels, high-temperature components, and helium circulation systems under pressures up to 2 MPa. Over its lifespan, it irradiated more than 250 fuel elements, demonstrating core stability at outlet temperatures around 750°C and contributing critical insights into fission product retention and thermal hydraulics that informed global HTGR designs. The project's success highlighted the viability of helium as a for achieving higher efficiencies than earlier carbon dioxide-cooled reactors. In the United States, early HTGR testing advanced to grid-connected operation with the Peach Bottom Unit 1 reactor in , which became critical in and supplied electricity from 1967 until its shutdown in 1974. This 40 MWe (115 MWth) facility, developed by Philadelphia Electric Company under the Atomic Energy Commission's initiative, was the world's first HTGR to deliver commercial power, using a helium-cooled core with hexagonal fuel blocks at outlet temperatures up to 760°C. It accumulated over 1,349 equivalent full-power days, testing fuel performance and steam cycle integration while confirming the technology's potential for baseload electricity with minimal operational incidents. Peach Bottom's data on core physics and component reliability directly supported larger-scale HTGR concepts. Germany pioneered the pebble-bed variant of HTGR with the AVR reactor at Jülich, operational from 1967 to 1988 as a 15 MWe (46 MWth) experimental unit that demonstrated continuous fuel recirculation. Cooled by helium at an average outlet temperature of 950°C—elevated from an initial 850°C in 1974—this design used spherical fuel elements containing thousands of coated uranium carbide particles, achieving high burnups and inherent safety through negative temperature coefficients. The AVR's 21-year operation provided essential validation of pebble-bed flow dynamics, dust management, and high-temperature materials, influencing subsequent modular HTGR architectures despite challenges like metallic impurity contamination. In the during the 1960s, HTGR research diverged from the dominant design—which employed but for dual civilian and plutonium production roles—toward helium-cooled concepts at facilities like the of . Early efforts focused on theoretical studies and small-scale loop tests for high-temperature gas cooling, aiming to leverage graphite's neutron economy for efficient power and process heat, though full-scale prototypes remained conceptual until the 1970s VGR-50 project. This parallel path underscored the Soviet emphasis on versatile graphite-based systems while addressing distinct challenges in helium purity and fuel fabrication.

Major Milestones and International Projects

The Fort St. Vrain (FSV) nuclear generating station in the United States represented a significant in scaling up prismatic-block high-temperature gas-cooled reactor (HTGR) to commercial power generation levels. Constructed by and owned by Company of Colorado, the 330 MWe (842 MWth) plant achieved initial criticality in 1974 and began commercial operation in 1976, operating until its permanent shutdown in 1989. Despite demonstrating the feasibility of -cooled prismatic fuel assemblies with thorium-uranium cycles, FSV encountered operational challenges, particularly repeated steam ingress events from the intermediate heat exchangers into the primary circuit, which led to oxidation and required extensive maintenance. These incidents provided critical lessons on material compatibility and system integrity under high-temperature conditions, influencing subsequent HTGR designs to prioritize advanced steam generators and ingress mitigation strategies. In , the marked the first full-scale deployment of pebble-bed HTGR technology for electricity production, advancing the modular concept toward commercialization. The 300 MWe (750 MWth) prototype, developed by Hochtemperatur-Kernkraftwerk GmbH (HKG) and commissioned in 1983 near Hamm, utilized thorium-highly fuel pebbles and achieved over 16,000 hours of operation before its shutdown on September 1, 1989. Key operational hurdles included difficulties with the continuous pebble refueling system, which experienced blockages and handling incidents that increased and maintenance costs. Although technically successful in validating pebble-bed core physics and safety margins, the plant's closure was precipitated by economic pressures and public opposition following these fuel handling issues, underscoring the challenges of integrating complex online refueling in commercial settings. Japan's High-Temperature Test Reactor (HTTR), a 30 MWth prismatic HTGR built by the Japan Atomic Energy Agency (JAEA) at the Oarai Research and Development Center, achieved criticality in 1998 and has since served as a cornerstone for high-temperature applications research. In June 2004, the HTTR first reached its design outlet temperature of 950°C, but a landmark milestone came in March 2010 with the completion of a 50-day continuous operation at this temperature, demonstrating long-term stability for process heat utilization such as and industrial . This achievement validated the reactor's features and fuel performance under prolonged high-temperature exposure, providing data that informed international HTGR safety standards and advanced the Generation IV very high-temperature reactor (VHTR) framework. The Pebble Bed Modular Reactor (PBMR) project in aimed to commercialize modular pebble-bed HTGRs for flexible, factory-assembled deployment, drawing on German AVR and heritage. Initiated in 1993 by and the PBMR Pty Ltd consortium, the design targeted 165 MWe per module with helium outlet temperatures up to 900°C, emphasizing economic viability through and high fuel burnup. However, escalating development costs, coupled with the global and inability to secure international customers or firm orders, led to progressive funding cuts by the South African government—first in March 2010 and fully in September 2010—resulting in the project's cancellation. In November 2025, the South African government announced plans to revive the project, lifting it from care and maintenance. Despite the termination, the PBMR effort yielded valuable insights into modular , fuel qualification, and economic modeling, which have influenced subsequent pebble-bed initiatives elsewhere. China's , a 10 MWth pebble-bed test reactor at the Institute of Nuclear and New Energy Technology (INET) in , , became operational in 2000 and has played a pivotal role in reviving and advancing HTGR technology on an industrial scale. Achieving full-power operation in 2003, the HTR-10 demonstrated safe shutdown without active systems during loss-of-coolant tests and provided essential data on pebble-bed neutronics, thermohydraulics, and fuel integrity at 750°C outlet temperatures. This experimental success directly paved the way for the demonstration project, a 210 MWe twin-module plant at Shidao Bay that entered commercial operation in December 2023, by validating the scalability of modular pebble-bed designs for commercial power and .

Safety Characteristics

Inherent Safety Features

High-temperature gas-cooled reactors (HTGRs) incorporate several features that rely on physical properties and design principles to prevent core damage and fission product release without the need for active intervention or operator action. These features stem from the reactor's core materials, form, and characteristics, ensuring self-regulation and passive management during normal operation and postulated accidents such as loss of or loss of power. A key inherent safety mechanism in HTGRs is the strong negative temperature coefficient of reactivity, which provides automatic self-regulation of the reactor power. As the core temperature rises, the reactivity decreases due to the of neutron absorption resonances in the and thermal expansion effects in the graphite moderator, leading to a reduction in fission power without control rod insertion. This coefficient remains negative across the full range of operating and accident temperatures, ensuring the reactor shuts down passively even in scenarios with limited flow. The TRISO (tristructural-isotropic) coated particle fuel represents another fundamental element, designed to maintain integrity and retain fission products under extreme conditions. Each particle consists of a kernel surrounded by multiple layers of and , which act as a robust and diffusion barrier. This fuel form retains fission products with very low release fractions—typically less than 0.1% for key isotopes—up to temperatures of 1600°C, well beyond the of conventional fuels, preventing significant radioactive release even if overheats. HTGRs operate at a low core power density, typically 3–5 kW/L, which is approximately 20–30 times lower than the 100 kW/L in light-water reactors (LWRs). This low density, combined with the large thermal mass of the graphite moderator, allows for extended passive removal of decay heat following shutdown, as the heat generation rate per unit volume remains manageable without forced cooling. The design ensures that fuel temperatures stay below critical limits for prolonged periods, facilitating natural dissipation to the environment. The core structure further enhances through its excellent thermal conductivity and high , enabling passive heat dissipation via conduction and after a (LOCA). In the absence of active cooling, conducts through the graphite matrix to the vessel and is radiated to surrounding structures or convected by residual gas, maintaining peak temperatures below 1600°C indefinitely. This passive mechanism eliminates the risk of core meltdown, as demonstrated in safety tests on the AVR reactor, where the core remained intact during simulated complete blackout and LOCA scenarios with no active systems.

Accident Mitigation Systems

High-temperature gas-cooled reactors (HTGRs) incorporate engineered accident mitigation systems to address beyond-design-basis events, ensuring and minimizing radiological releases through redundant, passive, and active features. These systems complement the inherent safety characteristics of HTGRs, such as high thermal margins and negative reactivity feedback, by providing additional barriers and cooling pathways during severe accidents like loss-of-coolant or external hazards. Containment structures in modular prismatic HTGR designs primarily rely on steel pressure vessels (RPVs) to enclose the reactor core, primary circuit, and associated components, maintaining integrity under accident conditions. The RPV serves as the final confinement barrier, designed to withstand internal pressures from primary coolant leaks or depressurization events. For instance, in designs like the General Atomics modular HTGR, the RPV holds helium pressures up to 6 MPa while accommodating thermal expansions and seismic loads, preventing uncontrolled release of fission products. This structure also integrates liners and insulation to limit heat transfer and corrosion, ensuring long-term stability during extended accident scenarios. Emergency cooling systems focus on passive decay heat removal to prevent core damage without relying on active power sources. Natural circulation loops utilize the helium coolant's buoyancy to transfer residual heat from the core to external heat exchangers during loss-of-forced-cooling events, maintaining fuel temperatures below 1600°C for extended periods. Complementing this, the reactor vessel auxiliary cooling system (RVACS) employs air-cooled heat exchangers surrounding the reactor vessel, dissipating up to 1-2% of full power as decay heat through natural convection and radiation, as demonstrated in analyses for modular HTGRs where RVACS alone suffices for indefinite cooling post-shutdown. These systems are integral to prismatic and pebble-bed configurations, providing multi-layered redundancy. Hydrogen management addresses potential generation from graphite moderator oxidation in steam-ingress accidents, using passive autocatalytic recombiners (PARs) to mitigate combustion risks. These devices, strategically placed within the , catalytically recombine and oxygen into without external power, maintaining concentrations below 4% to prevent . In advanced HTGR designs, PARs are integrated into the RPV or cavity, drawing on natural for effective distribution, as validated in severe simulations where they reduce hydrogen buildup by over 90% within hours. Seismic and flood protections are designed to withstand extreme external events, with HTGRs specifying a design basis typically of 0.2–0.3g (PGA), depending on site-specific hazards, with margins to higher loads such as 0.5g to ensure structural integrity of the RPV and core supports. Seismic isolation and damping features, such as base mats and flexible piping, limit accelerations to below equipment qualification thresholds, as seen in the prototype where analyses confirmed no core disruption under 0.5g horizontal loads. For flooding, elevated siting raises critical components above probable maximum flood levels—typically 5-10 meters above in coastal designs like —incorporating watertight barriers and drainage to prevent inundation of safety systems. Following the 2011 Fukushima-Daiichi accident, modern HTGRs like China's have implemented post-Fukushima upgrades, including enhanced venting systems with filtered paths to manage buildup and recombiners for severe . These incorporate advanced monitoring , such as real-time sensors and seismic detectors integrated into the and , ensuring early detection and automated response to multi-unit or external events. Such enhancements, verified through re-evaluations and commissioning tests in 2021–2023, confirm the HTR-PM's ability to maintain core integrity without off-site power for over 72 hours.

Advantages and Applications

Performance and Economic Benefits

High-temperature gas-cooled reactors (HTGRs) achieve thermal efficiencies of up to 50%, significantly higher than the approximately 33% typical of light-water reactors (LWRs), due to their high core outlet temperatures of 700–950°C that enable advanced Brayton or combined cycles. This elevated efficiency reduces fuel consumption per unit of electricity generated, with burnups reaching 80,000–150,000 MWd/tU, compared to 40,000–50,000 MWd/tU in LWRs, thereby minimizing the volume of spent fuel and by factors of up to four relative to LWRs through more complete fission of . Additionally, HTGRs support extended fuel cycles of 3–6 years between refuelings, facilitated by robust TRISO-coated particle fuel that withstands high temperatures without failure, allowing for operational flexibility in both prismatic and pebble-bed configurations. Economically, modular HTGR designs benefit from factory fabrication of standardized components, such as reactor vessels, heat exchangers, and fuel elements, which reduces on-site construction time to approximately 3 years—half that of traditional large-scale reactors—while improving and lowering labor costs through serial production. (LCOE) estimates for nth-of-a-kind modular HTGRs range from $60–80/MWh in 2019 USD (escalated to similar ranges for 2025 projections), competitive with or lower than LWRs at $70–90/MWh, owing to high capacity factors exceeding 90% and reduced outage durations of 30–60 days per cycle. These factors contribute to overall cost advantages, with potentially 20–30% lower than non-modular designs due to in . From an environmental perspective, HTGRs exhibit lifecycle of approximately 10 g CO₂ eq./kWh, among the lowest for nuclear technologies, primarily from and processing rather than operations. In deep-burn modes, where transuranic elements from LWR spent fuel are incorporated, HTGRs transmute long-lived actinides into short-lived fission products, eliminating most long-lived components and further reducing radiotoxicity and disposal burdens compared to conventional cycles.

Industrial and Hydrogen Uses

High-temperature gas-cooled reactors (HTGRs) are particularly suited for delivering process heat at temperatures up to 950°C, enabling decarbonization in energy-intensive sectors such as , , and chemical processing. These reactors can supply high-quality heat through intermediate heat exchangers, reducing reliance on fossil fuels for processes like reduction in (requiring 800–900°C) and lime in production (around 900°C). In chemical industries, HTGR heat supports endothermic reactions, such as synthesis or production, with modular designs like the Xe-100 providing 200 MWth of thermal output tailored for such applications. For example, in March 2025, Dow and submitted a construction permit application to the for a proposed Xe-100 HTGR project at Dow's site, aimed at providing power and steam for , with NRC review ongoing as of November 2025. A key application is via thermochemical , where HTGRs integrate with cycles like the sulfur-iodine (S-I) process to achieve efficiencies of 40–50% on a higher heating value basis. The S-I cycle involves high-temperature decomposition of , represented by the reaction: H2SO4→SO2+H2O+12O2\mathrm{H_2SO_4 \rightarrow SO_2 + H_2O + \frac{1}{2}O_2} at approximately 800°C, followed by lower-temperature steps for iodine recycling and generation, all powered by HTGR heat without direct electrical input. For instance, a 200 MWth HTGR module can support S-I cycle operations to produce significant volumes, with studies showing net thermal efficiencies up to 47.6% when coupled to reactors outlet temperatures of 950°C. This approach yields zero-carbon suitable for fuels, chemicals, and . HTGRs also enable for and seawater , leveraging their high outlet temperatures for efficient multi-purpose operation. The High-Temperature Test Reactor (HTTR) in demonstrated this capability by achieving 950°C outlet temperatures and testing steam generation for , producing up to 10 tons of fresh water per day in coupled systems. Such configurations allow simultaneous electricity, , and water production, with via multi-stage flash processes benefiting from HTGR's stable thermal supply at 80–150°C after heat cascading. For production (CO + H₂), HTGRs facilitate high-temperature steam methane reforming at 800–1000°C, enhancing over conventional methods by providing clean to endothermic reactions. This process supports downstream applications like Fischer-Tropsch synthesis for synthetic fuels, with nuclear-assisted reforming reducing CO₂ emissions by up to 35% compared to natural gas-fired plants. Market projections indicate that nuclear sources, including HTGRs, could supply up to 10% of heat demand by 2030, driven by decarbonization goals and the technology's ability to meet high-temperature needs in hard-to-abate sectors. By , initial deployments in pilots are expected to demonstrate scalability, with the market growing to support this transition.

Deployment and Status

Operational and Decommissioned Reactors

The High-Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) in , located at Shidao Bay in Province, represents the world's first commercial-scale modular HTGR deployment. This plant features two 250 MWth reactor modules coupled to a single , delivering a net electrical output of 210 MWe with a coolant outlet temperature of 750°C. It achieved initial grid connection on December 20, 2021, and entered full commercial operation in December 2023, demonstrating stable performance in and planned co-generation applications. With the successful operation of the HTR-PM, China has fully mastered the complete HTGR technology chain—from design and manufacturing to construction, commissioning, operation, and decommissioning—positioning it as a global leader in Generation IV HTGR designs. The High-Temperature Test Reactor (HTTR) in , operated by the Japan Atomic Energy Agency (JAEA) at the Oarai Research Establishment, is a prismatic-block HTGR dedicated to . With a thermal capacity of 30 MWth and no net electrical generation, it has been operational since achieving initial criticality in November 1998, reaching full-power tests at 950°C outlet temperature by 2004. As of 2025, the HTTR continues intermittent operations for validation, irradiation, and demonstrations, with restarts following regulatory approvals in 2021 and ongoing plans for facility expansions. Among decommissioned HTGRs, the Fort St. Vrain (FSV) plant in the United States, a prismatic-block design with a reactor vessel, operated from 1976 to 1989 as the only commercial-scale HTGR in . Rated at 842 MWth and 330 MWe, it faced challenges including helium circulator failures and leaks, achieving a lifetime below 20% before permanent shutdown in August 1989 due to economic factors. Decommissioning commenced in 1990, with fuel removal completed by 1992 and full site restoration achieved by 2011 using dry storage for spent fuel. The High-Temperature Reactor () in , a pebble-bed prototype, provided key operational data from its grid connection in 1985 until shutdown on September 1, 1989. With a thermal capacity of approximately 750 MWth and 300 MWe output, it utilized thorium-uranium fuel elements and operated for over 16,000 hours, generating about 1.67 billion kWh before closure due to financial and political pressures rather than technical failures. Decommissioning involved core defueling by 1995, transitioning to safe enclosure () status, with partial dismantling ongoing as of 2025 under German nuclear phase-out policies. Russia's VGR-50 prototype, a loop-type HTGR at the site, served as an early experimental platform integrating gas turbine elements during its operational phases from 1972 to 1987. Rated at around 50 MWth, it focused on high-temperature testing and services without commercial power production, contributing to subsequent designs like the VGM modular concept before decommissioning in the late . Globally, HTGRs have accumulated limited operational experience as of , totaling several GW-years across historical and current facilities in , process heat testing, and demonstrations, though constrained by historical scale and regional programs.
ReactorCountryTypeThermal Power (MWth)Electrical Output (MWe)Operational PeriodKey Notes
Pebble-bed500 (twin modules)2102021–presentCommercial modular deployment; co-generation capable.
HTTRPrismatic300 (test)1998–presentR&D focus on high-temperature applications.
Fort St. VrainPrismatic8423301976–1989Decommissioned; economic challenges.
THTR-300Pebble-bed~7503001985–1989Decommissioned; fuel testing.
VGR-50Loop-type~500 ()1972–1987Decommissioned; gas turbine integration tests.

Proposed and Under-Development Designs

China's HTR-PM600 project represents a significant expansion of the high-temperature gas-cooled reactor (HTGR) technology demonstrated by the initial units, featuring six 250 MWth pebble-bed modules connected to a single for a net capacity of 600 MWe. This design maintains the inherent safety features of the original while scaling up for commercial power generation and potential applications. The preliminary safety analysis report for the HTR-PM600 was submitted to regulators in , with ongoing reviews supporting commercialization efforts expected in the post-2030 timeframe. In the United States and , X-energy's Xe-100 is a pebble-bed HTGR (SMR) designed for 80 MWe per unit, with a standard four-unit configuration delivering 320 MWe for flexible deployment at industrial sites or power grids. The uses TRISO for high-temperature operation up to 750°C, enabling electricity production alongside process heat for applications like . A completed in September 2025 confirmed the Xe-100's suitability for repurposing coal sites in , with first-of-a-kind deployment targeted for the early 2030s through partnerships like the one with . In November 2025, X-energy announced a supply agreement for initial deployment and began vertical on its TRISO-X fabrication facility to support Xe-100 fueling. The Micro Modular Reactor (MMR), originally developed by Ultra Safe Nuclear Corporation, is a compact HTGR variant producing up to 15 MWe or 35 MWth, emphasizing process heat for remote or industrial uses such as and , with a focus on TRISO fuel encapsulation for enhanced safety. Following USNC's in 2024, the MMR technology was acquired by NANO Nuclear Energy in December 2024 and rebranded as the KRONOS MMR, with plans to advance demonstrations including a prototype at the and a U.S. Air Force contract. Although earlier proposals for an MMR deployment at Idaho National Laboratory were not pursued, the acquisition supports continued development toward operational prototypes in the late 2020s. As of 2025, the (IAEA) projects continued growth in advanced nuclear technologies, including HTGRs, with global nuclear capacity under construction reaching approximately 64 GW across all types and further expansions anticipated by 2030 to meet decarbonization goals; specific HTGR contributions are expected from ongoing projects in and , potentially adding several gigawatts in planning or early development phases.

Research and Innovations

Current R&D Efforts

In , the Institute of Nuclear and New Energy Technology (INET) at is conducting ongoing fuel qualification efforts for the reactor, including tests to validate performance at high levels exceeding 100 GWd/t, with recent analyses in 2024–2025 focusing on calibration and history dependencies to support extended fuel cycles. These tests build on prior qualifications and aim to confirm TRISO particle integrity under operational conditions up to 950°C, contributing to the reactor's commercial scalability. The Japan Atomic Energy Agency (JAEA) is advancing hydrogen production demonstrations using the High-Temperature Test Reactor (HTTR), with 2025 activities including preparations for integrating high-temperature helium gas at 950°C with steam methane reforming processes to produce clean hydrogen. This initiative, in collaboration with Mitsubishi Heavy Industries, targets operational tests by 2028 but involves 2025 regulatory applications and mock-up validations to demonstrate efficient heat transfer for decarbonized hydrogen generation. While primary focus is on steam reforming, exploratory synergies with high-temperature electrolysis are under consideration to enhance efficiency. Under the Euratom-funded SafeG project (2020–2024), research on the ALLEGRO gas-cooled fast reactor demonstrator incorporates synergies with high-temperature gas-cooled reactor (HTGR) technologies, particularly in for high-temperature gas cooling environments. Key outcomes, reported in publications dated 2025, include evaluations of cladding and structural materials like SiC composites to withstand coolant at temperatures above °C, drawing on HTGR expertise to improve resistance and thermal performance in gas-cooled systems. These efforts strengthen features such as removal, with project reports emphasizing cross-technology knowledge transfer for modular designs. The U.S. Department of Energy's Advanced Reactor Demonstration Program (ARDP) requested approximately $55 million in the FY 2026 budget (as of August 2025) for Advanced Reactor Technologies, including research and development in very high-temperature reactors (VHTRs)/HTGRs such as graphite and fuel qualification, with $10 million specifically for the Xe-100 HTGR demonstration and $58 million under Next Generation Fuels for TRISO fuel; this extends from 2025 planning and supports modeling via $28.6 million in Nuclear Energy Advanced Modeling and Simulation. FY2026 appropriations enacted in September 2025 provided $1.9 billion overall for the Office of Nuclear Energy, exceeding the request, though specific VHTR allocations remain consistent with the proposal as of November 2025. This funding supports TRISO fuel testing at high and multiphysics simulations for VHTRs, aiming to reduce deployment risks through virtual prototyping integrated with experimental data from facilities like the National Reactor Innovation Center. As of November 2025, reported continued advancements in the Xe-100 HTGR, focusing on fuel qualification and regulatory progress for 2027 deployment. The (IAEA) is leading a Coordinated Research Project (CRP) on modular high-temperature gas-cooled reactor safety design, active from 2024 through 2026, with a focus on multiphysics modeling to assess features under accident scenarios. Participants, including JAEA and INET, are developing integrated models for neutronics, thermal-hydraulics, and fuel behavior to propose updated safety criteria, with interim updates presented in early 2025 emphasizing in gas-cooled systems. This CRP facilitates international collaboration on validation benchmarks, ensuring robust computational tools for future HTGR deployments. In November 2025, Dutch engineering firm announced progress on adapting high-temperature gas-cooled reactor (HTGR) technology for small modular reactors (SMRs) in offshore applications, targeting integration with maritime industries to reduce global shipping emissions by up to 5% through high-temperature process heat for decarbonization.

Future Technological Advances

The Very High-Temperature Reactor (VHTR), a key Generation IV design, aims to achieve core outlet temperatures exceeding 950°C to enable high-efficiency and advanced process heat applications, with potential extensions beyond 1000°C in future iterations. This evolution supports closed fuel cycles incorporating thorium-uranium mixtures, enhancing resource utilization and waste reduction compared to traditional once-through approaches. Hybrid configurations integrating VHTR thermal spectra with fast-spectrum elements are also under conceptual exploration to optimize transmutation and fuel breeding. Advancements in structural materials, particularly silicon carbide fiber-reinforced silicon carbide (SiC/SiC) composites, are poised to enable VHTR operations at temperatures up to 1200°C by providing superior oxidation resistance and thermal stability in environments. These composites mitigate material degradation at elevated temperatures, potentially allowing reduced coolant pressures or purity requirements while maintaining core integrity. Integration of and for represents a transformative operational enhancement for HTGRs, with implementations anticipated between 2025 and 2035 to monitor component health in real-time and preempt failures. Such systems leverage sensor data to optimize maintenance schedules, drawing on ongoing demonstrations in advanced reactor prototypes. Proliferation-resistant features in future HTGR designs emphasize once-through fuel cycles using low-enriched , where buildup remains below 5% of heavy metal content due to high rates exceeding 100 GWd/t, rendering extracted material less suitable for weapons. The TRISO-coated particle architecture further complicates diversion by encasing in durable matrices. Market projections indicate substantial growth for HTGRs within the sector, with global investments forecasted to reach several billion dollars by the early 2030s, driven by demand for high-temperature heat in decarbonized industries. By 2040, HTGRs are expected to capture a notable share of SMR deployments, supported by international R&D collaborations targeting commercial viability.

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