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Very High Temperature Reactor (VHTR)

The Very High Temperature Reactor (VHTR), one of six Gen IV nuclear system candidates, is designed for electricity and heat cogeneration, and its high outlet temperature is ideal for hydrogen production and use in the chemical, oil, and iron industries. The VHTR features TRi-structural ISOtropic (TRISO) fuel, helium coolant, a specialized core layout, and low power density that supports passive decay heat removal. Drawing on operational experience from 47 gas-cooled reactors in five countries, the VHTR aims for outlet temperatures of 1,000°C to enhance hydrogen production efficiency. It also promises inherent safety, high thermal efficiency, process heat capability, low operational costs, and modular construction.

Presentation of the Very High Temperature Reactor System

Among the 6 candidates of the Gen IV nuclear systems in the technical roadmap of Gen IV International Forum (GIF), the Very High Temperature Reactor (VHTR) is primarily dedicated to the cogeneration of electricity and heat. Its high outlet temperature makes it attractive for hydrogen production as well as for the chemical, oil and iron industries.

The reference design considered by GIF utilizes the TRi-structural ISOtropic (TRISO) particles based fuel, helium coolant, as well as the dedicated core layout and lower power density to allow removal of decay heat through passive mechanisms.

The VHTR technology benefits from the operational feedback of 40 Gas Cooled Reactors (GCR, CO2 cooled) and 7 High Temperature Gas Reactors (HTGR, He cooled) gained in 5 different countries.

Original target of outlet temperature of 1 000°C from VHTR would support the efficient production of hydrogen by thermo-chemical processes. The VHTR has potential for inherent safety, high thermal efficiency, process heat application capability, low operation and maintenance costs, and modular construction.

Schematic view of a Very High Temperature nuclear Reactor (VHTR)

Schematic view of the Gen. IV Very High Temperature Reactor (VHTR) Nuclear Energy System

Attributes of the VHTR

The VHTR is a next step in the evolutionary development of high-temperature gas-cooled reactors (HTGRs). It is a graphite-moderated, helium-cooled reactor with thermal neutron spectrum. It can supply nuclear heat and electricity over a range of core outlet temperatures between 700 and 950°C, and potentially more than 1 000°C in the future.

The base elements of VHTR fuel are TRISO particles, each smaller than a millimeter in diameter. These particles contain a uranium oxycarbide kernel (about 0,5mm in diameter), enriched up to 17% U-235. Surrounding this kernel are layers of carbon and silicon carbide, providing stable containment for fission products at temperatures of 1600°C or higher. These particles can be either integrated in cylindrical compacts integrated in blocks resembling hexagonal 'prisms' made of graphite (prismatic-block type core) or encased in silicon carbide within graphite in pebbles of about 60mm in diameter (pebble-bed type core).

Although the shapes of the fuel compounds for these two configurations are different, the technical basis for both is the same: TRISO coated particles fuel in a graphite matrix, fully ceramic (graphite) core structure, helium coolant, and low power density, to achieve high outlet temperature and the retention of fission production inside the coated particle under normal operation condition and accident condition.

As mentioned before, to allow passive decay heat removal, each core options should be kept below the following power limits:

Below 625 MWth for prismatic-block type
Below 250 MWth for pebble bed type

The VHTR should be able to support alternative fuel cycles such as U-Pu, Pu, MOX, U-Th.

For electricity generation, a direct cycle with a helium gas turbine system directly placed in the primary coolant loop, or, at the lower end of the outlet temperature range, an indirect cycle with a steam generator and a conventional Rankine cycle can be used. A variety of different configurations of the reactor and balance of plant have been examined across the world. Today, the direct Brayton power conversion cycle is being pursued less aggressively in favour of an indirect Rankine cycle because of its lower technological risk and increased flexibility in terms of working fluid and mission of the reactor (electricity, process heat, and co-generation).

For nuclear heat applications such as process heat for refineries, petrochemistry, metallurgy and hydrogen production, the heat application process is generally coupled with the reactor through an intermediate heat exchanger (IHX), in what is often called an indirect cycle.

VHTR Reactor Parameters	Reference Value
Spectrum	Thermal
Core Outlet Temperature range	Graphite
Coolant	700–950°C with the goal to reach above 1000° temperatures at a later stage.
Primary pressure	He Gas
Power Range	10 MWth to 842 MWth
Fuel	Uranium containing (enriched in U235) TRISO based pebbles or compacts, can support alternative fuel cycles such as U-Pu, Pu, MOX, U-Th.

How does Very High Temperature Reactors meet Generation IV Criteria?

Type of fuel

TRISO particle-based fuels are highly resistant to neutron irradiation, corrosion, oxidation, and high temperatures. The triple-coated layers of each particle act as a self-contained system, guaranteeing the retention of fission products under various reactor conditions. Fission product retention has been demonstrated up to 1600°C.
The coated particle fuel form is the primary barrier to fission product release. The coatings, in particular the silicon carbide layer, effectively retain the bulk of the radionuclide inventory under all anticipated operating and accident conditions. The fuel performance envelope was largely established in the first generation HTGR programs in the US and Germany which specified limits on power density, temperature, fluence, burnup, and packing fraction. Recent fuel fabrication and qualification efforts have further improved the performance with respect to particle failures and radionuclide release. The performance of TRISO fuel is demonstrated through qualification testing (irradiations and post-accident heating) under conditions that fully envelope those of the reactor under all anticipated operating and accident conditions.

Management of waste

The reprocessing of TRISO particles fuel is still being investigated. Yet the TRISO particles themselves, notably before of the characteristics mentioned just above are great at encapsulating radioactive elements in a very stable envelope.

Radiation Protection

Internal radiation levels in the VHTR plant are generally lower than those in LWR because of the low circulating activity in the coolant, mainly in the form of graphite dust on which some radionuclides may be adsorbed or otherwise carried.

Other intrinsic safety features and very long grace periods:

The VHTR design makes use of the inherent safety features of a graphite-moderated core cooled by helium. The graphite possesses substantial thermal inertia, and the helium coolant is a single-phase, inert substance with no reactivity effects.

The VHTR core has a low power density, a high heat capacity, and a slender core shape (large height-to-diameter ratio). The transients resulting from Loss of Forced Cooling (LOFC) events develop and occur over tens or hundreds of hours; a very long ‘grace period’ relative to LWRs. If forced cooling of the VHTR core is lost, decay heat is transferred naturally via conduction and radiation from the core, through the outer reflector and reactor pressure vessel (RPV), and on to the ultimate heat sink. For VHTR, the removal of decay heat does not require active cooling with helium.

Benefits and Applications of Very High Temperature Reactors

While the original approach for VHTR at the start of the Generation IV program focused on very high outlet temperatures and hydrogen production, current market assessments have indicated that electricity production and industrial processes based on high temperature steam that require modest outlet temperatures (700-850°C) have the greatest potential for application in the next decade while also reducing the technical challenges associated with higher outlet temperatures. As a result, the focus has moved from higher outlet temperature designs such as in GT-MHR and PBMR reactors to lower outlet temperature designs such as HTR-PM in China and the NGNP in the US. These experiences should nevertheless pave the way for higher output temperature capable materials in the future.

Most of the benefits and advantages of VHTR designs have been listed in the above description of how those reactors meet the Generation IV criteria. Yet the potential, opened by very high outlet temperatures of VHTRs, for an expanded range of applications of nuclear heat cannot be understated.

Main Challenges for the Deployment of VHTR

Demonstrating the viability of the VHTR core requires solving several significant technical challenges. Fuels and materials must be developed that:

permit an increase of the core-outlet temperatures from around 800°C to more than 1 000°C for the entire plant lifetime;
permit the maximum fuel temperature under accident conditions to reach levels approaching 1 800°C;
permit maximum fuel burnup of 150-200 GWd/tHM;
avoid power peaking and temperature gradients in the core, as well as hot streaks in the coolant gas;
limit structural degradation from air or water ingress.

An important amount of work is needed regarding the use of the reactor heat for industrial processes and the coupling of a nuclear plant with other industrial facilities.

Process-specific R&D gaps need to be filled to adapt the chemical process and the nuclear heat source to each other regarding temperatures, power levels and operational pressures. Heating of chemical reactors by helium is a departure from current industrial practice and needs specific R&D and demonstration. The development of an intermediate heat exchanger, ducts, valves and associated heat transfer fluid is needed to deliver process heat to many of the chemical processes. The viability of using nuclear process heat to produce hydrogen needs further study. Any contamination of the product will have to be avoided. Development of heat exchangers, coolant gas ducts and valves will be necessary for isolation of the nuclear island from the production facilities. This is especially the case for isotopes like tritium, which can easily permeate metallic barriers at high temperatures. Over the past two decades, significant advances have been made in the key technologies necessary to deploy a VHTR. Further R&D is needed, and GIF is actively participating to deploy VHTR.

How is GIF working to solve those

In it’s 2014 Technology Roadmap GIF identified the following areas of R&D as priorities for the VHTR system:

Fuels and Materials
Reactor Systems
Balance of Plant
Fuel Cycle
Economics
Safety Objectives

To know more please consult the 2014 GIF Technology Roadmap.

GIF has established System Steering Committees (SSC) to implement the research and development (R&D) for each Generation IV Reactor Concept, with participation from GIF Members interested in contributing to collaborative R&D efforts on the said system. Each SSC plans and integrates R&D projects contributing to the development of a system.

Each SSC is governed by a System Arrangement (SA), provisional SSCs are created under a Memorandum of Understanding (MoU).

Several projects can be carried within an SSC, each project is government by a Project Arrangement (PA) and managed by a Project Management Board (PMB).

The GIF VHTR SSC was established in 2006. To effectively support the development of the VHTR technology/system and to overcome jointly the challenges identified on the path to do so, the GIF coordinates several international R&D activities related to VHTR system. They are now arranged by the VHTR SSC Signatories into four joint R&D projects.

GIF VHTR System Steering Committee and Projects

VHTR Projects History - Past projects

The VHTR technology benefits from the operational feedback of 40 Gas Cooled Reactors (GCR, CO2 cooled) and 9 High Temperature Gas Reactors (HTGR, He cooled) gained in 7 different countries (China, France, Germany, Japan, Spain, UK, US).

The High-Temperature Gas-Cooled Reactor (HTGR) has a history spanning over 50 years, marked by successes and challenges. The initial experimental plants included a 20 MWe unit in the UK and a 15 MWe unit in Germany, operational in 1966. The German project, intended to burn thorium, faced political and technical obstacles, leading to its abandonment in 1988.

In the U.S., two commercial power-producing HTR plants were constructed. The first, a 200 MWth/115MWth experimental unit at Peach Bottom, operated from 1967 to 1974, providing valuable insights. These findings contributed to the development of a larger 330 MWe plant at Fort St Vrain, active from 1979 to 1989. While this reactor demonstrated the technical feasibility of HTRs, further development was needed to establish their economic viability.

Slide presenting past HTGR VHTR nuclear reactors

Test Reactors and Prototypes - HTGRs/VHTRs

Project/Program	Country	Power Rating	Timeframe	Notes
High-Temperature Gas-Cooled Reactor (HTGR)	University of Chicago, US	N/A	1944 - 1947	Initial proposal of HTGR
Dragon Reactor	United Kingdom	21.5 MWth	1965 - 1976	Exploring use of TRISO fuel
Peach Bottom Unit 1 Reactor	United States	200 or 115 MWth	1966 - 1974	First HTGR to produce electricity
AVR Reactor	Germany	46 MWth	1967 - 1988	Exploring use of carbide BISO, and TRISO particle fuel
Fort St. Vrain Generating Station	United States	842 MWth	1979 - 1989	Load-following power plant
THTR-300	Germany	750 MWth	1983 - 1989	Prototype reactor using TRISO fuel

VHTR Current Developments

Operating VHTR Reactors

Three VHTR/HTGRs plants are currently operating with the HTR-PM being a commercial scale nuclear power plant with two reactors feeding one turbine.

Project Name	Country	Power Rating	Expected Deployment Date	Notes
High-Temperature Engineering Test Reactor	Japan	30 MWth	First criticality in 1998, operated 50 days in 2010, restarted in July 2021.	Outlet temperature of 950°C was reached at full thermal power in 2004 for the first time in the world. One of the current focuses is demonstrating elements related to coupling with H2 production.
HTR-10	China	10 MWth	First criticality in 2000	Prototype reactor of pebble-bed HTGR
HTR-PM (Shidao Bay 1)	China	250 MWth/reactor 200MWe net output for the whole plant with two reactors feeding one turbo-generator.	Criticality 2021 / Commercial operation December 2023	Scaled-up version of HTR-10 Two reactors feeding one turbine.

Project Name

Country

Power Rating

Expected Deployment Date

Notes

High-Temperature Engineering Test Reactor

Japan

30 MWth

First criticality in 1998, operated 50 days in 2010, restarted in July 2021.

Outlet temperature of 950°C was reached at full thermal power in 2004 for the first time in the world.
One of the current focuses is demonstrating elements related to coupling with H2 production.

HTR-10

China

10 MWth

First criticality in 2000

Prototype reactor of pebble-bed HTGR

HTR-PM (Shidao Bay 1)

China

250 MWth/reactor

200MWe net output for the whole plant with two reactors feeding one turbo-generator.

Criticality 2021 / Commercial operation December 2023

Scaled-up version of

HTR-10
Two reactors feeding one turbine.

VHTRs under development

Project/Program	Country	Power Rating	Expected Deployment Date	Notes
GTHTR300C	Japan Atomic Energy Agency, Japan	600 MWth 274 MWe	Conceptual	45~50% thermal efficiency
PMBR	Pebble Bed Modular Reactor (Pty) Limited / Eskom, South Africa	400 MWth / 200MWe	On Hold	International Cooperation design team, aimed for industrial application
Prismatic HTR	General Atomics, US	350 MWth 150 MWe	Under Design	TRISO particles, assembling in 2 steps.
SC-HTGR	Framatome, France/USA	625 MWth 272 MWe	Not known	Based on HTR-MODUL
U-Battery	Urenco, UK-Germany-The Netherlands	10 MWth 4 MWe	Cancelled in 2023	Based on Dragon reactor/Fort St. Vrain reactor
USNC MMR	Ultra Safe Nuclear Corporation , US	10~45 MWth 3.5~15 MWe	Pre-licensing steps carried out in Canada (VDR 1 done, VDR 2 ongoing, application for a license to prepare the site)	Micro reactors, Ceramic core design, Heat storage to allow for flexibility of output without having to cycle the reactor
Holos-Quad HTR	Hologen, US Military	22 MWth 10 MWe	Under development with ANL	Transportable in a 40-foot shipping container
Adams Engine	Adams Atomic Engines	10 MWe	Design shelved	Liquid nitrogen coolant
HTMR-100	Steenkampskraal Thorium Limited, South Africa	100 MWth 35 MWe	Conceptual	Very strong negative temperature coefficient
GT-MHR	General Atomics, US	600 MWth 285 MWth	On hold before construction	Multi-layer ceramic coating fuel
MIGHTR - Modular Integrated Gas High Temperature Reactor	Boston Atomics, US	20-600MWth	In development since 1984

GIF VHTR Related Publications

GIF has produced several reports and conducted analysis on VHTR Systems produced by cross cutting methodological working groups (Risk and Safety WG, Proliferation Resistance & Physical Protection Working Group). GIF's annual reports, technology roadmap and R&D Outlooks provide more information on the progress made by GIF's VHTR System Steering Committee and Project Management Boards.