The Very High Temperature Reactor

Benjamin Kallman
March 19, 2013

Submitted as coursework for PH241, Stanford University, Winter 2013

Fig. 1: Refueling the prismatic fuel blocks at the Fort Saint Vrain helium-cooled reactor. (Source: Wikimedia Commons)

The Very High Temperature Nuclear Reactor (VHTR) is one of 6 technologies classified by the Generation IV International Forum as a promising reactor type likely to power our world in the coming decades. The fourth generation reactor designs are expected to be deployed between 2015-2025, and are identified as being economical, safe, efficient, and proliferation resistant. [1] The fundamental physics governing the operation of the VHTR is identical to the current generation of fission reactors. The nuclei in a fissile fuel are split via neutron bombardment, converting nuclear binding energy into thermal energy (~200 million electron Volts per fission event) capable of being converted to useful work or process heat. [2] The energy released by fission events is transferred out of the reactor core via a working fluid. The defining characteristic of a VHTR is the very high temperature of this working fluid, capable of running an efficient power cycle or being used as a high temperature input for a chemical transformation process (e.g. hydrogen production).

In the following sections, I will outline the benefits, design traits, and challenges of a VHTR.

Benefits of the VHTR

The VHTR offers two advantages to modern day generation III reactor designs. The high temperature of the coolant exiting the reactor core enables high thermal efficiency for electricity generation, and can serve as process heat for hydrogen production.

Efficient Power Generation

Most nuclear power plants currently in operation harness the thermal energy released during fission events by boiling water to create high pressure, high temperature steam. This steam is used to drive a steam turbine in a Rankine cycle, whose efficiency is governed in part by the temperature of the steam at the turbine inlet. Typical outlet temperatures for pressurized light water (PWR) reactors are less than 400°C, leading to thermal efficiencies (ratio of electric power generated to thermal energy input) around 34 percent. [3, 4] Gas-cooled reactors have been built, in which gas (carbon dioxide or helium) is used as the coolant, enabling higher coolant outlet temperatures (up to about 850°C) for use in a Brayton cycle. [5] The AVR reactor in Germany demonstrated Helium outlet temperatures above 900°C, but was plagued by contamination and only operated for four years. [6] The VHTR is the next generation of high-temperature gas-cooled reactors (HTGRs). The VHTR uses helium as coolant, with design temperatures breaking 1000°C (eventually slated to reach 1500°C). [5] Such high temperatures allow for higher thermal efficiencies (>50%), and therefore more electricity output per unit of nuclear fuel. [5]

Efficient Hydrogen Production

The high coolant temperature also allows for efficient hydrogen production. In a carbon-constrained world, hydrogen presents an attractive option as an energy carrier. Hydrogen combustion does not produce CO2 and Hydrogen is 3 times as energy dense as gasoline. [7] However, current hydrogen production methods rely on fossil fuels. The majority of hydrogen in the U.S. is produced using the steam-methane reforming process, in which natural gas and steam react in the presence of a catalyst (usually Nickel) at high temperatures (800 to 1000°C) to form syngas (a mixture of carbon monoxide and hydrogen). Additional hydrogen is extracted from the syngas through a lower temperature water gas shift reaction. [8] This process consumes 5% of the annual U.S. natural gas supply, and emits 74 million tons of CO2. [9] If we are to use hydrogen as an energy carrier in a carbon-constrained world, we must produce hydrogen in a clean and efficient manner.

Fig. 2: Table summarizing potential high-temperature applications of the VHTR. After reference [14].

Very High Temperature Reactors will supply process heat that can be used to generate the high temperatures required for clean hydrogen production. High temperature electrolysis and thermochemical water splitting offer two promising methods for nuclear powered hydrogen production. High temperature electrolysis works on the principal that the energy required to electrolyze water decreases with increasing temperature. The waste heat from a Brayton power cycle can be used as input to the electrolysis plant, enabling high overall hydrogen production efficiencies (45-50% thermal energy input to amount of hydrogen produced based on lower heating value of hydrogen). [10] Thermochemical water-splitting involves a series of thermally driven chemical reactions to separate water into hydrogen and oxygen. The Sulfur-Iodide cycle is a common thermochemical water-splitting cycle, in which heat is supplied as input to an endothermic reaction that dissociates sulfuric acid, eventually producing hydrogen and oxygen gas. Temperatures required for efficient S-I cycles are in excess of 950 °C and efficiencies have been reported above 50% (heat input to hydrogen output, again based on LHV). [9] Thermochemical water-splitting efficiencies are very temperature dependent. As temperature increases from 900 to 1000 °C, hydrogen production efficiency increases from about 50 to 60 percent. [9]

There are other hydrogen production methods that require high temperatures as input. Coal gasification presents an especially attractive option for hydrogen production in the U.S. given the abundance (484 billion tons of coal in reserves) of coal and current mining infrastructure. [11] Coal gasification requires a coal temperatures between 800 and 1800 °C, which can in part be reached using the high coolant temperatures from a VHTR. [12] Clean coal gasification will require advanced carbon capture systems.

In addition to hydrogen production, high temperature coolant can be used for other useful applications including desalinating water, municipal heating, iron and glass manufacturing, urea synthesis, cement manufacturing, and much more. (See Fig. 2.)

Design Traits of the VHTR

The VHTR represents the evolution of traditional gas-cooled reactors. The design of the VHTR is therefore similar to current generation III+ gas-cooled reactors, which fall into two general categories: prismatic block reactors, and pebble bed reactors. The two designs refer primarily to the fuel arrangement. Prismatic block reactors involve a fuel core surrounded by a hexagonal graphite reflector, while pebble bed reactor cores use mobile fuel spheres about 6 cm in diameter that cycle throughout the core. The reference design for the prismatic block VHTR is the GT-MHR (Gas-Turbine Modular Helium Reactor) by General Atomics. Much of the following discussion is based on the GT-MHR reactor. The VHTR will use a once-through fuel cycle with a thermal neutron spectrum meaning U-235 nuclei will be split, emitting neutrons which will be moderated (or thermalized) to a lower energy level (nine orders of magnitude less than the initial energy released through fission), and then permanently stored in a nuclear repository. [5,13].

of fuel rods, surrounded by a hexagonal graphite casing, which serves as a neutron reflector (see Fig. 1). The core will be approximately 26 feet tall and contain hundreds of fuel blocks, stacked on top of one another with a coolant jacket running around the outside of the core. [5] The fuel in each of the blocks will consist of graphite coated fuel particles (totaling about 10 billion pebbles for a full load) suspended in a graphite matrix. These particles act as their own mini pressure vessels, isolating fission products both during and after operation, therefore minimizing radioactive release. [5] The particles are also difficult to reprocess. A single fuel particle will contain a sphere of nuclear fuel only 650-850 microns in diameter. The 10 billion fuel pebbles will be difficult to separate from their graphite/SiC casings and fuel block assembly, thereby minimizing proliferation worries. [5] The fuel column will be built inside of a ceramic core. The ceramic structure will be able to withstand higher temperatures than the casing around the control rod sheaths, supposedly providing inherent safety in an accident scenario. [5]

Fig. 3: The American Society of Mechanical Engineers (ASME) provides maximum stress levels for different types of steels at different temperatures for use in pressure vessels. Note that the VHTR operates well beyond the 600°C point. After reference [15].

The coolant will be helium, which is attractive due to its negligible neutron absorption cross-section (and therefore ability to remain non-radioactive), and its single phase during operation. [13] The helium will directly power a Brayton cycle for power generation applications and will use an intermediate heat exchanger (IHX) for hydrogen production applications. The IHX simplifies the hydrogen production plant, since it will not have to be built to nuclear-standards, and provides a thermal buffer between the direct helium coolant line and the sensitive chemical processes in the hydrogen production process. [5]

Technical Challenges

Many of the challenges of the VHTR have been addressed in prior plants. The Fort Saint Vrain helium cooled gas reactor operated in Colorado for 12 years, providing valuable insight into the challenges of a helium-cooled reactor. The Thorium High Temperature Reactor in Germany operated for 4 years, and Japan's High Temperature Engineering Test Reactor demonstrated 850°C outlet temperatures. The primary challenges of the VHTR are enumerated below:

As we look to the future of nuclear reactor design, we seek a technology that can both power our cities via electricity generation, but also provide an energy source for transportation fuels. The Very High Temperature Reactor marks the next evolutionary step in reactor design, capable of powering our cities with greater than 50% thermal efficiency, and producing clean-burning hydrogen via carbon-free processes. [5] It is worth noting here that the majority of documentation on the VHTR is published by generation IV supporters, likely vying for funds for reactor development. The design of the VHTR is largely based on current gas-cooled reactor designs, but materials research and hydrogen production research need to be conducted before large-scale adoption of the VHTR.

© Benjamin Kallman. The author grants permission to copy, distribute and display this work in unaltered form, with attribution to the author, for noncommercial purposes only. All other rights, including commercial rights, are reserved to the author.


[1] "A Technology Roadmap for Generation IV Nuclear Energy Systems," Generation IV Intl. Forum, GIF-002-00, December 2002.

[2] H. J. Arniker, Essentials of Nuclear Chemistry, 4th Ed. (New Age Intl.,1996), p 248.

[3] P. MacDonald et al., "Feasibility Study of Supercritical Light Water Cooled Reactors for Electricity Production," Idaho National Engineering and Environment Laboratory, INEEL/EXT-04-02530, January 2005.

[4] " Nuclear Reactor Types," Institution of Electrical Engineers, November 1995.

[5] D. Chapin, S. Kiffer and J. Nestell, "The Very High Temperature Reactor: A Technical Summary," MPR Associates Inc., June 2004.

[6] Moorman, R. "AVR Prototype Pebble Bed Reactor: A Safety Re-evaluation of Its Operation and Consequences for Future Reactors," Kerntechnik 74, No. 1-2, 8 (2009).

[7] Lower and Higher Heating Values of Gas, Liquid and Solid Fuels, in Biomass Energy Data Book, Oak Ridge National Laboratory, ORNL/TM-2011/466, September 2011.

[8] L. Barelli, et al., "Hydrogen Production Through sorption-Enhanced Steam Methane Reforming and Membrane Technology: A Review," Energy 33, 554 (2008).

[9] K. R. Schultz, L. C. Brown and G. E. Besenbruch, "Large-Scale Production of Hydrogen by Nuclear Energy for the Hydrogen Economy," General Atomics, GA-A24265, February 2003.

[10] M. G. McKellar, J. E. O'Brien and J. S. Herring, "Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types," Idaho National Laboratory, NL/EXT-07-13575, September 2007.

[11] Annual Energy Review 2011. U.S. Energy Information Administration DOE/EIA-0384(2011) , Table 4.8, p 103.

[12] M. Blesl and D. Bruchof, Syngas Production from Coal," Int. Energy Agency, May 2010.

[13] G. Roberts, Nuclear Reactor Basics and Designs for the Future," Physics 241, Stanford University, Winter 2013.

[14] "Generation IV Roadmap: Description of Candidate Gas-Cooled Reactor Systems Report," Generation IV International Forum, GIF-016-00, December 2002, p. 83.

[15] B. A. Pint, J. R. DiStefano and I. G. Wright, "Oxidation Resistance: One Barrier to Moving Beyond Ni-Base ssuperalloys," Mat. Sci. Eng. A 415. 255 (2006).