Very high temperature reactor

The Very High Temperature Reactor is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle. This reactor design envisions an outlet temperature of 1000°C. The reactor core can be either a “prismatic block” or a “pebble-bed” core. The high temperatures enable applications such as process heat or hydrogen production via the thermo-chemical sulfur-iodine cycle.

The reactors are intended for use in nuclear power plants to produce nuclear power from nuclear fuel.

History

This design was formerly known as "High Temperature Gas-cooled Reactor" or HTGR, originally developed in the 1950's. The 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). [http://www.iaea.org/inisnkm/nekr/htgr/countries/abst_iwggcr1_01.html] HTGRs have also existed in Germany, Japan and China, and are promoted in several countries by reactor designers. [ [http://www.nukeworker.com/pictures/thumbnails-485.html HTGR - High Temperature Gas-cooled Reactor _ Nuclear Pictures - NukeWorker.com ] ] More recently, this reactor design type has been updated and is now more commonly known as the Very High Temperature Reactor.

Nuclear reactor design

Neutron moderator

Some United States and Russian designs refer to a prismatic block core configuration, where hexagonal graphite blocks are stacked to fit in a circular pressure vessel. Pebble bed designs are also being studied and have been used at lower temperatures than those envisioned for the VHTR. Pebble bed designs usually have a core where the pebbles are in an annulus, and there is a graphite center spire.

Nuclear fuel

The fuel is usually referenced to be uranium dioxide in a TRISO configuration, however, uranium carbide or uranium oxycarbide are also possibilities. 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.

Coolant

Helium

This helium cooled reactor type is the dominant one being studied; its primary design uses a 600-MW thermal core with a helium outlet temperature of 1,000°C. Helium has been used in most high temperature gas reactors (HTGR) to date. Helium is an inert gas, so it will not react with any materials except through its stored heat. Additionally, exposing helium to radiation does not make it radioactive, unlike most other possible coolants.

Molten salt

The molten salt cooled variant, the LS-VHTR, similar to the advanced high temperature reactor (AHTR) design, uses a molten salt for cooling in a prismatic core. It is essentially a standard VHTR design that uses molten salt as a coolant instead of helium. The molten salt would pass through holes drilled in the graphite blocks. The LS-VHTR has many attractive features, including: the ability to work at very high temperatures (the boiling point of most molten salts being considered are >1,400°C), low pressure cooling that can be used to more easily match hydrogen production facility conditions (most thermo-chemical cycles require temperatures in excess of 750°C), better electric conversion efficiency than a helium-cooled VHTR operating at similar conditions, passive safety systems, and better retention of fission products in case an accident occurred. Because it is relatively untested, this proposed version requires somewhat more research.

Control

In the prismatic designs, control rods would be 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

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 process heat.

References

* [http://nuclear.inl.gov/gen4/vhtr.shtml Idaho National Lab VHTR website]
* [http://gif.inel.gov/roadmap/pdfs/p_grns_june_25-27_presentation_gp32-00.pdf VHTR presentation]
* [http://www.gen-4.org/Technology/systems/vhtr.htm Generation IV International Forum VHTR website]
* [http://neri.inel.gov/program_plans/pdfs/appendix_1.pdf INL VHTR workshop summary]
* [http://www.raphael-project.org The European VHTR research & development programme: RAPHAEL]

ee also

* Generation IV reactor
* Pebble bed reactor
* List of nuclear reactors
* Next Generation Nuclear Plant
* Nuclear physics
* Nuclear reactor physics

External links

* [http://www.iaea.org/inis/aws/htgr/ IAEA HTGR Knowledge Base]
* [http://www.ornl.gov/info/ornlreview/v37_1_04/article_02.shtml ORNL NGNP page]
* [http://www3.inspi.ufl.edu/icapp06/program/abstracts/6208.pdf INL Thermal-Hydraulic Analyses of the LS-VHTR]
* [http://www.lcrc.anl.gov/~talamo/alby.php Alberto Talamo Technical Articles on the GT-MHR]