GOKHAN YESILYURT
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MISCELLANEOUS
HIGH TEMPERATURE GAS COOLED REACTORS
What are they and how do they work?
The HTGR is a inherently safe, modular, underground helium-cooled nuclear reactor technology, The reactor and the nuclear heat supply system (NHSS) is comprised of three major components: the reactor, a heat transport system and a cross vessel that routes the helium between the reactor and the heat transport system. The NHSS supplies energy in the form of steam and/or high temperature fluid that can be used for the generation of high efficiency electricity and to support a wide range of industrial processes requiring large amounts of heat or steam. Development of the HTGR in the U.S. is funded by the Energy Policy Act of 2005. The program for the Next Generation Nuclear Plant (NGNP) Project is managed by the Idaho National Laboratory with funding through the Department of Energy.
The NHSS design is modular with module ratings from 200 MWt to 625 MWt, reactor outlet temperatures from 700 °C to 850 °C and heat transport systems that provide steam and/or high temperature fluids. The range of power ratings, temperatures and heat transport system configurations provides flexibility in adapting the modules to the specific application. Safety at the highest levels is designed into the HTGR. No harmful release of radioactive material under any conditions is assured by design.
Multiple assured barriers to the release of radioactive material are provided. These barriers include multiple layers of ceramic coatings on the nuclear fuel, the carbon encasement and the graphite core structure. Additional barriers include the reactor vessel and the reactor building. The high temperature and robust structural capabilities eliminate concerns of fuel damage that could lead to significant release of radioactive materials from the nuclear fuel. The ceramic coated nuclear fuel provides the primary containment for radioactive materials rather than depending on a containment building.
Reactor power levels are limited and the nuclear reactor shuts down if reactor temperatures exceed intended operating conditions. Inherent to the nuclear reactor design is suppression of the nuclear reaction if the operating temperature increases. Complete shutdown is achieved through automatic insertion of control rods into the reactor core by gravity.
No actions by plant personnel or backup systems are required to either ensure shutdown of the reactor or ensure cooling. Conversely, actions of plant personnel cannot achieve conditions that cause the reactor fuel to lose its ability to contain radioactive material.
No power and no water or other cooling fluid is required. Heat removal from the reactor occurs naturally and directly to the earth if normal heat transport systems are not available.The low energy density of the reactor core combined with the large heat capacity of the graphite structure results in the reactor taking days to reach maximum temperatures (still well below temperatures that could cause fuel degradation), even if normal cooling systems are not functional.
Reactor materials including the reactor fuel will not chemically react or burn to produce heat or explosive gases. Helium is inert and the fuel and materials of construction of the reactor core and the nuclear heat supply system are chosen to preclude such reactions.
Intrusion of water or air into the reactor systems does not result in substantive degradation of the capability to contain radioactive materials and maintain a shutdown condition. The presence of water will enhance the heat removal path.
Spent or used fuel is stored in casks or tanks in underground dry vaults that can be cooled by natural circulation of air and shielded by steel plugs and concrete structure. No water is required for either cooling or radiation shielding and no active cooling system is required.
The VHTR is a type of high-temperature reactor that conceptually can reach higher outlet temperatures (up to 1000 °C); however, in practice the term “VHTR” is usually thought of as a gas-cooled reactor, and commonly used interchangeably with “HTGR” (high-temperature gas-cooled reactor).
There are two main types of HTGRs: pebble bed reactors (PBR) and prismatic block reactors (PMR).The prismatic block reactor refers to a prismatic block core configuration, in which hexagonal graphite blocks are stacked to fit in a cylindrical pressure vessel. The pebble bed reactor (PBR) design consists of fuel in the form of pebbles, stacked together in a cylindrical pressure vessel, like a gum-ball machine. Both reactors may have the fuel stacked in an annulus region with a graphite center spire, depending on the design and desired reactor power.
History
The HTGR design was first proposed by the staff of the Power Pile Division of the Clinton Laboratories (known now as Oak Ridge National Laboratory) in 1947. Professor Dr. Rudolf Schulten in Germany also played a role in development during the 1950s. The Peach Bottom 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). HTGRs have also existed in the United Kingdom (the Dragon reactor) and Germany (AVR reactor and THTR-300), and currently exist in Japan (the HTTR 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 HTR-PM, each with 100 – 195 MWe of electrical production capacity are under construction in China as of November 2009, and are promoted in several countries by reactor designers.
Nuclear reactor design
Neutron moderator
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
The fuel used in HTGRs is coated fuel particles, such as TRISO 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. 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 concept conceived at Argonne National Laboratory has been used to better manage the excess of reactivity.
Coolant
Helium has been the coolant used in most HTGRs to date, and the peak temperature and power depend on the reactor design. Helium is an inert gas, so it will generally not chemically react with any material. Additionally, exposing helium to neutron radiation does not make it radioactive, unlike most other possible coolants.
The molten salt cooled variant, the LS-VHTR, similar to the advanced high-temperature reactor (AHTR) design, uses a liquid fluoride salt for cooling in a pebble core. It shares many features with a standard VHTR design, but uses molten salt as a coolant instead of helium. The pebble fuel floats in the salt, and thus pebbles are injected into the coolant flow to be carried to the bottom of the pebble bed, and are removed from the top of the bed for recirculation. The LS-VHTR has many attractive features, including: the ability to work at high temperatures (the boiling point of most molten salts being considered are > 1,400°C), low-pressure operation, high power density, 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.
Control
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.
Materials challenges
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. Reactor is designed for 60 years of service.
Pebble-bed reactor
The pebble-bed reactor (PBR) is a design for a graphite-moderated, gas-cooled nuclear reactor. It is a type of very-high-temperature reactor (VHTR), one of the six classes of nuclear reactors in the Generation IV initiative. The basic design of pebble-bed reactors features spherical fuel elements called pebbles. These tennis ball-sized pebbles are made of pyrolytic graphite (which acts as the moderator), and they contain thousands of micro-fuel particles called TRISO particles. These TRISO fuel particles consist of a fissile material (such as U235) surrounded by a coated ceramic layer of silicon carbide for structural integrity and fission product containment. In the PBR, thousands of pebbles are amassed to create a reactor core, and are cooled by a gas, such as helium, nitrogen or carbon dioxide, that does not react chemically with the fuel elements.
This type of reactor is claimed to be passively safe; that is, it removes the need for redundant, active safety systems. Because the reactor is designed to handle high temperatures, it can cool by natural circulation and still survive in accident scenarios, which may raise the temperature of the reactor to 1,600 °C. Because of its design, its high temperatures allow higher thermal efficiencies than possible in traditional nuclear power plants (up to 50%) and has the additional feature that the gases do not dissolve contaminants or absorb neutrons as water does, so the core has less in the way of radioactive fluids.
The concept was first suggested by Farrington Daniels in the 1940s, said to have been inspired by the innovative design of the benghazi burner by British desert troops in WWII, but commercial development did not take place until the 1960s in the German AVR reactor by Rudolf Schulten. This system was plagued with problems and political and economic decisions were made to abandon the technology. The AVR design was licensed to South Africa as the PBMR and China as the HTR-10, the latter currently the only such design operational. In various forms, other designs are under development by MIT, University of California at Berkeley, General Atomics (U.S.), the Dutch company Romawa B.V.,Adams Atomic Engines, and Idaho National Laboratory.
Design
A pebble-bed power plant combines a gas-cooled core and a novel packaging of the fuel that dramatically reduces complexity while improving safety.
The uranium, thorium or plutonium nuclear fuels are in the form of a ceramic (usually oxides or carbides) contained within spherical pebbles a little smaller than the size of a tennis ball and made of pyrolytic graphite, which acts as the primary neutron moderator. The pebble design is relatively simple, with each sphere consisting of the nuclear fuel, fission product barrier, and moderator (which in a traditional water reactor would all be different parts). Simply piling enough pebbles together in a critical geometry will allow for criticality.
The pebbles are held in a vessel, and an inert gas (such as helium, nitrogen or carbon dioxide) circulates through the spaces between the fuel pebbles to carry heat away from the reactor. Pebble-bed reactors need fire-prevention features to keep the graphite of the pebbles from burning in the presence of air if the reactor wall is breached, although the flammability of the pebbles is disputed. Ideally, the heated gas is run directly through a turbine. However, if the gas from the primary coolant can be made radioactive by the neutrons in the reactor, or a fuel defect could still contaminate the power production equipment, it may be brought instead to a heat exchanger where it heats another gas or produces steam. The exhaust of the turbine is quite warm and may be used to warm buildings or chemical plants, or even run another heat engine.
Much of the cost of a conventional, water-cooled nuclear power plant is due to cooling system complexity. These are part of the safety of the overall design, and thus require extensive safety systems and redundant backups. A water-cooled reactor is generally dwarfed by the cooling systems attached to it. Additional issues are that the core irradiates the water with neutrons causing the water and impurities dissolved in it to become radioactive and that the high-pressure piping in the primary side becomes embrittled and requires continual inspection and eventual replacement.
In contrast, a pebble-bed reactor is gas-cooled, sometimes at low pressures. The spaces between the pebbles form the “piping” in the core. Since there is no piping in the core and the coolant contains no hydrogen, embrittlement is not a failure concern. The preferred gas, helium, does not easily absorb neutrons or impurities. Therefore, compared to water, it is both more efficient and less likely to become radioactive.
Safety features
An advantage of the pebble-bed reactor over a conventional light-water reactor is in operating at higher temperatures. A technical advantage is that some designs are throttled by temperature, not by control rods. The reactor can be simpler because it does not need to operate well at the varying neutron profiles caused by partially withdrawn control rods.
Pebble-bed reactors are also capable of using fuel pebbles made from different fuels in the same basic design of reactor (though perhaps not at the same time). Proponents claim that some kinds of pebble-bed reactors should be able to use thorium, plutonium and natural unenriched uranium, as well as the customary enriched uranium. There is a project in progress to develop pebbles and reactors that use MOX fuel, that mixes uranium with plutonium from either reprocessed fuel rods or decommissioned nuclear weapons.
In most stationary pebble-bed reactor designs, fuel replacement is continuous. Instead of shutting down for weeks to replace fuel rods, pebbles are placed in a bin-shaped reactor. A pebble is recycled from the bottom to the top about ten times over a few years, and tested each time it is removed. When it is expended, it is removed to the nuclear waste area, and a new pebble inserted.
When the nuclear fuel increases in temperature, the rapid motion of the atoms in the fuel causes an effect known as Doppler broadening. The fuel then sees a wider range of relative neutron speeds. Uranium-238, which forms the bulk of the uranium in the reactor, is much more likely to absorb fast or epithermal neutrons at higher temperatures. This reduces the number of neutrons available to cause fission, and reduces the power of the reactor. Doppler broadening therefore creates a negative feedback because as fuel temperature increases, reactor power decreases. All reactors have reactivity feedback mechanisms, but the pebble-bed reactor is designed so that this effect is very strong. Also, it is automatic and does not depend on any kind of machinery or moving parts. If the rate of fission increases, temperature will increase and Doppler broadening will occur, decreasing the rate of fission. This creates passive cooling.
Because of this, and because the pebble-bed reactor is designed for higher temperatures, the reactor will passively reduce to a safe power level in an accident scenario. This is the main passive safety feature of the pebble-bed reactor, and it makes the pebble-bed design (as well as most other very-high-temperature reactors) unique from conventional light water reactors which require active safety controls.
The reactor is cooled by an inert, fireproof gas, so it cannot have a steam explosion as a light-water reactor can. The coolant has no phase transitions—it starts as a gas and remains a gas. Similarly, the moderator is solid carbon; it does not act as a coolant, move, or have phase transitions (i.e., between liquid and gas) as the light water in conventional reactors does.
A pebble-bed reactor thus can have all of its supporting machinery fail, and the reactor will not crack, melt, explode or spew hazardous wastes. It simply goes up to a designed “idle” temperature, and stays there. In that state, the reactor vessel radiates heat, but the vessel and fuel spheres remain intact and undamaged. The machinery can be repaired or the fuel can be removed. These safety features were tested (and filmed) with the German AVR reactor. All the control rods were removed, and the coolant flow was halted. Afterward, the fuel balls were sampled and examined for damage and there was none.
PBRs are intentionally operated above the 250 °C annealing temperature of graphite, so that Wigner energy is not accumulated. This solves a problem discovered in an infamous accident, the Windscale fire. One of the reactors at the Windscale site in England (not a PBR) caught fire because of the release of energy stored as crystalline dislocations (Wigner energy) in the graphite. The dislocations are caused by neutron passage through the graphite. At Windscale, a program of regular annealing was put in place to release accumulated Wigner energy, but since the effect was not anticipated during the construction of the reactor, and since the reactor was cooled by ordinary air in an open cycle, the process could not be reliably controlled, and led to a fire. The 2nd generation of UK gas-cooled reactors, the AGRs, also operate above the annealing temperature of graphite.
Berkeley professor Richard A. Muller has called pebble-bed reactors “in every way … safer than the present nuclear reactors”.
Containment
Most pebble-bed reactor designs contain many reinforcing levels of containment to prevent contact between the radioactive materials and the biosphere.
Most reactor systems are enclosed in a containment building designed to resist aircraft crashes and earthquakes.
The reactor itself is usually in a two-meter-thick-walled room with doors that can be closed, and cooling plenums that can be filled from any water source.
The reactor vessel is usually sealed.
Each pebble, within the vessel, is a 60 millimetres (2.4 in) hollow sphere of pyrolytic graphite.
A wrapping of fireproof silicon carbide
Low density porous pyrolytic carbon, high density nonporous pyrolytic carbon
The fission fuel is in the form of metal oxides or carbides
Pyrolytic graphite is the main structural material in these pebbles. It sublimates at 4000 °C, more than twice the design temperature of most reactors. It slows neutrons very effectively, is strong, inexpensive, and has a long history of use in reactors and other very high temperature applications. For example, pyrolytic graphite is also used, unreinforced, to construct missile reentry nose-cones and large solid rocket nozzles. Its strength and hardness come from anisotropic crystals of carbon.
Pyrolytic carbon can burn in air when the reaction is catalyzed by a hydroxyl radical (e.g., from water). Infamous examples include the accidents at Windscale and Chernobyl—both graphite-moderated reactors. However, all pebble-bed reactors are cooled by inert gases to prevent fire. All pebble designs also have at least one layer of silicon carbide that serves as a fire break as well as a seal.
Production of fuel
All kernels are precipitated from a sol-gel, then washed, dried and calcined. U.S. kernels use uranium carbide, while German (AVR) kernels use uranium dioxide. German produced fuel-pebbles release about three orders of magnitude (1000 times) less radioactive gas than the U.S. equivalents, due to these different construction methods.
Germany (AVR)
A 15 MWe demonstration reactor, Arbeitsgemeinschaft Versuchsreaktor (AVR translates to experimental reactor consortium), was built at the Jülich Research Centre in Jülich, West Germany. The goal was to gain operational experience with a high-temperature gas-cooled reactor. The unit’s first criticality was on August 26, 1966. The facility ran successfully for 21 years, and was decommissioned on December 1, 1988, in the wake of the Chernobyl disaster and operational problems. During removal of the fuel elements it became obvious that the neutron reflector under the pebble-bed core had cracked during operation. Some hundred fuel elements remained stuck in the crack. During this examination it became also obvious that the AVR is the most heavily beta-contaminated (strontium-90) nuclear installation worldwide and that this contamination is present in the worst form, as dust. In 1978, the AVR suffered from a water/steam ingress accident of 30 metric tons, which led to contamination of soil and groundwater by strontium-90 and by tritium. The leak in the steam generator, leading to this accident, was probably caused by too high core temperatures (see criticism section). A re-examination of this accident was announced by the local government in July, 2010.
The AVR was originally designed to breed uranium-233 from thorium-232. Thorium-232 is about 400 times as abundant in the Earth’s crust as uranium-235, and an effective thorium breeder reactor is therefore considered valuable technology. However, the fuel design of the AVR contained the fuel so well that the transmuted fuels were uneconomic to extract—it was cheaper to simply use natural uranium isotopes.
The AVR used helium coolant. Helium has a low neutron cross-section. Since few neutrons are absorbed, the coolant remains less radioactive. In fact, it is practical to route the primary coolant directly to power generation turbines. Even though the power generation used primary coolant, it is reported that the AVR exposed its personnel to less than 1/5 as much radiation as a typical light water reactor.
The localized fuel temperature instabilities mentioned above in the criticism section resulted in a heavy contamination of the whole vessel by Cs-137 and Sr-90. Some contamination was also found in soil/groundwater under the reactor, as the German government confirmed in January, 2010. Thus the reactor vessel was filled with light concrete in order to fix the radioactive dust and in 2012 the reactor vessel of 2100 metric tons will be airlifted to an intermediate storage. There exists currently no dismantling method for the AVR vessel, but it is planned to develop some procedure during the next 60 years and to start with vessel dismantling at the end of the century. In the meantime, after transport of the AVR vessel into the intermediate storage, the reactor buildings will be dismantled and soil and groundwater will be decontaminated. AVR dismantling costs will exceed its construction costs by far. In August 2010, the German government published a new cost estimate for AVR dismantling, however without consideration of the vessel dismantling: An amount of 600 million € ( $750 million) is now expected (200 million € more than in an estimate of 2006), which corresponds to 0.4 € ($0.55) per kWh of electricity generated by the AVR. Consideration of the unresolved problem of vessel dismantling is supposed to increase the total dismantling costs to more than 1 bn €. Construction costs of AVR were 115 million Deutschmark (1966), corresponding to a 2010 value of 180 million €. A separate containment was erected for dismantling purposes.
Germany (THTR)
In general, thorium does not have quite enough neutrons to achieve a self-sustaining nuclear reaction. Small amounts of neutron rich material such as Uranium 235, or so called nuclear “waste,” would be specially processed together with thorium to supply the extra needed level of neutrons. Such nuclear “waste” from atomic weapons decommissioning and atomic power plants has long been stockpiled in the U.S. specifically with the intention of recycling it some day to generate safer and cheap electricity in “thorium” powered atomic power plants; thus reducing the stockpile of nuclear “waste,” while taking advantage of vast deposits of thorium throughout the Earth’s outer crust. The Traveling wave reactor endorsed by TerraPower, is perhaps the easiest to understand type of thorium reactor.
Following the experience with AVR, a full scale power station (the thorium high-temperature reactor or THTR-300 rated at 300 MW) was constructed, dedicated to using thorium as the fuel. THTR-300 suffered a number of technical difficulties, and owing to these and political events in Germany, was closed after only four years of operation. One cause for the closing was an accident on 4 May 1986 with a limited release of the radioactive inventory into the environment. Although the radiological impact of this accident remained small, it is of major relevance for PBR history. The release of radioactive dust was caused by a human error during a blockage of pebbles in a pipe. Trying to restart the pebbles’ movement by increasing gas flow led to stirring up of dust, always present in PBRs, which was then released, radioactive and unfiltered, into the environment due to an erroneously open valve.
In spite of the limited amount of radioactivity released (0.1 GBq 60Co, 137Cs, 233Pa)a commission of inquiry was appointed. The radioactivity in the vicinity of the THTR-300 was finally found to result 25% from Chernobyl and 75% from THTR-300. The handling of this minor accident severely damaged the credibility of the German pebble-bed community, and they lost a lot of support in Germany.
The overly complex design of the reactor, which is contrary to the general concept of self moderated thorium reactors designed in the U.S., also suffered from the unplanned high destruction rate of pebbles during the test series at the start up, and the resulting higher contamination of the containment structure. Pebble debris and graphite dust blocked some of the coolant channels in the bottom reflector, as was discovered during fuel removal some years after final shut-down. A failure of insulation required frequent reactor shut-downs for inspection, because the insulation could not be repaired. Further metallic components in the hot gas duct failed in September 1988, probably due to thermal fatigue induced by unexpected hot gas currents. This failure led to a long-term shut-down for inspections. In August, 1989, the THTR company almost went bankrupt, but was financially rescued by the government. Because of the unexpected high costs of THTR operation, and this accident, there was no longer any interest in THTR reactors. The government decided to terminate the THTR operation at the end of September, 1989. This particular reactor was built, despite strong criticism at the design phase. Most of those design critique’s by German physicists, and by American physicists at the National Laboratory level, went ignored until it was shut down. Nearly every problem encountered by the THTR 300 reactor were predicted by the physicists that criticized it as “overly complex.
China (HTR-PM)
China has licensed the German technology and has developed a pebble-bed reactor for power generation. The 10 megawatt prototype is called the HTR-10. It is a conventional helium-cooled, helium-turbine design. The Chinese are, as of 2015, building a 250 MW demonstration pebble-bed reactor: HTR-PM.
South Africa (PBMR)
In June 2004, it was announced that a new PBMR would be built at Koeberg, South Africa by Eskom, the government-owned electrical utility. There is opposition to the PBMR from groups such as Koeberg Alert and Earthlife Africa, the latter of which has sued Eskom to stop development of the project. In September 2009 the demonstration power plant was postponed indefinitely. In February 2010 the South African government stopped funding of the PBMR because of a lack of customers and investors. PBMR Ltd started retrenchment procedures and stated the company intends to reduce staff by 75%.
On the September 17, 2010 the South African Minister of Public Enterprises announced the closure of the PBMR. The PMBR testing facility will likely be decommissioned and placed in a “care and maintenance mode” to protect the IP and the assets.
- Gokhan YesilyurtLead Nuclear Engineer @ X Energy, LLC in Greenbelt, MD 20770