I have claimed on here that thorium, if efficiently utilized in a liquid-fluoride reactor, is an energy source of such magnitude that it is not difficult to conceive of an entire planet powered by thorium.
So let me lay out the calculations upon which I base my argument. In a fission reaction, thorium-232 (having been transmuted to uranium-233) will release roughly 190 MeV of energy per fission reaction. Assuming that the original thorium had a mass of 232 atomic units (u), then that is equivalent to 190 MeV/232 u = 820 keV/u.
How much energy is that? If converted to electricity at 50% efficiency (which can be achieved through the use of a helium gas turbine power conversion system), 820 keV/u is equivalent to 11 billion kilowatt-hours per metric ton of thorium. (Note that a billion kilowatt-hours [BKWH] is equivalent to a terawatt-hour [TWH].)
In 2003, it was estimated that the world produced 16.5 trllion kilowatt-hours of electricity. If this had all been produced by liquid-fluoride thorium reactors, this would have required 1500 metric tonnes of thorium. Future energy projections foresee electrical production reaching 21.4 trillion kilowatt-hours by 2015. To bring the entire world’s population up to the level of the average American’s electrical consumption would require 80 trillion kilowatt-hours.
Is 1500 metric tonnes a lot? Well, consider that until recently, the United States had 3216 metric tonnes of thorium nitrate in storage. Recently, this thorium was deemed worthless by the government and buried at the Nevada Test Site. Thorium is a very dense material, and 1500 metric tonnes of thorium metal would only occupy 130 cubic meters of volume, or about the volume of a room 23 ft on a side and 9 feet high.
How does this compare to conventional (solid-core uranium) nuclear? In 2002, the world’s nuclear reactors produced 2.56 trillion kilowatt-hours of electricity and consumed 67,000 metric tonnes of uranium to do it. That is equivalent to only 0.038 billion kilowatt-hours per metric tonne of uranium. Why such a disparity between thorium at 11 TWH/MT and uranium at 0.038 TWH/MT? Because today’s nuclear only utilizes the small (0.7%) amount of uranium that is fissile (U-235), and only ends up using about half of that. Light-water reactors burning scarce U-235 and converting it to electricity at 33% efficiency are just less efficient than liquid-fluoride thorium reactors at 50% efficiency.
It is worth considering for a moment that the thorium required to fuel the entire world’s electrical needs would fit in a reasonably sized room, and the thorium required would only be about 2% of the mass of uranium mined today.
Recent statements at the federal level indicate that the concept of “reprocessing” nuclear fuel is ready to be looked at again. This controversial decision was originally made by President Ford in the twilight days of his administation in October 1976, and then reiterated by President Carter shortly thereafter–that civilian reprocessing of nuclear fuel was to be ended.
A recent article about controversies over reprocessing of spent nuclear fuel.
The favored technique for reprocessing spent solid uranium oxide fuel is called PUREX (plutonium-uranium-extraction) and flowcharts of this process can be found here, here, here, and an image of the PUREX plant at Hanford, Washington.
On one hand, advocates of reprocessing will point to the fact that there is still significant amounts of unburned U-235 and Pu-239 remaining in spent fuel, and reprocessing could extract these unburned actinides and use them again. They also often point to a future of fast-breeder reactors where the U-238 could even be consumed (by conversion to Pu-239). Lately, another driving factor for reprocessing advocates has been the dim prospect of another federal repository beyond Yucca Mountain, based on the difficulties with opening that site.
Foes of reprocessing come in a lot of stripes. Some are simply against nuclear power in all forms, others do not think reprocessing makes economic sense. One of the basic arguments against reprocessing (that holds some water, in my opinion) is the fact that reprocessing of spent nuclear fuel involves conversion of the solid fuel rod into a liquid stream. The fuel must be chopped up, declad, dissolved by nitric acid, and then go through a complicated series of steps to separate actinides (uranium, neptunium, plutonium) from fission products. Then further steps separate the actinides from one another. At the end of this separation, the liquid stream must be reconverted back to solid fuel again. But that solid fuel is significantly less desirable to a reactor operator because it is much more radioactive than fresh, unirradiated fuel.So none of the reactors in our country utilized recycled fuel. And as I learned when I studied reactor theory, using recycled fuel is more difficult from an analytical prediction perspective because it consists of a melange of different isotopes (U-235, U-236, Np-237, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, etc.) that each have a different nuclear response (especially in resolving absorption resonances). Contrast that with fresh fuel which consists of only two isotopes (U-235 and U-238). It is not a great surprise to me that this “mixed oxide” fuel is not widely used.
And even if it was, to what end? Pu-239 cannot sustain conversion of U-238 in a thermal specturm reactor, so using recycled fuel in thermal reactors doesn’t get you very far. It might take three reactors to create enough fuel to be recycled to fuel one reactor, and less and less each generation. It’s a losing game (which is why the nuclear industry is always trying to tell you about the brave new world of fast-spectrum reactors).
Liquid-fluoride reactors running on thorium are something else entirely–they can sustain conversion of thorium to fissile U-233 in a thermal-spectrum reactor. They are already in liquid form, and reprocessing of fuel consists of two easy steps (fluorination and distillation) that are done right there near the reactor, during normal reactor operation. And the waste stream, rather than being inflated as in typical reprocessing by dissolution in acid, is actually deflated through distillation, to a concentrated mass of liquid fission-product fluorides (no actinides).
Another very real concern of conventional reprocessing is the need to transport spent fuel from the reactors to the reprocessing plant, transfer separated fuel from the reprocessing plant to the fuel fabrication plant, and transfer “recycled” fuel (which is significantly more radioactive than fresh fuel) back to the reactors. This problem absolutely bedeviled early efforts at civilian reprocessing (the facilities at West Valley, NY, Morris, IL, and Barnwell, SC). This problem would be exacerbated if the reprocessing facility was located in a state that had no nuclear plants of its own, because politics and NIMBYism would inevitably get involved.
Reprocessing liquid-fluoride fuel is so simple that it would easily be located at the plant site and operate while the reactor was online. No need to transport fuel and no need to fabricate new fuel. Further, since fluorination and distillation are essentially impervious to radiation, there is no need to “cool” spent fuel like in the PUREX process (where high gamma radiation can disrupt the organic solvents used in the dissolution/separation).
Personally, I think reprocessing of conventional solid spent nuclear fuel doesn’t buy you much and will cost you a lot–money that would be far better spent getting liquid-fluoride reactor technology fielded.
When talking to folks about thorium, I often mention as one of the basic advantages the fact that you can “burn” thorium in a thermal spectrum reactor, and don’t need a fast spectrum reactor. Which usually elicits the question, “What the heck is a “thermal spectrum reactor” and why should I care that you can burn thorium in one?”
It’s a good question, and worth taking a little time to explain.
The basic idea behind nuclear fission is that you can use an electrically neutral particle, the neutron, to destabilize a nucleus and cause it to split. This is a big deal because it’s very difficult to get charged particles, like protons and electrons, anywhere near the nucleus–they’re repelled by electrical forces. That’s the basic reason why nuclear fusion is so difficult.
But with the neutron, it’s a different story. It just waltzes right up to a nucleus and hits it, and the nucleus never saw it coming.
Here’s an animated gif of how fission works, and a little movie too.
So you want slowed-down neutrons to maximize fission. And then from fission comes more neutrons, which continue the reaction. Well, mostly right. Actually, the neutrons borne from fission are going really fast. Really, really fast. And they have to slow down to have a good chance of causing fission. That’s where the moderator comes in.
The moderator in a nuclear reactor is the material whose job it is to slow down neutrons without absorbing them. This slowing-down is done by neutrons bouncing off the nuclei of the atoms in the moderating material. For most reactors, moderation takes place in the water that also cools the reactor. For a high-temperature reactor like the liquid-fluoride reactor, graphite (carbon) is used as the moderator.
The neutrons are born from a fission reaction, bounce around in the moderator, slow down, and then cause another fission reaction. This “bouncing-around” process is also called “thermalizing” the neutrons, because by bouncing around in the moderator, the neutrons are brought to the point where they have the same thermal energy as the surrounding material.
This graph shows how likely a fission reaction is based on the speed (kinetic energy) of the neutron that strikes the nucleus is. Cross-section is a concept that corresponds to the probability of interaction–the larger the cross-section, the more the probability of interaction. The energy of the thermalized neutron corresponds to temperature. If a neutron were at the same temperature as the room you’re in (~300 K), it would have an average energy of 0.025 eV. Not very much. If the neutron instead were at the same temperature as the hot fluoride salt in the center of a liquid-fluoride reactor (~1000 K) its average energy would be 0.086 eV. Not much more.
When neutrons are born from the fission reaction, they have energies around 2,000,000 eV, which corresponds to a temperature of 20 billion degrees! That’s much hotter than the center of the Sun! But like hot water poured into snow, when neutrons are that much hotter than their surroundings, they lose energy fast. And most all of that energy is lost through collisions with the nuclei of the moderating material.
So a “thermal-spectrum” reactor is a reactor that has been arranged in such a way so as to optimally “cool” the neutrons so they can cause fission. And as can be seen from the graph, fission is hundreds of times more likely when neutrons are “cooled” down by thermalization/moderation than when they’re “fast”.
So it’s logical to ask at this point, why would anyone want to build anything but a thermal-spectrum reactor? It would seem to have the minimum amount of fuel requirement for a reactor, and it would seem to maximize your chances of getting nuclear reactions. And indeed it does. But there is more to the story.
Uranium is an interesting substance, consisting overwhelmingly (99.3%) of an isotope, uranium-238, that is not fissile. But if uranium-238 captures a neutron it becomes plutonium-239, which is fissile. One more neutron into the plutonium and you get a fission reaction and energy. So you can imagine that it takes two neutrons to “burn” uranium-238.
But there is a very small amount of uranium (0.7%) that consists of the isotope uranium-235, which is fissile and only requires one neutron to fission. Despite constituting such a small fraction of uranium, this U-235 is where nearly all of our nuclear energy comes from today. And the fact that we are burning up this small resource is one of the basic reasons that our nuclear infrastructure is not sustainable. It’s also one of the basic reasons that today’s reactors make so much nuclear waste.
So couldn’t we just burn up the U-238 after the U-235 is gone? Well, to do that, we need to make sure that the fission of Pu-239 (which is what U-238 turns into after it absorbs a neutron) gives off at least two neutrons–one to convert a new U-238 into Pu-239, and another to fission that Pu-239. So how many neutrons does the fission of Pu-239 give off? Well, it all depends on the energy of the neutron that the Pu-239 absorbs. Here’s a graph showing the relationship.
Now this graph shows two lines. One is the line in purple that shows how many neutrons are given off from a fission in Pu-239. As you can see, it’s pretty constant across energies–nearly three neutrons emitted per fission. That seems to indicate there will be plenty of neutrons for fission, conversion, and even some to spare. But the blue line tells a different story. The blue line is the number of neutrons given off per absorption in Pu-239. Why are they different? Because Pu-239 has the unpleasant habit of sometimes just absorbing the neutron that struck it, and not fissioning. This happens more often when the neutron it absorbs is at the slowed-down, thermal energies.
The fact that plutonium-239 likes to eat thermal neutrons and not fission has tremendous implications for our energy future. At thermal neutron energies, the effective number of neutrons given off per absorption isn’t enough to sustain “burning” of U-238. You can see the line dip and weave around the magic 2.0 number at thermal energies (the energies at the left-hand side of the plot). When you account for neutron losses and a number of other things that real reactors must deal with, there’s just not enough neutrons to go around.
Here is the point where the road forks, where two paths present themselves, and one was taken, and the other effectively ignored. One path is thorium, the other path is the plutonium fast-breeder.
The path that was taken, or at the very least, the path that the nuclear community has wanted to take for the last sixty years, is the path to the plutonium fast-breeder. Confr
onted with the data that you can’t get enough neutrons from a thermal-spectrum reactor to “burn” U-238, they began to investigate what happens if you use a “fast-spectrum” reactor. At “fast” energies (the energies on the right-hand side of the plot) things start to look a lot better for plutonium. It makes significantly more neutrons per absorption than 2, and so the “burning” of U-238 looks to be quite feasible. But now you have a different problem, that of building a fast-spectrum reactor.
But before I go too far, let’s talk about the path not taken–thorium. Thorium is about three times more common than uranium and consists of only one isotope, thorium-232. It has no naturally fissile isotope like U-235, and thorium is not fissile in and of itself. But like U-238, it can be converted into a fissile isotope (U-233) by absorbing a neutron. One more neutron absorption in U-233 causes fission. So again, we ask the question, how many neutrons does the fission of U-233 give off? Is it more than 2? More to the point, is it more than 2 per absorption?
Yes, U-233 not only gives off more than two neutrons per absorption at thermal energies, it gives off significantly more than 2, which is enough to account for the inevitable losses that will occur in a real reactor. This means that a thermal-spectrum reactor can “burn” thorium in a sustained manner and doesn’t need to go to a fast-neutron spectrum. And that has tremendous advantages for safety, economy, and nuclear proliferation.
Twenty years ago, the worst nuclear power accident in history took place in the Ukraine at the Chernobyl Nuclear Power Station. The details of the accident are described extensively on a number of different websites, and I will not go through them here. But I will take the opportunity to talk about nuclear safety.
One of the basic differences between the energy in a coal or gas-fired power plant and the energy in a conventional nuclear power plant is that the nuclear power plant contains all the energy in its core that it will liberate over its lifetime. In fact, the core of a nuclear reactor contains considerably more energy than it will liberate over its operation. It is very important that the rate of this energy release be carefully controlled, or an accident can result, as was the case at Chernobyl.
Now we imagine a nuclear reactor—a unit that carries a vast amount of fissile material and we imagine and wonder if it could explode like a bomb. The quick answer is no—-at least most kinds of reactors can’t. The reason is in the design. All reactors in the United States are designed to have a “negative temperature coefficient of reactivity”, which is a complicated way to explain a simple principle. When the temperature in the reactor increases, the rate of nuclear fission in the reactor must always decrease. Conversely, when the temperature decreases, the rate of fission can increase, under the right conditions. But in all conditions, no matter what, if the temperature increases, the fission rate must decrease. This is the basis for the safe design of all nuclear reactors.
Permit me to spend a little time talking about the temperature coefficient of reactivity, and why it must always be negative. It is the single most important quantity in any reactor design, and the reason why can be understood from a thought experiment. Let us imagine, for a moment, that instead of being negative, the temperature coefficient of reactivity were positive. This would mean that if the temperature in the reactor were to increase, then the fission rate would increase. If the fission rate increased, then the reactor would generate more power and heat, which in turn would lead the temperature to rise, which in turn would lead to more fission, until boom! The reactor would either explode, like a nuclear bomb, or explosively disassemble itself. This of course is not acceptable.
So reactors must be built where the temperature coefficient of reactivity is negative. But how do we do this? Actually, it’s much easier than you think. In a convetional light-water reactor, the water both cools the nuclear fuel and moderates (slows) the neutrons, increasing the probability of fission. The temperature coefficient in these reactors is negative, because if the temperature increases, the water heats up and expands. If it expands, there’s less water in the core, which means there’s less hydrogen (bound in the water) to moderate the neutrons. The energy of the neutrons increases, and fission becomes less likely (as we can see from the earlier chart that shows the probability of fission as a function of neutron energy). Hence, the temperature coefficient controls the temperature of the reactor—too high and it brings it down, too low and it brings it up. It is a very nice operational feature—a built-in natural throttle.
What about a LFTR (Liquid Fluoride Thorium Reactor)? How does it maintain a negative temperature coefficient? There’s no water to moderate the neutrons, so how does it work? Well, assuming that the liquid-fluorides are circulating through a lattice of solid graphite (graphite is chosen as a moderator material because it is stable at high temperatures whereas water is not, although it is an inferior moderator to water), then fission will only take place when the neutrons generated from fission are moderated in the graphite. The negative temperature coefficient comes about because as the salt heats, it expands, and as it expands, there is less of it in the core. That means that the neutrons generated from fission tend to “leak” out of the core more when the salt gets hotter, which has a strong influence on the fission rate, causing it to fall, and reducing the rate of fission. So unlike the light-water reactor, which depends upon the thermal expansion of the moderator to reduce fission rate, the liquid-fluoride reactor depends upon the expansion of the fuel itself.
This actually gives the liquid-fluoride reactor a significant advantage over the light-water reactor in nuclear response. Even though the water expansion is very effective in generating a negative temperature coefficient, there is a time dependence built into using moderator expansion instead of fuel expansion—it takes time for heat generated in an “excursion” to conduct through the fuel, gas gap, and clad, to heat the moderator (water) and cause the restoring effect. In some classes of nuclear accidents, it is possible to melt the fuel so quickly that the heat hasn’t even made it outside of the fuel element to the moderator to stop the accident from occurring.
Nevertheless, in general, water-cooled and moderated reactors can rather easily achieve negative temperature coefficients of reactivity, and reactors with fluid fuels (such as the liquid-fluoride reactor) can achieve VERY strong negative temperature coefficients due to fuel expansion. But a basic assumption in both cases is that the reactor is in its most reactive configuration during operation.
At Chernobyl, through a combination of bad design, bad operation, and bad understanding of the response, the reactor got into a situation where it briefly had a positive temperature coefficient. And in a few seconds, the power generated by the reactor exponentially increased to many times its design rating, the water coolant flashed to steam, and the reactor disassembled itself in a steam explosion. Without a containment building, there was nothing to hold the force of the explosion in, and then the core began burning from unremoved decay heat.
It didn’t have to happen, and shouldn’t have, but it will always stand as a lesson of why the temperature coefficient of reactivity must always be strongly, strongly negative in a reactor.
The nuclear materials of a properly-designed, two-fluid (LFTR) Liquid Fluoride Thorium Reactor are extremely easy to process. This is because the fuel (one-fluid) and the fertile blanket (the other fluid) are already liquids. This immediately eliminates the first step in typical solid-fuel reprocessing, which is converting the fuel to a liquid.
This process is well-described in a document prepared by ORNL (Oak Ridge National Laboratory) scientists and engineers in 1967: Fuel and Blanket Processing Development for Molten-Salt Breeder Reactors (ORNL-TM-1852, PDF, 4.1M).
“One of the most attractive features of a two-region, [liquid-fluoride] reactor is the ease with which the fuel and fertile streams can be processed for removal of fission products and recovery of bred material. The fluid streams are easily removed from and returned to the reactor without disturbing operations, and the processing methods are relatively simple and straightforward. On-site processing is a primary requirement of a [thorium] reactor, which would be at an extreme disadvantage if a sizable inventory of fissile material were held up in decay cooling and transit.
“Four major operations are needed to sufficiently decontaminate the fuel stream of an [LFTR]. These are fluorination, sorption of UF6, vacuum distillation, and salt reconstitution. The fertile stream requires fluorination, sorption of UF6, and for maximum effectiveness includes protactinium removal. These operations represent the most straightforward processing for achieving a high-performance [liquid-fluoride thorium reactor]. The technology for fluorination and sorption is well-developed (through the operation of the fluoride volatility pilot plant at ORNL); the other operations have been demonstrated in small engineering experiments and/or in the laboratory. The process for each stream is capable of economically recovering more than 99.9% of the uranium, 94% or more of the LiF-BeF2 in the fuel carrier, and more than 99% of the LiF in the fertile stream.”
When I first was studying nuclear energy, it didn’t take me long to realize that what was really needed was a breakthrough in reprocessing so that thorium fuel could be used. What I didn’t realize at the time was that solid nuclear fuel would never be as easy to reprocess as liquid-fluoride fuel. I truly believe that the liquid-fluoride reactor is the reactor that makes the thorium option not only viable, but compelling.
When I learned that the entire National Defense Stockpile of thorium (3216 metric tonnes) was slated for burial in the Nevada desert, that was bad enough. But the destruction of our U233 really breaks my heart and hurts so much worse.
Uranium-233 is the ideal fuel to start a liquid-fluoride reactor, and there is a very little bit in the world, left over from different attempts to get a thorium-powered future going. Now the DOE is taking great pride in the fact they are going to throw it away. I can only comfort myself with the idea that if they knew how valuable this material is for starting a liquid-fluoride reactor, they would never do this.
It gets even worse–the $128 million that they plan to spend to “blend” down this little bit of U-233 could be used to progress liquid-fluoride reactors, which currently get about $40K a year under the DOE Gen-4 program. The fellow that gets the money tells me it’s enough to “answer the phone”.
And once blended with U-238, the U-233 will be unrecoverable (I’m sure this is what they want). We could not isotopically separate it like natural uranium, since it will be far, far too radioactive to introduce into a diffusion plant. So it’s gone–thrown away when it could have started a thorium reactor.
What a tragic loss and waste…
Sign the petition at White House.gov (click here)
Another semi-recent story concerns the use of thorium as a “fuel” in typical, light-water reactors. I put “fuel” in quotations because the only way to truly utilize the thorium resources of the world is to follow the three-step process I have outlined. First, convert the thorium to protactinium-233, isolate the protactinium until it decays to uranium-233, then introduce the uranium-233 back into the reactor to fission and produce the neutrons to convert additional thorium.
As has been mentioned, this process is not only possible in thermal-neutron reactors, but attractive. However, there are more and less attractive ways to do this, and a recent article Thorium Fuel for Nuclear Energy certainly falls in what I would call the latter category.
Like all other plans that involve using Thorium in a solid form, this plan misses the crucial step where protactinium is isolated from neutrons. Protactinium-233 has a large “cross-section” for thermal neutrons, meaning that it really wants to gobble one up. And if it does, it becomes protactinium-234 and then decays to uranium-234, which is not fissile and can’t continue the process.
Some solid-fueled reactors have attempted to beat this problem by reducing the neutron flux levels in the reactor to the point where Pa-233 has a chance to decay to U-233, rather than gobble up a neutron. But by reducing neutron flux, you’re also reducing the fission rate and the amount of thermal power that the reactor can generate. This means less electricity and less reason to have the reactor in the first place.
But the proposal discussed in this article doesn’t even get that far. The proposers say they want to burn weapons-grade nuclear fuel in thorium, so that as the plutonium burns, additional plutonium isn’t formed. Seemingly admirable, until you realize that again the goal isn’t sustained nuclear energy production but just the destruction of this material that we labored so hard to produce in the first place–material that could be utilized in the right reactor to produce much much more energy.
So rather than taking the short-sighted view and mixing a little thorium with your plutonium, let’s get to work on the liquid-fluoride and chloride reactors, so we can destroy the plutonium, make the U-233, and get our thorium reactors running as soon as possible.
In 2003, an article in the Bulletin of the Atomic Scientists, Between MOX and a Hard Place, described the difficulty of disposing of the vast amounts of weapons-grade plutonium and uranium left over from the Cold War. “Weapons-grade” material generally means plutonium that consists predominantly of the isotope plutonium-239, and uranium that consists predominantly of the isotope uranium-235.
The plan is to “blend” these materials with standard uranium, and form what is called in the nuclear industry “MOX”, for “mixed oxides of uranium”. Then the MOX fuel would be introduced into typical light-water nuclear reactors as a fuel. In the course of power production, the Pu-239 and U-235 would fission and release neutrons, but due to the poor performance of Pu-239 at thermal (slowed-down) neutron energies, in about a third of the instances when Pu-239 absorbed a neutron, it would not fission and rather become Pu-240, whose characteristics make it much less desirable for nuclear weapons.
While I am no fan of the uranium-plutonium fuel cycle, I recognize that there are some far better, and ironically, easier options to dispose of the plutonium fuel. If a liquid-chloride reactor, with its very fast neutron spectrum, were constructed, the plutonium could be completely consumed and the neutrons generated could be captured in a blanket of thorium tetrafluoride, leading to the production of large amounts of uranium-233, which would easily be removed from the blanket by fluorination to uranium hexafluoride. This 233UF6 could then be reduced to 233UF4, which could be used as the “start charge” for many, many liquid-fluoride reactors. Once started, these reactors would not require additional U233, but would be able to sustain themselves solely on thorium, which is abundant and inexpensive.
Much like seeds that could either be eaten and provide a moment’s nutrition, or planted and yield so much more, this Cold War plutonium, purchased at such great cost from the reactors at Hanford and Savannah River, could be used to create the U233 to start many, many liquid-fluoride reactors.
The LFTR is an innovative design for a thermal breeder reactor that was developed from the 1950s through the 1970s at ORNL Oak Ridge National Laboratory in Oak Ridge, Tennessee. The reactor utilized a fluid-fuel form, with uranium and thorium fluoride salts dissolved in a matrix of lithium and beryllium fluoride salts. The melted salt was pumped throughout the reactor vessel and generated energy in an interesting manner. As the salt passed through the “core” region of the reactor, moderation provided by solid graphite elements led to neutron thermalization and fission reactions that produced heat. Then as the salt accumulated in a plenum and was pumped out of the core, fission ended and the salt passed through an external heat exchanger where it was cooled and transferred its heat to a secondary salt, and ultimately to a working fluid.
In 1970s, a steam-Rankine cycle was the basic power conversion technique considered for the liquid-fluoride reactor. However, there were a number of problems with this approach, mainly stemming from the fact that the natural temperature range of the fluoride salt was significantly above the typical operating temperatures of steam systems used for nuclear reactors.
More recent work on the liquid-fluoride reactor has focused on using the helium-Brayton (gas-turbine) power conversion cycle for electrical generation. This cycle would offer higher conversion efficiencies (~50% efficiency) through the salt’s unique abilities to take advantage of multiple reheating steps in the Brayton cycle. The high-temperature attributes of the salt also enable other unique applications, such as the thermochemical generation of hydrogen directly from nuclear heat.
The unique attributes of the liquid-fluoride reactor are a consequence of its fuel form. Salts of fluorine and alkali metals are exceptionally stable since they are formed from the most electronegative of elements (fluorine) and the most electropositive (lithium, beryllium, sodium). Due to their exceptional chemical stability, these fluoride salts have low vapor pressures at high temperature (enabling high temperature operation at low pressure) and they do not react with air or water, unlike molten metal coolants such as sodium. The favored combination for a neutronically-efficient liquid-fluoride reactor is a combination of lithium fluoride (highly enriched in the lithium-7 isotope) and beryllium fluoride. Through a proper ratio of these two salts, a solvent with a low melting point can be constructed. The minimum melting temperature of these salts is achieved when a composition of 52 mole % LiF and 48 mole % BeF2 is used. This combination will melt at 356°C. Typical compositions of base salts that have been used in liquid-fluoride reactors are 66 mole % LiF and 34 mole % BeF2.
In the late 1940s, excitement and enthusiasm about all things “atomic” was common among military planners. Having “harnessed” the energy of the atom for nuclear weapons, naturally they began to imagine how this energy could be used to drive other military activities. About this time a young Navy captain, Hyman Rickover, was beginning to think about the possibilities of nuclear energy (reactor) for powering submarines, and the Air Force, not to be left behind, was imagining long-range bombers that could fly indefinitely, powered only by nuclear energy.
The difficulties of building a nuclear aircraft were vastly greater than building a nuclear submarine. Central among them was the need to build a reactor that could reliably provide heat at the high-temperatures needed to drive a turbojet. In a conventional turbojet engine, cold ambient air is drawn in the intake, compressed to high pressures in the compressor, and then heated to high temperature in the burner by mixing and combusting a small amount of jet fuel. The hot gas then expands through a turbine, generating the shaft power to drive the compressor, and is exhausted through the nozzle, creating thrust.
To build a nuclear-powered aircraft, the heat generated by combustion had to be replaced with heat generated by a nuclear reactor. But the typical water-cooled reactors favored for submarine proplusion could not provide heat at nearly the temperatures needed for aircraft propulsion. Beyond the high temperature requirements, the reactor needed to be extremely simple and easy to operate, since most of the crew would be occupied in flying the aircraft.
Even since the days of the Manhattan Project, some nuclear engineers had wondered if a low-melting point liquid might be a better form for nuclear fuel than a high-melting point solid. Their reasoning had mostly been centered around the ease of reprocessing a liquid-fuel form, but there were other important advantages as well. A liquid-fuel would expand when heated, creating a strong negative temperature coefficient. It could be easily drained in the event of a loss-of-coolant into a passively-cooled form. And the fuel concentration in the liquid could be altered much more readily than a solid fuel form.
Different fluid-fuels had been considered, most of them based on uranium compounds that could be dissolved in water, such as uranyl sulphate. But water-based reactors couldn’t reach the temperatures needed for aircraft propulsion, even under extreme pressure. A fluid that was stable at high temperatures was needed, and stability at high temperatures implied chemical stability. Thought was given to using hydroxides as a solvent fluid, but hydroxides had limited stability at high-temperatures and were extremely corrosive to most metal structures.
In 1951, Ray Briant was a chemist working on nuclear aircraft propulsion at Oak Ridge National Laboratory. At that time, a beryllium-oxide moderated, sodium-cooled reactor with solid fuel elements was favored, but temperatures that would be attained in the reactor (1600°F) made it difficult to conceive that the fuel elements would survive long. Briant believed that such a reactor would have fuel elements that would look like “a bunch of spaghetti”.
He tried to conceive of a reactor that could operate stably at such temperatures and naturally began to think about a fluid fuel form. Briant’s colleagues, Vince Calkins and Ed Bettis, proposed the fluorides of the alkali- and alkaline-earth metals as solvents, but the behavior of uranium fluoride in these salts was unknown. At first blush, however, the fluoride salts had many advantages. They were extremely chemically stable, and thus could attain very high temperature operation. But could they be used in a reactor?
The possibility of a high-temperature, high-power density reactor was very tempting, and so an effort to prove the concept of the liquid-fluoride reactor began. A small research reactor that was being designed for the Aircraft Nuclear Program was modified to serve as a testbed for the liquid-fluoride concept. Since blocks of beryllium oxide had already been ordered for the previously-favored concept, the decision was made to use them and flow the fluoride salt through Inconel tubes in and out of the beryllium oxide block to simulate reactor performance. Thus the Aircraft Reactor Experiment was born.
The ARE went critical for the first time on November 3, 1954 using a mixture of sodium fluoride, zirconium fluoride, and uranium tetrafluoride. It operated for a total of 100 hours at a maximum temperature of 1600°F and a maximum power of 2.5 MW (thermal). Heat generated in the fluoride salt was removed by a liquid sodium coolant loop and then dumped in an air-cooled heat exchanger. The ARE showed that not only was the UF4 chemically stable in the solvent, but also that the fission products generated by fission formed stable fluorides in the salt mixture and did not plate out on surfaces. Another surprise was that gaseous fission products were removed essentially automatically by the pumping action of the reactor, accumulating in the pump bowl above the reactor. The fluid fuel had a very strong negative temperature coefficient, and the reactor could easily be started and stopped by changing the power demand on the reactor, without control rods. Despite its success, the engineers were not anxious to run the reactor for an extended period, since the “in-and-out” tubular configuration could not drain the salt from the core in the event of an accident. After 8 days the reactor was shut down.
Flushed with success from the ARE, ORNL engineers proposed the liquid-fluoride reactor as the baseline for the Aircraft Reactor Program and it was selected. Plans were made to build a “real” liquid-fluoride reactor that would operate at 60 MWt and would be of a flight-like configuration. This reactor was to be called the Aircraft Reactor Test, but the engineers referred to it as the “Fireball”
. The Fireball was a reflector-moderated design that used the NaF-ZrF4-UF4 fuel of the ARE, but was moderated by beryllium metal and cooled by liquid sodium-potassium eutectic (NaK). The NaK was planned to carry the fission heat to the turbojet engines that would provide thrust to the aircraft in flight.
Despite the technical triumph of the liquid-fluoride reactor, the Aircraft Nuclear Program faced severe technical difficulties from the weight of radiation shielding (necessary to protect the pilot and crew) and the advent of alternative forms of nuclear weapons delivery, such as the intercontinental ballistic missile. After Kennedy took office in 1960, the ANP was quietly discontinued after the expenditure of $880 million.
ORNL interest in the liquid-fluoride reactor did not wane, however. The high-temperature performance of the reactor coupled with its neutron economy and operational stability led ORNL engineers to propose the LFR as a civilian power reactor. At first, the LFR as a converter reactor was the proposed application, but further investigation of the properties of uranium-233 led engineers to propose the LFR as a thermal breeder reactor.
1. U-233 has the highest ? (neutrons emitted per neutron absorbed) at thermal energies of any nuclide; but it was significantly less than the ? expected from Pu-239 at fast energies. Thus, exceptional neutron economy was very important, and the lack of internal components in the graphite-moderated LFR led to very high neutron economy.
2. Xe-135 could be removed continuously from the LFR, significantly improving neutron economy and eliminating the issues with xenon transients that dogged reactor startup and shutdown.
3. After Th-232 absorbs a neutron, it becomes Th-233 and then decays (with a half-life of 22.3 min) to protactinium-233, which has a half-life of 27.0 days and a sizeable thermal neutron cross-section. Once formed, Pa-233 should be sequestered from the reactor and allowed to decay to U-233; otherwise it will absorb a neutron and form Pa-234 then U-234, which is not fissile. The fluid fuel nature of the LFR allows newly formed Pa-233 to be isolated from the fuel (or blanket), allowed to decay, and then reinserted into the fuel. This remarkable process is simply not possible in a solid-fueled thorium reactor—they must rely on low neutron flux to avoid protactinium destruction, which severely penalizes performance.
4. The low breeding margin for thorium-uranium means that breeders cannot afford to waste neutrons on control rods and burnable poisons. The strong negative temperature coefficient of the LFR allows stable operation at a large variety of power settings with very little absorptive-type control.
5. Thorium forms a tetrafluoride that is stable and dissolves high concentrations in the lithium-beryllium fluoride mixtures used.
The Molten-Salt Reactor Program (MSRP) was begun at ORNL under H.G. “Mac” MacPherson in 1958 and came to include many of those who had worked on the ARE. The MSRP won permission from the AEC to build a small reactor on the condition that it have less than 10 MW of thermal power.
Design and construction of the Molten-Salt Reactor Experiment (MSRE) began in 1961. It was a “true” liquid-fluoride power reactor. It utilized a lithium7-beryllium fluoride solvent into which was dissolved zirconium and uranium tetrafluorides. The goal of thorium breeding was deferred since the favored design at the time was a two-region liquid-fluoride breeder. The MSRE was designed to simulate the “core” of that future reactor.
The MSRE went critical on June 1, 1965 and operated for 4.5 years until it was shut down in December 1969. The MSRE was the first (and probably only) reactor to operate on all three fissile fuels: U-233, U-235, and Pu-239. During its operation, uranium was completely removed from the salt through fluorination by bubbling gaseous fluorine through the salt. The fluorine caused the uranium tetrafluoride to convert to uranium hexafluoride, which is gaseous, and could then be removed. In 4 days, 218 kg of uranium was separated from the intensely radioactive fission products and its activity was reduced by a billionfold. The reactor was then loaded with U-233 that had been made by early runs of thorium fuel at the Indian Point reactor in New York.
When restarted, the MSRE was operating on U-233 and the Pu-239 that remained from the previous operation on 20% enriched uranium.
Despite its success, the AEC was heavily committed to the sodium-cooled fast breeder and the military was very interested in the high-quality, weapons-grade plutonium that would be generated by future fast breeders. The thermal-breeder operating on thorium simply could not compete on this count, and the AEC moved to cancel the MSRP in 1972. They commissioned a report (WASH-1222) that was highly critical of the liquid-fluoride reactor concept and praised the liquid-metal fast breeder. Ironically, this report omitted nearly all of the inherent safety of the liquid-fluoride reactor, its fast response to transients, its neutron economy, proliferation-resistance, and reprocessability. Instead, it focused on a few minor issues that had cropped up during MSRE operation, such as tritium generation, tellurium cracking, and graphite replacement. The program was subsequently cancelled in January 1973.
It 1974, the program was briefly restarted and solutions were pursued to tellurium-cracking and tritium isolation. These were basically solved to the satisfaction of the engineers, but a follow-on the MSRE was not approved by the AEC and the program was terminated again in 1976. The AEC’s heavy commitment to LMFBRs ended up being a great failure, with the cancellation of the Clinch River
LMFBR and the subsequent end of most LMFBR research programs around the world over the last 30 years.
In retrospect, many of the reasons that the LFR was originally terminated would be selling points for the reactor today.
1. Inherent safety. The strong negative temperature coefficient of the fluid fuel, its response to transients, the stability of fission products in the salt, and the ability to drain the core into a passively-cooled configuration have led many to conclude that the liquid-fluoride reactor is probably the safest reactor ever designed. Such issues of passive safety were not of primary concern when the LFR was compared to the LMFBR in the early 1970s. Typical passively-safe nuclear reactor designs usually involve drastic performance reductions to the reactor, such the PIUS concept where the reactor is isolated in a pool of highly borated water. The LFR does not compromise performance for safety since the safety is inherent in the fuel form.
2. High performance. The LFR can operate at the high-temperatures and low pressures needed for high-efficiency electrical production from gas turbines or high-temperature thermochemical hydrogen production. Such high temperatures were almost considered a nuisance when the LFR was coupled to a steam system in the old ORNL designs.
3. Fuel cycle. The neutron economy of the LFR allows it to breed thorium to uranium and essentially run forever. Thorium is plentiful and the resources available would fuel planetary energy production for thousands of years. The DOE recently disposed of a stockpile of 3216 metric tonnes of thorium nitrate that if burned in liquid-fluoride reactors would provide all US energy (electricity and transportation) needs for five years. Fission products can be isolated from the salt and disposed in a geological repository, where their activity would drop below background levels in ~300 years. Actinides would be retained in the core and not end up in the geological repository. The generation of trans-uranic nuclides from the thorium-uranium cycle is essentially zero.
4. Operability and reliability. The LFR can be refueled continuously and easily while online, which would improve the competitiveness of utilities by eliminating refueling shutdowns. The composition of the salt is continuously re-homogenized by pumping the salt through the core. There are no “hot channels” or local burnup in a liquid-fluoride core due to this action, and not need for fuel reshuffling. Fuel can be removed easily by draining the core. The strong negative temperature coefficient allows the reactor to “follow the load” without operator intervention, and to reduce power generation extremely rapidly in response to “loss of load” accidents.
5. Response to accidents or sabotage. A properly-designed LFR can withstand accidents of tremendous magnitude such as a breach of vessel and containment, whether intentional or accidental. If the fuel salt were inadvertently exposed to the outside environment through a combined breach of containment and vessel, the salt would freeze and occlude fission products in the salt as stable fluorides. Gaseous fission products are removed from the salt in normal operation and would not comprise much of the fission product inventory. In the event of complete power loss and no backup power or cooling, the reactor would melt a plug of frozen salt in the bottom of the reactor and drain into a passively-cooled, noncritical configuration. Thus reactor operators could conceivably turn off all power and walk away from a full-power reactor and it would passively “safe” itself without incident.