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.
As a breeder, the LFR had some distinct advantages over other breeder concepts.
1. U-233 has the highest η (neutrons emitted per neutron absorbed) at thermal energies of any nuclide; but it was significantly less than the η 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.
24 comments:
I have a few questions about these reactors.
1. Since there is essentially no excess reactivity, each time a neutron is lost, a replacement neutron must be provided from outside the Uranium/Thorium system. Aside from replacing burned Thorium, it seems like the reactor is going to need fuel shipments to keep going.
Furthermore, most Thorium reactors have expensive sounding systems for injecting excess neutrons (accelerators being the most expensive sounding suggestion I've heard so far). The numbers of neutrons needed is usually enormous -- the kind you only see in a nuclear reactor! How would the molten U/Th reactor gain excess reactivity?
2. The motivation for the expensive chemical seperation facility is unclear. Also, when you gave recovery percentages, you didn't say what happens to the stuff that's not recovered. Where does it go?
It appears you have the seperation step to reduce the Thorium stockpile on site. But if the stuff is cheap and nonradioactive, why bother? You could have a mix of Uranium and Thorium salts. The mix could spend a little time (hours?) in the high-flux region, burning some Uranium and producing some Protactinium. Then the mix might be stockpiled in a low-flux region for a few months until sufficient Pa-233 has decayed to U-233, at which point the mix can be recycled back into the high-flux region.
Admittedly, the above cycle implies a radioactive stockpile hundreds of times the mass of the core. The fluorination/distillation system you describe appears to have the advantage of reducing the size of the Thorium stockpile, but it seems it would not reduce the size of fissile stockpile. And, of course, it requires online chemical processing of very hot highly radioactive fluids, which is a very serious drawback since it raises the issue of handling a spill.
3. You don't describe how the fission products, mostly fluorides, would be seperated, except to say that Xenon comes out relatively quickly as a gas. That sounds good. Are there other physical processes for removing other fission products, especially easy ones that might take out a few products with high cross-sections? Given the difference in molar mass between the products and the fuels, and the months of time the stuff must be stockpiled, one might think that the products would simply rise to the surface after a month or two of sitting around.
Question 1:
"No excess reactivity" means that the LFR has the minimum amount of uranium-233 in solution to maintain criticality (i.e. each fission leads to another fission). Since each fission of U-233 gives off about 2.5 neutrons, where do they go? Well, you must insure that one neutron makes it to another U-233 nucleus (to maintain criticality) and a little more than one neutron makes it to a thorium nucleus (to get it converted to U-233).
There are inevitable neutron losses (leakage, absorption in moderator and fission products, etc.) but the overall goal is a "conversion ratio" of at least 1.0--sometimes called a "breakeven breeder". When you reach this state, all the thorium that goes in the system eventually becomes U-233, and all U-233 eventually gets fissioned. So zooming back, you're basically "burning" thorium (with U-233 being the intermediate state) and the only feed the reactor needs is thorium. Of course, you need your original "shot" of U-233 to get everything started...
I've never understood why Carlo Rubbia took a perfectly good liquid-fluoride reactor and complicated it so terribly by trying to add a particle accelerator, and then marketed it to the world as an "energy amplifier". All he did was reduce the energy "gain" by a few orders of magnitude and increased the costs. If safety was his argument, then he should have studied how strong the negative temperature coefficient of reactivity can be in a properly-designed liquid fluoride reactor and he would realize the accelerator was superfluous.
Question 2:
Why reprocess the salt? Two reasons--first you need to extract bred U-233 from the thorium blanket and inject into the core salt, and second, you want to get neutron-hungry fission products out of the core salt. The worst offender, by the way, is xenon-135, and it's also the easiest to get out. It's gaseous and simply comes out of solution. That means the LFR needs about 3% less reactivity right off the top in order to achieve criticality, plus it doesn't have xenon transients.
The amount of fissile material in the reprocessing system is not hundreds of times the mass of material in the core. Quite the contrary, it's probably closer to one-hundredth the mass of fissile that's in the core, which is one of the tremendous benefits of continuous reprocessing. Only solid-core reactors have to reprocess in such a heavy-handed way.
As far as mixing thorium and uranium in a single salt--that's exactly what ORNL proposed to do later in the molten-salt program, and I think it was a big mistake. You see, when you mix the thorium and uranium in a single salt, not only does reprocessing become more difficult, but you kill off a lot of your negative temperature coefficient. The reason why is the main basis of the negative coefficient is salt expansion, which reduces the amount of fissile U-233 in the core. This really kills reactivity and gives you a strong negative coefficient. But when the thorium is mixed with the uranium (fuel) salt, expansion removes thorium from the core too. To the reaction, thorium is just a big neutron-absorbing poison (but we know it will be fuel, which is why we put it in the reactor to begin with). So when salt expands through heating out of the core, uranium and thorium both go and tend to cancel out the desired effect. Sure there's less uranium fuel, but there's also less thorium "neutron poison" and the net result is a weakly negative temperature coefficient.
We want the most negative temperature coefficient we can get--no parameter is more important to nuclear safety than that number. So we want two-fluid LFRs, primarily for the safety, but also for the ease of reprocessing. ORNL could never figure out an attractive core geometry for the two-fluid LFR, but a few years ago I think I figured out the solution...
Question 3:
It's not so much that you remove fission products from the salt, it's that you remove salt from the fission products. In a two-fluid LFR, the core salt is lithium fluoride-beryllium fluoride-uranium-233 fluoride. After you "run" it for awhile, there's fission product fluorides in there too. All the gaseous fission products just come right out of the salt through pumping (xenon, krypton, etc.)
So to clean up the salt, you take out of the salt what you want to keep. Uranium is really easy to remove--just sparge the salt with fluorine gas and the uranium turns from a tetrafluoride to a hexafluoride that is gaseous, and it comes right out. Now you have barren salt (LiF-BeF2) and fission products. Distillation takes advantages of the fact that the fission products are heavier than the solvent and the solvent will vaporize at a lower temperature. So you put the barren salt in a vacuum still, crank up the heat (or just insulate it and let fission product decay heat heat it up) and the LiF and BeF2 will turn vaporous and leave the solution. Of course, it's not perfect, some LiF and BeF2 stays in the still, but most makes it out. Then you condense the LiF and BeF2, add the 233UF6 and reduce it back to 233UF4 using hydrogen, and you're ready to go again. All your fission products and a little solvent salt (LiF-BeF2) stays behind in the still and that's your waste. After 300 years it will be at background levels of radiation and you can spread it on your lawn if you want. Contrast that with the mixed fission product-transuranic waste generated by solid-core uranium-plutonium reactors, which will be dangerously radioactive for tens of thousands of years.
Heck, people live in houses older than 300 years...
Kirk, it's awfully nice of you to answer my questions. Thank you.
Re: question 1
"...each fission of U-233 gives off 2.5 neutrons..."
I have read of U/Th reactors before, and the previous discussions all talked about how the reactors had to be designed for extremely high neutron efficiency. I had presumed that was because the U/Th core doesn't have much or any excess reactivity. Is there some other reason other, solid U/Th cores, like the Shippingport design, were so careful with neutrons? In particular, is the difference that the molten salt core has a smaller inventory of poisons after a long period of operation, and the solid core has a much larger inventory, sapping it's excess reactivity?
Re: question 2
Good answer.
Re: question 3
If there are two seperate fluid systems, are there two seperate reprocessing systems? You've described the system for the core fluid. But how do you seperate the UF4 from the ThFx in the blanket fluid, so you can inject the UF4 into the core fluid?
Okay, two more questions. How much UF4 gets dissolved in the LiF-BeF2 solvent? And, what happens to all these F, Li, and Be nuclei? Do they somehow not absorb neutrons? What do they turn into?
Hi Ambivalent,
In any reactor where you intend to "burn" thorium (i.e. convert it to U-233 on a sustainable basis) you must be very careful with your neutrons. Because you get 2.5 neutrons per fission, you actually only get about 2.2 neutrons per absorption, and that's the number that really matters. You know where two of those neutrons have to go, the whole art is make sure your losses do not exceed 0.2 neutrons. In a light-water reactor like the Shippingport LWBR, that was a real challenge, since the hydrogen in light-water has a significant neutron appetite (which is why heavy-water reactors like CANDU can reduce those losses by using deuterated water, which has almost no neutron absorption).
I'm not really sure how much excess reactivity was present in the Shippingport LWBR core, because theoretically, if you're making new fuel as fast as you're using it, you wouldn't need much excess reactivity, but in the LWBR the new fuel was being generated in the solid thorium blanket rather in the core regions--you need to reprocess the fuel and refabricate new fuel rods to get the bred fuel from the blanket to the core.
When ORNL was trying to develop a liquid-fluoride thorium breeder, they anticipated a conversion ratio of 1.07, which is probably about the highest thermal breeding ratio anyone ever achieved. I am shooting for a conversion ratio of 1.0, which is just enough to keep everything going--I don't want to breed excess U-233. Instead, we can produce new U-233 in liquid chloride reactors while we're destroying transuranic waste.
As far as the "two separate reprocessing systems", that shouldn't give you too much concern, because we're talking about an extremely simple reprocessing system--nothing like the complicated stuff they have to do for solid fuel. To reprocess the core salt, you fluorinate and distill. To reprocess the blanket salt, you could simple fluorinate the bred U233 to UF6, which will then come out of solution as a gas. Or if you want even better performance, you remove the protactinium from the blanket, store it in a holdup tank for the few weeks it takes to decay to U233, then remove it by fluorination. Both fluorination and distillation are simple, radiation-hard processes based on physical chemistry. They can be done remotely and in continuous processes.
The amount of UF4 dissolved in the solvent (LiF-BeF2) is just a few percent. The Li-7, Be-9, and F-19 all have very very low neutron absorption cross-sections and are essentially impervious to the neutron flux. Occasionally, the Li-7 will take a fast neutron and turn into He-4 and tritium, which is removed by the off-gas system.
Kirk,
So if I'm getting you correctly, you want to remove the Pa-233 from the high flux region for a while so that most of it becomes U-233 instead of Pa-234 then U-234. So, how do you separate out the Pa-234? You mention in your last comment removing the Pa from the blanket, but how do you do that? Does it also form a hexafloride like Uranium does? Does thorium? Just curious--I'm one of those guys who knows just enough about nuclear processes to make a fool of himself occasionally.
~Jon
Hi Jon,
You've got the right idea--you want to get the protactinium out of the neutron flux so that it can decay to U-233. The basic advantage of chemical separation (over isotopic separation) is that thorium and protactinium are two chemically distinct elements, and you can use chemical processes to isolate them.
ORNL investigated a large number of different chemical processes to accomplish this separation. Some were more advantageous for two-fluid reactors (where there was little to no fission products in the blanket) while others were more advantageous for one-fluid reactors (where everything was mixed in there together). The favored process they had when they stopped working on LFRs in the mid-1970s was to isolate protactinium from the fluoride salt by using liquid bismuth as an extraction medium--a technique called reductive extraction.
I will go back and look at some of the documents, because they liked the reductive extraction process at the same time they liked one-fluid LFRs (which I don't like very much). I remember reading about other ideas like selective oxidation, where they put a little oxygen in the salt, which protactinium tended to mop up preferentially, then it would precipitate out as a solid oxide. I'll dig up some of the papers on the topic and send them to you if you'd like.
Kirk,
Yeah, I'd be interested in reading more about that, as that seems to be the key to getting this whole thing to work. If you can't extract the Pa-233 from the stream to sequester it and let it decay to U-233, then the only other method would be to have the whole fertile stream spend most of its time outside of the reactor in order to let the Pa-233 mutate to U-233 and get separated before going back into the reactor.
If you want to email me some documents, you can send them to jongoff (at) gmail (dot) com.
Thanks,
~Jon
Kirk, you are espousing a two-fluid LFR, where Pa-233 is chemically
extracted from the blanket fluid, allowed to decay to U-233, then
injected into the core fluid. Simultaneously, the core fluid is run
through a floridation/distillation system to extract U-233 and solvent
and concentrate the fission products away from the core.
Can you explain the problems with a one-fluid LFR, where the fluid
spends time in the core building up Pa-233, then time in holding
converting Pa-233 to U-233, then back into the core again. As I noted
above, the primary problem seems to be big holding tanks compared to
the size of the core.
But there appear to be really big advantages: no online reprocessing.
Since Krypton and Xenon seperate from the fluid physically, you have
an easy mechanism to extract some of the poisons. If you don't try to
remove the rest, and just let them build up until you lose your excess
reactivity, what happens? Does the reactor run colder over time (at
any given load level)? At what level of burnup does it grind to a
halt?
I realize I'm sort of asking the same question again. But in your
earlier response, you assumed I had misunderstood your proposal. I
haven't. I'm making a counterproposal. What's the flaw?
Because... big holding tanks don't seem like a big problem if you'd
like to have 50 years of thorium fuel on hand, because refuelling is
never going to happen, and waste product buildup isn't a big problem.
Hello Iain,
The basic reason I am so much more interested in a two-fluid LFR than a one-fluid LFR is safety. The two-fluid version will be able to achieve a very strong negative temperature coefficient, which will in turn govern the stability and response of the reactor. In a sense, both LFRs that were previously built (the ARE and MSRE) were like two-fluid LFRs in the sense that they didn't have any thorium in the "core".
On the time scale of the nuclear reaction, thorium is a "poison"--it drinks up neutrons that could otherwise cause fission reactions in the fissile fuel (be it U-233, U-235, whatever). If the thorium suddenly isn't drinking up those neutrons, then there's a lot more to cause fission, leading to more reactivity. So removing thorium from the core, somehow, is kind of like pulling a control rod out of the reactor--the fission rate will increase.
In a one-fluid LFR, when the salt expands due to heating (and I believe that the expansion wave would propagate at the speed of sound in the material, which is very fast in liquids), it would be removing not only fuel (which would reduce the rate of fission) but also thorium "poison" (which would increase the rate of fission). So these two effects would tend to cancel each other out. We really want to keep the first effect, because it leads to a strong negative temperature coefficient, and ditch the second effect, because it tends to undo that.
In a two-fluid LFR, transient heating would cause the fuel salt to expand and reduce reactivity, but the blanket thorium salt would lag in its response to the increased heating, tending to preserve the strongly negative temperature coefficient. Thus the reactor could operate without control rods (which are just big neutron sinks) and rely on the temperature coefficient for control.
It took me a long time to get used to the idea that a reactor that could operate without control rods was actually safer than one that needed them, but it can be shown to be true.
So that's the basic reason I like the two-fluid reactor. There's more reasons like ease-of-reprocessing, reduced core inventory, simplified neutronics, but safety is the biggest reason.
As far as your idea of a very long loop one-fluid reactor, apart from the safety issues I mentioned, it could work, but the real drawback would be the HUGE amount of fissile fuel you would need to keep the reactor running. You see, if you're isolating the salt to let the protactinium decay, and there U-233 in the salt, it's not doing any work (fission) while it's out of the core--it's just sitting there. So you would need a lot of fissile material to keep the thing running because only a small fraction of the fissile would be working (fissioning) at any given time.
Good thoughts! Keep em coming!
Kirk,
I hear your safety point. But it seems that once the heat has propagated from core fluid to the blanket fluid, the blanket will expand too, so you are back to the temperature reactivity coefficient of the one-fluid LFR. So that had better be safe enough, regardless of how fast the heat moves from core to blanket.
If you wanted a very strong negative reactivity, a phase change to vapor in the core would do it (a Boiling Salt Reactor?). In fact, your LFR already has that feature if it were to suffer a large positive reactivity injection, although there is the trouble that the solvent salts vaporize first, potentially concentrating the fuel that is presumably creating the excess reactivity in the first place. I guess you would then want the molten salt to exit the bottom of the reactor.
Another question about a one-fluid LFR: Is there any difference between removing the fluid from the core to allow the Pa-233 to decay, and then reinjecting that fluid, and simply having a bigger core with lower neutron flux? In other words, with the same volume of fluid and same thermal megawatt production, does it matter if we move a fission hot spot around in the fluid, or just let the whole thing bake at the same rate?
What is the minimum ratio of U-233 to Pa-233 in the core, before too many neutrons are lost to the U-234 path? I understand that in a single-fluid LFR, this same ratio exists in the decay tanks, so that if the ratio is 100:1, you'd have to keep enough U-233 in the mix to burn for 100s of months even without breeding. But what is the actual ratio?
Kirk,
Let me expand on my point preceding:
How much U-233 and total fluid do we need in a single-fluid LFR?
The thermal production rate sets the fission rate and the rate of Pa-233 production (which are about the same). Since the decay rate of Pa-233 must match the production rate, this sets the Pa-233 inventory size.
Using Kirk's earlier figures, 2 GW (thermal) plant fissions 90 g Thorium per hour. If we produce 90 g Pa-233 per hour, we must decay 90 g per hour, and if the half-life is 27 days, we'll need about 60 kg of Pa-233 on hand to see that decay rate (maybe a little less).
Now, I don't know the relative cross-sections of Pa-233 and U-233 to 1600 F neutrons, but lets set that to X. I'll guess X=100 for now. So if all the fluid in the core has the same Th-232/Pa-233/U-233 ratio, the molten salt must dissolve 6000 kg of U-233 along with 60 kg of Pa-233. At the specified burn rate, that's 7.6 years of fuel.
If the fluid is not well mixed, but divided into different streams, some with smaller concentrations of Thorium (core fluid) and a lot more with more (blanket fluid), you can arrange for X to be smaller in the blanket fluid, since it doesn't sustain a chain reaction.
If the UF4 is just a few percent (by mass, not by mol, right?), then a single-fluid 2 GWt reactor might need 265 metric tons of fluid, and would have enough Thorium on hand to operate for seven centuries, at which point the fluid will be leaking freely through corrosion holes in the reactor pipe walls. These numbers are absurb, of course. But X will not equal 100. Any idea what X is, Kirk?
I realize you don't like the single-fluid reactor, but you are really going to have to kill it dead in order to motivate online reprocessing. Frankly, mixing fluorine gas at 1600 F with radioactive anything, or precipitating stuff which is decaying into fissile material, seems incredibly scary to me. From some of the documents on your site, it sounds like just containing the molten salt for decades in metal tubing is not a solved problem. And I haven't yet found a discussion of the chemical action of the fission products on the fluid
containment.
Hi Iain,
Wow, you've got me thinking...let me go look up the thermal cross-sections of U-233 at Pa-233 and get back with you.
Until then, take a look at ORNL-4865 to learn more about fission product behavior in the reactor and ORNL-4575 to learn more about the plans to build a fluorinator. I don't know if this will alleviate your concerns about fluorination, but fluorination is done all the time with uranium in the process of preparing it for isotopic separation. It's strange to think, but uranium has very low levels of radioactivity -- even U-233. It's the fission products that have all the radioactivity. When they fluorinated the original uranium out of the MSRE, a few hours later, Alvin Weinberg held the specimen in his hand. Earlier it had been mixed with stuff that would have killed him almost instantly from gamma radiation, but once separated, it was almost harmless.
Your comments are giving me excellent insights into how the fluorination process might be percieved publicly...
"Fluoridation, Mandrake...the most monstrously conceived and dangerous commie plot we have ever had to face..."
Kirk,
In "The Design And Performance Features Of A Single-Fluid Molten-Salt Breeder Reactor", they say that the Lithium and Beryllium in the salt are fine moderators, but more moderation is needed. Why is that?
In the design presented in that paper, they have a 240 tonne graphite moderator that has to be replaced every four years. The thing that comes out is the ultimate tar baby -- literally dripping wet with molten highly radioactive salt, and swollen and distorted from radiation damage. They suggest the thing be stuck in a cask and left to cool off for a few centuries. Having this component seems like a fairly serious drawback to any LFR design, and so it seems one might be willing to put up with some other drawback, like needing a larger core holding more dilute Uranium, if one could get rid of the tar baby.
The paper was pretty interesting. It's fascinating how attitudes have changed. Nowadays the reactor has to be safe with all equipment failed and no intervention of any sort. Back then, they seem to have been comfortable with requirements that at least some pumps work, as well as radiators, etc.
It seems to me that a very interesting design point would be one where the entire high-flux portion of the core is liquid fuel. If no structural, containment, or control bits are in the high-flux region, the problem of reactor longevity is greatly alleviated. To get this to work, you need the chain-reaction sustaining fluid to be surrounded by something nonstructural that will not sustain the chain reaction. One wonders what maintains the boundary between the two... perhaps it is a liquid to solid transition, with the blanket cooled to freezing by steam generators or helium heat exchangers or somesuch.
> fluorination is done all the time with uranium in the process of preparing it for isotopic separation
Um, yeah, but that's just natural uranium... low enough radioactivity that you can touch it. Doing the same thing at very high temperatures, with a highly radioactive melt, remotely (so no one gets a dose), for decades, with equipment that has to be maintained remotely... It seems like you've taken several giant steps in a bad direction.
Not that I'm a chemical engineer, of course.
Hi Iain,
More moderator is needed because the lithium and beryllium in the salt just doesn't give enough moderation. In that same document you see that the salt cross-section was about 12% of the total cross-section--the rest was graphite. You probably also noticed that your previous idea was anticipated by the ORNL engineers in the design of the one-fluid reactor--they made a "blanket" and a "core" out of the same salt by just changing the concentration of moderator. By putting less moderator in the blanket, the salt was a lot less reactive, and the effect was like having a lower uranium fraction in the salt, and favoring absorption by thorium instead of U-233.
Graphite moderator was the life limiter in the design you read about. They were going for a high core flux, so as to improve the economic performance of the reactor--mind you they were not only trying to "burn" thorium, but to produce excess U-233 to start other LFRs! That's a tall order.
I'm just shooting for an LFR that can "burn" thorium--in other words, achieve a conversion ratio of 1.0. We'll get the "start charge" of U-233 from somewhere else. What is the minimum amount of processing you need to do to achieve a CR of 1.0?
Not sure--but we can imagine dropping steps to achieve greater simplicity at the cost of CR. We could skip distillation and fluorination of the core salt, leaving fission products in the core salt, and that will reduce CR, because some of the fission products will absorb neutrons. We could then just go to a "batch" scheme for core salt reprocessing, probably once every 3-5 years.
That would leave only the fluorination and protactinium extraction system for the blanket, which operates at orders-of-magnitude lower radioactivity that the core salt (no fission products). If we skip the Pa removal step and just fluorinate blanket salt then we accept some losses of U-233 due to neutron absorption in Pa-233, but it gets simpler. The simplest system we could imagine that still "burns" thorium would only have a blanket fluorinator--could we achieve a CR of 1.0 in that configuration?
Not sure...
Kirk,
Do I understand correctly that varying moderator can change the relative absorptivity of U-233 and Th-232? Does it also vary the relative absorptivity of U-233 and Pa-233? If so, does that let us reduce the fissile load the reactor has to carry around?
The simplest system I can envision has no blanket fluorinator either -- the blanket fluid is removed from the flux to decay, and then cycled back in again, until the U-233 proportion has gotten so large that it's nearly critical even without moderator. This fluid is then mixed into the core fluid to add reactivity.
Ideally the core fluid has a fluid moderator mixed in, that the blanket fluid does not have.
Kirk and Iain,
Hope this get through to you both. Why has this discussion just come to an end? Did "Men in Black" get to both of you?
I remember hearing about this some 15 years ago based on research some team at the University of South Australia were doing (could be wrong), and just how small these reactors could be (about the size of a refrigerator) and how they eat radioactive and how they could be used as a nuclear incinerator for destroying pesticide and toxic waste products.
Then everything went quite. I've been looking for information once a year ever since, and this is the first time I've struck gold.
In the current climate (no pun intended) of need for greenhouse clean power systems, why hasn't this been put on the forefront of every political policy.
We have an election coming up here in Australia in Nov 07, as do America in 08, and 6 months ago, our Government was advocating a pro-fast breeder solid core nuclear policy of 25 reactors. They have since gone very quite on the idea, instead advocating the totally unproven and extremely expensive concept of "Clean Coal" and CO2 substration as an answer to our dirty power needs.
Our main opposition are totally against nuclear poser, even though we are the uranium producers in the world, and have the largest stocks of unmined uranium.
Then there are the Greens, (similar to the European Greens parties) that have an environmental agenda for political policy, who are even anti-uranium mining.
This technology sits right in the middle of all of them. It's also a solution to the Global terrorism, and the Korean and Iranian problem if you believe that there nuclear industry isn't weapons based as they and the UNNC say.
It is brilliant in it's simplicity, it's ecology and economic viability, and should be shoved under the nose of everybody who can bring influence to bear on climate change or energy policy.
I would like to start getting this blog out into the social network, and get some support... which would make you (esp. Kirk) the center of attention. Especially people like:
Dr. David Suzuki, http://www.davidsuzuki.org/About_us/Contact_us.asp
Dr. Tim Flannery
http://www.theweathermakers.com/about/
Al Gore
http://algoresupportcenter.com/contactal.html
Dr Helen Caldicott
http://www.helencaldicott.com/index.htm
I can go on...
We need these guys to advocate that this will work or not.
The systems will eventually have to be picked up by a General Electric, or Westinghouse to build the them, but they aren't going to move until somebody makes them.
Couple of things I'd like cleared up.
Oak Ridge National Laboratory
"take a look at ORNL-4865 to learn more about fission product behavior in the reactor and ORNL-4575". Where can we take a look at these? There is references up on the www.osti.gov, but it isn't clear how you can get a copy of the reports, and under what statutes the information is available.
"Fluoridation, Mandrake...the most monstrously conceived and dangerous commie plot we have ever had to face..." where is this quote from, because this guy's an idiot.
"The Design And Performance Features Of A Single-Fluid Molten-Salt Breeder Reactor" http://www.osti.gov/energycitations/product.biblio.jsp?osti_id=4710728
you need to include a bibliography in this blog! Where is R. C. Robertson these days if he is still alive?
Hi Luke,
Most of our discussion these days is going on over in the discussion forum.
The documents you're looking for should be in the document repository.
And the quote's from the movie "Dr. Strangelove". If you haven't seen it you've got to rent it!
I'm preparing a presentation on new nuclear power technologies for my class, rethinkingnuclearpower.googlepages.com. I am covering PBMR, GT-MHR, 4S,and the 6 candidate Gen IV technologies. I'd like to do a few slides on thorium molten salt reactors. I've read some of your materials, but I wonder if you could clarify some questions that I may be asked?
1. Why isn't thorium fuel considered in the Gen IV MSR?
2. What's the function of the high purity Li-7?
3. What's the function of the Zr?
4. Where does the removed Xenon go?
5. Some U-234 will be formed before the Pa-233 escapes the neutron flux; will U-234 keep accumulating in the fluid?
6. Won't some actinides be formed? Do they just accumulate?
7. So how often do you have to shut down and recharge the MSR to get rid of U-234, actinides, and such?
Hello Dr. Hargraves,
I'll try to do my best to answer your questions:
1. Why isn't thorium fuel considered in the Gen IV MSR?
I think the main reason that thorium fuel isn't considered extensively for Gen-4 is because the overwhelming paradigm in the nuclear community is solid fuel, and even more specifically solid oxide fuel. Thorium can be used in solid oxide form, but it has some serious disadvantages in this configuration.
As a fluoride however, especially when the thorium fluoride is kept separate from its conversion-product uranium-233, thorium has some fantastic advantages over uranium.
This point is poorly understood within the nuclear community, and so with their fixation on solid fuel I think they tend to overlook what thorium can really do.
2. What's the function of the high purity Li-7?
Thorium fluoride and uranium fluoride by themselves have melting points too high to be used effectively. The melting points of these fluorides can be reduced tremendously by mixing them with a fluoride "solvent". This solvent needs to have attractive nuclear properties of its own. A mixture of lithium and beryllium fluorides can be an excellent solvent if one is careful to insure that the lithium is almost completely the more abundant lithium-7 isotope.
This is because lithium-6 (the rarer form of natural lithium) has a strong tendency to absorb thermal neutrons, whereas lithium-7 has almost no such tendency.
3. What's the function of the Zr?
In the Molten-Salt Reactor Experiment, zirconium served as an oxygen "getter" in the event oxygen made its way to the fuel. Oxygen was a concern because, although uranium tetrafluoride is more chemically stable than uranium oxide, there is some equilibrium amount of uranium oxide that will form in uranium tetrafluoride when exposed to oxygen.
This uranium oxide, being a solid, could precipitate in unexpected places in the fuel lines with troublesome results over long-duration operation.
The formation of zirconium oxide is chemically favored over the formation of uranium oxide, so any oxygen that makes its way to the salt is "gettered" by the zirconium before it can form uranium oxide.
Since then, however, ORNL chemists have devised ways to prevent uranium oxide formation without resulting to zirconium inclusion in the fuel salt.
4. Where does the removed Xenon go?
To a decay tank where it decays rather rapidly (in about a week) to cesium.
5. Some U-234 will be formed before the Pa-233 escapes the neutron flux; will U-234 keep accumulating in the fluid?
It reaches an equilibrium condition where it is being destroyed as fast as it is being created. The U-234 will eventually absorb a neutron forming U-235, which is fissile and will usually fission upon absorbing another neutron.
6. Won't some actinides be formed? Do they just accumulate?
Assuming the reactor has been started with a pure thorium/U233 blanket/fuel combination, the highest-mass actinide that will form will be neptunium-237, which is chemically distinct from uranium and can be removed easily through fluorination (to NpF6) and distillation. Depending on flux, neutronics, and the fluorination system of the reactor, probably much less than 1% of the original thorium will ever make it this far.
7. So how often do you have to shut down and recharge the MSR to get rid of U-234, actinides, and such?
You don't ever get "rid" of the actinides in a thorium/U233 MSR--you just keep burning them essentially forever. When you ultimately shut one reactor down, you simply take its fuel load and transfer it to the next fluoride reactor to start up. The formation of U234 and its equilibrium conditions does not present an undue burden to the system.
Accumulation of fission products in the salt must be ultimately dealt with--there are a number of options for this, including continuous or batch reprocessing, that have different effects on the operating performance of the reactor.
Kirk Sorenson,
Thank you for your prompt and helpful response. It will help me present the thorium molten salt reactor more credibly.
The MSR described in the Gen IV roadmap, http://nuclear.inl.gov/gen4/docs/gen_iv_roadmap.pdf
, is indeed a liquid fuel reactor. I received an email from Dr Ralph Bennett at INL indicating that the MSR and LFR are the least advanced of the 6 projects. I suggest that now is a good time for you to try to influence the project direction to investigate thorium fuel as well as U and Pu.
Dr. Hargraves,
I invite you and Dr. Bennett to participate in ongoing, detailed discussions on this technology in the Thorium Discussion Forum. There are sections of that discussion that are only available to registered forum members that would shed much more light on these implementation issues.
The first, 2007, Annual Report of the International Generation IV research co-operative was published today, at
http://www.gen-4.org/PDFs/annual_report2007.pdf.
I am reading there about the status of the MSR.
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