The Molten Salt Reactor Family: Two Fluid Reactors

The two fluid Molten Salt Reactor uses separate fluids for fuel and carrier salts. In the two fluid desige, two seperate salt fluids are present in the MSR core. The first is the fuel carrier, that is it contains one or more fissionable isotope. U-233, U-239 or Pu-239. A second fluid carries a fertile material, in MSRs always Th-232. There were from the start a numbr of advantages to the two fluid approach, as David LeBlanc points out,

Advantages

* Much more practical fission product processing without thorium in fuel salt
* Have choice of Vacuum Distillation or Simplified Liquid Bismuth Extraction
* Strongly negative temperature/void fuel salt reactivity constants
* Pa removal easily avoided by simply increasing volume of blanket salt Neutron leakage near zero

There were also a number of disadvantages,

Disadvantages

* Interlacing of fuel and blanket salt within core is the “Plumbing Problem”
* Blanket salt has positive temperature/void coefficients
* Need for extra heat transfer loop for the blanket salt (5-10% of heat load)

The carrier salt is referred to as the blanket salt. David LeBlanc pointed out that one advantage of a two fluid design is that protactinium need not be removed immediately from the blanket salt, while its presence effects both the nuclear process and the chemistry of the one fluid design.

Protactinium 233 is the 27 day half life intermediate between fertile Thorium and fissile U233. The problem is that it has a moderately high cross section for absorption, such that if it stays in the reactor loop, it may capture a neutron and not become U233. The average neutron flux the Pa sees is the main factor on how many neutrons will be lost. By increasing the volume of salt that carries the Thorium, one can lower these losses and skip this processing step. However, in a Single Fluid design if one increases the salt volume (by a lower power density core etc.) then this also increases the amount of fissile U233 needed at the same time.

In a 2 Fluid design with Thorium only in the blanket salt, one can increase its volume which does not increase how much U233 is needed. This is a financial burden in terms of carrier salt and Thorium but it is by no means excessive.

Skipping the Pa removal step is important for two main reasons. First, the process must be done very quickly, on the order of processing the whole volume of carrier salt in3 to 10 days. This is an enormous technical and economic challenge. Second is that this introduces a unique proliferation risk. When U233 is produced while in the reactor, significant quantities of U232 are also present. This is highly radioactive and leads to an extremely strong gamma ray being emitted. This would make working with the material extremely difficult if not impossible and also make detection of illicit material easily detectable. However, if the Pa is removed and allowed to decay outside the reactor it produces relatively clean U233.

Despite the disadvantages of the single fluid reactor in the late 1960's ORNL leaders felt that they were in competition with Liquid Metal Fast Breeder Reactors, championed by Argonne National Laboratory. The LMFBR had some significant disadvantages, but it did feature a high breeding ratio, much higher, in fact than the breeding ratio of thermal cycle thorium breeders. Thus the slight theoretical breeding advantage of the single fluid MSR was attractive to them. Then ORNL scientists discovered

Liquid Bismuth reductive extraction process.

And although difficult, indeed quite possibly extremely difficult, it offered a route to staying in the breeder game for a little while. The breeder competition was itself evidence of the incompetence of the American Nuclear establishment. The LMFBR was a failure, that ended up costing billions of dollars only to end up being too expensive to build even a prototype. It could not compete with the MSR as far as safety or reprocessing technology. Eventually scientists at Argonne National Laboratory were to scrap the old LMFBR concept, and develop a new sodium cooled fast breeder design, the Integral Fast Reactor, that came much closer to matching many MSR advanced features. By that time the MSR was nothing more than a large series of research reports, and a shutdown prototype awaiting decommissioning.

At any rate, the ORNL two fluid MSR design from the 1960's was a very advanced reactor design that still looks very futuristic. reactor design. The design featured a 1000 MW power station, powered by 4 modular 250 MW two fluid MSRs. The motive for the modular design was not the cost advantage of factory production although ORNL researchers were no doubt aware of that. Rather, there intention was

replacement of an entire reactor vessel assembly after the core graphite received its allowable exposure to neutrons. [After] . . . about eight years of full-power operation.

ORNL 3996 stated,

An important factor in low power costs is the ability of the power plant t o maintain a high plant-availability factor. Thus design features
that can improve this factor are desirable if these features do not themselves introduce compensating disadvantages.

A four Modular Molten Salt Breeder Reactor (MMSBR) facility will clearly more reliable than a one reactor facility. If one reactor is down, 750 MWs of electrical power would still be available, while in a single reactor one GW plant the down reactor would mean that no power would be available.

Like all two fluid MSRs, the OENL Modular Molten Salt Breeder Reactor two fluid reactor design divided its core into reactive and blanket regions. A blanket suggests that the core would have an inner region for fissionable materials and an outer region for fertile materials. This was not in fact the case in the ORNL modular design. The graphite in the core was divided up into tubes and core and blanket tubes were interlaced. This feature was not viable. Graphite shrinks and then swells under neutron bombardment. This caused ORNL MSR designers one huge headache, because the core plumbing would vary in size as the graphite evolved, this would effect regional power output within the core in unpredictable ways, and introduce an unacceptable level of uncertainty into reactor operations. ORNL-4528 noted,

The major concern was whether mechanical failure of graphite tubes in the reactor core would cause the effective lifetime of the core to be significantly less than the eight years imposed by the effects of irradiation on the graphite.

ORNL-4528 added,

The change in radial dimensions presented a more difficult problem. Densification of the graphite to produce a 2.5% reduction in distance across the flats of the hexagonal tubes would cause the fraction of the cross section of the core occupied by fuel cells to decrease by 5%, and the space occupied by the blanket salt would increase correspondingly. For the reactor with an average power density of 20 kw/liter, the volume fractions in the core would change from 0.802 to 0.762 for graphite, 0.134 t o 0.127 for fuel salt, and 0.064 to 0.111 for blanket salt. Changes of equal magnitude, but opposite in direction, would occur during the expansion phase. The rates of change of dimensions would vary with local power density, so at no time during the life of a core would the volume fractions corresponding to the maximum contraction or expansion exist throughout the core. At the end of life the graphite at the center of the core would have reached its maximum volume; graphite in the regions of average power density would be about at its minimum volume, and graphite in the outer fuel cells would be about halfway into the contraction stage.

ORNL 4528 explained,

Under irradiation the isotropic graphite being con-sidered at the time of these studies would decrease in volume by 7.5% during the contraction stage and then would increase in volume by as much as 7.5% over its initial volume by the end of its useful life. These changes in volume correspond to changes in linear dimensions of M.5% over the initial dimensions and create several design problems.

The MMSBR was intended to be a true breeder and depending on core design could breed at about a 1.05 or 1.06 ratio, with a fissile inventory as low as 220 kgs, or about a ton of U-233 per GW of rated electrical output. This was spectacular, but a larger core would mean longer graphite life and would nearly double the fissile inventory of the MMSBR. In order to capture every spare neutron, 150 times as much thorium was to be pumped through the MMSR core. The fissile inventory performance of the MMSBR is quite impressive when compared to that of recent French Fast Molten Salt Reactor designs that require as much as a six times larger fissile inventory. Sodium cooled fast breeders require a fissil inventory of as much as ten times as large.

Thus we have seen that the two fluid ORNL MMSBR although promising posed some significant materials and design problems. There was no advance in two fluid MSR design for nearly 40 years, until Canadian Physicist David LeBlanc proposed a radical change in two Fluid MSR core design. The old two fluid design had a diameter of 10 feet or more, while LeBlances new design had a diamiter of one meter.

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The LeBlanc tube core represents a potential breakthrough in reactor core design. It is extremely simple and manufacturing costs would be mainly material costs. The core itself could be built in a day. It can either be graphite moderated, coolant moderated or fast.

A recent paper by Reactor researchers from the Czech Republic, Jan Frybort and Radim Vocka argued that some problems of the ORNL MMSBR could be solved, while pointing out a new and previously unrecognized safety problem. The Paper, titled Neutronic Analysis of Two-Fluid Thorium Molten Salt Reactor (See Kirk Sorensen's discussion). Kirk summarizes,

In part 4A, they found that the temperature coefficient was strongly negative and that the breeding factor was good. In part 5, they looked at changing the design to improve it, and found that by making the fuel channels bigger than the original design, they improved nearly all parameters. This is an important result, since it’s not often that you change a parameter and find improvements in nearly all outputs from that parameter.

And then notes,

In part 7 of the paper, they mention the second key issue with a two-fluid reactor–the problem of the blanket void coefficient. Since the original ORNL design had the problem, and since they modeled only parametric variations on that original design, it’s no surprise that the problem still shows up. It must be fixed, probably through a new design approach to the two-fluid reactor. I have some ideas, most all of them based around physical situations where a loss of blanket fluid leads to a loss of moderation. I anticipate that this could be done by floating moderator elements (graphite) in the blanket salt, so that as the level of the blanket salt falls, the moderation decreases more than the absorption decreases from the loss of blanket. These ideas definitely need more modeling, but I think they are essentially sound.

Would the same effect apply to David LeBlanc's two shell design? Probably not. The Blanket void, positive coefficient of reactivity problem stems from the interlacing of blanket and fissionable salts in the MMSBR. David LeBlanc does not interlace blanket and core salts, thus a blanket void would not seem to increase core neutrons. A simple solution to the problem in the ORN design would be to keep core salt tubes in the center of the tube array, while blanket salt tubes form a ring around them on the outside. I suspect that Kirk has something like this in mind as his solution to the blanket salt void problem. At any rate, the two blanket solution appears to be alive and a very promising path toward a nuclear future.