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PostPosted: Feb 03, 2010 10:44 pm 
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Here is the idea I had for continuous distilling.

First I presume the critical item in distilling is the surface area.
Imagine a flat plane of liquid, perhaps 1-2 cm deep, 15 cm wide, and 1.2m long or a total surface area of 0.18 m^2 and a fuel volume of 0.0018 m^3.
Picture a switchback like series of baffles with the fluid starting at the back left traveling forward 14 cm, going to the right 1cm then traveling back 14cm, etc. snaking its way to the exit at the front right. The plate is held flat. Let the walls of the baffles and edges of the plate be 3cm tall.

Over the plate is a half-pipe with a 15cm radius which continues down a meter or two where the pipe is cooled and the condensate collects.

Now picture a plumbing traps at the start, end, and where the condensate exits which serve to keep our vacuum.


The liquid is added to the left side of the trap and thus raises the liquid level on both sides of the start trap. The right side of the start trap is connected to a hole in the back/left side of the evaporative plate. As UF4 evaporates it goes up and then down where the cooling of the pipe condenses it. The liquid UF4 then enters the condensate trap raising its level and causing an equivalent amount of liquid UF4 to exit the condensate trap. The fuel salt travels the 16.6m (14cm * 119 switchbacks) to get to the exit of the plate. Going through a hole at the exit the fuel salt goes into the exit trap and eventually out the other side.

Since the entire apparatus is sealed with liquid the vacuum pump is only needed to start the process and to make up for any gas leaks.
Level sensors are needed to know how much fuel salt to add. If desired, a mechanical gate can be used to slow the flow.

Electrical heating and cooling pipes should be provided in the plate to allow bringing the fuel salt up to temperature.

Four such are required for (U), (Zr, Cs), (Li,Be), and (Th) respectively. Each can be independently controlled for both temperature and vacuum. The flow rates are linked.

There should be no precipitates from the (U), and (Zr,Cs) steps. I'm less clear what happens when we pull out the (Li, Be) as these are the primary solvents. The temperature should be quite high at this point (1150-1200C) to keep the ThF4 molten. I'm not sure about the solubility of the fission products and PuF3 in ThF4.

The fourth stage to recover some of the Th is optional - and could easily be put off for a while to let things cool. This is not a critical item since the Th is fairly cheap (and as Jaro would say it is needed anyway - though ORNL/MSBR did plan on storing the fission products w/o Th). The primary need is to recover the U. The second goal is to recover the 7Li. Anything else is much less important.


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PostPosted: Feb 04, 2010 9:10 am 
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Lars wrote:
....Continuous distillation design....
Neat. See also falling film evaporator. My favourite approach at the moment is double falling film. Two concentric cylinders, axis vertical, the outer one of hastelloy, the inner of something exotic - tungsten or graphite or vitreous carbon. The outer one runs at ordinary (for LFTRs) temperatures of ~500C, and has a falling film of clean salt on its inner surface. The inner one is electrically heated to >1000C, except for its ends, by which it can be held onto the outer cylinder. Dirty salt is pumped/poured onto its outer surface, and runs down while being heated by the hot surface. The whole thing is in a vacuum chamber, which doesn't have to be particularly hot. Volatile components jump from the hot salt to the cold. The receiver salt keeps everything diluted, so no solids (UF4) can form on the condenser. Chain together as many of these units as required to o

I've not said anything further on this thread because it really comes down to a materials of construction problem. If we can build a fluoride resistant structure to run at 1200C, there are several ways to arrange the kit to do the job, although keeping the high temp parts as simple as possible will help.


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PostPosted: Feb 04, 2010 12:08 pm 
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Luke wrote:
[If we can build a fluoride resistant structure to run at 1200C, there are several ways to arrange the kit to do the job, although keeping the high temp parts as simple as possible will help.

That's a great attitude ! .....I like it !

.....call it "KISS with performance" :O)


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PostPosted: Mar 16, 2010 9:12 am 
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I noticed that plutonium has many fluoride states. Normally in the fuel salt it is PuF3. In ORNL's vacuum distillation experiments they used PuF6 at 650C with good success - though it was tough to get the Pu to be in PuF6 state. The PuF4 state has a similar boiling point to ThF4. I was wondering, if we bubble fluorine in the liquid salt and then apply vacuum distilling at 1200C will that pull most of the plutonium out together with the thorium to return to the reactor locally?


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PostPosted: Mar 17, 2010 8:32 am 
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Lars wrote:
....The PuF4 state has a similar boiling point to ThF4. I was wondering, if we bubble fluorine in the liquid salt and then apply vacuum distilling at 1200C will that pull most of the plutonium out together with the thorium to return to the reactor locally?
We have 3 options:-
No fluorination. 1-2% of the U is UF3, most of the Np is NpF3, all the Pu is PuF3. All the trifluorides follow the fission product trifluorides, you need a 2nd processing step, such as the electrochemical route used for IFR pyroprocessing, to recover the U. This will probably pick up the TRUs as well.

Some fluorination. Push all the U to UF4 so it is recovered in the still. Most of the Np will follow the U, you can leave the PuF3 with the fission products and deal with it later.

More fluorination. It's much easier to make PuF4 than PuF6, and it is less corrosive. Distill it out and return it to the core as you suggest. However, with the Pu put back, you will get more Am, Cf..... so you still need some secondary treatment to avoid sending them to waste.

Unfortunately, I don't know whether UF4 --> UF6 or PuF3 --> PuF4 is easier, so it's possible you might have to do the standard UF4 --> UF6 --> UF4 process as well as the high-temp still, with the U and most of the Np bypassing the still. Literature data on standard oxidation/reduction potentials suggests that this would be the case, but the data I have to hand (CRC handbook) is all for aqueous systems so its relevance is dubious.


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PostPosted: Mar 17, 2010 11:51 am 
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If we leave the Pu in locally it means we generate 20kg Pu per GWe-yr to the spent fuel for later recovery. This is fine by me. The pu238 is a great fertile material but losing it won't hurt our neutronics. Although if the Np is primarily NpF3 then I suppose we really don't have Pu238 but rather Np237 - still the same amount. But this actually would help our local neutronics and make the neutronics worse back at the central reprocessing site.

If we process the salt on a 3 year schedule it means we leave 3 kg u233/u235 (1500kg)*(1% UF3)/(3 years)(60% of the uranium is either u233 or u235). This hurts a little bit.



If we can manage to do mild fluorination then perhaps we would extract most of the UF3, NpF3, and PuF3 on-site. This would reduce our spent fuel actinides down to around 1kg/GWe-yr.

This isn't a major deal to me as we should like to central processing anyway and it has a minor effect on the neutronics for the local site. However, it is possible (probable?) that we never get around to the central site - in which case a 20 fold reduction is attractive.


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PostPosted: Apr 06, 2010 10:42 am 
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The waste flow from the Lithium Homogeneous Reactor (LHR) will contain only trace amounts of thorium since TH232 and U233 will remain confined to the slurry particle.

This is a consequence of the fission recoil separation effect (FRSE).


When fission occurs in a slurry particle, the byproduct waste isotopes of the fission are almost always blasted out of the particle into the surrounding liquid suspension medium. It is highly unlikely that the fission products will reenter or physically recombine with the slurry particle population. The fission products form a particulate suspension separate from the original slurry. The waste products float in the lithium as separate atoms or as sub nano-particles whereas the feedstock U233 and thorium are retained in the slurry particles.



In the LHR, since the fission process physically separates its own waste products and the slurry and the waste can then be separated by centrifugal means, it is only necessary to chemically process the absolute minimum amount of waste material.

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PostPosted: May 31, 2010 4:01 pm 
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Luke wrote:
Lars wrote:
....The PuF4 state has a similar boiling point to ThF4. I was wondering, if we bubble fluorine in the liquid salt and then apply vacuum distilling at 1200C will that pull most of the plutonium out together with the thorium to return to the reactor locally?
We have 3 options:-
No fluorination. 1-2% of the U is UF3, most of the Np is NpF3, all the Pu is PuF3. All the trifluorides follow the fission product trifluorides, you need a 2nd processing step, such as the electrochemical route used for IFR pyroprocessing, to recover the U. This will probably pick up the TRUs as well.

Some fluorination. Push all the U to UF4 so it is recovered in the still. Most of the Np will follow the U, you can leave the PuF3 with the fission products and deal with it later.

More fluorination. It's much easier to make PuF4 than PuF6, and it is less corrosive. Distill it out and return it to the core as you suggest. However, with the Pu put back, you will get more Am, Cf..... so you still need some secondary treatment to avoid sending them to waste.

Unfortunately, I don't know whether UF4 --> UF6 or PuF3 --> PuF4 is easier, so it's possible you might have to do the standard UF4 --> UF6 --> UF4 process as well as the high-temp still, with the U and most of the Np bypassing the still. Literature data on standard oxidation/reduction potentials suggests that this would be the case, but the data I have to hand (CRC handbook) is all for aqueous systems so its relevance is dubious.


Suppose we do this in two steps:
1) at 600C add some F2 to convert the UF3 to UF4 then vacuum distil. We will extract the uranium and some of the Zr. This step will recover 90% of the value of the fuel salt very quickly. It will also extract any NpF4 present.

We can then let the salt sit and cool for a while (and let the PaF4 convert to UF4).

2) At 1200C add some F2 to convert the remaining NpF3 to NpF4 and the PuF3 to PuF4. There will be very little uranium left (basically just what results from decaying Pa) so whether it is UF4, UF5, or UF6 makes little difference. If it is truely painful, then we can do some vacuum distillation at 600C (after waiting long enough for the Pa to decay) before raising the temperature up to 1200.

When we reinject the extracted good stuff (U,Np, Pu, Th, Li, Be) we add thorium (not ThF4 but Th) to rebalance the fluorine environment (making up for the fluorine added in the vacuum distillation stage).


Earlier you expressed concern about forming too much UF5 and UF6 - has splitting the process into two steps alleviated that concern?


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PostPosted: May 31, 2010 6:09 pm 
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Lars wrote:
Suppose we do this in two steps:
1) at 600C add some F2 to convert the UF3 to UF4 then vacuum distil. We will extract the uranium and some of the Zr. This step will recover 90% of the value of the fuel salt very quickly. It will also extract any NpF4 present.

We can then let the salt sit and cool for a while (and let the PaF4 convert to UF4).

2) At 1200C add some F2 to convert the remaining NpF3 to NpF4 and the PuF3 to PuF4. There will be very little uranium left (basically just what results from decaying Pa) so whether it is UF4, UF5, or UF6 makes little difference. If it is truely painful, then we can do some vacuum distillation at 600C (after waiting long enough for the Pa to decay) before raising the temperature up to 1200.

When we reinject the extracted good stuff (U,Np, Pu, Th, Li, Be) we add thorium (not ThF4 but Th) to rebalance the fluorine environment (making up for the fluorine added in the vacuum distillation stage).


Earlier you expressed concern about forming too much UF5 and UF6 - has splitting the process into two steps alleviated that concern?

I think this would work. My only concern is that if you've gone to the trouble of handling fluorine at all, would you also want to do high temp distillation of UF4. It might be simpler overall to just do standard recovery of UF6 and NpF6, which would also convert PuF3 to PuF4. Then after allowing for decay you do one distillation and get PuF4, any remaining UF4 and NpF4, and the ThF4

There are many variations and tweaks one could try with this process. I haven't tried to define it further because the speculations aren't constrained by data. We need reliable values for vapour pressures of salt mixtures, and how interactions between the salt components changes the vapour composition. For example, ORNL work and other reports show that the vapour pressure of ZrF4 over NaF/ZrF4 mixtures is markedly less than a simple calculation from the pure ZrF4 vapour pressure and the dilution ratio would suggest. ZrF4/LiF is probably similar, and (Th,U,Np,Pu)F4/LiF as well, but I haven't been able to find anything on these. Relative reactivity to fluorine for NpF4 and UF4 to NpF6/UF6, against PuF3 to PuF4 would also help. The measurements aren't that hard, but someone has to do them, and pay for the time and equipment.


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PostPosted: May 31, 2010 8:16 pm 
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Agreed experimental data is needed, but brainstorming the likely results and how they might be used helps to shape the experiments as well.

Vacuum distilling is attractive to me since we likely need it anyway to recover the 7Li.
The fissile uranium is 90% of the value of the fuel salt.
The 7Li is almost all the rest of the value.
For economic reasons I'd like to recover the uranium rapidly, on-site, either continuously or in very small batches. This also plays well for anti-proliferation efforts - there is no extra fissile in the reactor and none to transport.
If we can pick up the Np, Pu, and Th as well this plays well for waste minimization.


Bubbling F2 at modest temperatures through the salt is a much simpler effort than the falling drops at 600C in fluorine gas that ORNL did for converting PuF3 to PuF6. So if the same vacuum distilling to garner the 7Li can also pick up PuF4 then I think this will be simpler.


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PostPosted: Jun 02, 2010 7:05 pm 
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(Replying here to try and avoid further pollution of Kirk's cheap salt thread). In that thread
Lars wrote:
..... Oxygen is generally bad for the reactor - are you confident that ALL of the oxygen will go to Zr and not to U?

Not without some more checks. Add it to the ever-growing experiments-to-do list. There is some data in ORNL reports
ORNL-3215_Semiannual_Aug_1961, page 125 wrote:
.... typical case of ZrO2 precipitation, without the formation of UO2, was afforded by a melt which became contaminated with oxide from insufficiently pure helium used as a protective atmosphere. In this case ZrO2 accumulated in a layer at the bottom of the melt, and no UO2 was found......
This is from a section discussing the proposed MSRE fuel composition LiF-BeF3-ZrF4-ThF4-UF4 70-23-5-1-1 mole%. I think that sets an upper bound to the Zr concentration, but it is too high, that composition has as much Zr by weight as U and Th combined.
ORNL-3936_Semiannual_Feb_1966, page 15 wrote:
..... as ZrF4 (present in in MSRE fuel) is added to 2LiF-BeF2 (flush salt), the oxide tolerance at first falls, reaches a minimum, and then rises.....
From the plot on the next page, the minimum is at about 0.15 mole Zr/kg or 1.4% w/w or 0.65 mol%, and Zr is as good an oxide scavenger at 0.03 mol/kg, 0.27% w/w, 0.13 mol% on the low side of the minimum as it is at 5 mol%, 10.5% w/w on the high side.

From which I tentatively conclude that at low U concentrations typical of thermal spectrum reactors it will be possible to selectively precipitate out ZrO2 free of UO2 and keep the Zr concentration below 1.5% w/w. For fast spectrum machines with lots of fissile, UO2 precipitation could be a problem. If the UO2 solubility curve is well behaved and follows the ideal (U conc) * (oxide conc)^2 = constant, a first guess for the limit would be 4 mole% U, from the fact that the minimum oxide concentration in the ZrO2 curve is about half the value for 5 mole% Zr, which we know to be OK with 1 mole% U. However, the fact that the ZrO2 curve has a minimum is a warning that such simplistic calculations are no more than guesses.

Lars, do you have authority to move posts around? Kirk must be swamped with starting a new job, and we've got off-topic stuff all over. It's fine for those taking part now, but trying to find stuff a year from now will be tricky.


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PostPosted: Jun 04, 2010 4:01 pm 
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Thanks for the answers. No I don't have authority to move stuff. Individuals can delete their wrongly placed messages. Does anyone else have authority? Rick?


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PostPosted: Jun 04, 2010 6:39 pm 
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Nah, only admins, global mods, or mods for the forum in question can move stuff around.

We've only got essentially one of the above - Comrade General Secretary Sorenson, an admin.

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PostPosted: Jun 05, 2010 8:35 pm 
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Isn’t there some way that we can use this approach for waste processing? It could save a lot of time, money, and work.

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PostPosted: Jun 06, 2010 4:56 am 
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Axil wrote:
Isn’t there some way that we can use this approach for waste processing? It could save a lot of time, money, and work.
That is the method developed for IFR processing that we have mentioned a few times up thread. My (and I think Lars's) plan is to use this as a back-end to the reprocessing system, as it has the best selectivity for removing actinides from the shorter 1/2-life fission products of any known method, and all the development has been done already, at least to pilot plant scale. Ideally, this would be done at a central facility, so the equipment at each reactor is as simple, and therefore cheap, as possible. People have been distilling things - mostly aqueous ethanol -for >500 years. It is a fundamentally simple process, which I hope will let us keep cost/reactor down. I did try to work out an all-electrochemical reprocessing flow, but gave up and went for the still idea instead.


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