Energy From Thorium Discussion Forum

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PostPosted: Apr 22, 2011 11:34 am 
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Jess,
From some initial looks it appears that hot (1200C) vacuum distillation can do a decent job of separating thorium from the salt seekers. ORNL only considered metal vessels for vacuum distillation and did not consider doing it at such a high temperature. This seems a natural for a graphite crucible. Do you foresee a problem using vacuum distillation at 1200C using a graphite pot?

The heat load in the salt seekers (post vacuum distilling) after 5 or 6 years is dominated by 238Pu and 90Sr. I would guess we separate the 238Pu using a liquid metal extraction but I'm not sure how to extract the 90Sr. Any ideas?


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PostPosted: Apr 22, 2011 2:04 pm 
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Strontium fluoride is very stable so you could try adding some metal (sodium?) that forms a slightly less stable fluoride, reducing most (all?) other salt-seekers to metal and leaving the strontium as fluoride.

Perhaps the low temeperature high or open magnetic gradient seperation technique can also be used. It has been used before to seperate radioactive stuff (uranium traces I think).


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PostPosted: Apr 22, 2011 6:09 pm 
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Location: Oak Ridge, TN
Lars wrote:
Jess,
From some initial looks it appears that hot (1200C) vacuum distillation can do a decent job of separating thorium from the salt seekers. ORNL only considered metal vessels for vacuum distillation and did not consider doing it at such a high temperature. This seems a natural for a graphite crucible. Do you foresee a problem using vacuum distillation at 1200C using a graphite pot?

The heat load in the salt seekers (post vacuum distilling) after 5 or 6 years is dominated by 238Pu and 90Sr. I would guess we separate the 238Pu using a liquid metal extraction but I'm not sure how to extract the 90Sr. Any ideas?



Lars - Not being a chemist I don't know. I suspect that there are ways in which this can be done, but it will be driven by the economic trade off between performing these separations and just purchasing fresh salt.


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PostPosted: Apr 22, 2011 6:23 pm 
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IF that is the case it will never be done since fresh thorium is very cheap. But the tradeoffs in publicity for waste generation could be worth much more than the costs to recycle it. It makes for a pretty attractive story to be able to say that we recycle the thorium until it is consumed and in the process use 200 times less mined fuel than current reactors.


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PostPosted: Apr 22, 2011 7:09 pm 
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The high-temperature still needs something in the boiler to keep the salt-seeker fluorides molten, provide some vapour pressure at 1200C, and dilute the heat-generating materials enough to allow them to be kept cool. The initial proposal was to use ThF4 as it is there already. The waste stream is then ~400 kg/year of salt seekers + ~4000 kg/year of thorium, all as fluorides. After a few years it might be worth pulling some of the Th out to reduce the volume of material to be disposed of, but even then total Th usage would be more like 3 Te/year than 1 Te/year. It's not going to be economic to mine for Th in a radioactive mess compared to obtaining it from rare earth mines - much the same mix of elements, but all stable isotopes.


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PostPosted: Apr 22, 2011 7:19 pm 
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Lars wrote:
IF that is the case it will never be done since fresh thorium is very cheap. But the tradeoffs in publicity for waste generation could be worth much more than the costs to recycle it. It makes for a pretty attractive story to be able to say that we recycle the thorium until it is consumed and in the process use 200 times less mined fuel than current reactors.


Lars - I agree, it's a combined cost of thorium and disposing of the discard salt. It makes a good story, until we look at this in more detail we will not know the extent to which we can utilize the thorium. As I pointed out, even with the relatively large discard rate of the MSBR, we would be doing quite well, exceeding 10% utilization, which is 15x better than current LWR uranium utilization (but less that the 166x better with full utilization).


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PostPosted: Apr 23, 2011 10:04 am 
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Jess Gehin wrote:
Note that the assumption of the 1 tonne of thorum per GW is the theoretical minimum. It's not clear that all of the thorium can be recycled without losses or eventually the need to discard the thorium because of the buildup of isotopes that are just too hard to separate from thorium. The ORNL MSBR design just assumed that the salt/thorium would be replaced on a 5 year cycle. My calculation on the thorium feed rate for the MSBR single fluid design was 6 tonnes per GWe-year.


Besides economic reasons, does this figure change a bit if a two fluids MSR is considered ?


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PostPosted: Apr 23, 2011 11:45 am 
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It is easier to remove BeF2 and LiF from the fission products than thorium so the separation job is easier. If the very high temp vacuum distillation works then I don't think there is a significant difference.

The desire to keep things liquid for easier chemical processing is still there and the rare earth fission product fluorides don't melt at a reasonable temperature so they need to be dissolved in something. The 7Li is pretty expensive so we do want to recover it.

Still mulling this one over. The challenges are:
1) if you have just fission products the heat load is significant. From the salt seekers most decay away pretty fast so waiting 5-6 years before doing your final reduction substantially reduces the problem. The 238Pu (87.7 yr half life) and 90Sr (28.8 yr half life) are still serious heat contributors though.
These nuclides could be quite useful. If they are removed then the remaining heat load is very managable.

2) It is much easier to process the waste in liquid form.

I am not familiar with the process needed to put the waste into final form.

Since even after our best processing there will still be residual amounts of plutonium in the waste I presume we still will need to put it into a robust chemical form and then bury it deep. If it is not giving off significant heat then burying it is much easier. Most deep disposal sites are limited by the heat generation. I did a hand calculation on the heat generated by 144Ce (likely the next biggest heat contributor) and after 10 years the waste from 1 GWe-yr is down to 100 Watts. Likely the heat from the residual 90Sr, 238Pu, etc. will be larger - but the conclusion is that within a short time the heat will be reduced to what is generated by the residual amounts of 137Cs, 90Sr, and 238Pu.

It may well be that separating these is too painful and we just have to use intermediate storage and dilution to get the specific heat down far enough for permanent disposal. But maybe it isn't so painful.


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PostPosted: Apr 23, 2011 12:55 pm 
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Alex P wrote:
Jess Gehin wrote:
Note that the assumption of the 1 tonne of thorum per GW is the theoretical minimum. It's not clear that all of the thorium can be recycled without losses or eventually the need to discard the thorium because of the buildup of isotopes that are just too hard to separate from thorium. The ORNL MSBR design just assumed that the salt/thorium would be replaced on a 5 year cycle. My calculation on the thorium feed rate for the MSBR single fluid design was 6 tonnes per GWe-year.


Besides economic reasons, does this figure change a bit if a two fluids MSR is considered ?


Yes it could. As mentioned in my previous post, at some point the build of parasitic fission product absorbers in the blanket salt may become a problem unless you remove them. Since the goal is to have a low fission rate in the blanket, you can likely live with a lower level of salt cleanup or discard than in a single fluid system.


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PostPosted: Apr 23, 2011 1:03 pm 
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Lars wrote:
It is easier to remove BeF2 and LiF from the fission products than thorium so the separation job is easier. If the very high temp vacuum distillation works then I don't think there is a significant difference.

The desire to keep things liquid for easier chemical processing is still there and the rare earth fission product fluorides don't melt at a reasonable temperature so they need to be dissolved in something. The 7Li is pretty expensive so we do want to recover it.

Still mulling this one over. The challenges are:
1) if you have just fission products the heat load is significant. From the salt seekers most decay away pretty fast so waiting 5-6 years before doing your final reduction substantially reduces the problem. The 238Pu (87.7 yr half life) and 90Sr (28.8 yr half life) are still serious heat contributors though.
These nuclides could be quite useful. If they are removed then the remaining heat load is very managable.

2) It is much easier to process the waste in liquid form.

I am not familiar with the process needed to put the waste into final form.

Since even after our best processing there will still be residual amounts of plutonium in the waste I presume we still will need to put it into a robust chemical form and then bury it deep. If it is not giving off significant heat then burying it is much easier. Most deep disposal sites are limited by the heat generation. I did a hand calculation on the heat generated by 144Ce (likely the next biggest heat contributor) and after 10 years the waste from 1 GWe-yr is down to 100 Watts. Likely the heat from the residual 90Sr, 238Pu, etc. will be larger - but the conclusion is that within a short time the heat will be reduced to what is generated by the residual amounts of 137Cs, 90Sr, and 238Pu.

It may well be that separating these is too painful and we just have to use intermediate storage and dilution to get the specific heat down far enough for permanent disposal. But maybe it isn't so painful.



A waste form that should be considered is fluorapatites. You'd want to recover any enriched lithium but the rest could be combined with phosphates and converted for fluorapatites, placed in canisters and discarded. The long-term behavior of these materials containing fission products and actinides need to be looked at. Most disposal options would assume preliminary decay storage of perhaps 50-75 years.


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PostPosted: May 10, 2011 1:22 am 
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Darryl Siemer, who posts here quite a bit, has done quite a lot of work figuring out how to passivate nuclear waste using iron phosphate glass. It looks promising, and cheap.


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PostPosted: May 13, 2011 7:41 am 
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Thorium is available in rare earth ores in high concentrations of upto 4-10%. It should be cheaper than uranium.
Thorium can be used in the uranium nuclear fuel in a dual role-consumable poison and fertile fuel. Indian Th-LEU AHWR fuel uses both these roles effectively, getting 39% of energy from burning of U233 produced in situ.
I think that the idea should be developed further. Pebble fuel can be fabricated by using 20% LEU or recovered plutonium as fissile feed, a thorium shell as fertile fuel and BeO as moderator. A thorium metal casing with a thickness of 10% of OD will be 48.8% thorium by volume and 51.2% space to be filled with fissile fuel and moderator. It can be coated with standard pebble SiC to contain any escape of fission products. It could be cooled with a cost-effective liquid salt like the proposed design of AHTR as it is not required to dissolve thorium. Costly 99.995% Li7 can be avoided. A safer liquid metal coolant can also be used.
After use, thorium casings can be electro refined to obtain U233 and to recover thorium for reuse.


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