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PostPosted: Feb 16, 2010 10:25 am 
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Warning: blasphemy, no thorium in this cycle! :cry:

In a U/Pu cycle like Jaro's CANDU-MSR, running nearly breakeven, what kind of processing would be preferred? Assume there is only straight UF4 or UF4/UF3, no enrichment, in a high temp reactor.

Do we:

- fluorinate basically all the U out
- distill the Pu
- distill no further and simply send the fission products to temporary storage (5-10 years), U and Pu go back.

That should be pretty simple reprocessing. Am I missing something?


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PostPosted: Feb 16, 2010 10:34 am 
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Yes. The tough job is to separate the Pu from the fission products.
So far as I know the PuF3 does not distill out and neither do the salt-seeking fission products.
With very aggressive fluorination ORNL was able to remove PuF5 but it was really hard on the container.
They pursued coating the container with fuel salt (by cooling the walls while heating the fluorine gas) - I'm not sure of the final result of that approach.
The third approach is the liquid bismuth exchange process. Lots of hope here but the actual lab results were disappointing (1-50% of the calculated throughput). This approach is the one most people continue to assume will do the job. Switching from bismuth to aluminum seems to hold better promise but I haven't seen lab results for this.

Second, you will need to make the wall out of something. What temperature do you need and what material are you thinking of?


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PostPosted: Feb 16, 2010 11:55 am 
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It would be nice if we could just fluorinate the Pu out. Maybe then, with no carrier salt (or only a little sacrificial ZrF4, also for eutectics management and melting point depression), we'd not need a vacuum still at all for the online reprocessing?

But I thought the Pu would come out as PuF6, not PuF5?

http://en.wikipedia.org/wiki/Plutonium_hexafluoride

Seems like PuF6 wants to become PuF5 but saturation with fluorine should deal with that.

As for the corrosion issue, there are a number of coatings, such as (noble) metal and carbon based, as well as high temperature composites, that could deal with the hot fluorine and fluorides environment long enough to work reliably. Other than that the aluminium reduction thing sounds promising.

The wall is easy, reference Jaro's siphon design, graphite tubes most likely, lasts long enough with free creeping and is easy to replace fully robotically.


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PostPosted: Feb 16, 2010 11:56 am 
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Temperature probably 1000 - 1050 C peak. If this is a problem, add more ZrF4.


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PostPosted: Feb 16, 2010 12:15 pm 
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I haven't been around much, but I'm still finding time to work on nuclear things.

I will refer at some point in the future to the Japanese "Fluorex" process which is quite neat.


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PostPosted: Feb 16, 2010 12:21 pm 
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You are right it is PuF6 my memory failed me.
In concluding fluorination is the right path you should be forewarned that the experiments ORNL did that produced excellent results concluded that this was not likely to be useful for fluorinating Pu. So while the results seemed very encouraging to me, they were worried about something. Higher temperatures (650C) definitely helped the Pu extraction but were harder on the vessel walls. Graphite walls would seem like the natural answer but they did not mention them. See ORNL-4224.

The second issue, is that I suspect the US will actively work to prevent public development and deployment of technologies to isolate plutonium from uranium and fission products. So while I think we could develop a cost effective means for on-site removal of plutonium from the spent fuel stream I have changed my approach to leaving the plutonium in the spent fuel stream and eventually shipping it back to a central repository for processing. I may be wrong on this - but the impression I get is that the folks who worry about proliferation would be active opponents of reactors that deploy on-site plutonium isolation.


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PostPosted: Feb 16, 2010 12:24 pm 
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If you are running UF4/{UF3,PuF3} eutectic, you are already well over 1000C in the reactor. If you have suitable exotic materials available for that, you can build a still out of them and pull the UF4 out under vacuum. The remaining UF3/PuF3/fission product mix is just like IFR fuel, so use their methods- electrolysis in a bath of NaF/KF to pull out mixed actinides, leaving the fission products dissolved in the salt until they've cooled off enough for disposal.

A still is a much simpler/cheaper device than a fluorinator/defluorinator setup, especially if you are trying to handle PuF6, and the associated systems for dealing with radioactive contaminated HF from the defluorinator.


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PostPosted: Feb 16, 2010 12:29 pm 
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Can't you do it with electro-refining? Copper is refined industrially by electrical refining in an aquous solution. Aluminium in fluoride salts.


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PostPosted: Feb 16, 2010 1:33 pm 
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Thanks for your comments. Looks like no fluorinator in this design then! The fluoride bath sounds useful also as temporary storage for fission products, doing double duty as spent 'fuel' storage pool. We sure wouldn't need much heating for the bath! And a salt to air cooler to prevent overheating is easy and efficient (MSRE had one).

As for proliferation, our case is better than IFR since we don't breed net plutonium, in fact it is inherently impossible to breed strongly in a bi-modal spectrum using U/Pu, and we start with natural unenriched uranium. However if this is still a problem we could leave the TRU with the fission products. We'd have to add more fresh natural uranium or LEU of course but that's ok, the point about this design is to get some experience with fluid fueled reactors and their continuous reprocessing, and do it with low political profile (ie natural uranium startup).


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PostPosted: Feb 16, 2010 1:37 pm 
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Quote:
blasphemy, no thorium in this cycle!


FYI

The thermal neutron capture cross section for Th-232 (7.4 barns) is almost three times higher than for U-238 (2.7 barns) : Quicker buildup of fissile U-233

The fast fission cross section of Th-232 is almost five times smaller than that of U-238: Smaller contribution from fast fission

Take away

CANDUs should breed U233 from thorium.

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PostPosted: Feb 16, 2010 1:48 pm 
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A variant on this thought is to use 3-5% enriched uranium and add thorium to balance the fissile/fertile ratio.
This will mean most (80%?) of the fissile generated by the reactor is uranium-233 rather than plutonium-239.
The uranium-233 is denatured by the u238. The plutonium quality is lowered by the pu238 we generate plus the slow neutron burn-up of 239Pu.

I wonder what the melting point of ThF4/UF3/UF4 is?

You could use a vacuum still for removing the uranium and leaving behind the salt-seekers (and thorium and plutonium) for the spent fuel pool.
This should result in a higher burn-up.

I don't imagine you could get to unity breeding with this so periodically you would need to add fresh uranium but it should be pretty good. Jaro may have a better feel for how such a reactor would perform.

It is going to be a pretty hot reactor though and that will catch the attention of political opponents.


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PostPosted: Feb 16, 2010 1:52 pm 
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Axil wrote:
Quote:
blasphemy, no thorium in this cycle!


FYI

The thermal neutron capture cross section for Th-232 (7.4 barns) is almost three times higher than for U-238 (2.7 barns) : Quicker buildup of fissile U-233

The fast fission cross section of Th-232 is almost five times smaller than that of U-238: Smaller contribution from fast fission

Take away

CANDUs should breed U233 from thorium.


Unfortunately this is a 'single fluid' design and putting a large amount of thorium in there leads to a number of complications, including

- impossibility of startup with natural uranium (precisely because thorium is so good at capturing neutrons!).
- requirement for Pa separation (large amounts absorb too many neutrons) with unknown political proliferation profile, or a large buffer for the fuel salt (might work well actually but still a bit of a bother) or very little thorium but that's not worth it IMHO.
- thorium reprocessing issues with single fluid designs.

Also, this design is bi-modal so gets advantage from fast breeding U-238. Its goal is to get some experience with fluoride fuelled reactors and their reprocessing, at the lowest political profile possible.


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PostPosted: Feb 16, 2010 1:57 pm 
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Lars wrote:
A variant on this thought is to use 3-5% enriched uranium and add thorium to balance the fissile/fertile ratio.
This will mean most (80%?) of the fissile generated by the reactor is uranium-233 rather than plutonium-239.
The uranium-233 is denatured by the u238. The plutonium quality is lowered by the pu238 we generate plus the slow neutron burn-up of 239Pu.

I wonder what the melting point of ThF4/UF3/UF4 is?

You could use a vacuum still for removing the uranium and leaving behind the salt-seekers (and thorium and plutonium) for the spent fuel pool.
This should result in a higher burn-up.

I don't imagine you could get to unity breeding with this so periodically you would need to add fresh uranium but it should be pretty good. Jaro may have a better feel for how such a reactor would perform.

It is going to be a pretty hot reactor though and that will catch the attention of political opponents.


That sounds like a good compromise. I'd still prefer the natural uranium, no thorium version, for a first development.


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PostPosted: Feb 16, 2010 2:54 pm 
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If you choose that path, are you going to try plutonium separation on-site?
If not, how long can you go before the fission product buildup forces you to replace the salt?


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PostPosted: Feb 16, 2010 3:11 pm 
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Lars wrote:
If you choose that path, are you going to try plutonium separation on-site?
If not, how long can you go before the fission product buildup forces you to replace the salt?


Didn't consider that option yet. Xenon removal alone should give a nice bonus though even with the fission products left in. Supposedly the fission products will corrode stuff badly at high temperatures, but we can run with more ZrF4 in to reduce melting temperature a bit.

I thought if Pu separation is a political problem we can just leave the plutonium alone with the fission products; only distill the UF4 back. We'd want smaller channels to reduce Pu breeding, and more topping fresh fuel, but that's fine.


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