Energy From Thorium Discussion Forum

It is currently Jan 23, 2018 9:10 am

All times are UTC - 6 hours [ DST ]




Post new topic Reply to topic  [ 17 posts ]  Go to page 1, 2  Next
Author Message
PostPosted: May 30, 2009 11:07 pm 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
Normally, we add ThF4 as fuel to the reactor and we use BE to maintain UF4/UF3 balance.

Could we instead add ThF4 and Th in proper balance to keep the UF4/UF3 balance?
This would avoid the use of Be - which I understand requires additional safety precautions.


Top
 Profile  
 
PostPosted: Jun 01, 2009 4:54 am 
Offline

Joined: Jul 28, 2008 5:01 am
Posts: 462
Location: Teesside, UK
Lars wrote:
Normally, we add ThF4 as fuel to the reactor and we use BE to maintain UF4/UF3 balance.

Could we instead add ThF4 and Th in proper balance to keep the UF4/UF3 balance?
This would avoid the use of Be - which I understand requires additional safety precautions.

The fuel cycling for reprocessing has a much higher flow rate (1-200 kg heavy metal /day for TMSR) than the 3-4 kg/day flow for input Th, so the ratio in the returning fuel dominates. You need to take some of the returning UF4 and reduce it to UF3, which you can then dose back into the reactor as needed. Alternatively, it can be done by forcing the potential of the salt with an electrode. There was a discussion of the electrical method earlier this year but I can't find the thread now.

Luke


Top
 Profile  
 
PostPosted: Jun 01, 2009 11:56 am 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
From ORNL7207 page 78
As indicated previously, the fission process occurring with UF4 is
significantly oxidizing. During the 30-year reactor lifetime (22.5 full -
power years assumed) with 70% of the fissions occuring in uranim isotopes,
nearly 6.12 x 10^4 moles of uranium will have been fissioned. If
the fissioning uranium is 95% UF4, as much as 5.8 x 10^4 moles of UF3 might
be oxidized [at a generally unifom rate of 7 moles (2.1 kg) per fullpower
day] during the reactor lifetime. Its reduction would require some
2.9 x 10^4 moles (261 kg) of metallic beryllium.

This reactor is running primarily on U235 (hence 70% fissions in Uranium). If we run primarily on u233 then we have fission of u233 (90%) and u235 (70%) so that 97% of fissions are in uranium. This will scale up the 7 moles/day to 9.7 moles/day of UF3 converted to UF4. If we add Th (not ThF4) to the fuel salt will we get the reaction
Th + 4 UF4 -> ThF4 + 4 UF3

If we assume we get 100kg/GWe-yr of UF4 from the blanket and we consume 1000kg/GWe-yr of U233 this is 900kg/yr or 2.47kg/day (10.58 moles/day) of u233.
To generate the u233 we need to add 10.58 moles/day of thorium.
Seems like we want to add 2.45 moles (0.57kg) of Th and 8.16 moles (2.51kg) of ThF4 each day.


Top
 Profile  
 
PostPosted: Feb 15, 2010 6:25 am 
Offline

Joined: Jul 14, 2008 3:12 pm
Posts: 5057
I'm wondering if other metals would work, like lithium-7 or zirconium.


Top
 Profile  
 
PostPosted: Feb 15, 2010 1:01 pm 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
I would imagine so. We do burn a bit of 7Li so we will need to add makeup 7Li. In this case the LiF will become 2 atoms of He + a fluorine. So we would just balance this effect if we add straight Li. It is similar for 9Be(n,2n).

We also lose (I suspect a much larger amount) salt with the waste flow from extracting the salt seeking fission products. In this case the LiF leaves together so the right makeup would be LiF.


Once the fuel salt contains a stable amount of fission products we can see that the offgas system will remove fission products but virtually no fluorine. Likewise, the noble metals will be extracted with almost fluorine. In general, the salt seekers will be extracted together with some fluorine, some mono, some di, and some tri fluorides plus ZrF4 (are there other salt-seekers with 4 fluorines?) So overall it makes sense that we lose less the four fluorines per fission. So we ought to add less than four fluorines per thorium atom added. I don't imagine it matters whether the missing fluorines are in the added thorium, lithium or Be.


Top
 Profile  
 
PostPosted: Feb 15, 2010 5:02 pm 
Offline

Joined: Jul 14, 2008 3:12 pm
Posts: 5057
Thanks Lars. Perhaps getting rid of 7LiF is useful because of it's high boiling point (and reduced tritium of course). ZrF4-BeF2 looks about as good as FLiBe in terms of absorbtions though obviously it moderates a bit less. But we can recover it easier than LiF? It's certainy cheaper so we can afford to lose some of it in processing.


Top
 Profile  
 
PostPosted: Feb 15, 2010 5:15 pm 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
We don't really know how the salt seekers are going to be removed from the fuel salt. There are several ideas and some lab experiments but nothing yet that works so well that there is agreement on choosing that path. As such it is hard to say how much of the LiF will go with the waste flow of salt seekers. In general, there isn't much concern about a modest portion of the salt going off to the waste flow. I don't think the cost here is significant.

You are right that the use of lithium does mean a pretty significant tritium production - similar to a heavy water reactor. Containing that tritium will impose additional costs of unknown magnitude. One of the development risks with LFTR. Switching to ZrF4 instead I presume would make the tritium production go way down and make it much easier to contain the tritium produced. This would be the my main interest in avoiding litium - not the cost of 7Li, not the flow to the waste stream, but avoiding the tritium production and the associated costs of containing it.

I'm imagining an onsite vacuum distillation process to isolate the salt-seekers (+Pu) temporarily (call it spent fuel?) and then later (perhaps decades) so fancy electrolysis process at an approved, central location to pull out the plutonium. In such a scheme either ZrF4 or LiF will come out quite nicely. It is mostly that I would prefer the temporary spent fuel to remain a liquid for ease of handling during processing and we may need to keep some of the LiF to accomplish this.


Top
 Profile  
 
PostPosted: Feb 15, 2010 5:31 pm 
Offline

Joined: Jul 14, 2008 3:12 pm
Posts: 5057
Good points. If I understand things correctly then LiF boiling point is the highest distillation temperature, right? So with a lower boiling point carrier, we can save costs and make things easier by lower temperature distillation?


Top
 Profile  
 
PostPosted: Feb 15, 2010 5:45 pm 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
ZrF4 is more volatile than LiF so yes the ZrF4 comes out at a lower temperature (or higher pressure).
ThF4 is less volatile than either and will require the higher temperatures. It is also much less valuable so there is less urgency (interest expense) in removing it and in any event we would need to let the fuel cool some before pulling out much of the thorium.

I haven't taught myself enough about the vacuum distilling process to feel comfortable with the temperature profiles needed to extract ZrF4 versus LiF. My impression though is that without the requirement for neutron transparency and with no neutron damage our selection of materials to construct the still is wide enough that this isn't a major problem.

If we pull out the UF4/3, LiF, ZrF4, CsF, and BeF2 then I think the main solvent left in our salt is ThF4 with a melting point of 1110C. So, I'm thinking the operating temperature of the final stage of the still will need to be in the range of 1200C to keep the ThF4 fluid.


Top
 Profile  
 
PostPosted: Feb 15, 2010 6:06 pm 
Offline

Joined: Jul 14, 2008 3:12 pm
Posts: 5057
Ah, but you're thinking of a 1 or 1 1/2 fluid design. Reducing carrier salt boiling point compared to LiF won't be as useful then.

However, I'm thinking about a pure two fluid design with a fat thick blanket to avoid Pa separation. Just fluorinate out the U-233, leave the thorium in there (except for physically circulating through the HX for cooling). Then, it seems useful to have ZrF4 and BeF2 rather than LiF and BeF2. We'd only have to fire the still to BeF2 boiling point.


Top
 Profile  
 
PostPosted: Feb 15, 2010 6:35 pm 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
Is the residual a liquid or a solid at that point? I think it will keep the machinery simpler if the residual is liquid.

Beside, it would be a pretty intense heat source to have just fresh fission products.

If the fuel salt doesn't contain any 7Li then I don't imagine there is much of value left in the fuel salt once you have removed the uranium. In this case, I'd probably just remove the uranium and leave the rest for a holding tank for three years. Then I'd pull out some of the salt (since much of decay power is gone now and we can manage reducing the volume of waste without create an excessive heat load problem).

But the biggest problem I have with the pure 2 fluid design is the wall lifetime.

I'm willing to tackle tougher processing in order to reduce requirements on the reactor wall.

In the 1.5 fluid design only 5-10% of the neutrons make it to the wall. In a two fluid design 50% of them have to go through the wall.


Top
 Profile  
 
PostPosted: Feb 16, 2010 4:51 am 
Offline

Joined: Jul 14, 2008 3:12 pm
Posts: 5057
If you want to keep the fission product fluorides liquid then yes you may have to go higher. The highest ones have melting points like 1500 C or something. Though they will form some kind of weird nasty cocktail eutectic/peritectic, that would have a considerably lower melting point. Close to average fission product fluoride melting point? Much of the stuff wouldn't want to form eutectics maybe? Maybe we have to leave a bunch of ZrF4 in there. It would sure be a nasty goo.

I checked LiF boiling point, it is rather steep at over 1600 C @ 1 atm, so recovering all that definately means higher temperature distillation. Not sure if that's such a big problem, just wondering.


Top
 Profile  
 
PostPosted: Feb 16, 2010 4:55 am 
Offline

Joined: Jul 14, 2008 3:12 pm
Posts: 5057
As for the wall life, it sounds plausible that the 1 1/2 fluid has a longer wall lifetime but this really requires testing, with various different materials. It may not be a big enough difference to warrant more complicated reprocessing.


Top
 Profile  
 
PostPosted: Feb 16, 2010 10:44 am 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
Cyril R wrote:
If you want to keep the fission product fluorides liquid then yes you may have to go higher. The highest ones have melting points like 1500 C or something. Though they will form some kind of weird nasty cocktail eutectic/peritectic, that would have a considerably lower melting point. Close to average fission product fluoride melting point? Much of the stuff wouldn't want to form eutectics maybe? Maybe we have to leave a bunch of ZrF4 in there. It would sure be a nasty goo.

I checked LiF boiling point, it is rather steep at over 1600 C @ 1 atm, so recovering all that definately means higher temperature distillation. Not sure if that's such a big problem, just wondering.


While the boiling point is high at 1 atm it is much more reasonable in a near vacuum. ORNL used a Hastalloy-N vessel for their vacuum distillation experiments. I think they were running around 950C.


Top
 Profile  
 
PostPosted: Feb 16, 2010 10:51 am 
Offline

Joined: Jul 28, 2008 10:44 pm
Posts: 3069
Cyril R wrote:
As for the wall life, it sounds plausible that the 1 1/2 fluid has a longer wall lifetime but this really requires testing, with various different materials. It may not be a big enough difference to warrant more complicated reprocessing.

Maybe not but this is the area that I'm most concerned about. It appears it is OK for a graphite filled reactor like MSBR. The last I saw from the French was that the wall in their fast 1 1/2 fluid design had to be periodically replaced.

With a 1 1/2 fluid the worst case in the processing is that we have a spent fuel pool for decades. The spent fuel pool would contain the salt-seeking fission products + 238Pu + a small amount of 239Pu + tonnes of thorium. It would be a job for the future to separate the stuff. The thing of particular value would be to extract the plutonium. The thorium would have modest value and likely would be extract primarily to reduce the volume of waste storage.

In my view the wall lifetime is a show stopper for some designs. The spent fuel pool is not, though as a policy I would want to guarantee the future extraction of the Pu and funding it as it is generated.


Top
 Profile  
 
Display posts from previous:  Sort by  
Post new topic Reply to topic  [ 17 posts ]  Go to page 1, 2  Next

All times are UTC - 6 hours [ DST ]


Who is online

Users browsing this forum: No registered users and 1 guest


You cannot post new topics in this forum
You cannot reply to topics in this forum
You cannot edit your posts in this forum
You cannot delete your posts in this forum
You cannot post attachments in this forum

Search for:
Jump to:  
Powered by phpBB® Forum Software © phpBB Group