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PostPosted: Mar 02, 2009 6:02 pm 
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Potentialities of the fast spectrum molten salt reactor concept : REBUS-3700
http://cat.inist.fr/?aModele=afficheN&cpsidt=17946152

Abstract :
"...a description of the system design, as well as a justification of its operating parameters is given. The performed neutronics studies demonstrate that REBUS-3700 gathers a positive breeding gain and strong negative salt expansion reactivity feedback. The safety analysis demonstrates an excellent reactor behavior with respect to the majority of single and combined unprotected events"

Have you got the full paper? Do you know anything more?


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PostPosted: Mar 02, 2009 9:51 pm 
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I have an electronic version on my work computer I believe and a paper copy here at home. The study found that you could still use natural chlorine (as opposed to enriched) and still breed as long as you made the fluid core big enough to limit neutron leakage. Their design was a large core of 3.8 meters wide by 3.25 high and a relatively low power density of 100 MW/m3. It needs a massive 18 tonnes of fissile transuranics (Pu239+Pu241 etc) which means about 30 tonnes of total transuranics (Pu,Am,Np). It is for about 1500 MWe so about the same fissile needs as a sodium cooled fast breeder. The salt is 45%(U + TRU) + 55% NaCl where the heavy metals are about 2/3 U and 1/3 TRUs. The melting point is 600 C.

Even with the big core, it still loses 0.39 neutrons per fission to leakage. Natural chlorine would capture 0.17 n/fission and the fission products 0.1 (removed on a 1 year time frame). It just breaks even so once started it produces its own fuel but none extra. They mention how adding a blanket could make it a breeder (catching some of those leakage neutrons) but this is a big step. They don`t really know what materials will be best but they suggest another Hastelloy alloy or using titanium.

Interesting design but a great deal of unknowns in terms of materials and their corrosion behavior along with much R&D needed for processing. As mentioned on other threads, using natural chlorine produces sulfur by neutron capture that can be a corrosion problem. My view anyhow is if we can do what is needed with fluoride designs it is probably not worth the R&D investment to try the chloride route.

David L.


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PostPosted: Mar 03, 2009 6:36 am 
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David wrote:
...
Interesting design but a great deal of unknowns in terms of materials and their corrosion behavior along with much R&D needed for processing. As mentioned on other threads, using natural chlorine produces sulfur by neutron capture that can be a corrosion problem. My view anyhow is if we can do what is needed with fluoride designs it is probably not worth the R&D investment to try the chloride route.


I wonder they decided not to pursue chlorine enrichment, it doesn't seem a really difficult task and easy the life in terms of corrosion, neutron economy and fissile load

Which are the issues in processing fuel salts, I guess in a fast reactor we don't need so high decontamination factors in separating the fissile to back to the reactor from the fission products; or am I wrong ?


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PostPosted: Mar 03, 2009 9:58 am 
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Alex P wrote:
I wonder they decided not to pursue chlorine enrichment, it doesn't seem a really difficult task and easy the life in terms of corrosion, neutron economy and fissile load

Which are the issues in processing fuel salts, I guess in a fast reactor we don't need so high decontamination factors in separating the fissile to back to the reactor from the fission products; or am I wrong ?


Chlorine enrichment may not be too hard - Kirk has a (translated) Russian paper Chlorine Isotope Separation by Liquid-Phase Thermal Diffusion in the archives giving one method for doing it - but it's not an established large-scale process either, and is bound to be an added cost. Whether or not to do it is one of the many things that would have to be investigated by an R&D effort on chloride reactors.

Fast reactors can cope with higher concentrations of fission producs, and can therefore accept some combination of less reprocessing/less-efficient reprocessing compared to thermal reactors. Even for thermal reactors, though, the difficult problem is getting all of the actinides out of the waste products, not getting the actinides pure.

Luke


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PostPosted: Mar 03, 2009 10:27 am 
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In any fast reactor the fissile & fertile loading will be much higher (5-8x) and so a neutron is much more likely to find a productive target than fission products. We can use this to process the fuel less frequently. I assume we have the same leakage of TRUs into the waste flow each time we process the fuel. This means a faster reactor will likely leak less TRUs into the waste stream. Although I hasten to point out that fast or slow if we can pull off what is talked about here we will dramatically (>100x) reduce the waste flow. This is true whether the fast reactor is chloride based for fluoride based.


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PostPosted: Mar 04, 2009 2:19 pm 
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Intersting enough, the reactor achieves a breeding gain of 0,25 even without chlorine enrichment, absolutely higher than that experienced at Superfenix sodium fast reactor (very curious what that number can become with chlorine enrchment!)


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PostPosted: Mar 05, 2009 8:45 am 
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Have you got any idea about the potential cost of construction at max of a first of a kind chloride fast breeder prototype in a range of 100 MW thermal of power, including the begining R&D phase?


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PostPosted: Mar 05, 2009 12:09 pm 
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You might start with how much was spent on the fluoride reactor and then recall that they chose fluorine in part because it was easier.


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PostPosted: Mar 05, 2009 5:49 pm 
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Lars wrote:
You might start with how much was spent on the fluoride reactor and then recall that they chose fluorine in part because it was easier.


Sure, but my question is more generic, in particular if one government (Japan, China, Europe or Usa, etc..) decides to pursue an ambitious program based on MSR technology how much should it spend (including beginning R&D phase), for example, to build three 10-100 MW thermal reactors like these : two liquid fluorides, one graphite moderated thermal spectrum and a non moderated fastish spectrum, and a chloride fast breeder ?


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PostPosted: Mar 08, 2009 7:05 pm 
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Even if I realized that you don't like so much the chlorides option even on a R&D level, an issue about it isn't clear to me : in a possible "symbiosys" between chlorides and fluorides reactors (for example, the uranium 233 produced from the former is used as fuel seed for the second) is it particular difficult to change with no loss the chemical spent fuel waste from a compound to an other, chlorides to fluorides and/or viceversa, or is pratically a non issue?


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PostPosted: Mar 08, 2009 8:09 pm 
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The only issue for me is R&D money. If I were in charge of spending the stimulus money we would do both (assuming we were staff or facility limited). How much symbiosis is there between them? Certainly either can benefit from u233 produced by the other. I don't think it is particularly difficult to switch between fluoride and chloride forms of the fuel. But given limited resources I choose the path where we already have $1B worth of work completed.


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PostPosted: Mar 09, 2009 1:40 am 
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If you do not bother to use light nuclei like Li or Be, even a fluoride salt based reactor could be a fast spectrum one. SnF2 is an interesting possibility with its liquid range being 213-850c. ThF4 solute shall extend it both ways for a convenient operating range. While more volatile fission products like Xe, ZrF4, and TcF6 can be constantly removed in a gas stream, the rare earth fluorides have boiling points even higher than ThF4. Once you take care of 233UF6, you can distill the salt and fuel and leave Strontium and rare earth fission products as residue. Only cesium falls between SnF2 and ThF4 and has to be removed fractionally.


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PostPosted: Mar 09, 2009 1:53 am 
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First thing to look at is the capture cross-section. In the most interesting range (100 ev to 100kev) (excluding HW-MSR) the cross section for Lithium varies from 0.01 to 1e-5. For Sn it varies from 5 to 0.1 - or more than 100x. We will lose 100x as many neutrons to capture in the salt. It will never be critical. You can see the capture cross-section data (n, gamma) at www.nndc.bnl.gov/sigma


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PostPosted: Mar 10, 2009 7:35 am 
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The cross-sections I used are attached. They seem to be for 2200m/s neutrons. I'd love to have some equally easily understood updated data.
Attachment:
X-sec.odt [34.74 KiB]
Downloaded 352 times


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PostPosted: Mar 10, 2009 10:09 am 
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Alex P wrote:
Intersting enough, the reactor achieves a breeding gain of 0,25 even without chlorine enrichment, absolutely higher than that experienced at Superfenix sodium fast reactor (very curious what that number can become with chlorine enrchment!)


Forgot to mention that in a chloride fast reactor with a thorium blanket the 0,25 breeding gain cited is very likely not achievable at all, a non-negligible fact


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