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PostPosted: Jun 08, 2010 8:57 pm 
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The toxicity of nuclear waste is proportional to the quality of the chemistry used and the scrupulousness of the engineers that design the salt purification process. In the figure “impact of loss fraction” as follows, it shows that the fraction of actinides removed from the salt will determine the ultimate safety of the MSR waste stream.

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PostPosted: Jun 09, 2010 5:26 am 
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what is the relative toxicity normalised to, in that graph ? (Thnx)


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PostPosted: Jun 09, 2010 6:36 am 
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Googling the graph title brings up this pdf of a presentation as the likely source of the plot (on p29), and this paper as the source of the data (p6-7), where they state that the radiotoxicity is normalised to that of the uranium ore from which the waste is ultimately derived.

These documents are reasonably interesting, but are unfortunately missing a discussion of the best option...


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PostPosted: Jun 24, 2010 10:25 pm 
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Kirk's question on samarium removal on the blog prompted me to look at how fast we could process if we really wanted to. I didn't think we were that desperate for breeding gain these days, but it's an interesting exercise. From my back-of-envelope calc on the blog, we need to process the core salt, assumed to be 20 m^3, in < 5 days. Target is therefore 3 litres/minute.

Conditions are as ORNL distillation - hastalloy construction, 1000°C, 0.5 mmHg pressure. Unfortunately I don't have much artistic talent, so I hope this sketch and a description are sufficient.

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High throughput still.png
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Still element
Start with a flat box section in 1 mm hastalloy sheet, 1 m long by 10 cm wide by 1 cm high. Inside this put an electric heater, tungsten wire on an insulating spacer. Silica would do, or zirconia if you want to put the price up. The heater power needs to be about 10 kw.
One long edge of the box is welded to a split pipe which distributes molten salt onto its surface. This edge of the box is also fixed to the still wall, with the box dipping away from the wall.

Still
Stack up 10 still elements on a wall, and another 10 facing them. Fix ends plates on, sealing the still elements at one end, the other end aligning with slots to give access for connecting the heater power.
At the top of the still, running along its length, a large split pipe. It is set to create a 'waterfall' of relatively cool molten salt. This is the condenser that can't be blocked with frozen UF4 (or ZrF4). The vapours dissolve as they condense.
Drains in the floor of the still separately collect the undistilled salt, and the condenser salt.

Still array
Stack 10 stills across the top of the reactor, plus another one to take their combined output and concentrate it further. Each still is sealed, but the whole array goes in a big vacuum vessel so that the mechanical stress of atmospheric pressure doesn't have to be taken by a structure at 1000°C

Plumbing
From the reactor outlet at 700°C, T off to take the 3 l/minute for processing, and distribute it to the still elements. From the primary heat exchanger outlet at 550°C, T off to take a much larger flow, 3 l/sec, for the 'waterfall' condensers. The condenser outflow drops the height of the reactor to get back to reasonable pressures by its own hydrostatic head, through a pump and back into the primary loop.
The heater power is set so that 90% of the feed is evaporated. The residual salt is collected and fed to one further still, where a further 90% is removed. THe output from this will be 30 ml/minute of lithium fluoride, with about 1% UF3 and 0.5%-0.7% salt seeking fission products. With the flow rate down to something sane, (electro)chemical processing to split UF3 from fission products from LiF will be practical.

Capacity
The main array has 20m^2 of evaporating area. I think you need that because at the very low pressures involved here even that area gives a vapour velocity of about 200 m/sec. If you can go hotter, the pressure goes up and the system can be smaller, but then you have to make the whole thing out of tungsten.

Power requirements
Heating 3 l/min of FLiBe from 700°-1000°C takes about 100 kw. Boiling it takes a MW. Radiative losses will be significant, but I don't know the emissivity of the materials. A 20m^2 black body at 1000C loses nearly 3 MW, but shiny hastalloy should be far less, so I'm guestimating 2 MW all in, 10 kw/element. The condenser salt gains 100°C as it falls

For more usual reprocessing rates, much simpler geometries will work, down to a pair of concentric pipes, boiling salt running down the outer surface of the inner pipe, which is heated from within, while condenser salt flows down the inner surface of the outer pipe.

As with all distillation reprocessing schemes, it can't remove cesium or zirconium very well. Fortunately, these have small neutron cross-secions so we can tolerate poor extraction.


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PostPosted: Jun 25, 2010 11:54 am 
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You propose boiling off 99% of the fuel salt w/o first converting the UF3 to UF4.
Is what is left still a liquid? Are you assuming any thorium in the fuel salt?
You mentioned posting in the blog the calculations that drove you to processing the entire fuel salt volume every 5 days. I couldn't find that. This processing rate seems rather extreme. (Not that it matters much but I think the power to raise the salt temperature from 700C to 1000C assumed 3 liters/sec rather than 3 liters/min).


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PostPosted: Jun 25, 2010 1:00 pm 
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This started as a question from Kirk, so I'm assuming a Sorensen reactor - 2-fluid, lots of graphite, thermal spectrum, no Th in the core, 1000 kg U in the salt, and 1% of that is UF3. The five-day schedule is estimated in the blog comment after yours (Lars), from cross-section ratios of 233-U and 149-Sm.

The reactor makes about 1 kg/day of salt seekers, and we take 1/5 of the 10 kg UF3 inventory with them, all diluted in 20 m^3 / 5 / 100 = 40 l of LiF. It will be liquid at 900°C, but pure LiF will freeze at 847. Any distillation process on FLiBe has to deal with this as the BeF2 is much lower boiling so you lose it.

I know this is extreme - about 1/10th this rate, in a single pass, would be more like it, but then all the 149-Sm will capture before you get it out. I was just wondering if it was possible at all. I think it is, but not remotely economic.

Heating power
5J/ml/K * 3000 ml/min * 300 K = 4,500,000 J/min = 75,000 J/sec = 75 kw. I rounded up to 100 kw, it's not accurate to 2 sig fig anyway.


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PostPosted: Jun 25, 2010 3:19 pm 
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It says a lot that it is even feasible. This is a tough way to earn 1% of our neutrons and I think we can afford to let the 149Sm get converted and keep the reactor simpler. Our salt seeker processing rate I think will be around 1% of the rate you proposed - which shows how feasible it would be.


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PostPosted: Jun 26, 2010 5:57 pm 
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Luke wrote:
Power requirements
Heating 3 l/min of FLiBe from 700°-1000°C takes about 100 kw. Boiling it takes a MW. Radiative losses will be significant, but I don't know the emissivity of the materials. A 20m^2 black body at 1000C loses nearly 3 MW, but shiny hastalloy should be far less, so I'm guestimating 2 MW all in, 10 kw/element.

Very different picture, I suspect, when normal reactor operating temperature is >1000°C and there's no FLiBe to muck up the works.....

Care to comment ? (Thnx)


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PostPosted: Jun 27, 2010 3:51 am 
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jaro wrote:
{Luke - 2MW power requirement}
Very different picture, I suspect, when normal reactor operating temperature is >1000°C and there's no FLiBe to muck up the works.....

Care to comment ? (Thnx)
No heating requirement, but that is trivial anyway compared to the latent heat. The ref up thread on UF4 gives the latent heat as 785 J/g at 1000C, so it's 13 kw do do 1kg/min. That's probably less than the decay heat.

Doesn't your high-temp proposal use UF3/UF4 eutectic, though? The UF3 will get left behind with the lanthanides. This is less of a problem for conventional designs with only 1% of the UF4 as UF3 - in fact, at moderate processing rates you just export the breeding gain with the fission products, then fish it back out at the central facility.


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PostPosted: Jun 27, 2010 11:03 am 
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Luke wrote:
jaro wrote:
{Luke - 2MW power requirement}
... That's probably less than the decay heat. ...


Now there is an idea! Can you use decay heat to provide most of the energy for the distillation? I don't know much about distillation systems but I think they are frequently designed recover a lot of the heat used to boil the input.


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PostPosted: Jun 27, 2010 2:23 pm 
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Posted: Jun 27, 2010 3:23 pm
We are talking apples and oranges here. Jaros proposal is premised on a very fast neutron spectrum which is ideal for burning U238 or ideally light water reactor waste. In fact, the UF3/UF4 eutectic is unsurpassed at the LWR waste burning mission; a uranium based waste elimination solution.

The HW-MSR can be highly optimized to meet the LWR waste consumption job over and above almost any other approach.

On the other hand, Kirk’s solution is a pure thorium approach that only produces a very small amount of waste compared to the HW-MSR uranium burner, mostly NP237, noble metals and rare earths.

The HW-MSR is primarily a fast reactor solution with a thermal blanket capability. Kirks approach is a mid neutron spectrum approach.

A completely thermal neutron spectrum deuterium slurry based approach could possibly use a wide collection of simple, well understood and straight forward particle separation strategies as a payoff for giving up fluoride salt chemistry.

Centrifuge particle separation, filtration, cold traps, getters, as well as chloride coolant chemical purification could make burning thorium in a thermal spectrum a lot easier.

Protactinium isolation is fast and simple when it is chlorinated since it boils off at 400 C. Samarium is pure of chloride at 700C. Most of the other rare earths will particulate out using the particle separation techniques at 1000C either before of after chlorination as appropriate.

From the fission recoil separation effect (FRSE), uranium and thorium will be largely confined in the slurry fuel form and therefore not be present in the waste stream, and only fission produces will be found in the lithium coolant; thus waste processing is greatly simplified.

Manufacture of the thorium, U232/U233 surrey will be a challenge because of U232 heavy radiation emissions. But keeping the surrey fuel form suspended in lithium is a common technology used widely in industry today.

High temperature cheap structural materials like nuclear qualified steels can then be used since there will be no fluoride or uranium corrosion problems to contend with.

Low corrosion means inexpensive silicon carbide (SiC) heat exchangers can be used and they will last a long time with long service lives and high availability.

One last thing to consider when thinking about this tradeoff, the problem of large amounts of graphite waste disposal could also be avoided.

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Last edited by Axil on Jun 28, 2010 11:51 am, edited 1 time in total.

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PostPosted: Jun 27, 2010 5:01 pm 
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Here is some info supporting the use of chlorides to separate most metals and rare earths from a slurry waste stream at low temperatures.

United States Patent
4,179,492
Kruesi
December 18, 1979
Quote:
Process for making rare earth metal chlorides
Abstract

A process for forming a metal chloride of a metal or its compound comprising forming a liquid fused salt bath mixture of at least two metal chlorides, the bath having the property of dissolving the formed metal chloride, and introducing the metal or compound into the liquid fused salt bath in the presence of chlorine to form the metal chloride and recovering the formed chloride from the liquid fused salt bath mixture. The metals which may be chlorinated are those from groups 1b, 2a, 2b, 3a, 3b, 4a, 5a and 8 of the periodic table and the rare earth metals. Compounds from which the metals may be chlorinated are the sulfides, oxides, carbonates and sulfates. Chlorine may be introduced as such or its source may be a chlorine donor such as ferric chloride or sulfur chloride. The chlorides for the liquid fused salt bath are those of alkali metals, alkaline earth metals, ammonia, zinc, and ferric iron. The chlorination can be performed within a temperature range of 150.degree. C.-1000.degree. C.

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PostPosted: Jun 30, 2010 7:28 am 
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785 J/g for liquid UF4, that means better volumetric heat capacity than FLiBe! I calculated around 5 J/CC @ 1300K. Impressive.

What about UF3, that should have even higher volumetric heat capacity right (higher melting temperature)?


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PostPosted: Jun 30, 2010 2:32 pm 
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Cyril R wrote:
785 J/g for liquid UF4, that means better volumetric heat capacity than FLiBe! I calculated around 5 J/CC @ 1300K. Impressive.

What about UF3, that should have even higher volumetric heat capacity right (higher melting temperature)?



In reference to our discussion in the other thread, lithium volumetric heat capacity is inconsequential in a lithium-thorium slurry approach.

The absence of fluoride in the pure uranium/thorium slurry will increase their volumetric heat capacity contribution by 25%; up from 785 J/g to an estimation of 981 J/g. With many tens of tons of these high dense metals in the reactor core, the contribution of pure lithium overall will be small. There is a good chance that taking fluoride salts out of the core will enable a smaller core size overall.

Furthermore, not having to use graphite will reduce the size of a lithium reactor to 1/3 the size of a graphite based fluoride thermal reactor. In the end, the Lithium Homogeneous Reactor may turn out to be the smallest liquid reactor configuration.

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PostPosted: Jul 01, 2010 3:09 am 
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Axil wrote:
Here is some info supporting the use of chlorides to separate most metals and rare earths from a slurry waste stream at low temperatures.

United States Patent
4,179,492
Kruesi
December 18, 1979
Quote:
Process for making rare earth metal chlorides


Forgive the question, but has this anything to do with chlorides fast breeders often cited in this site or it' s a totally different process that can be eventually implemented both in fluoride and chloride reactors ?


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