Energy From Thorium Discussion Forum

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PostPosted: Dec 02, 2010 3:00 pm 
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If we wanted to have a decent mass of U-233, couldn't we progressively breed more and more from Thorium? So, for instance, if I had a neutron source which was very small, but concentrated - from, say, Radium or a neutron lens, could I then place it in a small reflective sphere with a gram (whatever the mass is here) of thorium, and within 30 days, half of the thorium has become U-233, which is then placed along with the radium in a larger neutron-reflective sphere with more thorium, and in 30 days that has before half U-233, and a quarter of the initial thorium load has also become U-233, and repeat this process until enough U-233 was obtained? The neutron efficiency of thorium -> U-233 is 2.4 in theory, so in practical applications could a 2.0 neutron efficiency be hoped for?


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PostPosted: Dec 02, 2010 3:11 pm 
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2.4 would be the number of neutrons produced in an average fission of 233U.
You need to use one of these to cause a fission.
You need to use another to transmute a thorium atom into a 233U atom to replace the one you destroyed with the fission.
That leave you a budget of 0.4 neutrons per fission to spend on:
neutrons absorbed in the fuel salt and fission products
neutrons that escape the core and get absorbed in the reactor structure
neutrons that get absorbed by the 233Pa before it gets to decay to 233U
neutrons that get absorbed by 233U but do not cause a fission.

Finally, with any left over neutrons you can breed additional fuel (beyond replacement of the stuff you burned).

Best case you might get 0.2 neutrons for this purpose. More commonly paper designs show around 6% breeding gain. I think a good target is just breaking even.


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PostPosted: Dec 02, 2010 3:39 pm 
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OK Lars, what you're saying makes sense. I had misunderstood the meaning of the neutron efficiency figure. I had thought it was a NET generation of neutrons, when it is a GROSS figure. So, only .4 is left over, and most of that lost due to inefficiencies resulting from absorbtion of the walls, and so forth. But what if some good engineering were able to reduce the inefficiency so that 1/4 of what was left over, i.e. 1/10 neutron, was available. What would this do for doubling time? I'll have to sit down and calculate it, unless someone else on this forum can do it more quickly. What would the addition of neutrons from the lens (a LARGE one, placed in a neutron-rich environment, such as over the monazite sands of Idaho) provide?

But something occurs to me, that there may be people who would not like such a technology available, because of proliferation issues. This concern might kill development. It might even kill this discussion.


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PostPosted: Dec 02, 2010 5:51 pm 
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RWKeyes wrote:
OK Lars, what you're saying makes sense. I had misunderstood the meaning of the neutron efficiency figure. I had thought it was a NET generation of neutrons, when it is a GROSS figure. So, only .4 is left over, and most of that lost due to inefficiencies resulting from absorbtion of the walls, and so forth. But what if some good engineering were able to reduce the inefficiency so that 1/4 of what was left over, i.e. 1/10 neutron, was available. What would this do for doubling time? I'll have to sit down and calculate it, unless someone else on this forum can do it more quickly.

Typical doubling times are 10 to 50 years. Lots of issues here. I argue that if we can provide the start-up fissile using mined 235U then this is much cheaper and faster to deploy.

Quote:
What would the addition of neutrons from the lens (a LARGE one, placed in a neutron-rich environment, such as over the monazite sands of Idaho) provide?

But something occurs to me, that there may be people who would not like such a technology available, because of proliferation issues. This concern might kill development. It might even kill this discussion.

You need to run some numbers - how many neutrons per second are produced by the neutron-rich monazite sands ? How many atoms of 233U do you need for start-up?
I think you will find that it will take too long to generate fissile this way.


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PostPosted: Dec 02, 2010 6:10 pm 
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Yea, 10-50 years is a long doubling time. Especially on the upper end of the scale. It does seem that using mined & processed uranium for ignition of the thorium reactor is more practical than taking 120 years to breed up a decent supply.

For the neutron lenses, I have difficulty in coming up with figures. There's not a lot I can find out about these things. I'd guess that I'd make one as large as possible, given constraints for precision and cost. I'd guess a ten acre size if it's just steel in a fresnel lens formation. There would have to be several different types of construction, with the larger outer portion made of less costly, less precise construction and inner rings being made of more precise and higher cost construction. The advantage of the lens idea is that it's fairly simple and solid state, and consumes no fuel in operation.


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PostPosted: Dec 02, 2010 6:25 pm 
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This isn't going to get you anywhere. Off hand guess you would need 1 tonne of 233U or 1e6 grams or 1e6/233 moles which multiplied by 6.02e23 gives you the number of atoms you need.
2.6 e27 atoms.

Supposing the field generated 2.6e4 neutrons per second, then you would be talking 1e23 seconds to get enough to start one reactor or something like 1e16 years.

I'll let you try to dig up how many neutrons per second you can produce but I'm certain you will end up with a time scale that is out of this world.


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PostPosted: Dec 04, 2010 5:17 am 
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Lars wrote:
This isn't going to get you anywhere. Off hand guess you would need 1 tonne of 233U or 1e6 grams or 1e6/233 moles which multiplied by 6.02e23 gives you the number of atoms you need.
2.6 e27 atoms.

Supposing the field generated 2.6e4 neutrons per second, then you would be talking 1e23 seconds to get enough to start one reactor or something like 1e16 years.

I'll let you try to dig up how many neutrons per second you can produce but I'm certain you will end up with a time scale that is out of this world.


If you have one ton of curium you have about 1e13 neutrons per second. Still way too long. Plus curium is scary and expensive.

You need a chain reaction to make U233 effectively.


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PostPosted: Dec 04, 2010 5:23 am 
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Hmm. You wouldn't happen to have some californium-252 on you, would you now?

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

2e12 n/s/g. Of course this is very, very scary, and Cf-252 is very, very rare (this is a good thing).


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PostPosted: Dec 17, 2010 12:11 am 
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Cyril R wrote:
Lars wrote:
This isn't going to get you anywhere. Off hand guess you would need 1 tonne of 233U or 1e6 grams or 1e6/233 moles which multiplied by 6.02e23 gives you the number of atoms you need.
2.6 e27 atoms.

Supposing the field generated 2.6e4 neutrons per second, then you would be talking 1e23 seconds to get enough to start one reactor or something like 1e16 years.

I'll let you try to dig up how many neutrons per second you can produce but I'm certain you will end up with a time scale that is out of this world.


If you have one ton of curium you have about 1e13 neutrons per second. Still way too long. Plus curium is scary and expensive.

You need a chain reaction to make U233 effectively.


You guys are most likely right. But I'd still like to know about a neutron lenses. If anyone has some info, please post.


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PostPosted: Dec 17, 2010 12:42 am 
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Lars wrote:
This isn't going to get you anywhere. Off hand guess you would need 1 tonne of 233U or 1e6 grams or 1e6/233 moles which multiplied by 6.02e23 gives you the number of atoms you need.
2.6 e27 atoms.

Supposing the field generated 2.6e4 neutrons per second, then you would be talking 1e23 seconds to get enough to start one reactor or something like 1e16 years.

I'll let you try to dig up how many neutrons per second you can produce but I'm certain you will end up with a time scale that is out of this world.


I am trying to calculate the actual number of neutrons occurring at ambient conditions and can only get this measured in Sieverts: 40.50 nSv/h. That appears to be nanosieverts per hour. I don't know how to convert this into an actual number of neutrons, because Sv seems to be an indication of radiation exposure dose, which I am not sure but I believe that would be related to the energy of the neutron involved, and not their number. Does anyone know more about this?

Interesting is that in Ramsar, Iran the background radiation is 260 mSv yearly...


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PostPosted: Dec 17, 2010 2:50 am 
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1 Sv = 100 rem and yes it is a dose rate. To calculate the number of neutrons you would need to understand much about the radiation source. But you aren't going to generate fissile by placing thorium in a naturally radioactive area - otherwise we would find plutonium in uranium deposits.


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PostPosted: Jan 05, 2011 4:05 am 
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Some interesting thoughts here but nothing that says "Pick me !" I like two potential routes for breeding U233, one seems possible, the other is a very long way away.

Option 1 CANDU
By having some fuel elements in the fuel bundles with thorium inside, these are irradiated in the core and breeding occurs. The ability to vary the contents inside the the fuel bundles to have a higher enrichment allows the fertile/fissile balance to be maintained. Burning recycled Pu and getting U233 out of the process sounds like a very appealing swap to me, but pretty messy compared to once through fuelling on natural U with zero reprocessing.

Option 2 Molten Chloride Fast Reactors (MCFR)
MCFR's potentially have many spare neutrons that could be harnessed to breed U233 or do other useful things, but these reactors seem to have very few people interested in them and no programmes pursuing their development. Breeding ratios of 1.4 appear feasible, so that would see a a 1 GWe unit producing about 400 kg U233/year, a very useful quantity.

The trick is not to put the U233 into another MCFR, but into a soft spectrum core that needs modest quantities of fissile. Or use it to support a DMSR core that has been started on 20% LEU.


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PostPosted: Jan 05, 2011 11:34 am 
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Lindsay wrote:
Some interesting thoughts here but nothing that says "Pick me !" I like two potential routes for breeding U233, one seems possible, the other is a very long way away.

Option 1 CANDU
By having some fuel elements in the fuel bundles with thorium inside, these are irradiated in the core and breeding occurs. The ability to vary the contents inside the the fuel bundles to have a higher enrichment allows the fertile/fissile balance to be maintained. Burning recycled Pu and getting U233 out of the process sounds like a very appealing swap to me, but pretty messy compared to once through fuelling on natural U with zero reprocessing.

Option 2 Molten Chloride Fast Reactors (MCFR)
MCFR's potentially have many spare neutrons that could be harnessed to breed U233 or do other useful things, but these reactors seem to have very few people interested in them and no programmes pursuing their development. Breeding ratios of 1.4 appear feasible, so that would see a a 1 GWe unit producing about 400 kg U233/year, a very useful quantity.

The trick is not to put the U233 into another MCFR, but into a soft spectrum core that needs modest quantities of fissile. Or use it to support a DMSR core that has been started on 20% LEU.


Yes, the CANDU approach could be interesting with the existing fleet. Depending on enrichment costs (currently pretty cheap) there is in general an economic case for running slightly enriched uranium (SEU) in current CANDUs to get longer burnups and thus lower fabrication costs and less strain on the fueling machines. However, when you do this the central fuel elements in each bundle end up being too shielded from thermal neutrons to be of much use for power production so a logical choice is for the central elements to be thorium and outer ones SEU (to also balance out reactivity, i.e. enough fertile since there is more U235 present). Makes a pretty simple way to produce U233 (perhaps denatured). Not sure how much one could produce annually with the limited world fleet of CANDUs but it is likely a better fit than trying the same thing with LWRs which the French group currently propose as their future source of U233 for startups. The problem with LWRs is that thorium's absorption cross section is much lower than U238 so you don't get much in situ production. Conversely, with CANDUs colder neutrons thorium has a much larger cross section and can effectively suck up neutrons even in the lower flux of the central pin positions.

I'm not much a fan of developing chloride based reactors just to produce start charge fissile. I just think there are way too many other choices to justify the extra expense and new unknowns of chloride salts and very fast spectrums.

David LeBlanc


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PostPosted: Oct 24, 2011 12:58 pm 
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Lindsay wrote:
Option 2 Molten Chloride Fast Reactors (MCFR)
MCFR's potentially have many spare neutrons that could be harnessed to breed U233 or do other useful things, but these reactors seem to have very few people interested in them and no programmes pursuing their development. Breeding ratios of 1.4 appear feasible, so that would see a a 1 GWe unit producing about 400 kg U233/year, a very useful quantity.

The trick is not to put the U233 into another MCFR, but into a soft spectrum core that needs modest quantities of fissile. Or use it to support a DMSR core that has been started on 20% LEU.


The Indians are planning something similar.

Essentially they have a lot of separated plutonium, both from their weapons programmes and their civilian CANDU fleet. They plan to use this as the core material in sodium cooled fast breeder reactors, with thorium blankets , thus producing large quantities of U-233, which can then be used in their Advanced Heavy Water Reactors.

While this whole cascade of reactor technologies is kinda messy, and relies on fairly expensive sodium cooled reactors, it has a number of advantages:

1) They don't need to use the plutonium in a thermal spectrum , thus eliminating concerns with the accumulation of curium.

2) Sodium cooled reactors can have very high breeding ratios, allowing them to produce large quantities of U-233 for use in much cheaper thermal thorium reactors. This also counters the capital cost penalty of sodium cooled reactors, since only a few will be needed to produce thorium for much cheaper thermal designs.

3) Because there is no need for the fast reactor core to have a positive breeding ratio for plutonium ( the plutonium is produced by CANDU reactors ) , they can be optimised for destruction of TRU

4) The U-233 produced in a fast spectrum will be virtually free of Plutonium , U-238 and minor actinides, making it ideal for use in a thermal reactor.

5) The scheme provides for very efficient destruction of India's legacy waste, thus enabling a speedy transition to a thorium-based fuel cycle.

6) There is no need for uranium enrichment in any part of the fuel cycle. In fact, the only isotope separation required is the production of heavy water.


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PostPosted: Oct 25, 2011 10:03 am 
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If 20% LEU, used for experimental reactors and not treated as bomb material, is filled in 6 0f 37 pins of a CANDU fuel bundle in a multiple seed-blanket arrangement, the remaining 31 metallic thorium pins can be electro-refined for 233U. Indians have tried some thorium bundles without fissile feed in newly commissioned reactors as a power flattening arrangement. They have been subsequently processed for 233U. With focus on 233U breeding, the bundles can be designed accordingly.


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