Energy From Thorium Discussion Forum

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PostPosted: Jul 16, 2012 9:43 pm 
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Current theory is that moderator life is a function only of fluence and temperature.
Assuming fixed temperature, you can only get so much energy out of a kg of graphite.
From an economic point of view, everything else being equal, high power density
and short moderator life is cheaper than low power density and long moderator life.
You want to turn over your expensive inventory of graphite rapidly,

Reactor grade graphite costs up to $20/kg. It seems to me that the way forward
is high power density, short moderator life, and after a decay period recycle. Re-sintering
moderately radioactive graphite can't be that big a problem.


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PostPosted: Jul 16, 2012 10:43 pm 
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Re-sintering of graphite is a worthwhile objective for reduction of nuclear waste. The real problem, however, is accumulation of SNF. Many US states and European countries have put the nuclear programs on hold due to this problem.The only possibility of reducing, or at any rate putting a stop to increased accumulation of SNF, is by burning it in fast reactors.
MSR is a good design of a uranium burner as it allows escape of fission products Xe and Kr from fuel which then goes to a higher burn up before requiring removal of remaining (solid) neutron poisons. It has to be combined with fast reactors and periodic removal of neutron poisons to burn up more of U-238.


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PostPosted: Jul 17, 2012 8:18 am 
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Fast reactors are not the only possibility of burning SNF unless you include the uranium. The problem with the SNF are the TRUs and fission products not the uranium. No reactor fast or slow will do anything significant to the fission products. A slow reactor can consume the vast majority of the TRUs but not all of them. So, an alternative to large numbers of fast reactors would be large numbers of slow reactors combined with a modest number of fast reactors. Both plans would work to eliminate TRUs to the limits of the separation process. My thought is that the limits of the separation process is likely measured in xx grams of TRUs per tonne of salt-seeking fission products so at the end of the day either fast only or slow+a few fast reactors will consume almost all the TRUs and leave the same amount of trace TRUs escaping into the fission product flow.

A fast reactor has the advantage that it will have an easier neutron budget with TRUs than a slow reactor will have. Another way to state this is that the TRUs will capture fewer neutrons than a slow reactor will. The second advantage is that a fast reactor can tolerate a higher concentration of fission products. A disadvantage is that a fast reactor requires more fissile inventory. A second disadvantage is a PR problem - the fact that recriticality is possible if a reactor gets severely damaged. While we might engineer it to make the possibility remote it will not be zero and that will feed the media frenzy.


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PostPosted: Jul 17, 2012 10:46 pm 
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Anything to do with energy is playing with fire-physically and metaphorically. Nuclear fire is more concentrated and intense. Fast reactor is even more concentrated and intense than thermal ones, with all the advantages and disadvantages. The balancing factors are:-
1. All the uranium and thorium could eventually be used as fuel. This could include the used LWR fuel, Depleted Fuel and Thorium due to high conversion ratios to fissile isotopes.
2. Possible use of non-volatile coolants and safer, low pressure cores.
3. Reducing core and plant size due to non-requirement of moderator.
4. Freedom from graphite waste.


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PostPosted: Jul 18, 2012 12:40 am 
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jagdish wrote:
Anything to do with energy is playing with fire-physically and metaphorically. Nuclear fire is more concentrated and intense. Fast reactor is even more concentrated and intense than thermal ones, with all the advantages and disadvantages. The balancing factors are:-
1. All the uranium and thorium could eventually be used as fuel. This could include the used LWR fuel, Depleted Fuel and Thorium due to high conversion ratios to fissile isotopes.
2. Possible use of non-volatile coolants and safer, low pressure cores.
3. Reducing core and plant size due to non-requirement of moderator.
4. Freedom from graphite waste.

A thermal reactor can use all the thorium as fuel. It can also use non-volatile coolants and low pressure cores. Core size is pretty small regardless and plant size is likely determined by other things so I see no size advantage for fast or slow reactors. So the advantages of a fast reactor come down to:
1. Can use 238Uranium as fuel, including used LWR fuel and depleted uranium.
2. Freedom from graphite waste.
3. Ability to consume the even TRUs.

The disadvantages are likely higher cost, tougher materials science, much higher PR damage when there is trouble at one. So, I prefer to make the volume of reactors thermal and figure out what to do with the graphite waste (about one pencil's worth for each person's electricity for a year). A few fast reactors (about 2%) should be built to consume the even TRUs. The depleted uranium and used fuel uranium can be put back into the mines from whence they came. We can use the tailings from rare earth mines for windmills ;)


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PostPosted: Jul 18, 2012 7:34 pm 
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Lars wrote:
The depleted uranium and used fuel uranium can be put back into the mines from whence they came.
I guess that's what we're doing with Thorium today (DoE also intends to apply this solution to ORNL's stock of U233, after denaturing with U238).

Realistically, national & international policies forbid use of highly enriched fissile materials in civilian nuclear applications -- nuclear reactors in particular.

Consequently, any reactor using non-fissile thorium will be fuelled with low-enriched uranium, as a means for providing the required fissile material.
That means > 80% U238.

So regardless of whether we use thorium or not, we will be dealing with the U238/Pu fuel cycle.

In fact, the thorium in this fuel cycle complicates the reprocessing.

On the other hand, molten salt reactor technology simplifies the reprocessing.

Conclusion: keep the latter, dump the former (until national & international policies are changed).


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PostPosted: Jul 18, 2012 11:31 pm 
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Most of world population lives in South, East and South-east Asia, who except for Japan and China are low energy consumers. They have to live by rules that allow their survival. China is already doing so. Once they get their fast reactors running, they may quickly scale it up like they have done for coal plants and thermal reactors.


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PostPosted: Jul 31, 2012 4:07 pm 
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So would beryllium oxide last longer than graphite or less long? Or do we just not know?


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PostPosted: Jul 31, 2012 5:38 pm 
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What is the BeO inside of? If it is exposed to the fluoride salts it won't last at all (FLiBe contains BeF2 after all). If it isn't exposed then the question is what is the container holding it made of and how long will that last.


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PostPosted: Aug 01, 2012 6:28 am 
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Lars, that's a good point, but:

http://en.wikipedia.org/wiki/Aircraft_N ... Propulsion

Quote:
The US Aircraft Reactor Experiment (ARE) was a 2.5 MW thermal nuclear reactor experiment designed to attain a high power density for use as an engine in a nuclear powered bomber. It used the molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, was moderated by beryllium oxide (BeO) [...]


So if you use a different salt you can use BeO without containment. But I wonder if if lasts longer than graphite would or less long.


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PostPosted: Aug 01, 2012 8:45 am 
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Francis,

Yes the first Aircraft Reactor did use "canned" BeO as the moderator. However it was never designed to operate more that a few days and even their end goal of a reactor for an aircraft was not really expected to last more than a couple months (fly around that long then land and swap out the reactor!). So the question is what to use to clad or can the BeO that would both last a long time and not consume too many neutrons. For a breeder design where neutrons are precious, this would be extremely hard to do, even something like Silicon Carbide would have a lot of silicon absorbing neutrons. For convertor designs where you have more neutrons to spare it likely is an interesting area to at least investigate. We know graphite works so well already though.

David LeBlanc


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PostPosted: Aug 02, 2012 4:23 am 
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David wrote:
Yes the first Aircraft Reactor did use "canned" BeO as the moderator. However it was never designed to operate more that a few days and even their end goal of a reactor for an aircraft was not really expected to last more than a couple months (fly around that long then land and swap out the reactor!). So the question is what to use to clad or can the BeO that would both last a long time and not consume too many neutrons. For a breeder design where neutrons are precious, this would be extremely hard to do, even something like Silicon Carbide would have a lot of silicon absorbing neutrons. For convertor designs where you have more neutrons to spare it likely is an interesting area to at least investigate. We know graphite works so well already though.

David LeBlanc


Not so sure about that, David. After all, zircalloy is more absorptive than silicon carbide, and CANDUs are the most neutron efficient reactors today...


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PostPosted: Aug 02, 2012 4:33 am 
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Francis wrote:
Lars, that's a good point, but:

http://en.wikipedia.org/wiki/Aircraft_N ... Propulsion

Quote:
The US Aircraft Reactor Experiment (ARE) was a 2.5 MW thermal nuclear reactor experiment designed to attain a high power density for use as an engine in a nuclear powered bomber. It used the molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, was moderated by beryllium oxide (BeO) [...]


So if you use a different salt you can use BeO without containment. But I wonder if if lasts longer than graphite would or less long.


It should last much longer, because of much less swelling. Eventually you get some tritium in the BeO. That would cause some cracks possibly. If you design the reactor to have mechanically decoupled moderator - eg Jaro's HW-MSR concept - then there's not much structural demand on the moderator. It just has to "sit there". So it can be cracked and still be good for its function.

BeO is a quite nice moderator. Better than graphite. It actually produces more neutrons than it absorbs itself, oddly. So it adds slightly to your breeding ratio (graphite deteriorates it).


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PostPosted: Aug 02, 2012 10:17 am 
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djw1 wrote:
Current theory is that moderator life is a function only of fluence and temperature.
Assuming fixed temperature, you can only get so much energy out of a kg of graphite.
From an economic point of view, everything else being equal, high power density
and short moderator life is cheaper than low power density and long moderator life.
You want to turn over your expensive inventory of graphite rapidly,

Reactor grade graphite costs up to $20/kg. It seems to me that the way forward
is high power density, short moderator life, and after a decay period recycle. Re-sintering
moderately radioactive graphite can't be that big a problem.


It's not just sintering. I believe it will need to be cooked for a while in vacuum to get any adsorbed xenon and krypton out and then re-densified with chemical vapor deposition to get the outer surface to be smooth and nonporous again.

Cooking radioactive stuff in high pressure flammable gas? Sounds like fun. Should be feasible, but it's not going to be too cheap.


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PostPosted: Aug 02, 2012 10:33 am 
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Cyril R wrote:

It should last much longer, because of much less swelling. Eventually you get some tritium in the BeO. That would cause some cracks possibly. If you design the reactor to have mechanically decoupled moderator - eg Jaro's HW-MSR concept - then there's not much structural demand on the moderator. It just has to "sit there". So it can be cracked and still be good for its function.

BeO is a quite nice moderator. Better than graphite. It actually produces more neutrons than it absorbs itself, oddly. So it adds slightly to your breeding ratio (graphite deteriorates it).

Cyril, I don't think BeO can exist in contact with the salt. So the question isn't how long the BeO lasts but how long the container it is put in will last. That containment is now in the middle of the highest flux which is quite a different story than the hastalloy at the boundary of the core.


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