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Liquid Fluoride Thorium Reactor

by gcarlin
Tags: fluoride, liquid, reactor, thorium
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mesa
#163
Nov26-12, 07:46 PM
P: 546
Quote Quote by mheslep View Post
...Unlike enrichment of uranium, the chemicals in a LFTR will have strong gamma and beta emitters, and the process will necessarily be in close proximity to an operational reactor.
Okay, is it that the gamma and beta radiation will interfere with the chemical reactions? or are there problems with shielding for personel? both? or something else I am completely missing?
wizwom
#164
Nov26-12, 07:47 PM
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Quote Quote by mheslep View Post
I'm not sure why it must be so that material lasts the life of the reactor, when the design specifies the fluoride salt can be drained away from the fission core / moderator area at any time, allowing replacement of the core material (graphite?) at whatever schedule desired.

Yes there will need to be thorough certification process for material in contact with the salt (Hastelloy-N?), but then again that effort should be seen in the context of the conditions which the LFTR would replace: a PWR with 153 atm water at 300C and fuel reaching 600C in zircalloy, also w/ 10^15 n/cm^2/s.
I suppose one could do a tubesleeve system, although swelling issues are significant, and then replace the tubes when they seem to have lost cohesion. The lack of significant pressure will also alleviate the material concerns; you can be more brittle when your hoop stresses are lower.

Since ZrF4 was used as fluoride salt component in various MSRs, I'd expect Zircaloy is right out as a tubing material; The temperature range puts us into SiC or ZrC ranges; but they are non-ductile. I expect ODS alloys will be the likely tubing.

Of course, this is assuming we can't separate the corrosion resistance and ductility under radiation problems; if we can work out a reasonable method for SiC coating parts that stands up to radiation and thermal changes then almost all corrosion difficulties can be ignored, and the structural material can be chosen on retention of ductility alone.
mesa
#165
Nov26-12, 07:57 PM
P: 546
Quote Quote by wizwom View Post
I suppose one could do a tubesleeve system, although swelling issues are significant, and then replace the tubes when they seem to have lost cohesion. The lack of significant pressure will also alleviate the material concerns; you can be more brittle when your hoop stresses are lower.

Since ZrF4 was used as fluoride salt component in various MSRs, I'd expect Zircaloy is right out as a tubing material; The temperature range puts us into SiC or ZrC ranges; but they are non-ductile. I expect ODS alloys will be the likely tubing.

Of course, this is assuming we can't separate the corrosion resistance and ductility under radiation problems; if we can work out a reasonable method for SiC coating parts that stands up to radiation and thermal changes then almost all corrosion difficulties can be ignored, and the structural material can be chosen on retention of ductility alone.
Are there issues with using Hastelloy N that were not adressed during the running of the research reactor at ORNL or are these simply better choices (cost, durability, etc.) by comparison due to material advancement in the last half century?
mheslep
#166
Nov26-12, 10:51 PM
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Quote Quote by wizwom View Post
I suppose one could do a tubesleeve system, although swelling issues are significant, and then replace the tubes when they seem to have lost cohesion. The lack of significant pressure will also alleviate the material concerns; you can be more brittle when your hoop stresses are lower.

Since ZrF4 was used as fluoride salt component in various MSRs, I'd expect Zircaloy is right out as a tubing material; The temperature range puts us into SiC or ZrC ranges; but they are non-ductile. I expect ODS alloys will be the likely tubing.

Of course, this is assuming we can't separate the corrosion resistance and ductility under radiation problems; if we can work out a reasonable method for SiC coating parts that stands up to radiation and thermal changes then almost all corrosion difficulties can be ignored, and the structural material can be chosen on retention of ductility alone.
We may be talking about two different things.

In the case of a liquid molten salt reactor, it seems to me there are two primary materials to select. The first is the moderator, which will suffer the neutron flux, but has little structural support responsibility. The ONR experiment used a graphite block w/ channels through which the salt was pumped. I assume that's still the first choice for a moderator. The second material is for structural containment. It receives relatively small neutron flux, high radiation, and must structurally contain the ~700C salt. ONR used Hasteloy N.
mheslep
#167
Nov26-12, 11:59 PM
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Quote Quote by mesa View Post
Are there issues with using Hastelloy N that were not adressed during the running of the research reactor at ORNL or are these simply better choices (cost, durability, etc.) by comparison due to material advancement in the last half century?
Yes there's a problem that was recognized but not yet addressed (AFAIK) in an operation. It is mentioned in the video interview link you provided and on the MSR experiment wiki page. They found that Tellurium, a fission product, causes cracking presence of radioactivity in the alloy ONR used. This would not be trivial thing to test.
mesa
#168
Nov27-12, 06:07 PM
P: 546
Quote Quote by mheslep View Post
Yes there's a problem that was recognized but not yet addressed (AFAIK) in an operation. It is mentioned in the video interview link you provided and on the MSR experiment wiki page. They found that Tellurium, a fission product, causes cracking presence of radioactivity in the alloy ONR used. This would not be trivial thing to test.
Agreed, but certainly possible.

Here is another interview with Dick Engel where he discusses this exact problem, it rings deeper than just the Tellurium (which it seems the material scientists had a solution for)

http://www.youtube.com/watch?v=ENH-j...layer_embedded

The question is raised at 17:25 and goes to 20:56, although (once again) I really found the discussion as a whole very interesting.

I like Dick Engals take on how to test materials for future reactors, same link but starting at time frame 19:41.
mheslep
#169
Nov27-12, 07:32 PM
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Quote Quote by mesa View Post
...

The question is raised at 17:25 and goes to 20:56, although (once again) I really found the discussion as a whole very interesting.
Yes I'd seen it previously. I just watched it again and Engel raised an obvious point that I missed before. He points out that if the Te corrosion problem is solved, the overall corrosion problem may not be solved because another element might cause trouble. The larger point being that fission of course means a large chunk of the periodic table would be present, everything from gallium to hafnium, including the very reactive alkali and halogen groups. Does this mean a chemical analysis the interaction of most of the elements in the periodic table against Hasteloy N must be done under LFTR conditions?

One of the advantages of LFTR is supposed to be that high burnup and low waste is possible in part because fission poisons, esp. xenon, can be chemically removed from a liquid fueled reactor, unlike a solid fueled reactor which must have the fuel replaced every couple years. But while targeting the removal of some elements is surely feasible, I doubt it is so easy to remove most of the periodic table.

It may be that in the case of long term corrosion the issue turns in favor short turn fuel supplies, as while fission also generates products in the solid fuel Zirc alloy rods, they're pulled out of service while LFTR is intended to keep going for 30 years or so.
mheslep
#170
Nov27-12, 08:09 PM
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I've just read some of the critical final report on the Molten Salt Reactor Experiment written by the AEC in 70's after the shutdown. I knew of its existence but had avoided it given the politics of the time heavily favoring light water reactors and liquid metal breeders, and I thought it biased.

The structural material discussion starts on page 30. The argument seems valid, if overly absolute ("not suitable").

http://www.energyfromthorium.com/pdf/WASH-1222.pdf
Astronuc
#171
Nov28-12, 07:54 AM
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Quote Quote by mheslep View Post
I've just read some of the critical final report on the Molten Salt Reactor Experiment written by the AEC in 70's after the shutdown. I knew of its existence but had avoided it given the politics of the time heavily favoring light water reactors and liquid metal breeders, and I thought it biased.

The structural material discussion starts on page 30. The argument seems valid, if overly absolute ("not suitable").

http://www.energyfromthorium.com/pdf/WASH-1222.pdf
On page 32 of WASH-1222 is the statement:
"In addition to the intergranular corrosion problem, the standard Hastelloy-N used in the MSRE is not suitable for use in the MSBR because its mechanical properties deteriorate to an unacceptable level when subjected to the higher neutron doses which would occur in the higher power density, longer-life MSBR."

Note that MSRE was only 8 MW (without the electrical generation system, and with off-line batch processing), while the MSBR was planned for 2250 MWt (1000 MWe) and use of on-line continuous processing. See Table III, p. 21 of WASH-1222.

Nickel is problematic in any neutron environment. It absorbs neutrons and becomes active (producing Co-58 and some Co-60) and suffers from an (n,α) reaction. An alloy with lower Ni content would be preferable, something more along the lines of more Cr-Mo (Hastelloys are Ni-Cr-Mo).

Several other technical issues are mentioned. The MSBR concept proposes high temerpature steam cycle, and that presents a challenge, particularly with respect to the heat exchanger, which basically can't be allowed to fail (leak), and then there is the materials compatibility issues between the steam and salt loop. The chemical separation part of the plant would also be challenging. Storage of Xe, Kr, I would be challenging, as well as ultimate disposition of the other fission products (ostensibly they would be converted to oxides and vitrified).

A large scale LFTR would not be a trivial undertaking.
mheslep
#172
Nov28-12, 09:18 AM
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Quote Quote by Astronuc View Post
On page 32 of WASH-1222 is the statement:
"In addition to the intergranular corrosion problem, the standard Hastelloy-N used in the MSRE is not suitable for use in the MSBR because its mechanical properties deteriorate to an unacceptable level when subjected to the higher neutron doses which would occur in the higher power density, longer-life MSBR."

Note that MSRE was only 8 MW (without the electrical generation system, and with off-line batch processing), while the MSBR was planned for 2250 MWt (1000 MWe) and use of on-line continuous processing. See Table III, p. 21 of WASH-1222.

Nickel is problematic in any neutron environment. It absorbs neutrons and becomes active (producing Co-58 and some Co-60) and suffers from an (n,α) reaction. An alloy with lower Ni content would be preferable, something more along the lines of more Cr-Mo (Hastelloys are Ni-Cr-Mo).
Yes, apparently He production inside the hastelloy is also a concern. However, it seems to me the neutron flux can be held to some arbitrarily low limit for the outer, structural support holding the salt, where no fission need occur, and with an arbitrary amount of salt or other neutron stops been the graphite-salt-core and the containment. So before WASH stated the material was "unsuitable" without caveat they might have demonstrated how a large neutron flux on the containment was unavoidable.

Several other technical issues are mentioned. The MSBR concept proposes high temerpature steam cycle, and that presents a challenge, particularly with respect to the heat exchanger, which basically can't be allowed to fail (leak),
The high temperature (~700C) is different from a PWR, but I don't know that such temperatures are more challenging than those encountered by any existing Brayton cycle system (e.g. jet engine)

and then there is the materials compatibility issues between the steam and salt loop. The chemical separation part of the plant would also be challenging. Storage of Xe, Kr, I would be challenging, as well as ultimate disposition of the other fission products (ostensibly they would be converted to oxides and vitrified)....
Yes, removal and storage of most every element below U would be required over the lifetime of reactor. There must be some kind of overall chemical architecture to address that issue, as working piecemeal against each and every fission product and their daughter products seems intractable.
mesa
#173
Nov28-12, 09:59 AM
P: 546
Quote Quote by mheslep View Post
Yes I'd seen it previously. I just watched it again and Engel raised an obvious point that I missed before. He points out that if the Te corrosion problem is solved, the overall corrosion problem may not be solved because another element might cause trouble. The larger point being that fission of course means a large chunk of the periodic table would be present, everything from gallium to hafnium, including the very reactive alkali and halogen groups. Does this mean a chemical analysis the interaction of most of the elements in the periodic table against Hasteloy N must be done under LFTR conditions?

One of the advantages of LFTR is supposed to be that high burnup and low waste is possible in part because fission poisons, esp. xenon, can be chemically removed from a liquid fueled reactor, unlike a solid fueled reactor which must have the fuel replaced every couple years. But while targeting the removal of some elements is surely feasible, I doubt it is so easy to remove most of the periodic table.

It may be that in the case of long term corrosion the issue turns in favor short turn fuel supplies, as while fission also generates products in the solid fuel Zirc alloy rods, they're pulled out of service while LFTR is intended to keep going for 30 years or so.
What really matters is to what degree these elements formed, your statement is too broad and your link is for fissioning U235 not U233 but lets take a closer look anyway. You mentioned Gallium, pretty nasty stuff when in contact with most metals however it isn't even showing on your distribution scale for the graph you linked.

You had asked:
"Does this mean a chemical analysis the interaction of most of the elements in the periodic table against Hasteloy N must be done under LFTR conditions?"

No, its imperative to be more careful about allowing opinion to get in the way before looking at the facts.
mesa
#174
Nov28-12, 10:33 AM
P: 546
Quote Quote by mheslep View Post
I've just read some of the critical final report on the Molten Salt Reactor Experiment written by the AEC in 70's after the shutdown. I knew of its existence but had avoided it given the politics of the time heavily favoring light water reactors and liquid metal breeders, and I thought it biased.

The structural material discussion starts on page 30. The argument seems valid, if overly absolute ("not suitable").

http://www.energyfromthorium.com/pdf/WASH-1222.pdf
I have heard about this report, seems to be fairly well known amongst the LFTR community; their opinions of it are less than favorable but a closer look should be done before forming an opinion on this matter.

On page 30 I noticed the report starts by talking about how the ORN scientists wanted to 'freeze' the salt along the walls to prevent corrosion in the flourinator, seems like a pretty clever idea, but is it feasable? The author thinks no but this report was not written by the scientists actually working on the project who had experience working with similar techniques; Time frame 8:16:

http://www.youtube.com/watch?v=ENH-j...layer_embedded

So now come the questions, how hard would it be? how much energy does it use? what is the cost of a system like this? From my own experience in refrigeration I don't think this would be difficult to add on. What is your opinion?
mesa
#175
Nov28-12, 11:51 AM
P: 546
Quote Quote by Astronuc View Post
A large scale LFTR would not be a trivial undertaking.
No doubt, but does it look promising enough to justify further develop?
Do you see the materials for the reactor as being the largest obstacle?

Quote Quote by mheslep View Post
Yes, removal and storage of most every element below U would be required over the lifetime of reactor. There must be some kind of overall chemical architecture to address that issue, as working piecemeal against each and every fission product and their daughter products seems intractable.
I thought removal of these elements is part of the design with much talk (amongst advocates, so it should be investigated more thoroughly) about how they have high value for the industrial and research markets.

On another note, I don't see why we have to remove 'most every element below U' if their concentration is almost undectectable and there is little to no effect on the reactor itself. It would seem more important to concentrate on the elements that effect the lifecycle of the reactor i.e. materials longevity, efficiency, waste, etc.
mheslep
#176
Nov28-12, 01:32 PM
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Quote Quote by mesa View Post
What really matters is to what degree these elements formed,
Yes, though operation for 30 years means everything has time to accumulate, unlike in solid fuel reactors.

your statement is too broad and your link is for fissioning U235 not U233
Why is the slight difference between 233 and 235 products relevant to the point, which is that a broad swath of periodic table is dumped into the salt over time via fission products?


You had asked:
"Does this mean a chemical analysis the interaction of most of the elements in the periodic table against Hasteloy N must be done under LFTR conditions?"

No, ..
Why not? In addition swath of fission products, there are other paths for the introduction of elements in elemental form, including the elements from the salt itself - lithium, beryllium, fluorine - then higher Z elements formed from neutron capture of those elements, the decay daughter products of the fission products, carbon from the moderator, and so on.
mheslep
#177
Nov28-12, 01:41 PM
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I thought removal of these elements is part of the design with much talk (amongst advocates, so it should be investigated more thoroughly) about how they have high value for the industrial and research markets.
Yes, that is my understanding, that in fluid fueled reactors it is feasible to remove fission products so that, by removing Xenon via chemical processing, poisoning can be stopped allowing high burn-up. This capability is not feasible in solid fuel designs.

I'm suggesting that along with the advantage comes a problem. While the dispersal of fission products throughout the reactor makes them removable, if the chemical means are put in place, it also means the reactor structural containment must accommodate contact with all of those products which accumulate over long periods.

Quote Quote by mesa View Post
...
On another note, I don't see why we have to remove 'most every element below U' if their concentration is almost undectectable and there is little to no effect on the reactor itself. It would seem more important to concentrate on the elements that effect the lifecycle of the reactor i.e. materials longevity, efficiency, waste, etc.
Again, a light water reactor w/ solid fuels would have very similar fission products in the short term. The difference with MSRs is that the fuel salt stays in the reactor for the life of the reactor, as I understand it. So that in a solid fuel reactor the minor products might only accumulated at trace levels, while in the MSR they have 30 years to accumulate. After that much time would minor products still be "undetectable"? I don't know that to be the case.

PS One speculative idea that comes to mind: After a high fuel burnup, dump the salt, say, every ten years. The MSR is designed for this for safety reasons in any case. Give it some decay time (short because of the low concentration of actinides in a Thorium cycle), then bury/dispose?

The idea might be a step in the wrong direction, i.e. away from passive, walk away safety. As it implies a design that it a *dump* maintenance is neglected the structural containment is at threat of failure.
mesa
#178
Nov28-12, 02:03 PM
P: 546
Quote Quote by mheslep View Post
Yes, though operation for 30 years means everything has time to accumulate, unlike in solid fuel reactors.



Why is the slight difference between 233 and 235 products relevant to the point, which is that a broad swath of periodic table is dumped into the salt over time via fission products?


Why not? In addition swath of fission products, there are other paths for the introduction of elements in elemental form, including the elements from the salt itself - lithium, beryllium, fluorine, then higher Z elements formed from neutron capture of those elements, the decay daughter products of the fission products and so on.
That it does but for most of the elements it looks like that accumulation is still trivial even after 30 years. Do you know of a good source of data on fission byproducts that we could use to make actual calculations? Otherwise this is just a circular arguement.

As far as your graph link I was simply pointing out it was for the wrong fissile material, your new link is much better, thanks for posting it.
mheslep
#179
Nov28-12, 02:11 PM
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Quote Quote by mesa View Post
That it does but for most of the elements it looks like that accumulation is still trivial even after 30 years. Do you know of a good source of data on fission byproducts that we could use to make actual calculations? Otherwise this is just a circular arguement...
I think the information is roughly available from that products graph.

For instance, for every mole of U233 consumed, 2% of a mole of some fission product (with atomic weight 85) is produced, 7% Zr, 6% Cs and so on. Burn another mole of U233, get another 2%, 7%, 6%, ... which the remains in the reactor, unless it has a fast decay path thus becoming something else, or unless it happens to have a high neutron capture cross section thus becoming something else, ...
mesa
#180
Nov28-12, 03:06 PM
P: 546
Quote Quote by mheslep View Post

I think the information is roughly available from that products graph.

For instance, for every mole of U233 consumed, 2% of a mole of some fission product (with atomic weight 85) is produced, 7% Zr, 6% Cs and so on. Burn another mole of U233, get another 2%, 7%, 6%, ... which the remains in the reactor, unless it has a fast decay path thus becoming something else, or unless it happens to have a high neutron capture cross section thus becoming something else, ...
Well lets look back at your first link (similar enough to U233) since it shows a bit more of the dropoff at atomic masses of less than 75 at a rate of .0001% of all fissions and falling off drastically from there. Gallium (like you had mentioned earlier) is 69.723amu so what percentage of fission products produce this element? We need to be careful as well and take into account all isotopes.

With this data we can simply calculate the accumulation of this element of the course of say a 30 year life cycle based off of anticipated (MWt energy of a reactor)/(energy per fission)*time for a rough estimate.

Astronuc, can you point us in the direction of a source with more detail of the fission products from U233?
Nevermind, found it, here is a link for anyone interested in running some calculations:
http://www-nds.iaea.org/relnsd/vchart/


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