Liquid Fluoride Thorium Reactor

 Quote by mheslep 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.

Recognitions:
Gold Member
 Quote by Astronuc 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.

 Quote by mheslep 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.

"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.

 Quote by mheslep 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:

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?

 Quote by Astronuc 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 by mheslep 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.

Recognitions:
Gold Member
 Quote by mesa 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.

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.

Recognitions:
Gold Member
 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 by mesa ... 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.

 Quote by mheslep 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.

Recognitions:
Gold Member
 Quote by mesa 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, ...

 Quote by mheslep 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/

Recognitions:
Gold Member
 Quote by mesa ... 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. ...

PWR typical burnup is around 50 GWdays/ton, or 5% of the fuel. Up to 500 GWdays/ton is expected in an experimental reactor, says the wiki. LFTR supposedly will have very high burnup, so optimistically assume 500 GWdays/ton, or ~120GWdays per 1000 moles of U, or given a 33% efficient reactor, 40GWe-days/1000 moles, or ~11GWe-years/1e5 moles U.

So for every 11 years of operation, and again following the fission products curve, a 1GWe reactor produces 7e3 moles of Zr, 6e3 moles of Cs, etc, for the high probability products. Or, all products with amu's from 82 to 105, and 127 to 150 would accumulate 5e2 moles, or higher, in 11 years. Those concentrations will change through decay or neutron capture.

The consequence of the result would depend on chemistry of the particular element in contact with the alloy which is beyond me.

 Quote by Astronuc 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.
 Quote by mheslep PWR typical burnup is around 50 GWdays/ton, or 5% of the fuel. Up to 500 GWdays/ton is expected in an experimental reactor, says the wiki. LFTR supposedly will have very high burnup, so optimistically assume 500 GWdays/ton, or ~120GWdays per 1000 moles of U, or given a 33% efficient reactor, 40GWe-days/1000 moles, or ~11GWe-years/1e5 moles U. So for every 11 years of operation, and again following the fission products curve, a 1GWe reactor produces 7e3 moles of Zr, 6e3 moles of Cs, etc, for the high probability products. Or, all products with amu's from 82 to 105, and 127 to 150 would accumulate 5e2 moles, or higher, in 11 years. Those concentrations will change through decay or neutron capture. The consequence of the result would depend on chemistry of the particular element in contact with the alloy which is beyond me.
Interesting approach, I did it this way using Astronuc's thermal value above for a commercial generating facility of 2250MWt:
2250MWtx24hoursx365daysx30years/((MeV per fission)x(4.4504902416667x10^(-17))) = total number of fissions for the life cycle of the reactor. From here we can just multiply by the Cumulative Fission Yields to get:

4.9x10^21 Ga atoms produced, or .0081mols
Using your method I get .0025mols Ga for the same time frame.

If we are correct Gallium will not be an issue. Granted we could also account for Ga production from U235 since small amounts will also appear in this reactor but that lowers our values since they are an order of magnitude less in production of Ga in the thermal spectrum. Also, as Astronuc pointed out in the other thread, 8-10% of fission in LFTR will be fast neutrons, however this value is comparitively insignificant as well for this particular case.

 Quote by mheslep PWR typical burnup is around 50 GWdays/ton, or 5% of the fuel. Up to 500 GWdays/ton is expected in an experimental reactor, says the wiki. LFTR supposedly will have very high burnup, so optimistically assume 500 GWdays/ton, or ~120GWdays per 1000 moles of U, or given a 33% efficient reactor, 40GWe-days/1000 moles, or ~11GWe-years/1e5 moles U. So for every 11 years of operation, and again following the fission products curve, a 1GWe reactor produces 7e3 moles of Zr, 6e3 moles of Cs, etc, for the high probability products. Or, all products with amu's from 82 to 105, and 127 to 150 would accumulate 5e2 moles, or higher, in 11 years. Those concentrations will change through decay or neutron capture. The consequence of the result would depend on chemistry of the particular element in contact with the alloy which is beyond me.
We should go visit Borek in the Chemistry section and see what his thoughts are on this.

As for the remainder, calculations for the rest of the elements produced along with their constituent isotopes (and variations) would be helpful but improvement is needed on how calculations are performed to get decent sig figs.

Any thoughts?

 Admin * Independent fission yield (%): number of atoms of a specific nuclide produced directly (not via radioactive decay of precursors) in 100 fission reactions * Cumulative fission yield (%): total number of atoms of a specific nuclide produced (directly and via decay of precursors) in 100 fission reactions From http://www-nds.iaea.org/publications...ecdoc-1168.pdf These may not include activation (n-capture). -------------------------------------------------- Fission product pairs for U (Z, 92-Z; A, 234-A for U235 or 232-A for U233), assuming 2 neutrons released per fission. The neutrons affect A, not Z. Code: Z A 92-Z 234-A for U-235; 232-A for U-233 63 Eu 29 Cu 62 Sm 30 Zn 61 Pm 31 Ga 60 Nd 32 Ge 59 Pr 33 As 58 Ce 34 Se 57 La 35 Br 56 Ba 36 Kr 55 Cs 37 Rb 54 Xe 38 Sr 53 I 39 Y 52 Te 40 Zr 51 Sb 41 Nb 50 Sn 42 Mo 49 In 43 Tc 48 Cd 44 Ru 47 Ag 45 Rh 46 Pd 46 Pd -------------------------------------------------- Another factor to consider is the delayed neutron precusors that leave the core. Delayed neutrons are important with respect to control the reactor as well as irradiating the structure and piping outside the core. Reactivity control is another consideration, so a large MSBR may require use of control elements. The graphite must be supported, so there is a core support plate (not graphite), which will receive a neutron flux.Differences in thermal expansion between graphite and the structural alloy will have to be investigated. Hideout of the molten salt could be an issue. Note the MSRE operated 4 years and surface defects of 7 mils were found. Larger defects may propagate. Also, a 40 to 60 year lifetime is preferable. The numerous technical issues should be listed and discussed separately.

 Quote by Astronuc * Independent fission yield (%): number of atoms of a specific nuclide produced directly (not via radioactive decay of precursors) in 100 fission reactions * Cumulative fission yield (%): total number of atoms of a specific nuclide produced (directly and via decay of precursors) in 100 fission reactions From http://www-nds.iaea.org/publications...ecdoc-1168.pdf These may not include activation (n-capture). -------------------------------------------------- Another factor to consider is the delayed neutron precusors that leave the core. Delayed neutrons are important with respect to control the reactor as well as irradiating the structure and piping outside the core. Reactivity control is another consideration, so a large MSBR may require use of control elements. The graphite must be supported, so there is a core support plate (not graphite), which will receive a neutron flux.Differences in thermal expansion between graphite and the structural alloy will have to be investigated. Hideout of the molten salt could be an issue. Note the MSRE operated 4 years and surface defects of 7 mils were found. Larger defects may propagate. Also, a 40 to 60 year lifetime is preferable. The numerous technical issues should be listed and discussed separately.
Agreed.

I recieved an email from FliBe energy giving a link to the pdf files of the ORNL research program on the MSR. There is a substantial amount of information:

http://energyfromthorium.com/pdf/

Recognitions:
Gold Member
 Quote by Astronuc Reactivity control is another consideration, so a large MSBR may require use of control elements.
The MSRe had a *negative* temperature reactivity coefficient. The salt expands with temperature, density falls, reactivity falls. Is there some reason that control method must change with large reactor?

 Quote by mheslep The MSRe had a *negative* temperature reactivity coefficient. The salt expands with temperature, density falls, reactivity falls. Is there some reason that control method must change with large reactor?
Here is Chris Holdens reason for it @6:16 in his presentation for his reactor design, here is a link: