## Liquid Fluoride Thorium Reactor

 Quote by Astronuc 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.
Rusty Holden had an interesting idea about a different moderator @ 3:12:

What is 'hideout'? Are you referring to areas in the reactor where flow rates of the salt drop significantly?

"Also, a 40 to 60 year lifetime is preferable."
That would seem reasonable.

Recognitions:
Gold Member
 Quote by mesa Here is Chris Holdens reason for it @6:16 in his presentation for his reactor design, here is a link: http://www.youtube.com/watch?v=ZbtVk8r6-3U ... This is already a proven technology and it would seem the question is whether it is needed or not; it is reasonable to assume regulatory agencies could insist on such measures as they are a standard today even if shown to be unneccesary for LFTR.
Regulatory agencies could insist on anything they like, just because that's the way it has been done. But that's not technically relevant. No MSR is going to see approval in the US by the NRC for decades to come. The design will have to be built abroad, so I don't see tailoring a design to NRC inertia without valid technical reasons, driving up cost, as particularly wise.

 Quote by mesa 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?
It sounds stupid, wasteful in terms of energy on the one hand and on the other I do not see how you could control the quality of the salt layer. The interface would surely see a lot of stress and cracks and whatnot. Would they propagate to the walls? How could you tell if they did? And so on.

 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?
The negative temperature and void coefficients are useful for limiting a reactivity excursion, which is the case in LWRs. However, they are not suitable for power maneuvering a reactor. The delayed neutrons determine the period or rate at which power increases for a given insertion of positive reactivity (e.g., increase in fuel enrichment or removal of a neutron poison). The objective is to maintain control of the power level, and to avoid a rapid increase in reactor power.

Another matter to consider is the guide structure in the core. Control rods are positioned at the edge of the core for rapid insertion. The control rod and guide structure materials must be able to resist the high fluence and fluoride salt interaction.

A lot of the issues mentioned in this thread are also being explored in the Gen-IV MSR program.
 Recognitions: Gold Member As I recall the ONR MSR ~7MWth experiment mainly used load following to control the reactor. Increase the load which removes heat faster, the salt cools, reactivity increases to meet the load.

 Quote by zapperzero It sounds stupid, wasteful in terms of energy on the one hand and on the other I do not see how you could control the quality of the salt layer. The interface would surely see a lot of stress and cracks and whatnot. Would they propagate to the walls? How could you tell if they did? And so on.
One of the big issues with this type of reactor is the materials reacting with the salt and byproducts of fission; keep in mind that rates of reaction go up drastically with temperature and solids are no where near as reactive as liquids so this idea, (that came from the scientists at ORNL/MSR), seems to have some validity.

Either way we should look through the documents first to see what their proposed approach was before attempting to invalidate/validate this idea with arguement. Here is the link if you missed it:

http://energyfromthorium.com/pdf/

 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
Okay, thank you.

 Quote by Astronuc 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 --------------------------------------------------
This information is very useful but just for clarification what column is Z and which is A, or are the columns just not lined up?

 Quote by mesa This information is very useful but just for clarification what column is Z and which is A, or are the columns just not lined up?
The Z is over the atomic number (number of protons in the nucleus). The A and 234-A are over the letters designating the element (nuclide) corresponding to the Z.

If one fission produces Eu (Z=63, A=158) then the other fission product is necessarily Cu (Z=29, A = 234-158 = 76) + 2 neutrons. If Eu-159 was the fission product, then Cu-75 would be the other fission product + 2 neutrons. If 3 neutrons are released during fission, then the pair would be Eu-158, Cu-75 or Eu-159, Cu-74.

When U-233/U-235 absorbs a neutron and becomes an excited U-234/U-236 nucleus and fissions, the atomic numbers of the fission products, Z1 and Z2 must sum to 92 (or Z, 92-Z). The atomic numbers, A1 and A2, sum to 232/234 if 2 fission (prompt) neutrons are released (or A2 = 232-A1, or 234-A1), or 231/233 if 3 fission (prompt) neutrons are released. Some fission products release 'delayed' neutrons as well - usually fractions of a second up to 60 to 80 seconds later. The fraction of delayed neutrons with U-233 is less than for U-235.

 Quote by Astronuc The Z is over the atomic number (number of protons in the nucleus). The A and 234-A are over the letters designating the element (nuclide) corresponding to the Z. If one fission produces Eu (Z=63, A=158) then the other fission product is necessarily Cu (Z=29, A = 234-158 = 76) + 2 neutrons. If Eu-159 was the fission product, then Cu-75 would be the other fission product + 2 neutrons. If 3 neutrons are released during fission, then the pair would be Eu-158, Cu-75 or Eu-159, Cu-74. When U-233/U-235 absorbs a neutron and becomes an excited U-234/U-236 nucleus and fissions, the atomic numbers of the fission products, Z1 and Z2 must sum to 92 (or Z, 92-Z). The atomic numbers, A1 and A2, sum to 232/234 if 2 fission (prompt) neutrons are released (or A2 = 232-A1, or 234-A1), or 231/233 if 3 fission (prompt) neutrons are released. Some fission products release 'delayed' neutrons as well - usually fractions of a second up to 60 to 80 seconds later. The fraction of delayed neutrons with U-233 is less than for U-235.
Okay, I understand; I thought your chart represented something else, but it is still good for quick reference.

I would like to put together a data table on fission products that have high cross sectional areas for capturing thermal neutrons in the Th/U233 breeder cycle and see which are of biggest concern (like zenon 135).

It would also be good to run through the fission products and see which will have a high likelyhood for rate of reactivity/concentration (like tellurium) with the Hastelloy N. This part will likely prove difficult to compute without experimentation; hopefully there is sufficient information in the ORNL documents.
 Admin One would have to do some calculations based on flux and fuel composition, or find detailed tables that list specific nuclides and their decay chains, for example - Ba147 -> La147 -> Ce147 -> Pr147 -> Nd147 -> Pm147 -> Sm147 (stable), but each nuclide can absorb a neutron (but with different cross sections). Sm is a moderate neutron poison. And there are heavier nuclides, e.g., Pm155 -> Sm155 -> Eu155 -> Gd155, where Eu and Gd are stronger neutron poisons, but their fractional yields are quite low. Meanwhile, these can provide some idea of the FP vector. http://www.doitpoms.ac.uk/tlplib/nuc..._processes.php http://en.wikipedia.org/wiki/File:Th...ssionYield.svg http://en.wikipedia.org/wiki/Fission...ts_(by_element) http://en.wikipedia.org/wiki/Fission....2C_127_to_132 http://en.wikipedia.org/wiki/Fission..._152.2C_154.29 http://upload.wikimedia.org/wikipedi...nYield.svg.png There are preferred nuclides, i.e., those with high yield fractions. Also of interest - http://en.wikipedia.org/wiki/Fluoride_volatility

 Quote by Astronuc One would have to do some calculations based on flux and fuel composition, or find detailed tables that list specific nuclides and their decay chains, for example - Ba147 -> La147 -> Ce147 -> Pr147 -> Nd147 -> Pm147 -> Sm147 (stable), but each nuclide can absorb a neutron (but with different cross sections). Sm is a moderate neutron poison. And there are heavier nuclides, e.g., Pm155 -> Sm155 -> Eu155 -> Gd155, where Eu and Gd are stronger neutron poisons, but their fractional yields are quite low. Meanwhile, these can provide some idea of the FP vector. http://www.doitpoms.ac.uk/tlplib/nuc..._processes.php http://en.wikipedia.org/wiki/File:Th...ssionYield.svg http://en.wikipedia.org/wiki/Fission...ts_(by_element) http://en.wikipedia.org/wiki/Fission....2C_127_to_132 http://en.wikipedia.org/wiki/Fission..._152.2C_154.29 http://upload.wikimedia.org/wikipedi...nYield.svg.png There are preferred nuclides, i.e., those with high yield fractions. Also of interest - http://en.wikipedia.org/wiki/Fluoride_volatility
Yes, this will take some time.

It shold be fairly straightforward to find the products that need the most attention, we need to set up a formula to account for concentration (based on fission products/decay chains) and 'poisoning/absorbance' via cross secions, pretty straight forward.

To keep things simple a strictly Th232/U233 breeder cycle should be considered including U235 and other fissile isotopes formed in meaningful concenrations for calculations.
Any thoughts?

http://www-nds.iaea.org/relnsd/vchart/
This interactice chart has a comprehiensive list of the nucleotide products and their decay chains, although the data has some minor conflicts with other sources (like we saw with Ga) and so there will have to be discussion before number crunching.

 Quote by mesa To keep things simple a strictly Th232/U233 breeder cycle should be considered including U235 and other fissile isotopes formed in meaningful concenrations for calculations. Any thoughts? *Here is a link that may also be helpful: http://www-nds.iaea.org/relnsd/vchart/ This interactive chart has a comprehensive list of the nuclide products and their decay chains, although the data has some minor conflicts with other sources (like we saw with Ga) and so there will have to be discussion before number crunching.
I believe the approach is to start MSR (MSBR) with U-235 in Th-232 until sufficient U-233 is available - then perhaps wean the system from U-235 to U-233.

There is also the consideration of neutron spectrum, e.g., thermal, epi-thermal or even fast. One current MSR concept is for a graphite free core, which might imply more moderation from Be. In addition, the Li in the LiF should be depleted in Li-6 to minimize tritium production.

Here is a somewhat relevant report - http://www.princeton.edu/sgs/publica...df/9_1kang.pdf

 Quote by Astronuc I believe the approach is to start MSR (MSBR) with U-235 in Th-232 until sufficient U-233 is available - then perhaps wean the system from U-235 to U-233.
Should we be so concerned with the initial injection of fissile U235 or concentrate on the Th232/U233 breeder cycle as the majority of operational time will go to that? On another note, perhaps the easiest way to set this up would be based on MWt generated since we can directly calculate fissions per U233 (and small amounts of U235 created from the breeder cycle)
 Quote by Astronuc There is also the consideration of neutron spectrum, e.g., thermal, epi-thermal or even fast. One current MSR concept is for a graphite free core, which might imply more moderation from Be.
That sounds reasonable, calculations will have to include all neutron energies that bring significant concentration of daughter nuclie(s) of interest (neutron poison).

On the graphite free core; very interesting idea employing beryllium in the salt as the moderator although the design requires Hastelloy 'tubes' for these salts, that could be technically difficult as the materials are one of the largest obstacles and this system would require a vast increase in surface area while being at minimum thickness for optimal heat transfer.

Perhaps we should just pick up where ORNL left off and assume for a graphite core in the interim. Other considerations can be taken into account after getting these initial values.

Also if we just base the calculations strictly off of MWt then they could be 'adjusted' to any of these systems estimated MWt.
 Quote by Astronuc In addition, the Li in the LiF should be depleted in Li-6 to minimize tritium production.
I have seen interviews of the scientists from ORN suggesting that removal of tritium is not an issue, also considering the difficulty in isotopic seperation of Li6 could add a great deal of expense to a reactor on commercial scale when considering the large quantities of salt required.

Also current worldwide production of tritium is remarkably small:
"According to the Institute for Energy and Environmental Research report in 1996 about the U.S. Department of Energy, only 225 kg (500 lb) of tritium has been produced in the United States since 1955. Since it continually decays into helium-3, the total amount remaining was about 75 kg (170 lb) at the time of the report" link here:
http://en.wikipedia.org/wiki/Tritium
At current market value of almost \$30,000/g I would assume this is an asset, not a liability.

We are covering an aweful lot of ground here, what are your thoughts as far as where we should focus our energy for now?

 Quote by Astronuc

 Quote by mesa Should we be so concerned with the initial injection of fissile U235 or concentrate on the Th232/U233 breeder cycle as the majority of operational time will go to that? On another note, perhaps the easiest way to set this up would be based on MWt generated since we can directly calculate fissions per U233 (and small amounts of U235 created from the breeder cycle)
I think one has to start with U-235/Th-232 until one produces enough U-233.
 I have seen interviews of the scientists from ORN suggesting that removal of tritium is not an issue, also considering the difficulty in isotopic seperation of Li6 could add a great deal of expense to a reactor on commercial scale when considering the large quantities of salt required.
Laser isotopic enrichment/selection is very advanced.

 We are covering an aweful lot of ground here, what are your thoughts as far as where we should focus our energy for now?
There are a lot of technical issues in design a nuclear power system. Just take a look at the DC process. Adding a chemical separation plant in parallel just adds to the complexity (and I'm not sure that is not addressed in current licensing bases). Core and fuel design are a somewhat small but significant part of the system.

Note in WASH-1222 (TABLE I), the proposed Specific Fissile Fuel Inventory for the MSBR is 1.5 kg/MWe. So that for a 1 GWe plants, the inventory would be 1.5 Mt. If that's just the fissile content, then at 3% (by mass), the fertile inventory is about 49 Mt. It also proposes 72% LiF, 16% BeF2, 12% ThF4 and 0.3% UF4 (based on moles?).

If a MSR was to be built, I'd recommend a 200 MWt system, rather than attempting a larger full scale system.

FYI - some options - http://www.gen-4.org/GIF/About/docum...-8-Renault.pdf

 Quote by Astronuc I think one has to start with U-235/Th-232 until one produces enough U-233.
Okay, but I have an objection; I think we should get comfortable with calculations of products off of the Th232/U233 breeder cycle first since fissile ratios will be for the most part consistent before jumping into changing mixtures of fissile that we will see in the primary reactions as U235 is replaced by U233.

I understand that the U235 cycle comes first in operation but it would be nice to have more familiarity with running figures before diving into the shallow end of the pool.

Either way, I am ready to crunch.
 Quote by Astronuc Laser isotopic enrichment/selection is very advanced.
Any idea where the costs would be? Is removal of Li6 critical for operation? If not is there value in the tritium production?
 Quote by Astronuc There are a lot of technical issues in design a nuclear power system. Just take a look at the DC process. Adding a chemical separation plant in parallel just adds to the complexity.
This is your field of expertise, so if you have an idea of where would be best to focus, you have my attention.

It would seem like a good idea to run numbers on the neutron poisons along with fission products that will cause issues with the materials however number of fissions of fissile must be know first (U235/Th232/U233 or the Th232/U233 cycle). What are your thoughts?

 Quote by Astronuc Note in WASH-1222 (TABLE I), the proposed Specific Fissile Fuel Inventory for the MSBR is 1.5 kg/MWe. So that for a 1 GWe plants, the inventory would be 1.5 Mt. If that's just the fissile content, then at 3% (by mass), the fertile inventory is about 49 Mt. It also proposes 72% LiF, 16% BeF2, 12% ThF4 and 0.34 (based on moles?). If a MSR was to be built, I'd recommend a 200 MWt system, rather than attempting a larger full scale system. FYI - some options - http://www.gen-4.org/GIF/About/docum...-8-Renault.pdf
For some reason this part of your post wasn't showing before I replied :/

That is a lot of material. We will probably have to read through the ORN documents to get a better idea on whether these figures are based on mols, mass, etc. since the WASH-1222 doc came from that data.

"A 200MWt system rather than attempting a larger full scale system", it's funny to think of 200MWt as 'small'. Can you elaborate why this is a reasonable target for a test reactor?