Liquid Fluoride Thorium Reactor

[mheslep]

...

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.

 

[mesa]

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.

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.

 

[mesa]

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?

 

[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/tecdocs/iaea-tecdoc-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.

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.

 

[mesa]

* 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/tecdocs/iaea-tecdoc-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 received 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/

This should be helpful.

 

[mheslep]

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?

 

[mesa]

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:



Calculating for if they are necessary would be good, however there are many things Astronuc suggested that seem like viable avenues to look at. 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.

 

[mesa]

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.

 

[mheslep]

Here is Chris Holdens reason for it @6:16 in his presentation for his reactor design, here is a link:



... 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.

 

[zapperzero]

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.

 

[Astronuc]

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.

 

[mheslep]

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.

 

[mesa]

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/

 

[mesa]

* 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.

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.

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?

 

[Astronuc]

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.

 

[mesa]

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.

 

[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/nuclear_materials/nuclear_processes.php

http://en.wikipedia.org/wiki/File:ThermalFissionYield.svg

http://en.wikipedia.org/wiki/Fission_products_(by_element )

http://en.wikipedia.org/wiki/Fission_products_(by_element)#Tellurium-125.2C_127_to_132

http://en.wikipedia.org/wiki/Fissio...7.2C_and_samarium-149.2C_151.2C_152.2C_154.29


http://upload.wikimedia.org/wikiped...onYield.svg/750px-ThermalFissionYield.svg.png


There are preferred nuclides, i.e., those with high yield fractions.

Also of interest - http://en.wikipedia.org/wiki/Fluoride_volatility

 

[mesa]

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/nuclear_materials/nuclear_processes.php

http://en.wikipedia.org/wiki/File:ThermalFissionYield.svg

http://en.wikipedia.org/wiki/Fission_products_(by_element )

http://en.wikipedia.org/wiki/Fission_products_(by_element)#Tellurium-125.2C_127_to_132

http://en.wikipedia.org/wiki/Fissio...7.2C_and_samarium-149.2C_151.2C_152.2C_154.29


http://upload.wikimedia.org/wikiped...onYield.svg/750px-ThermalFissionYield.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?

*Here is a link that may also be helpful:
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.

 

[Astronuc]

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 - www.princeton.edu/sgs/publications/sgs/pdf/9_1kang.pdf

See also - https://www.physicsforums.com/showpost.php?p=2546513

 

[mesa]

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)

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.

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 separation 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?

 

[mesa]

See also - https://www.physicsforums.com/showpost.php?p=2546513

Great link!

 

[Astronuc]

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 separation 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/documents/30-Session2-8-Renault.pdf

 

[mesa]

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.

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?

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?

 

[mesa]

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/documents/30-Session2-8-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?

 

[mheslep]

... In addition, the Li in the LiF should be depleted in Li-6 to minimize tritium production.

Some are proposing a sodium based salt rather than lithium for just that reason.

 

[sliebsch]

I actually chose to write a paper on this topic in undergrad. Very very interesting topic. They come with a unique set of problems, but nothing insurmountable. They use familiar, abundant materials (fluoride and thorium), and have a reliable passive safety feature(the salt plug). However, nobody in my class had ever even heard of this reactor type, even though it appears to be competitive with or even superior to LWRs in size and safety. I think more research needs to be done, that's just my two cents.

 

[sliebsch]

If I remember right, they didn't become popular because LWR technology had already been developed for submarines, so there was a great deal of research and investment momentum in that direction, even though it might not be the best option for commercial power production.

 

[Steve Brown]

Thorium reactor also produces waste. Less transuranics than uranium cycle, but about the same amount of fission products. This waste needs to be disposed off just the same.

The mass of fission products created is about the same, but it is many times smaller than the total mass of spent fuel rods. About 98% of the mass of a spent fuel rod is unreacted fissile material. On the other hand, a LFTR continuously processes the core salt to remove fission products, while leaving the fissile material in the core salt, where it can react to provide energy. There is no spent fuel assembly to dispose of, and there is no waste of valuable fissile material. In other words, there is no turning valuable fissile material into radioactive waste without deriving energy from it.

 

[nikkkom]

The mass of fission products created is about the same, but it is many times smaller than the total mass of spent fuel rods. About 98% of the mass of a spent fuel rod is unreacted fissile material.

I know that, and I already said in other thread that IMO spent fuel should be reprocessed, not buried as-is.

On the other hand, a LFTR continuously processes the core salt to remove fission products, while leaving the fissile material in the core salt, where it can react to provide energy. There is no spent fuel assembly to dispose of, and there is no waste of valuable fissile material.

This is verging on being a blatant PR.

LFTR in this regard is not better than other reactors, because processing of highly radioactive core salt is neither easy nor cheap - roughly on par with cost and difficulty of spent fuel reprocessing for LWRs.

LWR proponents can easily do the same and portray it as a weakness of LFTR: "every LFTR requires a small reprocessing plant on-site, whereas LWRs can use a common reprocessing plant, utilizing economies of scale."

In other words, there is no turning valuable fissile material into radioactive waste without deriving energy from it.

LWRs don't do it either, at least in France.

 

[Steve Brown]

This is verging on being a blatant PR.

You wrote that in response to the following statement of facts:

"On the other hand, a LFTR continuously processes the core salt to remove fission products, while leaving the fissile material in the core salt, where it can react to provide energy. There is no spent fuel assembly to dispose of, and there is no waste of valuable fissile material."

I can't help it if you don't like facts, but calling them "PR" does not make them any less true.

LFTR in this regard is not better than other reactors, because processing of highly radioactive core salt is neither easy nor cheap - roughly on par with cost and difficulty of spent fuel reprocessing for LWRs.

That sounds more like opinion than fact. Processing of solid fuel rods requires shutting down the reactor, physically removing and transporting them to a reprocessing facility. There, the rods have to be disassembled, the solid material has to be converted to liquid or gas phase in order to separate fission products and transuranic isotopes from the fissile material. Then, new fuel rods have to be fabricated at great expense, transported back to the reactor, and installed. Processing of molten core salt obviates all the steps of shutdown, removal, transport, disassembly, conversion to liquid or gas phase, fabrication, transport, installation, and reactor startup. Not only that, but continuous processing keeps the level of neutron absorbers such as xenon-135 low, whereas these poisons build up in fuel rods, necessitating replacement of the rods, for that and other reasons, after only a small fraction of the fissile material is reacted. On balance, the solid fuel cycle entails costly and wasteful inefficiencies that the molten salt reactor avoids.

I get that you don't like the molten salt reactor concept, or that you simply like to argue, but just as you insisted on staying on topic, I insist that you stick to discussing facts instead of characterizing them as "PR" or anything else.