Get rid of transuranians in Liquid Fluoride Thorium Reactors?

[Moniz_not_Ernie]

The latter implies, what, for a thorium machine? That the design would continuously have its neutron count boosted by addition of uranium (not just at start up)? If so it once again becomes dependent on an enriched fuel cycle, discarding a major advantage of a thorium design.

So three neutrons per fission event can be thrown away without concern?
232-Th+n ->233-Pa -> 233-Pa +n -> 234-Pa -> beta + 234-U -> 234-U+n -> 235-U

The first neutron is required to produce 233-U anyway.
233-Pa has a capture cross-section of 39.5, and a half-life of a month. It is a bit of a race, but I think far more will decay before it absorbs. I don't know the math needed to prove that. I just figure a fuel atom "lives" in the core for maybe a year without fissioning. I get that from the following logic: If new fuel is 5% enriched, and (say four years later) old fuel is 1%, an individual atom has 20% chance of surviving four years.
The fuel has a fission cross-section of 530. Throw in the thinner neutron flux in the blanket, compared to the core. Might that come to only 1% losses to 234-U? (Granted, two neutrons are lost by that route.)

 

[Astronuc]

Demonstrated? Even if a fast reactor enables consumption of other actinides, how does a solid fueled fast reactor dispense with poisons, or cladding degradation, or any of the other problems that are likely to demand fuel removal before completion of a high burn up?

Fast reactor fuel uses stainless steel cladding in a liquid metal coolant. The end of life issues are different that those of a Zr alloy in a water-cooled environment. Fast reactor fuel is also enriched to about 20% fissile, which is usually Pu-239/240/241 and U in (U,Pu)O₂ or (U,Pu)C or (U,Pu)N. One can return a fuel element to the core and drive it with fresh fuel nearby, so the burnups achieved can be quite high. Xe-135/135m, which is a strong poison in an LWR thermal neutron flux is not so strong in a fast flux.

In LWRs, used fuel assemblies can be 'driven' to higher burnups by face adjacent fresh fuel. In addition, high burnup fuel rods can be placed into locations in fresh fuel assemblies and those older fuel rods can be 'pushed' to higher burnups. However, several key issues for high burnup LWR fuel include fission gas release and rod internal pressure, oxidation of the cladding (leading to metal wall loss), hydriding of the cladding (leading to brittle cladding or enhanced oxidation, and growth of the fuel rod in an assembly not designed for increased growth.

Some LWR test fuel has gone above 60 GWd/tU up to ~100 GWd/tU, but that was done under special test conditions not in a commercial reactor. Commercial reactors have much more severe limits due to safety concerns.

 

[Astronuc]

The first neutron is required to produce 233-U anyway.
233-Pa has a capture cross-section of 39.5, and a half-life of a month. It is a bit of a race, but I think far more will decay before it absorbs. I don't know the math needed to prove that. I just figure a fuel atom "lives" in the core for maybe a year without fissioning. I get that from the following logic: If new fuel is 5% enriched, and (say four years later) old fuel is 1%, an individual atom has 20% chance of surviving four years.
The fuel has a fission cross-section of 530. Throw in the thinner neutron flux in the blanket, compared to the core. Might that come to only 1% losses to 234-U? (Granted, two neutrons are lost by that route.)

This is only part of the story. In an LWR, with 5% enrichment, i.e., 5% U-235 and 95% U-238, some of that U-238 is converted to Pu-239, -240, -241, and other transuranic radionuclides that accumulate with time. So even though U-235 is depleted to 1% or so, at high burnup, the Pu-239 is being fissioned, and at some point, more so than the U-235, which is partially shielded by the Pu-239.

LWRs have a fast as well as thermal population of neutrons. Some fissions (~8-10%) are fast fissions, while the rest occur from thermal-neutron induced fissions. Pu-239 (and other transuranics) accumulates on the periphery of the fuel pellets, and during the course of a lifetime, the U-235 in the interior of the pellet is shielded by the Pu-239 on the exterior. If a fuel pellet has a burnup of 60 GWd/tU, then the rim (outer 10%) has a very high burnup of ~150-180 GWd/tU. This induces the so-called 'rim effect' seen in high burnup LWR fuel.

 

[Moniz_not_Ernie]

This is only part of the story. In an LWR, with 5% enrichment, i.e., 5% U-235 and 95% U-238, some of that U-238 is converted to Pu-239, -240, -241, and other transuranic radionuclides that accumulate with time. So even though U-235 is depleted to 1% or so, at high burnup, the Pu-239 is being fissioned, and at some point, more so than the U-235, which is partially shielded by the Pu-239.

LWRs have a fast as well as thermal population of neutrons. Some fissions (~8-10%) are fast fissions, while the rest occur from thermal-neutron induced fissions. Pu-239 (and other transuranics) accumulates on the periphery of the fuel pellets, and during the course of a lifetime, the U-235 in the interior of the pellet is shielded by the Pu-239 on the exterior. If a fuel pellet has a burnup of 60 GWd/tU, then the rim (outer 10%) has a very high burnup of ~150-180 GWd/tU. This induces the so-called 'rim effect' seen in high burnup LWR fuel.

Interesting geometry there. And I didn't know the fissions from fast neutrons was so high. But that depends on your definition of "fast". I have seen fission yield data for three speed bins, labeled 0.0253, 500K, and. 14 million eV. Do you include the middle energy bin in you 8-10%? What is your upper limit for slow neutrons?
I ask because I'm designing an MSR simulation. One basic question is whether we want fast or slow reactors. However, if the data comes in three speeds, So I may allow the user to choose reactor specs a little precisely. It will probably require aggregating ENDSF data into two or three bins. Wish I could contract that task out, but I'm on a tight budget ($=0).
I'm not a nuclear engineer. I'm a simulation designer, who got interested in this stuff through my hobbies; first astronomy (your username is intriguing), then cosmology and stellar evolution. I'll be asking some newbie questions here.

 

[Stanley514]

This one article states that Lithium or Beryllium are not consumed by LFTR.
http://www.2112design.com/blog/lftr/
Could somebody verify it?

 

[Astronuc]

Interesting geometry there. And I didn't know the fissions from fast neutrons was so high. But that depends on your definition of "fast". I have seen fission yield data for three speed bins, labeled 0.0253, 500K, and. 14 million eV. Do you include the middle energy bin in you 8-10%? What is your upper limit for slow neutrons?
I ask because I'm designing an MSR simulation. One basic question is whether we want fast or slow reactors. However, if the data comes in three speeds, So I may allow the user to choose reactor specs a little precisely. It will probably require aggregating ENDSF data into two or three bins. Wish I could contract that task out, but I'm on a tight budget ($=0).
I'm not a nuclear engineer. I'm a simulation designer, who got interested in this stuff through my hobbies; first astronomy (your username is intriguing), then cosmology and stellar evolution. I'll be asking some newbie questions here.

Cross section data are available over ~12 orders of magnitude of neutron energy, from about 1E-5 ev up to 10 MeV.

Many texts have the fission cross-sections for thermal (0.0235 eV), fast (~0.7 MeV), and 14.1 MeV (neutron energy from d,t fusion). These are arbitrary. For any simulation, e.g., steady-state core depletion or power distribution analysis in a reactor, one must consider the array of cross-sections from fissile and fertile isotopes, burnable poisons, fission products, coolant, and structural materials as functions of the energy spectrum. Fission neutrons are created in the MeV range (0.1 MeV to 10 MeV), so one has to consider the fission spectrum. In a thermal reactor, the thermal flux is somewhat less than the fast flux, but the thermal cross-sections are substantially greater, by about two to three orders of magnitude.

Data is compiled in various government issued databases.
http://www.nndc.bnl.gov/sigma/index.jsp?as=235&lib=endfb7.1&nsub=10

 

[Moniz_not_Ernie]

This one article states that Lithium or Beryllium are not consumed by LFTR.
http://www.2112design.com/blog/lftr/
Could somebody verify it?

From Wikipedia (Molten salt reactor or Liquid fluoride thorium reactor)
LFTR uses enriched Lithium, 99+% 7-Li. The remaining 6-Li will capture neutrons to a small extent, producing pesky Tritium. Minuscule amounts of Fluorine and Beryllium (and 7-Li) will also be transmuted. I would suggest not using the word "consumed" for this. Reserve that verb for the fuel. We'd get some sustenance from that.

 

[Astronuc]

This one article states that Lithium or Beryllium are not consumed by LFTR.
http://www.2112design.com/blog/lftr/
Could somebody verify it?

Li-7 would have a low cross-section for most n-absorption reactions, but Li-6 would experience (n, alpha)t reaction. I believe Be-9 has very low cross-section, so it should not be rapidly depleted.

 

[Moniz_not_Ernie]

Cross section data are available over ~12 orders of magnitude of neutron energy, from about 1E-5 ev up to 10 MeV.

Many texts have the fission cross-sections for thermal (0.0235 eV), fast (~0.7 MeV), and 14.1 MeV (neutron energy from d,t fusion). These are arbitrary. For any simulation, e.g., steady-state core depletion or power distribution analysis in a reactor, one must consider the array of cross-sections from fissile and fertile isotopes, burnable poisons, fission products, coolant, and structural materials as functions of the energy spectrum. Fission neutrons are created in the MeV range (0.1 MeV to 10 MeV), so one has to consider the fission spectrum. In a thermal reactor, the thermal flux is somewhat less than the fast flux, but the thermal cross-sections are substantially greater, by about two to three orders of magnitude.

Data is compiled in various government issued databases.
http://www.nndc.bnl.gov/sigma/index.jsp?as=235&lib=endfb7.1&nsub=10

Thanks. I didn't know the 14MeV data was for fusion. This simplifies things. I would use just two bins, fast and slow. The user would have some design in mind, and possibly some high-powered sims to estimate the energy distribution. That estimate would, with some more work, yield two numbers for input into my model: Design LFTR-22a produces 12% fast and 82% slow neutrons. (More input: Test Fuel - 2%Savannah2/1999+4.5LEU - see App. C for details. Test Cleaning: H-sparge only)
The user would set the end time, say four years, and start the sim. Results (isotopic composition of the salt) would also include the composition at two years, one year, half that, half that, etc, down to a day or so.
The idea is to test fast and slow designs and see how much of the Savannah waste one might consume. We could also try combinations of fast and slow in mixed fleets. The sim is really too simple for nuclear engineers, though they may be the user. They would run the sim for the real paying customer, politicians.

 

[mheslep]

Thank you, I'd not considered the lower cross section for fast spectrum. However:

However, several key issues for high burnup LWR fuel include fission gas release and rod internal pressure,

The volume expansion (or pressure increase) due to products would also be a problem in a fast reactor.

 

[Moniz_not_Ernie]

Thank you, I'd not considered the lower cross section for fast spectrum. However:

The volume expansion (or pressure increase) due to products would also be a problem in a fast reactor.

For a fast, solid fuel reactor, probably. This thread is about LFTR, a thermal liquid fueled reactor. There are fast liquid fueled reactor designs. Gaseous fission products bubble out of the liquid for either. Not a pressure problem, but piping the gasses off elsewhere has a number of other problems.
Also, IFR used solid metal, not oxides. Did those pellets have cladding? (Did they even have pellets?) If not, they may not have had a pressure problem.

 

[Astronuc]

Thank you, I'd not considered the lower cross section for fast spectrum. However:

The volume expansion (or pressure increase) due to products would also be a problem in a fast reactor.

Yes, swelling of austenitic stainless steels and dimensional distortions due to differential growth (cause by flux gradients) have been concerns. Newer alloys, types of ferritic or ferritic-martensitic steels (e.g., HT-9 or T91), have shown promise for higher exposures.

HT-9 - http://www.kns.org/jknsfile/v45/3Yiren_Chen.pdf

T91 - http://www.oecd-nea.org/pt/docs/iem/jeju02/session4/SessionIV-10.pdf

I understand that there is no significant experience with T91 in a fast reactor environment, unless it's been irradiated in Russia.

As for rod internal pressure, the fuel rod designs I saw has large plenum volumes.
The fuel rod design for Clinch River had a length of 2.9 m, a plenum length of 1.2 m, and fuel length of 0.91 m. The core diameter was about 2 m.

 

[mheslep]

As for rod internal pressure, the fuel rod designs I saw has large plenum volumes.
The fuel rod design for Clinch River had a length of 2.9 m, a plenum length of 1.2 m, and fuel length of 0.91 m. The core diameter was about 2 m. ...

The plenum is originally filled with what atop the solid fuel oxide? Surely not a vacuum? In any case such a design is sufficient to accumulate Xe and other gasses for the years required to achieve high burn-up?

 

[mheslep]

This thread is about LFTR...

Yes, about the various advantages and disadvantages of such a design. In response, there have been several comments asserting most of the proposed liquid fueled advantages can be obtained with solid fueled reactors, even existing PWR/BWRs, if operators chose to operate them in such matter, but currently do not.

Gaseous fission products bubble out of the liquid for either.

Exactly, made possible by the nature of liquid fuel designs. I fail to see how this is possible in PWR/BWR solid fuel designs without extraordinary measures.

 

[Astronuc]

Yes, about the various advantages and disadvantages of such a design. In response, there have been several comments asserting most of the proposed liquid fueled advantages can be obtained with solid fueled reactors, even existing PWR/BWRs, if operators chose to operate them in such matter, but currently do not.

Exactly, made possible by the nature of liquid fuel designs. I fail to see how this is possible in PWR/BWR solid fuel designs without extraordinary measures.

I believe the fuel rods had He and other gases (Ar, Ne, Kr, Xe for tagging), but I'm not sure of the pressure.

Design and manufacture of gas tags for FFTF fuel and control assemblies
http://inis.iaea.org/search/search.aspx?orig_q=RN:7220619

In the LFTR, the gases and other volatiles would have to be collected and allowed to decay.

 

[Moniz_not_Ernie]

I believe the fuel rods had He and other gases (Ar, Ne, Kr, Xe for tagging), but I'm not sure of the pressure.

Design and manufacture of gas tags for FFTF fuel and control assemblies
http://inis.iaea.org/search/search.aspx?orig_q=RN:7220619

In the LFTR, the gases and other volatiles would have to be collected and allowed to decay.

The collection point in the MSRE was the fuel input tank. It was partially filled with helium. They bubbled helium gas up through the fuel and this helped carry about 5/6ths of the xenon away. This paper from 1969 mentions that, and some of the research requirements for future MSR development - http://moltensalt.org/references/static/downloads/pdf/NAT_MSRintro.pdf

The gases trapped in solid fuel pellets are released when the pellets are chopped up. That seems to be the first step in most of the "advanced" reprocessing methods described here - http://www.inl.gov/technicalpublications/documents/5094580.pdf

 

[nikkkom]

I believe the fuel rods had He and other gases (Ar, Ne, Kr, Xe for tagging), but I'm not sure of the pressure.

I vaguely remember reading some rod design doc and it said that as a final manufacturing step rods are pressurized to 20 atm He and then sealed.

 

[Astronuc]

I vaguely remember reading some rod design doc and it said that as a final manufacturing step rods are pressurized to 20 atm He and then sealed.

LWR fuel is pressurized. Some PWR fuel designs have on the order of 20 atm of He, and some more, some less. The PWR primary system operates at about ~153 atm of pressure (~2250 psi). BWR fuel has lower pressure, on the order of 5 to 10 atm, since the BWR primary system operates at about ~72 atm (~1055 psi).

 

[Hologram0110]

The plenum is originally filled with what atop the solid fuel oxide? Surely not a vacuum? In any case such a design is sufficient to accumulate Xe and other gasses for the years required to achieve high burn-up?

Depends on the design. Current oxides fuel elements are typically filled with helium gas because of it is inert and has a high thermal conductivity. I believe that most fast reactor designs have large plenums extending from the reactor to accommodate the fission gas. I also heard of a fast reactor design which can periodically vent the gas from its fuel element to prevent over pressure (I believe TerraPower but I could be mistaken).

Some fast reactor design propose the fuel elements contain sodium which will melt and provide thermal bonding between the solid fuel and cladding. This allows the fuel to be undersized to accommodate a large degree of swelling while still maintaining good thermal contact to the coolant. It also means the fuel does not need to be manufactured with very tight tolerance greatly simplifying manufacturing (for recycled radioactive fuel where dust is a problem).

 

[Matte Patte]

Well, you can always deport them back to Transurania or force them to stop cross dressing and remain uranium-atoms...
[Sorry, could not help it...please don't ban me!?]

Molten salt reactors would have issues with the release of volatile fission products (Xe, Kr, I, Cs and so fourth). In todays' solid fuel the leakage is quite small and does not cause serious concern as most fission products are locked in the oxide matrix. For MSR's the solution around this would require some form of double cotainment to stop significant leakage of fission products outside the plant.

 

[Astronuc]

Well, you can always deport them back to Transurania or force them to stop cross dressing and remain uranium-atoms...
[Sorry, could not help it...please don't ban me!?]

Molten salt reactors would have issues with the release of volatile fission products (Xe, Kr, I, Cs and so fourth). In todays' solid fuel the leakage is quite small and does not cause serious concern as most fission products are locked in the oxide matrix. For MSR's the solution around this would require some form of double cotainment to stop significant leakage of fission products outside the plant.

Well, in theory, the primary system would be sealed so that the fission products are extracted from the coolant. The gases and volatiles would certainly have to be collected and held up to decay, and then that stream would have to be processed.

 

[Matte Patte]

Well, yes but you also have no secondary containment (fuel cladding). The primary circuit is (considered) the tertiary barrier in LWR's...

 

[Astronuc]

Well, yes but you also have no secondary containment (fuel cladding). The primary circuit is (considered) the tertiary barrier in LWR's...

I don't think we consider the fuel material a primary containment, since it doesn't necessarily retain all fission products, especially the Xe, Kr radioisotopes, or the volatiles, Cs, I, Te, etc. Fast reactor fuel allows for centerline void due to restructuring. In LWR and FR fuel, the primary containment is the cladding, the secondary containment is the primary or coolant circuit, and the tertiary containment is the containment building and ancillary buildings where coolant treatment system collect any fission products and activated corrosion products from the core.

In a liquid fuel system, the primary containment is the core and piping related to the liquid fuel transport and processing system. The benefit should be relatively low equilibrium activity in the core and primary circuit, except where the fission products accumulate.

 

[mheslep]

Well, you can always deport them back to Transurania or force them to stop cross dressing and remain uranium-atoms...
[Sorry, could not help it...please don't ban me!?]

Molten salt reactors would have issues with the release of volatile fission products (Xe, Kr, I, Cs and so fourth). In todays' solid fuel the leakage is quite small and does not cause serious concern as most fission products are locked in the oxide matrix. For MSR's the solution around this would require some form of double cotainment to stop significant leakage of fission products outside the plant.

Since MSRs would run at atmospheric pressure without water around the core which can flash to steam, some designers are arguing for no confinement.

 

[Astronuc]

Since MSRs would run at atmospheric pressure without water around the core which can flash to steam, some designers are arguing for no confinement.

Atmospheric pressure at the top of the highest point in the primary system, but going down to the bottom of the core under a few meters of the liquid fuel, the pressure will be greater by ρgh, so the bottom of the core will be several atmospheres. I imagine there will be some kind of containment to collect the radioactive gases and volatiles in the event of the break in the primary system, and particularly where the steam generator is located, since the steam pressure is likely to be on the order of 900 to 1000 psi. A steam-fluorine reaction would be problematic with respect to HF gas.

Note that a commercial scale has not yet been constructed let alone designed. As far as I know, none of the promoters/advocates in the US have designed or constructed any type of nuclear plant.

 

[mesa]

Atmospheric pressure at the top of the highest point in the primary system, but going down to the bottom of the core under a few meters of the liquid fuel, the pressure will be greater by ρgh, so the bottom of the core will be several atmospheres. I imagine there will be some kind of containment to collect the radioactive gases and volatiles in the event of the break in the primary system, and particularly where the steam generator is located, since the steam pressure is likely to be on the order of 900 to 1000 psi. A steam-fluorine reaction would be problematic with respect to HF gas.

Note that a commercial scale has not yet been constructed let alone designed. As far as I know, none of the promoters/advocates in the US have designed or constructed any type of nuclear plant.

Didn't Rusty Holden came up with a design? As I recall it was by no means a complete work of engineering but it seemed a like a reasonable start. Perhaps a few well placed emails will give a more clear answer on this.

 

[Astronuc]

Didn't Rusty Holden came up with a design? As I recall it was by no means a complete work of engineering but it seemed a like a reasonable start. Perhaps a few well placed emails will give a more clear answer on this.

I should probably qualify that statement further by saying a 'licensed and approved design'. Lots of folks can claim to have a design, but I'd want to see some details, particularly with respect to the nuclear design and core neutronics.

 

[mesa]

I should probably qualify that statement further by saying a 'licensed and approved design'. Lots of folks can claim to have a design, but I'd want to see some details, particularly with respect to the nuclear design and core neutronics.

An excellent point, and I would imagine these would be very difficult steps to complete. How would one go about doing this after coming up with a design?

 

[Matte Patte]

I don't think we consider the fuel material a primary containment, since it doesn't necessarily retain all fission products, especially the Xe, Kr radioisotopes, or the volatiles, Cs, I, Te, etc. Fast reactor fuel allows for centerline void due to restructuring. In LWR and FR fuel, the primary containment is the cladding, the secondary containment is the primary or coolant circuit, and the tertiary containment is the containment building and ancillary buildings where coolant treatment system collect any fission products and activated corrosion products from the core.

In a liquid fuel system, the primary containment is the core and piping related to the liquid fuel transport and processing system. The benefit should be relatively low equilibrium activity in the core and primary circuit, except where the fission products accumulate.

The fuel pellet is considered the first barrier, when the fuel is considered as part of the waste stream, I am pretty sure same goes for during operation but not 100% sure (one of five barriers, fuel pellet, cladding, primary circuit, containment and reactor building). Yes, the barrier is weak for some nuclides (so is the cladding) but it is pretty good at retaining more of the nasty ones.

Today, small cladding defects in a few fuel pins would bring a plant close to their annual release limits for radio-iodine, possibly xenon and krypton as well (PWR). At least in my neck of the woods. Running a complete core with only the primary circuit as containment for volatile fission products would require a pretty neat feat in convincing any regulatory body that those releases would be ok for the public to absorb (dose wise that is).

 

[Astronuc]

An excellent point, and I would imagine these would be very difficult steps to complete. How would one go about doing this after coming up with a design?

One would have to arrange for the NRC to develop a review plan and the regulatory structure with which to license and approve a design. It would likely be similar to the system now in place for current light water reactors, which includes NUREG-0800, Standard Review Plan.

http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/

The regulatory structure includes mandatory requirements concerning the design and operation of a nuclear power plant to ensure the safety of plant personnel and the public. Basically, the system must control the disposition of fission products, must ensure control of the reactor (reactivity control, with the requirement that the nuclear chain reaction can be terminated and the reactor maintained in shutdown), and must ensure coolability of the core/fuel in order to ensure that there is no release of fission products outside of the reactor system or damage to the reactor.

Some of the regulations are found in Title 10 of the Code of Federal Regulations, 10 CFR.
http://www.nrc.gov/reading-rm/doc-collections/cfr/

10 CFR 50 addresses Domestic licensing of production and utilization facilities
http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/
Appendix A provides the general design criteria

10 CFR 52 address Licenses, certifications, and approvals for nuclear power plants, which applies to new builds.
http://www.nrc.gov/reading-rm/doc-collections/cfr/part052/

There are also Regulatory Guides
http://www.nrc.gov/reading-rm/doc-collections/reg-guides/power-reactors/rg/

The process starts with a meeting with the US NRC, or corresponding body in another nation.