Hybrid/variable spectrum reactor

[armoredchestnut]

TL;DR Summary

Reactors that could start by burning natural U and transition to a fast spectrum after breeding enough fuel

Why have no Variable spectrum reactors been produced? A reactor that could start by burning natural U, like the CANDU or RBMK, and then transition to a fast spectrum seems to be ideal. Running nuclear fuel from raw material to nuclear ash all in one reactor seems to be the solution for people concerned about theft and proliferation of nuclear materials. Is there a engineering or physics trade off that I am unaware of?

 

[Alex A]

My understanding is that U-235 slow reactors cannot breed more fuel than they consume. Only fast reactors can. Thorium cycle reactors I've read can, but I don't know why. A fast spectrum reactor sounds ideal, but they are less easy to control and the construction is wildly different, coolant in one can be water in the other might be liquid sodium. There is then the problem that a fast reactor needs a much higher concentration of fuel than a slow reactor can be made to work with.

Since you need a fuel change, or a reprocessing step in between the stages, it doesn't make sense to use anything other than reactors designed specifically to be good at one thing.

Another idea that springs to mind is that no-one seems to want breeder reactors. We can enrich fuel and it is currently economical. It theoretically makes sense to breed fuel, and it theoretically makes sense to try to destroy waste but there is no appetite for it even before you get to the engineering issues.

 

[Astronuc]

TL;DR Summary: Reactors that could start by burning natural U and transition to a fast spectrum after breeding enough fuel

Running nuclear fuel from raw material to nuclear ash all in one reactor seems to be the solution for people concerned about theft and proliferation of nuclear materials. Is there a engineering or physics trade off that I am unaware of?

It's complicated, but there are significant differences among fast, epithermal and thermal reactors with respect to the fuel type, fuel geometry (fuel rods and fuel assembly), cladding design and materials, coolant (or heat transfer medium). Between the fuel and cladding, one has to limit fuel-cladding chemical interaction, while between cladding and coolant, one has to limit the cladding-coolant chemical interaction (e.g., corrosion). There are the matters of neutronics, reactivity control, power/burnup distribution and response of the fuel and core to reactivity and coolant transients. Generally, fast reactors use smaller fuel rods (smaller diameter) than CANDU/RBMK/LWRs, and most fast reactors are cooled with liquid metal; one does not simply take a fuel rods from one type of reactor and place it in another.

BWRs do use spectral shift, in which increased voiding in the upper part of the assembly is accomplished by reducing the flow. With a slightly harder spectrum, one transmutes more of the U-238 to Pu-239, and then burns the Pu-239. Utilities would use spectral shift to reduce feed enrichment.

Fast reactor fuel lattices are triangular or hexagonal lattices, which are tighter than square lattices typically used in LWR fuel. Cooling and thermal limits on heat flux are considerations.

Fast reactors typically use enrichments of about 20%, or slightly higher, in UO₂ or (U,Pu)O₂, while thermal reactors (LWRs) typically use enrichments less than 5%, although slightly greater enrichments are possible. CANDU have used U of natural enrichments, but some more recent designs have used slightly greater enrichment. CANDU and LWR fuel can use different enrichments in different fuel rods, and different enrichments in different fuel assemblies (split batches).

Another key factor is burnup and it's limits as it relates the production of fission gases, so rod internal pressure is an issue, as well as transient behavior. In addition to the gases, burnup accumulation causes the fuel to swell, and fission products compete with the fuel atoms for available neutrons.

Thorium cycle reactors I've read can, but I don't know why.

U-233 produces slightly more neutrons (on average) per neutron absorbed than U-235 (and a lower capture cross section), enough to permit a thermal breeder. U-233 is not natural, but is produced through the transmutation of Th-232 -> Th-233 -> Pa-233 -> U-233.

https://iopscience.iop.org/article/10.1088/1742-6596/1689/1/012031/pdf

 

[armoredchestnut]

It's complicated, but there are significant differences among fast, epithermal and thermal reactors with respect to the fuel type, fuel geometry (fuel rods and fuel assembly), cladding design and materials, coolant (or heat transfer medium). Between the fuel and cladding, one has to limit fuel-cladding chemical interaction, while between cladding and coolant, one has to limit the cladding-coolant chemical interaction (e.g., corrosion). There are the matters of neutronics, reactivity control, power/burnup distribution and response of the fuel and core to reactivity and coolant transients. Generally, fast reactors use smaller fuel rods (smaller diameter) than CANDU/RBMK/LWRs, and most fast reactors are cooled with liquid metal; one does not simply take a fuel rods from one type of reactor and place it in another.

BWRs do use spectral shift, in which increased voiding in the upper part of the assembly is accomplished by reducing the flow. With a slightly harder spectrum, one transmutes more of the U-238 to Pu-239, and then burns the Pu-239. Utilities would use spectral shift to reduce feed enrichment.

Fast reactor fuel lattices are triangular or hexagonal lattices, which are tighter than square lattices typically used in LWR fuel. Cooling and thermal limits on heat flux are considerations.

Fast reactors typically use enrichments of about 20%, or slightly higher, in UO₂ or (U,Pu)O₂, while thermal reactors (LWRs) typically use enrichments less than 5%, although slightly greater enrichments are possible. CANDU have used U of natural enrichments, but some more recent designs have used slightly greater enrichment. CANDU and LWR fuel can use different enrichments in different fuel rods, and different enrichments in different fuel assemblies (split batches).

Another key factor is burnup and it's limits as it relates the production of fission gases, so rod internal pressure is an issue, as well as transient behavior. In addition to the gases, burnup accumulation causes the fuel to swell, and fission products compete with the fuel atoms for available neutrons. U-233 produces slightly more neutrons (on average) per neutron absorbed than U-235 (and a lower capture cross section), enough to permit a thermal breeder. U-233 is not natural, but is produced through the transmutation of Th-232 -> Th-233 -> Pa-233 -> U-233.

https://iopscience.iop.org/article/10.1088/1742-6596/1689/1/012031/pdf

So the systems are just incompatible from an engineering and planning perspective because their requirements are too different? Alright thank you for your time. Though I will note that the problems with fuel swelling and gas buildup are obviated with a liquid fuel reactor as with the dual fluid reactor concept. I've just been on a nuclear reactor kick lately and that question has been bugging me.

 

[Astronuc]

So the systems are just incompatible from an engineering and planning perspective because their requirements are too different?

Essentially, yes. Current systems would be incompatible.

There was some thought that LWR fuel would be reprocessed and used in fast reactors, or FBR fuel would be reprocessed and used in LWR fuel. There is the matter of the different enrichments, and remote handling of the manufacturing processes given the radiological aspects.

One system I didn't mention is the Gen-IV concept of a Supercritical Water Reactor (SCWR). "SCWRs are high temperature, high-pressure, light-water-cooled reactors that operate above the thermodynamic critical point of water (374°C, 22.1 MPa). The reactor core may have a thermal or a fast-neutron spectrum, depending on the core design." Ref: https://www.gen-4.org/gif/jcms/c_42151/supercritical-water-cooled-reactor-scwr

There are various challenges regarding the assurance of mechanical integrity of the fuel and primary circuit. Current PWRs operate at about 15.5-15.8 MPa in the primary circuit at core exit temperatures of ~325-330°C.

Molten salt systems obviate some technical issues, but introduce new/different technical issues. Corrosion, thermochemical and thermomechancial challenges still have to be resolved. All systems have to address the disposition of fission products: radioisotopes of all the elements between Ga (Z=32) to Eu (Z=61) in their variable concentrations.A bit more on U-233 and the Thorium cycle. Article by Rod Adam. I'm familiar with the author and have met him before. He's generally knowledgeable and reasonable.
https://www.ans.org/news/article-1159/uranium-233-is-a-valuable-resource/