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Sun Aug 9, 2020, 08:05 PM

Liquid/Liquid Extraction Kinetics for the separation of Americium and Europium.

The paper I'll discuss in this post is this one: Liquid/Liquid Extraction Kinetics of Eu(III) and Am(III) by Extractants Designed for the Industrial Reprocessing of Nuclear Wastes (T. H. Vu, Jean-Pierre Simonin*, A. L. Rollet, R. J. M. Egberink, W. Verboomm AE Enschede,
M. C. Gullo, and A. Casnati, Ind. Eng. Chem. Res. 2020, 59, 30, 13477–13490.)

I have come to consider americium as a critical nuclear fuel if we are ever to get serious about addressing climate change, something about which clearly are not at all serious.

(Solar cells and wind turbines haven't cut it, aren't cutting it and won't cut it. The reason is physics, the extraordinary low energy to mass ratio of these systems which leads to them being purely and totally unsustainable, and in fact, environmentally odious.)

The advantages of americium nuclear fuel include these:

1) The melting point of the metal, while not as low as that of plutonium, is sufficiently low to be accessible for containment for long periods of time. The metal also has a very high liquid range, comparable to that of neptunium and gallium.

2) The critical mass in the fast neutron spectrum (which I regards as a superior spectrum) is much higher than it is for plutonium for the two most common isotopes, Am-241 and Am-243. I discussed these critical masses in this space here: Critical Masses of the Three Accessible Americium Isotopes. The continuous recycling of plutonium will lead to rising critical masses because of the increases in the proportion of Am-243 with respect to Am-241.

3) Over a long period of operation Americium will produce two valuable plutonium isotopes that are critical for denaturing weapons grade plutonium, Pu-238 and Pu-242.

4) Considerable inventories of Americium exist and are readily available owing to the environmentally disastrous fear and ignorance that have prevented the prompt recycling of nuclear fuels.

5) The accumulation of Curium-242, Curium-244, and Plutonium-238 in americium based fuels, besides generating heat, will also generate supplies of the important industrial gas helium, a gas which is rapidly being depleted from mined sources.

6) Over the long term, the decay of Pu-238 will produce important supplies of U-234, which will eliminate, largely, the need for isotopic enrichment of uranium (particularly when conducted in concert with the use of thorium based U-233.)

Besides these advantages, it is possible, but not known to my knowledge, that liquid americium metal will prove to be less corrosive than liquid plutonium. That would simplify things a bit, although in my opinion, the corrosive nature of plutonium is a surmountable problem.

The evolution of an americium based liquid metal fuel over a period of decades will result in an Americium-Plutonium-Uranium alloy. Here is a (computationally derived) ternary phase diagram of this alloy showing liquidus curves:

The caption:

Fig. 7. Predicted liquidus projection of the U-Pu-Am system. Along the dashed curve the liquid alloy decomposes to the U-rich and Am-rich b.c.c, phases..

Source: Ogawa, Journal of Alloys and Compounds, 194 (1993) pp. 1-7.

As the fuel evolves, the melting point decreases, and with it, the expectation of sufficiently liquid fuels to allow for in line fuel processing, with the caveat being the effect of fission products.

Europium is a fission product, a relatively minor fission product, but a fission product all the same. It is always to be expected to be present in used nuclear fuels, both from the capture of neutrons in samarium isotopes as well as a direct fission product. Overall, one would expect, qualitatively for europium to relatively depleted because the two natural isotopes Eu-151 and Eu-153, high neutron capture cross sections, at least in the thermal spectrum, as do the two fairly long lived radioactive isotopes and their nuclear isomers, Eu-152 and Eu-154, with half-lives respectively for their low energy isomers, of 13.5 and 8.6 years. Although small quantities of these isotopes probably represent burnable poisons, and may have some utility as such, it may be, and most likely is desirable to remove europium from americium.

This brings me to the paper under discussion. Liquid/liquid extraction (often abbreviated "LLE" ) has been the most common approach to reprocessing used nuclear fuels. I don't necessarily endorse these as being likely to be the best approach, but nobody cares what I think anyway. Almost always, these extractions require agitation and worse, the use of solvents obtained from dangerous fossil fuels. It has occurred to me in recent years, particularly in light of the development of low temperature ionic liquids, that there may be other ways to exploit mass transfer across liquid interfaces, and thus this paper is of potential interest for me as I develop my generally useless thinking.

From the introductory text of the paper:

The reprocessing of nuclear wastes resulting from spent nuclear fuel is a worldwide topic of utmost importance in the nuclear industry and for the society itself. Various processes, generally based on liquid/liquid (L/L) extraction stages, have been proposed with the aim of reducing the volume, heat, and radiotoxicity of highly radioactive waste (plutonium and americium in particular) for their disposal in a geologic repository.(1) These processes involve the separation of the most problematic radioactive elements in the wastes.

Various strategies have been developed worldwide for the reprocessing of used fuel. An overview of the main solvent extraction processes(2) (besides Europe) is presented in Table 1. References are indicated in the table that give more details on the policies of the countries in this domain.

Table 1:

The text continues:

In Europe, this topic has been tackled with determination through the financing of successive European EURATOM projects since the early 90’s: NEWPART, PARTNEW, EUROPART, ACSEPT, SACSESS,(7) and now GENIORS (GEN IV Integrated Oxide fuels Recycling Strategies).(8) The European approach was centered around the use of selective extractants and molecular diluents that would generate a minimal amount of secondary waste. A feature of this strategy is the use of chemicals that only comprise the C, H, O, and N atoms (often referred to as the CHON principle(9)), which makes them suitable for subsequent incineration.
Reference aqueous separation process routes have emerged from these in-depth studies. They are depicted in Figure 1.(10,11)

Figure 1:

The caption:

Figure 1. Main routes of the European partitioning process strategy envisaged for the recycling of actinides (An) from used fuel (Ln = lanthanides). EXAM = extraction of americium.

Some descriptive text:

One route uses the GANEX (Grouped Actinide EXtraction) process(12,13) in which uranium is separated from the waste in a first step, and then, transuranic actinide elements are isolated (Np, Pu, Am, and Cm) from all fission products.
In the other route, the PUREX (plutonium, uranium, reduction, and extraction) process,(14) first implemented in the Manhattan project, is employed for the separation of uranium and plutonium from other fission products by using tributyl phosphate (TBP) as the extractant. The COEX process is a modified version of PUREX. Then, the DIAMEX (DIAMide EXtraction) process developed at CEA (Commissariat à l’Energie Atomique) in France may be used. It consists of the co-extraction of trivalent minor actinides [MA’s, mainly composed of americium(III) and curium(III)] and lanthanides (Ln’s) from a PUREX raffinate by employing a malondiamide extractant. Although they constitute less than 0.1% of the initial spent fuel mass, the MA’s (especially neptunium, americium, and curium) will be the main contributors to the radiotoxicity (and heat generation) after a three-century storage of high-level radioactive liquid waste (obtained after the PUREX stage).

The authors here are referring to the "storage" of so called "nuclear waste." By contrast, I speak of the recovery of nuclear resources. They are also speaking of "once through" thermal fuel, largely, and not continuously recycled actinides.

These caveats aside - most of humanity has been trained to think in this way, of waste rather than resources and this is clearly a fatal way to think, fatal to the future.

In a continuous actinide recycling program, some calculations show that it is possible to obtain americium alone in concentrations of close to 1.5% (cf. Ref: Nuclear Reactor Physics, William E. Stacy, Wiley and Sons 2001. pg.234). In a world in which we we did not use any dangerous fossil fuels, where we let our rivers run free, where we did not convert our wilderness into industrial parks for wind turbines, destroy rain forests for biofuels, generate millions of tons of toxic electronic waste for solar cells, we would need, in order to produce 600 exajoules of energy per year that we use as of recent times, we would require the fission of about 7,500 tons of plutonium (or other actinides) per year. This implies about 100 MT of americium would be available per year, a significant quantity of potential industrial importance. There may not be a lot of americium, but for the reasons given above, it may prove a useful fuel.

The authors use a device known as "rotating membrane cell" which is pictured here:

The caption:

Figure 2. RMC technique. Left: View of the cell with the membrane glued at the bottom. Center: Cell rotating in the outer phase. Right: Sketch of the technique.

The technique is closely related to solvent extraction techniques used in the industry, with several extractants in a commercially available mixture of dodecane isomers known as "TPH" containing a small amount of n-octanol, whereupon the solvent is known as "TPH-O." The extractants utilized in these solutions are shown in the following figure:

The caption:

Figure 3. Chemical structures of the molecules used in this study (SO3-Ph-BTP in tetravalent ionic form, counterion: Na+).

The membranes employed are commercially available, one being a hydrophilic membrane, the other hydrophobic:

Two types of membranes were purchased from Merck Millipore: the hydrophilic Omnipore PTFE membrane (JHWP04700, manufacturer’s data: pore size 0.45 μm, porosity of 80%) was employed to contain aqueous solutions, and the hydrophobic Durapore PVDF membrane (HVHP04700, manufacturer’s data: pore size 0.45 μm, porosity of 75%) was employed for organic solutions. The porosity values were also measured by impregnating membranes (glued on a plastic cylinder) with TPH and by measuring the corresponding mass of diluent. The membrane thicknesses, L, were measured by using a digital micrometer.

There is considerable discussion in the paper of the technique, and there is not time to describe all of it in detail. However the separation efficiency and speed are significant.

The last figure in the paper shows distribution coefficients for one system evaluated:

The caption:

Figure 7. Aq/org distribution ratios (1/K, left scale), and separation factor [SF(Am/Eu), right scale], of Eu(III) and Am(III) for an organic 0.2 M TODGA solution in TPH-O and an aqueous 0.5 M HNO3 solution as a function of SO3-Ph-BTP concentration in the aqueous phase up to 40 mM. ( ● ) = Eu(III); ( ○ ) = Am(III).

The authors conclude, noting that the purpose of their work is to provide input for further modeling and development of new systems.

...Extraction and stripping of Eu(III) and Am(III) were studied for various concentrations of nitric acid and TODGA, in mixtures of CyMe4-BTBP with TODGA, and in the presence of the aqueous ligands SO3-Ph-BTP and PTD. It was somewhat striking to find that TODGA is not surface-active at the interface between nitric acid and TPH-O, which is in contrast with the case of TODGA in n-dodecane.

The kinetic data obtained in this work will be used as input parameters in simulation codes (such as, e.g., PAREX, developed at CEA(64,65)) for a modeling of separation processes carried out in extractors (e.g., centrifugal) that operate with a short contact time between the phases.

The experimental results, obtained with TODGA and the two aqueous stripping ligands, show that faster transfer kinetics are associated with higher partitioning for Am(III) over Eu(III). This favorable outcome bodes well for future efficient actinide/lanthanide separation in the nuclear reprocessing industry.

My personal feeling is that we need to move beyond nitric acid dissolution of used nuclear fuel, which is a feature of the chemistry herein.

I believe the cleaner option for future nuclear fuel reprocessing is to conduct some of it in line using techniques including but not limited to distillation. There are also possibilities, some of which were briefly explored in the 1950's, to utilize liquid/liquid extractions in line using inorganic liquids.

However, it may be that liquid/liquid interfaces may be important at some point in future techniques. We need to take a deeper look at molten salt based separations, included but not limited to organic ionic liquids. Much attention is being paid to these substances. Finally a driving force for both separations and dissolution may not need to involve mechanical forces. Electrochemical techniques, including those involving liquid membranes are worthy of consideration.

There are really a wealth of options for the treatment of nuclear fuels and the recovery and use of the radioactive and non-radioactive materials therein. These materials are probably, in my view, the best shot we have to save the world.

I hope your weekend was pleasant as much as it was safe.

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Reply Liquid/Liquid Extraction Kinetics for the separation of Americium and Europium. (Original post)
NNadir Aug 9 OP
hunter Aug 10 #1
NNadir Aug 13 #2

Response to NNadir (Original post)

Mon Aug 10, 2020, 01:06 PM

1. Americium is frequently mentioned as a fuel for tiny fission reactors.

These reactors might be useful as a neutron source for medicine or as power sources for space exploration.

Apparently 20 grams of Americium 242 is all you need for a 10-kW reactor:


Or maybe you could get to mars in two weeks riding this hellfire:


Americium is the element used in smoke detectors. One gram is enough to make 3 million smoke detectors.

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Response to hunter (Reply #1)

Thu Aug 13, 2020, 11:56 AM

2. Thank you for this interesting little paper. I have a few of this guy's papers in my files on Am.

I find that kind of critical mass a little bit surprising, even with a beryllium reflector and a homogenous reactor.

I'm not sure what library he used for it; some of these libraries are less reliable than others. The reactor is designed to produce neutrons for boron therapy; my guess is that there are better, simpler sources, in particular the old reliable Cf-252.

The author, Yigal Ronan, has written elsewhere on the isolation of Am-242m: Detailed Design of 242mAm Breeding in Pressurized Water Reactors. This paper refers on the difficult task of obtaining relatively pure Am-242m. He approaches it by arranging strong neutron absorbers around the Am-241 target.

This I think, is counter-productive, and doesn't exploit the high multiplicity which Americium isotopes provide.

Am-242 represents 4 captured neutrons. This is best achieved in breeder situations; otherwise the formation of Americium represents lost neutrons. Since the multiplicity of americium fission across the neutron spectrum is reportedly higher than 3, the use of americium fuel helps recover these neutrons for other uses.

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