Advanced partitioning and transmutation

The long-lived radioactive components of irradiated nuclear fuel serve to create a monumental challenge for managing the wastes from operation of nuclear power reactors. In the open fuel cycle option, the presence of alpha — emitting 239Pu (0.5%, tm = 24,100 yr), 237Np (0.05%, tm = 2,100,000 yr), and MAm (t1/2 = 432 yr, low primary yield, but of increasing importance with time in storage as the daughter of the в decay of 241Pu — 0.1%) demands that a geological disposal system must ensure retention of these isotopes within the repository environment for 240,000 years or more. Even with partial recycle (recovery of Pu isotopes to MOX fuels), the residues of Am and Np isotopes leave a considerable challenge for repository development. The need to qualify a repository for a span of time longer than that of all human civilization has been used to support arguments for further treat­ment (beyond PUREX processing) of dissolved nuclear fuels to enable transmutation of these elements in advanced reactors or by accelerator transmutation.

With the PUREX process, or an equivalent U/Pu management scheme that addresses concerns about the proliferation of nuclear weapons, the challenge of advanced nuclear fuel cycles to support transmutation of minor actinides ultimately focuses on the isolation of Np, Am and perhaps Cm from fission products. Of these, the easiest task is control of the partitioning of Np. In PUREX processing, Np can be made to partition to either the aqueous or organic phase based on its oxidation state. In HNO3 media less concentrated than about 4 m, NpO2+ is the dominant species, which is poorly extracted by TBP, hence Np remains with the fission products. At higher concentrations of HNO3, NpO2+ is susceptible to oxidation to the hexava — lent state (NpO22+), which readily co-extracts with UO22+. Other options feature the introduction of reducing agents that convert NpO2+ to extract­able Np4+. In modern adaptations of PUREX (e. g., UREX, COEX), mixtures of U/Np/Pu are created to increase proliferation resistance of the products.

Once choices are made regarding the partitioning of Np and Pu in PUREX/UREX/COEX processing schemes, the remaining challenge is recovery (for transmutation) of Am, with or without Cm. Following PUREX/ UREX, Am and Cm remain with the fission products in the raffinate. Am and Cm are most stable in acidic solutions as the trivalent cations, which is also the common oxidation state of the fission product lanthanides. This coincidence of stable oxidation states and limited reduction/oxidation (redox) adjustment options is unfortunate, as the trivalent actinides and lanthanides are also of similar size, and thus quite difficult to separate. Their mutual separation is an important issue, as some members of the lanthanide series have high neutron capture cross sections, thus they compete for the neutrons that might otherwise transmute actinides.

However, the tendency of cation radii to decrease across both series, combined with the slightly softer acidic nature of the actinide ions com­pared to the lanthanides in terms of the hard-soft acid-base theory (Pearson, 1963), offers an avenue to an aqueous processing pathway to support this separation and transmutation goal. The simplest solution to the separation of trivalent actinides from fission products is to develop soft donor extract­ant molecules that selectively remove trivalent actinides from the fission product mixture (i. e., rejecting the lanthanides). Unfortunately, soft donor extractants based on N or S donor functional groups have much higher affinity for selected transition metal fission products (noble metals) and corrosion products (Fe, Ni, Co, Zn); as a result a single-step isolation of the trivalent actinides from this mixture has thus far proven elusive. Advanced fuel cycle research efforts have as a result evolved toward a two-step process of extracting trivalent lanthanides and actinides away from the remainder of the fission residues followed by the mutual separation of the f-elements. The current state of this art has been described in detail previously (Nash et al, 2006), hence will be summarized only briefly here.

Motivated by the need to clean up residual wastes within the US nuclear weapons complex, research into the development of new reagents for total actinide removal from dissolved acidic waste solutions (similar to PUREX raffinates) arguably began with studies of extractants combining phosphine oxide and amide functional groups in the same extractant molecule. Based on the work of Siddall (1958, 1963, 1964), Horwitz and co-workers devel­oped a family of carbamoyl(methyl)phosphine oxide extractants designed to follow PUREX with selective partitioning of trivalent f-elements (along with any residues of tetra — and hexavalent actinides); the TRUEX process was developed based on these reagents (Schulz and Horwitz, 1988).

With the aim of eliminating the use of organophosphorus extractants from advanced reprocessing schemes, French researchers followed with the demonstration of similar separation systems based on tetra-alkyldiamides of malonic acid (malonamides; Madic et al.,1994), which performed simi­larly from more acidic media and with fewer complications arising from extractant degradation products. Sasaki and co-workers (Morita et al., 2002; Sasaki et al., 2001) subsequently produced tetra-alkyl diamides of diglycolic acid (e. g., tetraoctyldiglycolamide, TODGA) that introduced the interesting feature of maximum extraction of trivalent actinide nitrates under some conditions. In China, Zhu and co-workers developed a similar system based on commercially available trialkylphosphine oxide extractants (Zhu et al., 1983 ; Zhu and Song, 1992). This process required operation at reduced concentrations of HNO3 ; but similarly resulted in selective partitioning of trivalent actinides and lanthanides away from PUREX raffinates. Fuel cycle research continues to examine the potential for application of each of these classes of reagents.

With the f-elements thus separated from the problematic (from a separa­tions perspective) transition elements, attention turns toward the mutual separation of trivalent lanthanides and actinides. Several different approaches to pyrometallurgical separations offer electrochemical path­ways to a partial separation of the groups based on electrodeposition (Nash et al;, 2006). Though research continues on both aqueous and pyrometal — lurgical methods, it is clear at present that more efficient separations can be achieved using wet methods (at the cost of increased secondary waste volumes that are inherent to aqueous techniques). Both soft donor extract­ants and water soluble holdback reagents to selectively prevent actinide extraction have been studied for the purpose of separating these groups, each with some degree of success, but also with limitations. Though some have been characterized through the pilot stage using actual dissolved used fuel, to date none have been implemented on an industrial scale. The most mature options are described in brief in Table 5.1.

In the partial recycle options, U, Np, Pu mixtures are the products that must be managed for their possible recycle in (first) light water and (sub­sequent to the first cycle) advanced reactors. The wastes from this option that require disposal in a geological repository include most fission products (noble gases and potentially other volatile fission products like iodine would be managed separately) and the transplutonium actinides (Np, Am, and Cm). The likely form of this waste would be some formulation of boro — silicate glass. Repository retention times would be determined principally by the decay/release profiles of 241243Am, 129I, and 99Tc. In the more advanced fuel cycle options involving actinide transmutation, long-term radiotoxicity is reduced through the transmutation of Am, but 1 29I and 9 9Tc remain a long-term concern. The radiotoxicity of residual 129I and 99Tc is considered to be less than that of the original uranium ore; however, the environmental mobility of these isotopes remains a concern. The radiotoxicity is reduced

Table 5.1 Summary of aqueous processing options for separation of minor actinides from fission product lanthanides

Process

Comments

Status

TALSPEAK

DTPA complexes An(III) and HDEHP extracts Ln(III) from acidic streams.

Hot cell demonstration

‘Reverse’

TALSPEAK

An(III) and Ln(III) are both extracted. DTPA (and a carboxylic acid buffer) is then used to strip An from Ln

Complete partitioning system demonstrated with dissolved spent fuel in the CTH process

DIDPA

Modified reverse TALSPEAK. DIDPA used as the extractant, DTPA to strip Am, Cm.

Hot cell demonstration with actual waste

SETFICS

Modified TRUEX process. Uses DTPA complexant and modified CMPO extractant.

Not tested with actual waste

PALADIN

Malonamide co-extracts An(III) and Ln(III) from acid stream. HDEHP extractant and DTPA complexant selectively strips actinides.

Successfully tested

SANEX

processes

Selective extraction of An(III) from Ln(III):

Cyanex 301 uses R2PSSH as extractant. ALINA uses two extractants: a dithiophosphinic acid and trioctylphosphine oxide.

BTP/BTBPs and TMAHDPTZ w/ octanoic acid have been used as neutral extractants.

Demonstrated with Am, Ln mixtures and genuine wastes

DIAMEX-SANEX,

GANEX,

Combined diamide/cation exchanging (HDEHP) extractant with selective stripping of trivalent actinides using buffered solutions of aminopolycarboxylic acid complexants.

Demonstrated with Am, Ln mixtures and genuine wastes

DTPA — Diethylenetriaminepentaacetic acid, DIDPA — Diisodecylphosphoric acid, CMPO — Octyl. Phenyl-N, N-diisobutyl carbamoylmethyl phosphine oxide, HDEHP — Di-2-ethyl(hexyl)phosphoric acid, BTP/BTBP — bis-1,2,4-triazinyl pyri- dine/bipyridine, TMAHDPTZ — 4,6-di-(pyridin-2-yl)-2-(3,5,5-trimethylhexanoylamino)- 1,3,5-triazine.

to less than the uranium ore after the passage of ten half-lives of 137Cs and 90Sr.

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