TRUEX Process Research Articles

Group hexavalent actinide separation from lanthanides using sodium bismuthate chromatography

Advanced used nuclear fuel (UNF) reprocessing strategies are limited by the complex radiochemical separations and engineering required to achieve the separation of actinides (An) from neutron scavenging lanthanides (Ln). The accessibility of the hexavalent oxidation state for the actinides (U – Am) provides a pathway to achieving a group hexavalent actinide separation from the trivalent lanthanides and Cm. The solid oxidant and ion exchanger, sodium bismuthate (NaBiO3), has been demonstrated to quantitatively oxidize and separate Am from trivalent Cm in a column chromatographic system. This work expands on the use of NaBiO3 chromatography to characterize the adsorption, kinetic, and elution behavior of U, Pu, and Eu. Separation factors over 200 with rapid kinetics were observed at dilute nitric acid concentrations with a complete An/Ln separation achieved in under an hour. The adsorption and chromatographic behavior of key fission products present in various reprocessing raffinates was characterized which demonstrated potential application of a NaBiO3-based separation following a TRUEX process.

Hydrofluorocarbon Diluent for CMPO Without Third Phase Formation: Extraction of Uranium(VI) and Lanthanide(III) Ions

ABSTRACT The extraction behaviors of trivalent lanthanides (La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, and Lu) and hexavalent uranium with octylphenyl(,-diisobutylcarbamoylmethyl)phosphine oxide (CMPO) from a nitric acid medium were studied using hydrofluorocarbon (1,1,1,2,2,3,4,5,5,5- decafluoropentane: HFC-43) as a nonflammable diluent. It was revealed that HFC-43 exhibits high solubility of CMPO even in the absence of tributylphosphate (TBP), which has typically been necessary to prevent third-phase formation in the conventional TRUEX process using -dodecane as a diluent. Distribution ratios have been studied as a function of aqueous HNO concentrations, CMPO concentrations, and ionic radii of lanthanide(III) ions. The stoichiometry of the extracted coordination complex was examined based on the slope analysis, and it was determined that two to three CMPO molecules are coordinated to one lanthanide or uranium ion. The feasibility of the process is indicated by the distribution ratio, which is comparable to the conventional system using -dodecane as diluent. In addition, the safety benefits of applying the nonflammable solvent HFC to the TRUEX process are also presented.

A review on UREX processes for nuclear spent fuel reprocessing

This paper presents a review on the UREX and variants of UREX + processes used for reprocessing of the spent nuclear fuels. The UREX + process consists of nine versions from which seven product streams are obtained. Uranium and technetium are recovered as separate products with the yield > 99% in the first stage of the UREX process. Cesium and strontium are recovered from the UREX waste using CCD-PEG or FPEX process. Plutonium and neptunium are separated using NPEX process with high level of impurity. The NPEX raffinate is treated in the TRUEX process from which minor actinides along with rare-earth elements are separated. Purification of minor actinides from the rare-earth elements is carried out in the Cyanex 301 process. The technologies used for each reprocessing process are described in detail. The flowsheets and process chemistry of all the versions of UREX processes are also reported and discussed in detail. The key issues related to UREX processes for reprocessing the spent nuclear fuels have been summarized. The strategy of the non-proliferation resistance using the advanced reprocessing process is described in the present work.

A TRUEX-Based Separation of Americium from the Lanthanides

The inextractability of the actinide AnO2 + ions by the TRUEX solvent suggests the possibility of a separation of americium from the lanthanides using oxidation to Am(V). The only current method for the direct oxidation of americium to Am(V) in strongly acidic media is with sodium bismuthate. We prepared Am(V) over a wide range of nitric acid concentrations and investigated its solvent-extraction behavior for comparison to europium. While a separation is achievable in principal, the presence of macro amounts of cerium competes for the sparingly soluble oxidant and the oxidant itself competes for CMPO complexation. These factors conspire to reduce the Eu/Am separation factor from ∼40 using tracer solutions to ∼5 for extractions from the first-cycle raffinate simulant solution. To separate pentavalent americium directly from the lanthanides using the TRUEX process, an alternative oxidizing agent may be necessary.

Influence of ionic liquids on actinides extraction by diphenyl(dibutyl)carbamoylmethylphosphine oxide in different solvents from nitric acid solution

Influence of ionic liquids (ILs) addition (1–50 wt%) on extraction efficiency of actinides by diphenyl(dibutyl)carbamoylmethylphosphine oxide (Ph2Bu2) from 3 M HNO3 has been studied. Am(III) distribution ratios in two-phase systems 0.1 M Ph2Bu2 in either DCE or CHCl3–3 M HNO3 depending on the nature of additional ionic liquids: imidazolium-based ILs: [C4mim][PF6], [C4mim][BF4] and phosphonium-based ILs: PPF6, PBF4 and PCl were determined. The highest value of Am(III) extraction ratio change (1040) was found on addition of PPF6 to Ph2Bu2 in CHCl3. Extraction of Pu(IV) and U(VI) by 0.001 M Ph2Bu2 in the presence of [C4mim][PF6] in DCE, CHCl3 or meta-nitrobenzotrifluoride (NBTF) have been investigated. The greatest enhancement of extraction efficiency was observed using CHCl3, the least polar studied solvent. Using a mixture of conventional solvent and ionic liquid as a solvent for extractant enables one to increase distribution ratios and reduce viscosity of organic phase as compared with ionic liquid viscosity. The marked increase of Am(III), Pu(IV) and U(VI) extraction extent by Ph2Bu2 on addition of ionic liquids to the extent of 10 wt% permit one essentially to diminish amounts considerably more expensive carbamoylmethylphosphine oxide(the general name is CMPO) used in TRUEX process.

Extraction behaviour of actinides and lanthanides in TALSPEAK, TRUEX and NPEX processes of UREX+

Abstract Bench-scale studies to determine the extraction behavior of Pu, Np, Am and lanthanides with the organophosphorus extractants TBP, CMPO and HDEHP have been carried out. Based on the results obtained using actual spent nuclear fuel solutions, enhancements to the NPEX, TRUEX and TALSPEAK processes have been successful. In NPEX, >99.94% of both Np and Pu were separated from the fission products. In TRUEX, essentially complete recovery of the actinides (An) and the lanthanides (Ln) was achieved. In TALSPEAK, the complete separation of Pu, Np and Am from the lanthanides was demonstrated several times under various process conditions. The recovery of transuranics (TRU), including Am and Cm, is nearly 100% (below detection limit in the Ln stream), while the total recovery of Ln in the product stream exceeded 99.97%.

New uranium(VI) complexes with macrocyclic schiff bases and their derivatives

An important environmental problem faced by modern radiochemistry is processing of radioactive waste (RW) accumulated both in the manufacture of nuclear weapons and during operation of nuclear power plants. The experience in practical RW processing shows that liquid extraction is the most effective technique for preconcentration, separation, and isolation of radioactive elements. Considerable progress has been achieved in this field in recent years both in Russia and abroad: effective extraction processes for treatment of liquid nitric acid RW with the use of effective chelating agents based on carbamoylmethyl phosphonates (Truex process), malonamides (Diamex process), and other compounds have been developed [1‐3]. However, many countries still store substantial amounts of alkaline RW, which is most difficult to process by extraction. Despite the available data on the extraction of uranium, actinides, lanthanides, and other elements from alkaline solutions [4, 5], the search for effective extractants stable to alkalis is of considerable interest. Several types of polydentate ligands stable in alkaline solutions that selectively bind actinides to form stable complexes are known, in particular, “extended porphyrins” (containing several pyrrole nuclei) [6], macrocyclic Schiff bases (based on pyrrole and polypyrrole derivatives) [7], and salicylidenealdimine derivatives (salens) [8‐11]. However, macrocyclic azomethine I and its reduced analogue II , which we synthesized [12, 13], may prove to be the most promising ligands for actinide preconcentration.

Di(2‐Ethylhexyl) Pivalamide (D2EHPVA): A Promising Extractant for Selective Removal of Uranium from High Level Nuclear Waste Solutions

High concentration of uranium present in high‐level waste is responsible for third phase formation during actinide partitioning either by TRUEX or DIAMEX process. Di(2‐ethylhexyl) pivalamide (D2EHPVA) has been found to be a promising extractant for the selective removal of U over Pu, Np, and Am from simulated PHWR high level waste (SHLW). Role of reducing agents viz. hydroxylamine nitrate (HAN) and hydrazine nitrate (HN) has been evaluated for the adjustment of oxidation states of Pu and Np to inextractable state. Multiple extraction from SHLW containing 0.3 M HAN + 0.8 M HN at 4 M HNO3 using 1 M D2EHPVA in n‐dodecane revealed that ∼98% of U could be extracted in five contacts with <1% cumulative Pu extraction.

Effect of Diluents on the Extraction of Sr2+ from HNO3 Solutions with Dicyclohexano‐18‐crown‐6

Extraction of Sr2+ has been carried out with dicyclohexano‐18‐crown‐6 (DC18C6) in benzene from varying HNO3 concentrations, and the nature of species extracted has been established. The equilibrium constant values for the formation of Sr(NO3)2·DC18C6 have been evaluated, taking into account the aqueous complexing of Sr2+ with NO3 − and with aqueous‐soluble DC18C6, and also the uptake of HNO3 by DC18C6 in the organic phase. Effect of diluents apart from benzene, such as chlorobenzene, dichloroethane, chloroform, 1‐octanol, carbon‐tetrachloride, and nitrobenzene, and various mixtures of benzene and nitrobenzene on the extraction of Sr2+ with DC18C6 has also been studied. On the practical side, the extraction of Sr2+ has been examined with DC18C6 in solvent mixtures of tri‐n‐butyl phosphate (TBP), dodecane, and 1‐octanol. This solvent mixture, at least partially, avoids the disadvantages envisaged while using 1‐octanol alone for the separation and recovery of Sr2+ from minor actinide depleted TRUEX process raffinate.

R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC

Research and development(R&D) activities on partitioning and transmutation of trans-uranium nuclides (TRU) and LLFP and future R&D program in JNC were summarized. Feasibility design studies have been conducting to investigate the characteristics of a fast reactor core with TRU and LLFP transmutation. It was reconfirmed that the fast reactor has a strong potential for transmuting TRU and LLFP, effectively. R&D for establishing partitioning process of TRU apart from the high-level radioactive wastes have been carried out. By several counter-current runs of the TRUEX process using highly active raffinates, a process flow sheet capable of selective partitioning of actinides and fission products was newly developed. JNC has settled a new R&D program concerning partitioning and transmutation of long-lived radioactive waste based on recommendation of check & review for OMEGA program performed by the Ad Hoc Committee under the Atomic Energy Commission of Japan (AEC). The R&D program is composed of the design studies and development of element technologies (fabrication, irradiation) in the “Feasibility Studies” on commercialized fast reactor system and the basic studies with experiments (nuclear data, reactor physics, fuel property, etc.) to establish database and analytical tools for the TRU- and LLFP- containing fuel and core design.

Numerical Simulation of Extraction Behavior of Americium and Nitric Acid in the TRUEX Process.

Numerical Simulation of Extraction Behavior of Americium and Nitric Acid in the TRUEX Process.

A COUNTER CURRENT EXPERIMENT FOR THE SEPARATION OF TRIVALENT ACTINIDES AND LANTHANIDES BY THE SETFICS PROCESS

ABSTRACT The SETFICS process, a variation of TRUEX process, was developed for the recovery of Am and Cm from acidic waste solution and the separation of actinides (III) and light lanthanides. The process uses the general TRUEX solvent as the extracting reagent and a DTPA-sodium nitrate solution for selective stripping of actinides(III). The basic flow sheet is composed of four steps: extraction-scrubbing; acid stripping; actinide(III) stripping; and lanthanide stripping. To demonstrate the usefulness of the SETFICS process, a counter current experiment was conducted using a real TRUEX product solution. Americium and curium were successfully recovered with a solution of 0.05 M DTPA-4 M NaNO3 (pH 2.0). Although the actinide(III) product solution contained Sm and Eu, the decontamination factor of 144Ce/241Am was 72, and most of the light lanthanides, specifically La, Ce, Pr, Nd, were removed. At least 80 % of the lanthanides were separated from the Am and Cm end products. In order to minimize the acidity in the actinide(III) stripping step, the nitric acid which extracted with the trivalent metals was previously removed with a solution of 0.5 M NaNO3 (pH 2.0) in the “acid stripping” step.

Separation of actinides and fission products in high-level liquid wastes by the improved TRUEX process

With the aim of establishing an advanced fuel cycle enabling us to utilize all actinides in the spent fuels, separation of minor actinides in high-level liquid wastes (HLLW) has been studied using a CMPO/TBP mixture extractant. A novel flowsheet is being developed, by counter-current experiments using real HLLW from PUREX experiments, characterized by employing acidification for enhancing extraction of actinides with increasing RuNO 3+ decontamination, and a few mild salt-free reagents for fractional stripping and solvent clean-up steps. The flowsheet can give α- and salt-free HLLW providing a D F Am≥10 3. In the counter-current experiments for SETFICS a flowsheet has been developed to separate An(III) and Ln using DTPA, Am and Cm was recovered without light lanthanide, and the applicability of hydroxylamine nitrate as a salting-out reagent was confirmed.

Mercury Extraction By The TRUEX Process Solvent: III. Extractable Species And Stoichiometry†

ABSTRACT Mercury extraction from acidic aqueous solutions by the TRUEX process solvent, 0.2 M n-octyl(phenyl)N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), 1.4 M tributylphosphate (TBP) in n-dodecane, has not been extensively examined, Research is currently in progress at the Idaho Chemical Processing Plant to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Reactions for two mercury species, HgCl2 and HgCl4 −2, are reported. Classical slope analysis techniques were utilized to evaluate the stoichiometric coefficients of each Hg species independently for both CMPO and TBP. The slope analysis results indicate that even the HgCl4 −2 species extracts as HgCl2. The results also indicate that 1.9 to 2.2 moles of CMPO and 1.5 to 1.81 moles of TBP are required per mole of HgCl2 extracted. A generic equation for mercury extraction into CMPO or TBP has been shown to be: where x = 2 or 4, y = 0 or −2, E = CMPO or TBP, a = the experimentally determined CMPO or TBP stoichiometry, and over lines indicate organic species. Equilibrium constants for HgCl2 species were determined to be 295 when extracted by CMPO and 18.2 when extracted by TBP. Equilibrium constants for the HgCl4 −2 species were found to range between 5.8 and 12.0 for the CMPO reaction and 0.13 to 0.27 with TBP.

Sorption of plutonium, americium and fission products from reprocessing effluents using Rhizopus arrhizus

Rhizopus arrhizus biomass removed more than 95% of 239Pu, 241Am, 95Zr, 144Ce and 152+154Eu from different waste streams generated in Purex as well as Truex processes after suitable adjustment of pH. © Rapid Science Ltd. 1998

Modified TRUEX Process for the Treatment of High-Level Liquid Waste

The TRUEX process has been examined to recover Am and Cm from the high-level liquid waste of Purex reprocessing plant. Continuous counter current extraction and back-extraction experiments were carried out by a mixersettler using simulated waste solution for three process flowsheets, i.e. a process flowsheet developed by Argonne National Laboratory and other two process flowsheets which were modified in the scrub stage. The result indicates that the process flowsheet of Argonne National Laboratory cannot be applied for the high-level liquid waste containing high concentrations of lanthanide and actinide elements because of the formation of insoluble salts of these elements with oxalic acid, which is added to restrict the extraction of fission products such as Mo and Zr. A modified process flowsheet, which had only one scrub stage with high concentration of nitric acid, was found to be the best of three process flowsheets examined, where Nd as a simulated element of Am and Cm was sufficiently recovered and any precipitation of oxalate salt was not observed.

Modified TRUEX Process for the Treatment of High-Level Liquid Waste.

The TRUEX process has been examined to recover Am and Cm from the high-level liquid waste of Purex reprocessing plant. Continuous counter current extraction and back-extraction experiments were carried out by a mixersettler using simulated waste solution for three process flowsheets, i.e. a process flowsheet developed by Argonne National Laboratory and other two process flowsheets which were modified in the scrub stage. The result indicates that the process flowsheet of Argonne National Laboratory cannot be applied for the high-level liquid waste containing high concentrations of lanthanide and actinide elements because of the formation of insoluble salts of these elements with oxalic acid, which is added to restrict the extraction of fission products such as Mo and Zr. A modified process flowsheet, which had only one scrub stage with high concentration of nitric acid, was found to be the best of three process flowsheets examined, where Nd as a simulated element of Am and Cm was sufficiently recovered and...

Chemistry of nuclear fuel reprocessing: Current status

Current status on the chemical aspects of nuclear fuel reprocessing is presented with special emphasis on the Purex process which continues to be the process of choice for the last four decades. Better decontamination from fission products, new methods for uraniumplutonium partitioning and removal of actinides from high active waste are challenging areas in process chemistry. The development work on TRUEX and DIAMEX process for treating high active waste is briefly described. An overview of pyrochemical processes, which are important for Integral Fast Reactor Concept, is presented.

MERCURY EXTRACTION BY THE TRUEX PROCESS SOLVENT. II. SELECTIVE PARTITIONING OF MERCURY FROMCO-EXTRACTED ACTINIDES IN A SIMULATED ACIDIC ICPP WASTE STREAM1

ABSTRACT The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/1. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, 203Hg, was spiked into a simulated solution of SBW containing 1 g/1 mercury. Successive extraction batch contacts with the mercury spiked waste and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl2, from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HN03M and 0.077 with 0.25 M Na2C03. Because experiments described here show that mercury can be extracted from SBW and stripped from the solvent, a process has been developed to partition mercury from the actinides in SBW.

MERCURY EXTRACTION BY THE TRUEX PROCESS SOLVENT. I. KINETICS, EXTRACTABLE SPECIES, &amp; DEPENDENCE ON NITRIC ACID CONCENTRATION1

ABSTRACT Mercury extraction from acidic aqueous solutions by the TRUEX process solvent (0.2 M CMPO - 1.4 M TBP in n-dodecane) has not extensively been examined. Research at the Idaho Chemical Processing Plant (ICPP) is currently being conducted to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Preliminary experiments were performed involving the extraction of Hg203, added as HgCl2, from 0.01 to 10 M HNO3 solutions. Mercury distribution coefficients (DHg) range between 3 and 60 from 0.01 M to 2 M HNO3. At higher nitric acid concentrations, i.e. 5 M HNO3 or greater, DHg significantly decreases to values less than 1. These results indicate mercury is extracted from acidic solutions ˜5 M HNO3. Experimental results indicate the extractable species is HgCl2 from nitrate media, i.e., chloride must be present in the nitrate feed to e...