Recovery of palladium from reprocessing waste of spent nuclear fuel

A method has been investigated for high-speed and efficient recovery of palladium from reprocessing waste of spent nuclear fuel by mixing the matrix feedstock with a small amount of KI and an appropriate inert solvent (such as kerosene) as collecting agent. Equilibrium of the reaction can be obtained in less than 30 sec. Percent recovery of palladium is more than 97%. Decontamination coefficient is high. No loss of effectiveness of the system was observed below 1×10⁶ rad of irradiation.     

Recovery of palladium from basic reprocessing waste of spent nuclear fuel

The methods including collection method, extraction-collection method, and special extraction-collection method have been investigated for high speed and efficient recovery of palladium from high pH reprocessing waste of spent nuclear fuel. The equilibrium of the reactions can be obtained is less than 1 minute. The maximum percent recovery of Pd is about 89%, 96% and 97% for collection, extraction-collection, and special extraction-collection methods, respectively. Nearly 100% of back extraction of Pd in the organic phase can be attained by using 7.4M ammonia solution, with a phase ratio of 1:1. The purity of the Pd product is high. The percent recovery of Pd is constant, up to 5·10³ Gy of irradiation dose.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

New extractants for reprocessing of spent Th-U fuel

A study on solvent extraction of U(VI), Th(IV) and HNO₃ from nitric acid media by DEHSO is described. Extraction coefficients of U(VI), Th(IV) and HNO₃ as a function of aqueous HNO₃ concentration, extractant concentration and temperature have been studied. From the data the compositions of extracted species, equilibrium constants and enthalpies of extraction reaction have been evaluated. Back-extraction of U(VI) and Th(IV) from the organic phase by dilute nitric acid has also been tested. All studies on DEHSO are compared with TBP.     

Lanthanides extraction processes in molten fluoride media: Application to nuclear spent fuel reprocessing

This paper describes four techniques of extraction of lanthanide (Ln) elements from molten salts in the general frame of reprocessing nuclear wastes; one of them is chemical: the precipitation of Ln ions in insoluble compounds (oxides or oxyfluorides); the others use electrochemical methodology in molten fluorides for extraction and measurement of the progress of the processes: first electrodeposition of pure Ln metals on an inert cathode material was proved to be incomplete and cause problems for recovering the metal; electrodeposition of Ln in the form of alloys seems to be far more promising because on one hand the low activity of Ln shifts the electrodeposition potential in a more anodic range avoiding any overlapping with the solvent reduction and furthermore exhibit rapid process kinetics; two ways were examined: (i) obtention of alloys by reaction of the electroreducing Ln and the cathode in Ni or preferably in Cu, because in this case we obtain easily liquid compounds, that enhances sensibly the process kinetics; (ii) codeposition of Ln ions with aluminium ions on an inert cathode giving a well defined composition of the alloy. Each way was proved to give extraction efficiency close to unity in a moderate time.     

Improvements in the recovery of palladium from reprocessing waste of spent nuclear fuel reported earlier

A collection-precipitation (CP) method and an extraction-collection-precipitation (ECP) method have been investigated for high speed and efficient recovery of palladium from reprocessing waste of spent nuclear fuel. For the CP method, a matrix feedstock solution and a small volume of inert solvent, whose specific gravity is greater than that of feedstock solution, as collecting and protective agent, were powerfully mixed with an appropriate amount of potassium iodide solution at 15°C for 1 minute, recovering PdI₂ precipitate after centrifuging. For the ECP method, an extraction complex K(benzo-15-crown-5)₂I in the same organic solvent as for the CP method was mixed with feedstock solution. For both methods, the percent recovery of palladium is more than 99% and 95% for irradiation doses of 1·10³ and 5·10³ Gy, respectively. Decontamination of the palladium product is good.     

Characterization of zirconium complexes of interest in spent nuclear fuel reprocessing by electrospray ionization mass spectrometry

Liquid-liquid extraction of zirconium, one of the most important fission products, was followed using electrospray ionization mass spectrometry under conditions simulating reprocessing of nuclear spent fuel. Zr(IV) can precipitate from the organic phase after extraction by dibutylphosphoric acid (HDBP), the most common degradation product of tributylphosphate (TBP) radiolysis. Different complexes were detected with electrospray used in positive or negative ion modes, according to the extraction conditions such as the ligand/metal ratio. Stoichiometry of the Zr(IV) complexes was determined by combining isotopic labeling [H(15)NO(3)] of the aqueous phase in the extraction system and tandem mass spectrometry experiments. These results were compared with the species observed using other techniques reported in the literature. The mechanisms of ionization/desorption of these complexes are proposed depending on the organic ligand character (neutral (L) such as TBP, or acidic (HL') such as HDBP), and the ionization mode used. Copyright 2000 John Wiley & Sons, Ltd.     

Behavior of iodododecane in nuclear fuel reprocessing

By simulation experiments with a 10^–5 mol/l solution of iodododecane labeled with¹³¹I in n-dodecane the influence of various materials and conditions, which are possible in nuclear fuel reprocessing, has been investigated. The formation of decomposition products was detected via HPLC with a radioactivity monitor. By means of²⁵²Cf plasma-desorption mass spectrometry (PDMS) the decomposition products were identified. It was found that a temperature of 100°C favored the formation of iodoalkanes with chain lengths of C₁ to C₁₁. The presence of TBP(tri-n-butyl-phosphate) accelerated the decomposition of iodododecane. In pure TBP only iodobutane was formed as a decomposition product.     

N,N-Dialkyl amides as extractants for spent fuel reprocessing: an overview

Reprocessing of spent nuclear fuel is vital for the long-term global nuclear power growth and is the major motivation for developing novel separation schemes. Conventionally, PUREX and THOREX processes have been proposed for the reprocessing of U and Th based spent fuels employing tri-n-butyl phosphate (TBP) as extractant. However, based on the experiences gained over last five–six decades on the reprocessing of spent fuels, some major drawbacks of TBP have been identified. Evaluation of alternative extractants is, therefore, desirable which can overcome at least some of these problems. Extensive studies have been carried out on the evaluation of N,N-dialkyl amides as extractants in the back-end of the nuclear fuel cycle for addressing the issues related to the reprocessing of U and Th based spent fuels. Under advanced fuel cycle scenario, efforts are also being made by countries with a developed nuclear technological base to provide safe nuclear power to other countries and to minimize proliferation concerns worldwide. This paper presents an overview of studies carried out in our laboratory on different aspects of reprocessing of U and Th based spent fuels employing N,N-dialkyl amides as extractants.     

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.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Development of a data-mining methodology for spent nuclear fuel forensics

The purpose of this study is to categorize the type of spent nuclear fuels using simulation data-based classification methods. Considering the practical conditions making the full analysis of radioactive nuclides difficult, the classification methods were designed to be robust to noise and missing information. The strength and weakness of three classifiers, linear discriminant analysis, quadratic discriminant analysis and support vector classification were compared, which is developed by the history information such as burnup, enrichment, and cooling type generated from ORIGEN-ARP upon fuel assembly types. Auto-Associative Kernel Regression improved outlier management as a pre-processing technique.     

Magnesium phosphate clamps for spent nuclear fuel

Magnesium phosphate clapms are offered for conditioning spent nuclear fuel (SNF). It is experimentally shown that the starting material has a high fluidity which persists for not less one hour. It will allow reliable pouring tight SNF. The solid material with a density of 1.5–1.8 g cm⁻³ can function as a protective barrier: It is insoluble in water, and strontium and cesium radionuclides are strongly fixed in the material structure.     

Development of a high temperature treatment device for spent nuclear fuel

Novel reprocessing schemes and techniques are the focus of the Euratom FP7 project “Actinide Recycling for Separation and Transmutation” (ACSEPT), where the Paul Scherrer Institute (PSI) is represented in the pyrochemical domain. The subject of investigation is the selective separation of fission products (FPs) from spent nuclear fuel as a head-end step to either classical hydro based or pyro processes which are not yet applied on a large scale. The selective removal of FPs that are major contributors to the overall radiation dose or bear great potentials in terms of radiotoxicity (i.e. cesium or iodine), is advantageous for further processes. At PSI a device was developed to release volatile FPs by means of inductive heating. The heating up to 2,300 °C promotes the release of material that is further transported by a carrier gas stream into an inductively coupled plasma mass spectrometer for online detection. The carrier gas can be either inert (Ar) or can contain reducing or oxidizing components like hydrogen or oxygen, respectively. The development of the device by computer aided engineering approaches, the commissioning and evaluation of the device and data from first release experiments on a simulated fuel matrix are discussed.     

Identification of iodoorganic compounds in kerosene from nuclear fuel reprocessing

In kerosene samples from nuclear fuel reprocessing, iodoalkanes with chain-lengths from C₄ to C₁₃ have been identified. The kerosene samples were purified by means of solid-phase extraction. By this method other fission products like¹²⁵Sb and¹⁰⁶Ru were quantitatively removed from the solution. The only remaining radioactive nuclide was thus¹²⁹I. The iodoorganic compounds in the kerosene from the solvent were enriched from 6000 Bq/L to 100 000 Bq/L¹²⁹I by vacuum distillation. Chromatographic separation by HPLC, fractionation, and -measurement of the fractions showed that at least one polar and one nonpolar iodoorganic compound were present. Derivatisation of the iodoorganic compounds with, 1,4-diazabicyclo-2,2,2-octane to quatermary ammonium salts and²⁵²Cf plasma desorption mass spectrometry of the products revealed that the main iodoorganic constituents in the kerosene were iodobutane as polar and iodododecane as nonpolar compound in approximately equal concentrations.     

Effects of the shutdown of a nuclear fuel reprocessing plant on nearby streams

In a study conducted in 1971, levels of tritium were found in Cattaraugus-Creek, a stream in Western New York State. This material was attributed to the operation in West Valley, New York of the world's first commercial nuclear fuels reprocessing plant. Several fission fragment isotopes in addition to tritium were also observed in Buttermilk Creek, one of the tributaries of Cattaraugus Creek that runs through the reprocessing plant grounds. The plant ceased processing nuclear fuel in December 1971, and a new set of measurements in these streams were made to assess the effect of the ending of the plant's operation. Substantially lower concentrations of tritium and no fission produced isotopes have been observed.