Separation of U and Pu in spent nuclear fuel sample using anion-exchange-group-introduced porous polymer sheet for ICP-MS determination

Anion-exchange porous sheets were prepared by radiation-induced graft polymerization and subsequent chemical modifications. A diethylamino (DEA) group as an anion-exchange group was introduced into the polymer chain grafted onto a porous sheet. The DEA group-introduced porous sheet was cut into disks 13 mm in diameter and 3 mm in thickness to fit an empty cylindrical cartridge (DEA cartridge). The DEA sheet had a DEA group of 3.4 mol/kg of the DEA-group-containing porous sheet and a linear velocity of 46 m/h at a permeation pressure of 0.1 MPa at 298 K. The adsorption capacity of the DEA cartridge for FeCl₄⁻ as a model ion in equilibrium with 1 g-Fe(III)/L in 10 M HCl was 0.17 mmol-Fe(III)/DEA cartridge. No Pu leakage during the permeation of 5 mL of 10 M HCl-0.1 M HNO₃ containing Pu ionic species through the DEA cartridge was observed irrespective of the permeation rate ranging from 0.3 to 80 mL/min. A solution containing known amounts of ²³³U, ²⁴⁰Pu, and ²⁴¹Am in 10 M HCl-0.1 M HNO₃ was loaded onto the DEA cartridge. U and Pu were retained on the DEA cartridge, while Am was allowed to pass through the DEA cartridge. Subsequently, 7 M HNO₃ and 1 M HCl as eluents were permeated to elute U and Pu from the DEA cartridge, respectively. The decontamination factor of U in a Pu fraction, defined by dividing the activity of U in the feed solution by that of U in the Pu fraction, was 2.7 × 10⁵, which is desirable for the highly accurate ICP-MS determination of Pu for samples containing both U and Pu. The method using the DEA cartridge was validated by measuring isotopic compositions and quantities of U and Pu in a spent nuclear fuel sample by double-focusing magnetic sector ICP-MS.     

Impact of dissolution on the uncertainty of spent fuel analysis

One of the objectives of the French Alternative Energies and Atomic Energy Commission in the Marcoule Centre is to accurately quantify the composition of nuclear spent fuel, i.e. to determine the concentration of each isotope with suitable measurement uncertainty. These analysis results are essential for the validation of calculation codes used for the simulation of fuel behaviour in nuclear reactors and for nuclear matter accountancy. The different experimental steps are first the reception of a piece of spent fuel rod at the laboratory of dissolution studies, and then dissolution in a hot cell of a sample of the spent fuel rod received. Several steps are necessary to obtain a complete dissolution. Taking into account these process steps, and not only those of analysis for the evaluation of measurement uncertainties, is new, and is described in this paper. The uncertainty estimation incorporating the process has been developed following the approach proposed by the Guide to the Expression of Uncertainty in Measurement (GUM). The mathematical model of measurement was established by examining the dissolution process step by step. The law of propagation of uncertainty was applied to this model. A point by point examination of each step of the process permitted the identification of all sources of uncertainties considered in this propagation for each input variable. The measurement process presented involves the process and the analysis. The contents of this document show the importance of taking the process into account in order to give a more reliable uncertainty assessment to the result of a concentration or isotope ratio of two isotopes in spent fuel.     

Optimization of steam pretreatment of SO2-impregnated corn stover for fuel ethanol production

In this study, corn stover with a dry matter content of 20% was impregnated with SO₂ and then steam pretreated for various times at various temperatures. The pretreatment was evaluated by enzymatic hydrolysis of the solid material and analysis of the sugar content in the liquid. The maximum overall yield of glucose, 89% of the theoretical based on the glucan in the raw material, was achieved when the corn stover was pretreated at 200°C for 10 min. The maximum overall yield of xylose, 78%, was obtained with pretreatment at 190°C for 5 min.     

Direct estimation of 90Sr in water samples of spent fuel storage bay using partial counting rate technique

The direct estimation of ⁹⁰Sr by β counting from a mixture of other β and γ emitter is often difficult due to the efficiency variation among the β-emitters and the unknown nature of the sample. This paper deals with use of a combination of β and γ spectrometry measurements in estimating the activity of ⁹⁰Sr, pure β emitter from a mixture of other β–γ emitters in water samples. This procedure offers a simple, easy to use, rapid and a reliable method for ⁹⁰Sr estimation as an alternative to the tedious radiochemical separation procedure in this specific case.     

Critical evaluation of sample pretreatment techniques

Sample preparation before chromatographic separation is the most time-consuming and error-prone part of the analytical procedure. Therefore, selecting and optimizing an appropriate sample preparation scheme is a key factor in the final success of the analysis, and the judicious choice of an appropriate procedure greatly influences the reliability and accuracy of a given analysis. The main objective of this review is to critically evaluate the applicability, disadvantages, and advantages of various sample preparation techniques. Particular emphasis is placed on extraction techniques suitable for both liquid and solid samples. Figure Miniaturised extraction techniques allow sensitive analysis of also small sample volumes.     

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.     

Optimization of slitlike carbon nanopores for storage of hythane fuel at ambient temperatures

Carbons with slitlike pores can serve as effective host materials for storage of hythane fuel, a bridge between the petrol combustion and hydrogen fuel cells. We have used grand canonical Monte Carlo simulation for the modeling of the hydrogen and methane mixture storage at 293 K and pressure of methane and hydrogen mixture up to 2 MPa. We have found that these pores serve as efficient vessels for the storage of hythane fuel near ambient temperatures and low pressures. We find that, for carbons having optimized slitlike pores of size H congruent with 7 A (pore width that can accommodate one adsorbed methane layer), and bulk hydrogen mole fraction >or=0.9, the volumetric stored energy exceeds the 2010 target of 5.4 MJ dm(-3) established by the U.S. FreedomCAR Partnership. At the same condition, the content of hydrogen in slitlike carbon pores is approximately = 7% by energy. Thus, we have obtained the composition corresponding to hythane fuel in carbon nanospaces with greatly enhanced volumetric energy in comparison to the traditional compression method. We proposed the simple system with added extra container filled with pure free/adsorbed methane for adjusting the composition of the desorbed mixture as needed during delivery. Our simulation results indicate that light slit pore carbon nanomaterials with optimized parameters are suitable filling vessels for storage of hythane fuel. The proposed simple system consisting of main vessel with physisorbed hythane fuel, and an extra container filled with pure free/adsorbed methane will be particularly suitable for combustion of hythane fuel in buses and passenger cars near ambient temperatures and low pressures.     

Problems of modernization of spent nuclear fuel extraction processing

In order to try out the key process stages of prospective technologies and gain practical experience in the reduction of liquid process and non-process radioactive waste, works on establishment of the experimental and demonstration center have been initiated.     

Automation of sample pretreatment for liquid chromatography

The purpose of this review is to discuss the strategic problems of automating sample preparation (SP) for high performance liquid chromatography (HPLC). There is a general feeling that SP is the bottleneck of many HPLC procedures. Despite numerous reports of successful automation of SP, there are still many laboratories using manual or semiautomated SP procedures. This calls for a reevaluation of the present situation.     

Analysis of high burnup spent nuclear fuel by ICP-MS

Summary We have used inductively coupled plasma mass spectrometry (ICP-MS) as the primary tool for determining concentrations of a suite of nuclides in samples excised from high-burnup spent nuclear fuel rods taken from light water nuclear reactors. The complete analysis included the determination of ⁹⁵Mo, ⁹⁹Tc, ¹⁰¹Ru, ¹⁰³Rh, ¹⁰⁹Ag, ¹³⁷Cs, ¹⁴³Nd, ¹⁴⁵Nd,148Nd,147Sm, ¹⁴⁹Sm, ¹⁵⁰Sm, ¹⁵¹Sm, ¹⁵²Sm, ¹⁵¹Eu, ¹⁵³Eu, ¹⁵⁵Eu, ¹⁵⁵Gd, ²³⁷Np, ²³⁴U, ²³⁵U, ²³⁶U, ²³⁸U, ²³⁸Pu, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ²⁴²Pu, ²⁴¹Am, ^242mAm, and ²⁴³Am. The isotopic composition of fissiogenic lanthanide elements was determined using high-performance liquid chromatography (HPLC) with ICP-MS detection. These analytical results allow the determination of fuel burn-up based on ¹⁴⁸Nd, Pu, and U content, as well as provide input for storage and disposal criticality calculations. Results show that ICP-MS along with HPLC-ICP-MS are suitable of performing routine determinations of most of these nuclides, with an uncertainty of ±10% at the 95% confidence level.     

Adsorption of cesium from spent nuclear fuel basin water

Summary Batch equilibrium and kinetic measurements were performed for Cs⁺ exchange in silicotitanate zeolite (Ionsiv^® TIE-96) at 30 and 60 °C. The Langmuir isotherm equation provided a good fit of the equilibrium data. The heats of exchange reaction between Cs⁺ in the aqueous solution and Na⁺ in the zeolite structure were derived from the equilibrium data. The results indicate that the exchange mechanism is different from that of physical adsorption on heterogeneous materials. The apparent diffusion coefficients and activation energy were derived from the kinetic data and the values obtained for inter-diffusion of Cs⁺ and Na⁺ cations in the zeolite structure were 1.33 ^. 10^-12 and 1.04 ^. 10^-12 cm^{2 .} s^-1 at 30 and 60 °C, respectively. The activation energy for Cs⁺ was 1.7 kcal/mol, suggesting that the Cs⁺ cation can access easily all the sites in the zeolite framework. Thus, the exchange of Cs⁺ with Na⁺ in the zeolite was not hindered by ion-sieve effects.     

Electrochemical procedures in the treatment of spent nuclear fuel

The use of an electrochemical process for U/Pu partitioning has demonstrated a good performance and is a safe alternative for nuclear facilities. Its great advantages are the lack of introduction of foreign ions into the process and, especially, the minimization of the waste volume generated. For the introduction of electrochemical U/Pu partitioning in the 2nd Pu purification cycle, preliminary studies were carried out with a single mixer-settler unit. Based on the results, an 8-stage electrolytic mixer-settler (M-S MIRELE) was designed. Titanium was MIRELE's housing material (cathode) and platinum the anode, insulated with PTFE. The Pu recovery was higher than 99%, indicating the efficiency of this equipment.     

Determination of14C in spent HTGR fuel elements

Spent HTGR-fuel spheres were analysed chemically to determine their¹⁴C content.¹⁴C is mainly (∼96%) produced in the graphite matrix of the fuel. About 75% of the¹⁴C is derived from¹⁴N. More than 99% of¹⁴C is released as CO₂ during the combustion of the fuel elements. Assuming no¹⁴C retention, the maximum body burden at the critical region of a 50000 MW HTGR reprocessing plant will be ∼ 70 mrem/a. Paper presented at the 9th Radiochemical Conference, Piestany, CSSR.     

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.     

Leaching of spent nuclear fuel in the presence of siderophores

Metal species that are dissolved in water can be transported in the environment, because they can be mobile. Microorganisms can affect metal mobility by excreting organic ligands with high metal affinity. Siderophores are organic ligands with high affinities for Fe³⁺. They are also able to form complexes with other metals such as actinides. Many countries plan to deposit spent nuclear fuel in deep geological repositories. Microorganisms are present in these subterranean environments and could potentially affect the repository. In this study, the effect of microbial siderophores on the dissolution behavior of two fragments of a spent nuclear fuel pellet was investigated. The commercial hydroxamate siderophore, deferoxamine mesylate (DFAM), and pyoverdin siderophores, isolated from cultures of Pseudomonas fluorescens (CCUG 32456A), were used. DFAM and lyophilized pyoverdins were dissolved in synthetic groundwater to final concentrations of 10 μM and 2.5·10⁻² g·L⁻¹, respectively. The fuel pellet fragments were kept in sealed pressure vessels at 10 bars of H₂. The pyoverdin solution was first tested, followed by the DFAM solution and the pure synthetic groundwater. Samples were taken on 0, 1, 5, 9 and 14 days after changing the solution in the pressure vessels. The elemental composition of samples was analyzed by means of ICP-MS. The pyoverdin solution maintained significantly higher concentrations of Np and Pu than the pure synthetic groundwater. On the 14th day the concentrations of Np and Pu in the pure synthetic groundwater were 0.01 nM and 0.13 nM, respectively, compared to 0.02 nM and 0.31 nM in the pyoverdin solution. Furthermore, spent nuclear fuel samples were observed to release Ru in the presence of both pyoverdin and DFAM. Hence, it seems that siderophores can form complexes with elements present in spent nuclear fuel.     

金属有机骨架复合材料在样品预处理中的研究进展

金属有机骨架（MOFs）材料是近几年涌现出的一类新型多功能多孔材料,以金属离子或金属簇为配位中心,与含氧或氮的有机配体通过配位作用形成多孔骨架结构。相比于其他传统无机多孔材料,MOFs具有比表面积高、孔隙率大、热稳定性好和结构与功能多样化的特点,因而被广泛用于气体存储、催化、吸附和分离等领域。MOFs复合材料在样品预处理方面的应用引起了研究者们的极大兴趣和广泛关注。由于MOFs材料和不同功能材料如高分子聚合物、碳基材料以及磁性材料组装复合,使MOFs复合材料的性能优于原来的MOFs材料。综述了近年MOFs复合材料在样品预处理的研究应用,尤其是在固相微萃取、固相萃取以及磁性固相萃取等方面的应用。     

Electromigration method for destructive determination of spent nuclear fuel burnup

In this report we discuss the methods and results of VVER spent fuel burnup determination by¹⁴⁶Nd content and the correlation with accumulation of some transplutonium nuclides. For separation of trivalent rare earths and transplutonium elements the method of paper electrophoresis is used. For the quantitative determination of americum and curium isotopes a modification of α-spectrometric analysis is proposed with the chemical yield control of isolated elements using²⁴⁴Cm. The amount of¹⁴³Nd is determined by the isotopic dilution method combined with mass spectrometry with¹⁴²Nd as a tracer.     

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.