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.     

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.     

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.     

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.     

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.     

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.     

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.     

CARBEX process,a new technology of reprocessing of spent nuclear fuel

Advances in the CARBEX process, a new aqueous chemical method for reprocessing of spent nuclear fuel (SNF) in carbonate media, are considered. A review of carbonate methods for SNF reprocessing is given. The CARBEX process concept is presented and experimental data for every stage of the CARBEX process: high-temperature oxidation of spent fuel composition, its oxidative dissolution in carbonate aqueous solutions, extraction refining of U(VI) and Pu(VI), solid-phase re-extraction of carbonate complexes of U(VI) and Pu(VI), and obtaining of uranium and plutonium dioxide powders for fabrication of ceramic nuclear fuel, are discussed. It was shown that the CARBEX process can be more effective and safe than the well-known industrial PUREX process.     

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.     

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.     

New possibilities in isotope correlation analysis of spent nuclear fuel

Isotopic correlation analysis is believed to make possible quick and accurate determinations of nuclear fuel parameters for reactor operation, reprocessing, fuel management and nuclear safeguards. Correlation dependencies have been found between ratios of fission products on the one hand and isotope ratios of the heavy elements on the other hand. The use of the¹⁵⁴Eu/¹⁵⁵Eu ratio in correlation analysis was proposed by SMULEK. The scope of useful applications of this isotopic ratio has been further investigated. A quick and time-saving method to measure the¹⁵⁴Eu/¹⁵⁵Eu ratio has been elaborated. The atomic ratios have been found by internal calibration using the computer programme ABSINT. Beside this the atomic ratios of¹⁵⁴Eu/¹⁵⁵Eu as a function of nuclear fuel burn-up have been calculated using the computer programme ISOTOP. The correlation between the¹⁵⁴Eu/¹⁵⁵Eu ratio and nuclear fuel burn-up is best approximated by a quadratic function. Up to a burn-up of 1% fima a linear function can be used.     

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.     

Development of polyoxadiazole nanocomposites for high temperature polymer electrolyte membrane fuel cells

Novel nanocomposite membranes were prepared with sulfonated polyoxadiazole and different amounts of sulfonated dense and mesoporous (MCM-41) silica particles. It has been shown that particle size and functionality of sulfonated silica particles play an important role when they are used as fillers for the development of polymer electrolyte nanocomposite membrane for fuel cells. No significant particle agglomerates were observed in all nanocomposite membranes prepared with sulfonated dense silica particles, as analyzed by SEM, AFM, TGA, DMTA and tensile tests. The T_g values of the composite membranes increased with addition of sulfonated silica, indicating an interaction between the sulfonic acid groups of the silica and the polyoxadiazole. Constrained polymer chains in the vicinity of the inorganic particles were confirmed by the reduction of the relative peak height of tan δ. A proton conductivity of 0.034 S cm⁻¹ at 120 °C and 25% RH, which is around two-fold higher than the value of the pristine polymer membrane was obtained.     

Problems of spent nuclear fuel management in Russia and prospects for their solution

The policy of ROSATOM State Nuclear Energy Corporation in the field of spent nuclear fuel (SNF) management is reflected in the SNF Management Concept (2008) and is based on the following principles.     

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.     

Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel

The purpose of this study was to examine the nail response to concurrent intake of methyl mercury (MeHg) and Se. Weanling male Long Evans rats were fed diets containing Se and MeHg. The Se concentration in the nail was well correlated (r ²>0.90) with dietary Se in every MeHg supplementation group. Good correlation was observed between dietary MeHg and nail Hg concentration in every Se group. Exposure to high levels of MeHg resulted in a lower (p<0.03) measured nail Se concentration in the highest Se experimental group relative to control.     

Management of carbon dioxide released from spent nuclear fuel through voloxidation

Journal of Radioanalytical and Nuclear Chemistry - The management of radioactive carbon (C-14) from spent nuclear fuel (SNF) in a voloxidation process is vital to prevent radioactive contamination...