Inherent isotope dilution analysis of237Np in spent nuclear fuels

A combined -spectrometric and mass-spectrometric determination of²³⁷Np in spent nuclear fuels using²³⁹Np as an inherent spike has been developed. In spent fuel²³⁹Np is in radiochemical equilibrium with²⁴³Am. Neptunium in the tetravalent state is separated by TTA solvent extraction from the other actinides and most of the fission products and the -activity ratio of²³⁷Np to²³⁹Np is measured. The²³⁷Np activity is via²³⁹Np directly linked to the²⁴³Am mass in a sample, which is determined by mass-spectrometric isotope dilution analysis.     

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.     

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.     

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.     

Use of zirconium and ruthenium isotope correlations in the analysis of spent nuclear fuels

An approach is proposed to use the isotope ratios⁹³Zr/⁹⁶Zr,⁹¹Zr/⁹⁶Zr and¹⁰¹Ru/¹⁰⁴Ru for the determination of plutonium contriubtion to the total burnup of a spent nuclear reactor fuel of low initial enrichment in²³⁵U (3.6%). This approach was verified by determining the total burnup of the fuel from the¹³⁷Cs and¹⁴⁸Nd monitors and subtracting the partial burnup due to²³⁵U from the value obtained.     

The robotized chemical treatment of diluted spent fuel samples prior to isotope dilution analysis

A system made up by a Zymark robot and a separation automate preteats spent fuel samples and monitors a tri-n-octylphosphine oxide column extraction chromatographic procedure in order to isolate and purify uranium and plutonium present in the samples, prior to the spectrometric measurements. Up to 16 subsamples of spent fuel in dried or solution form are handled simultaneously in a completely unattended mode. The throughput of the robotized analytical procedure has increased by a factor of 3 compared to the earlier manual procedure without loss in the quality of the chemical treatment and of the mass- and -spectrometric measurements.     

Thermal analysis of reference repository for RBMK-1500 spent nuclear fuel in crystalline rocks

Total amount of RBMK-1500 type spent nuclear fuel (SNF) in Lithuania is approximately 22 thousands of fuel assemblies. All these assemblies should be stored for 50 years and then disposed of. International consensus prevails that SNF and long-lived high-level radioactive wastes are best disposed of in geological repositories using a system of engineered and natural barriers. Disposed nuclear waste induces a number of coupled thermo-hydro-mechanical processes around the repository. Thermal analysis of a deep geological repository could provide temperature distribution which is required for the repository’s design and for evaluation of thermal integrity of the engineered barriers. One of the most critical parameters for the repository is peak temperature at the outer surface of the canister. This temperature cannot exceed 100 °C; otherwise, unfavourable groundwater chemistry can adversely affect chemical stability of the engineered barriers. Thermal behaviour of the conceptual Lithuanian repository for RBMK-1500 type SNF in crystalline rocks was modelled using numerical codes ANSYS FLUENT and COMPASS. Very similar temperature distributions around the disposal canister were determined using both modelling tools. The modelling results revealed the importance of coupled heat and hydrodynamical processes for peak temperature in the engineered barriers, whereas the impact of mechanical processes evaluation was insignificant. It was also determined that peak temperature at the outer surface of the disposal canister does not exceed the permitted 100 °C.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.