Analytical techniques for the separation and determination of transuranium element ultratraces in depleted uranium ammunitions

It is well known that ammunition containing depleted uranium (DU) was used by NATO during the Balkan conflict. To evaluate the DU origin (natural uranium enrichment or spent nuclear fuel reprocessing) it is necessary to check the presence of activation products (²³⁶U, ^239+240Pu, ²⁴¹Am, ²³⁷Np, etc.) in the ammunition.

Every transuranium element (TRU) was separated from the uranium matrix by extraction chromatography with microporous polyethylene (Icorene) supporting suitable stationary phases. Plutonium was separated by tri-n-octylamine (TNOA). ²⁴¹Am was separated by TNOA and di(2ethylhexylphosphoric) acid (HDEHP). Neptunium also was separated by tri-n-octylamine using different conditions. After elution, the TRU elements were electroplated and counted by alpha spectrometry. The TRU decontamination factors from uranium were higher than 10⁶.

The final chemical yields ranged from 50 to 70%. The detection limit was 1?Bq?kg^?1 for 0.10?g ammunition; ^{239 + 240}Pu and ²⁴¹Am concentrations in two penetrators were 26 and 70?Bq?kg^?1 and <1 and 3.4?Bq?kg^?1, respectively; the ²³⁷Np concentration in one penetrator was 30.1?Bq?kg^?1.

The presence of these anthropogenic radionuclides in the penetrators indicates that at least part of the uranium originated from the reprocessing of nuclear fuel, although because of their very low concentrations, the radiotoxicological effect is negligible.     

Verification measurements of the IRMM-1027 and the IAEA large-sized dried (LSD) spikes

In the frame of the accountancy measurements of the fissile materials, reliable determinations of the plutonium and uranium content in spent nuclear fuel are required to comply with international safeguards agreements. Large-sized dried (LSD) spikes of enriched ²³⁵U and ²³⁹Pu for isotope dilution mass spectrometry (IDMS) analysis are routinely applied in reprocessing plants for this purpose. A correct characterisation of these elements is a pre-requirement for achieving high accuracy in IDMS analyses. This paper will present the results of external verification measurements of such LSD spikes performed by the European Commission and the International Atomic Energy Agency.

     

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.     

Preparation, certification and validation of a stable solid spike of uranium and plutonium coated with a cellulose derivative for the measurement of uranium and plutonium content in dissolved nuclear fuel by isotope dilution mass spectrometry.

A stable solid spike for the measurement of uranium and plutonium content in nitric acid solutions of spent nuclear fuel by isotope dilution mass spectrometry has been prepared at the European Commission Institute for Reference Materials and Measurements in Belgium. The spike contains about 50 mg of uranium with a 19.838% (235)U enrichment and 2 mg of plutonium with a 97.766% (239)Pu abundance in each individual ampoule. The dried materials were covered with a thin film of cellulose acetate butyrate as a protective organic stabilizer to resist shocks encountered during transportation and to eliminate flaking-off during long-term storage. It was found that the cellulose acetate butyrate has good characteristics, maintaining a thin film for a long time, but readily dissolving on heating with nitric acid solution. The solid spike containing cellulose acetate butyrate was certified as a reference material with certified quantities: (235)U and (239)Pu amounts and uranium and plutonium amount ratios, and was validated by analyzing spent fuel dissolver solutions of the Tokai reprocessing plant in Japan. This paper describes the preparation, certification and validation of the solid spike coated with a cellulose derivative.     

A novel rapid detection approach for the analysis of radionuclides in environmental samples using graphite MALDI mass spectrometry

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.     

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 approaches to processing and fabrication of oxide nuclear fuels

It was shown that, in contrast to the Purex process using aggressive and environmentally hazardous 8M HNO₃ solutions for dissolving spent oxide nuclear fuel (SNF), this fuel can be easily dissolved in aqueous subacid ([H⁺] ∼0.1 M) solutions of Fe(III) nitrate (chloride) with partial separation of uranium and plutonium from fission products (FP). The low acidity of the solutions obtained (pH ∼1) allows direct application of modern technologies of finishing processing of nuclear fuel by fluoride, carbonate, oxalate, or peroxide precipitation of uranium and plutonium. It was established that U(VI) is isolated from nearly neutral nitric acid solutions as a poorly soluble uranyl hydroxylaminate complex after adding hydroxylamine. It was shown that on thermal decomposition at 200–300°C under ambient atmosphere this compound converts into uranium dioxide. A similar approach was applied to obtain mixed oxide uranium-plutonium fuel (MOX fuel).     

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.     

Wet-oxidation of spent organic waste tributyl phosphate/diluents

Tributyl phosphate was used in reprocessing of spent nuclear fuel inthe Purex process. The amount of uranium retained in the organic phase dependson the type of TBP/diluent. Destruction of spent TBP is of high interest inwaste management. The use of the oxidative degradation of TBP diluted withkerosene, carbon tetrachloride, benzene and toluene using potassium permanganateas oxidant was carried out to produce stable inorganic dry particle residuewhich is then immobilized in different matrices. The different factors affectingthe destruction of spent waste were investigated. The uptake and decontaminationfactor for both ^{152, 154}Eu and ¹⁸¹Hf and the analysisof the final product have been studied.     

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0–8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0–8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5–7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing.     

The k0 and Q0 Values for the Zr-Isotopes: A Re-Investigation

The activities associated with the nuclear fuel cycle include uranium mining, enrichment and fuel fabrication, reprocessing and recycling of the spent fuel and management of nuclear waste. In this paper the nuclear fuel cycle strategies followed and envisages of the future directions are discussed.     

Alpha-spectrometry in nuclear fuel analysis

A survey is given of the analysis of actinide nuclides by means ofa-spectrometry in high-level radioactivity process solutions resulting from reprocessing of spent U-Th fuel. Separation of fission products and isolation of the nuclides²²⁸Th,²³¹Pa,²³²U,²³⁷Np and²³⁸Pu are performed by adsorption, ion-exchange, extraction chromatography and extraction techniques. After separation, samples for quantitative determination are prepared by electrodeposition and measured using a silicon-surface barrier detector combined with a multichannel analyzer. An error estimation is given.     

Determination of ultra traces amount of uranium in raffinates of Purex process by laser fluorimetry

A simple and rapid, laser fluorimetric method for the determination of uranium concentration in raffinate stream of Purex process during reprocessing of spent nuclear fuel has been developed. It works on the principle of detection of fluorescence of uranyl complex formed by using fluorescence enhancing reagent like sodium pyrophosphate. The uranium concentration was determined in the range of 0–40 ppb and detection limit of 0.2 ppb. The optimum time discrimination is obtained when the uranyl ion is complexed with sodium pyrophosphate. Need of preconcentration step or separation of uranium from interfering elements is not an essential step.     

UxTh1-x(C2O4)2 Solid Characterisation Studies

Many advanced reprocessing schemes under development are aimed at co-processing and co-conversion of actinides, unlike current reprocessing plants that produce separate uranium and plutonium products. The most well developed option for the co-conversion stage is probably oxalate co-precipitation, followed by the thermal co-conversion to a mixed oxide product. It is thus envisaged that future processes will avoid separation of plutonium from uranium and instead allow part of the uranium to flow with the plutonium, resulting in co-precipitation as the oxalate, and finally co-conversion to a mixed uranium-plutonium oxide (MOX), which can be fabricated into recycled nuclear fuel for further energy generation.The co-crystallisation of uranium (IV) and plutonium (III) into a single oxalate structure ensures the homogenous distribution of the two actinides at the molecular scale. The joint conversion of uranium and plutonium to the oxide form makes it possible to remove the complicated step of blending and grinding the two distinct oxide powders, as currently employed for the purposes of MOX fuel fabrication. This concept can also be extended to other actinides, including minor actinides from partitioning processes such as SANEX (Selective Actinide Extraction) and GANEX (Grouped Actinide Extraction) processes or even a thorium containing product from recycle of thorium based fuels.A selection of U_xTh_1-x(C₂O₄)₂ solids at varying concentrations of uranium and thorium were prepared by oxalate co-precipitation. Uranium (VI) was conditioned electrochemically at -0.7 V to uranium (IV), in the presence of hydrazine. The reduced uranium (IV) in nitric acid was mixed with thorium nitrate solutions at different concentration ratios with oxalic acid. The mixed tetravalent uranium-thorium oxalate solid products have been characterised by Raman and IR spectroscopies. The influence of thorium substituted into the uranium oxalate structure was evaluated. Several vibrational modes were found to be affected by the variation in ionic radius appearing to be metal sensitive and therefore, provide the initial indication in the evaluation of the chemical composition.     

Solvent Extraction Studies of Gadolinium in Tri-Butyl Phosphate

Fast reactor spent fuel reprocessing plants should be designed for inherent criticality safety due to high plutonium content. Addition of soluble neutron poison is one way to do that. Gadolinium is the best choice based on neutron absorption cross section and chemical compatibility. In this work, using classical thermodynamic approach, the distribution coefficient of gadolinium in tributyl phosphate has been calculated and compared with the experimental data. The influence of acidity and uranium at equilibrium on gadolinium distribution in tributyl phosphate has been investigated. The result establishes the feasibility of employing gadolinium as soluble neutron poison in fast fuel reprocessing.     

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied. The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.     

Feasibility studies for the detection of third phase during reprocessing of fast reactor fuel

A conductometric technique for the determination of third phase, a phase formed by splitting of organic phase during malfunction in operation, during reprocessing of fast reactor fuel using a high resolution conductivity monitoring instrument with pulsating sensor developed in-house was tested. Initial studies were carried out in laboratory using uranous salt solutions (U⁴⁺) in various synthetic samples containing different aqueous (A) to organic (O) phase ratios (A/O). Results show that there is appreciable increase in conductivity of third phase solution compared to the other organic phase (~100 to ~300 times higher). Such a large change in conductivity in third phase with respect to the other organic phase can be used as a deciding parameter to give first hand information about the occurrence of third phase.     

Environmental monitoring as an important tool for safeguards of nuclear material and nuclear forensics

Summary The use of environmental monitoring as a technique to identify activities related to the nuclear fuel cycle has been proposed by international safeguards organizations. The elements specific for each kind of nuclear activity, or “nuclear signatures”, inserted in the ecosystem can be intercepted by different live organisms. This work demonstrates the technical viability of using pine needles as bioindicators of nuclear signatures associated with uranium enrichment activities. Additionally, it proposes the use of HR-ICP-MS to identify the signature corresponding to that kind of activities in the ecosystem. Nitric acid solutions, used to wash pine needles sampled near nuclear facilities and containing only 0.1 mg ^. kg^-1 of uranium, exhibit a n(²³⁵U)/n(²³⁸U) isotopic abundance ratio of 0.0092±0.0002, while solutions originated from samples collected at places located more than 200 km far from activities related to the nuclear fuel cycle exhibit a value of 0.0074±0.0002. Similar results were obtained for sample solutions prepared using the acid leaching process. The different values of n(²³⁵U)/n(²³⁸U) isotopic abundance ratio obtained permit to confirm the presence of anthropogenic uranium and demonstrate the viability of using the methodology proposed in this work.     

Buffer gas effect on the detection of iodine by laser induced fluorescence

Laser-induced fluorescence (LIF) coupled with photon-counting technique to detect molecular iodine at ultratrace level is reported. Electronic quenching rate constants for N₂, NO₂ and H₂O, as well as for the mixture of NO₂ and H₂O has been measured. The application of the LIF method to monitoring¹²⁹I₂ in spent fuel reprocessing off-gas streams is evaluated.     

Current Progress in R&D of Fluoride Volatility Method

Fluoride Volatility Method is regarded to be a promising advanced pyrochemical reprocessing technology, which can be used for reprocessing mainly of oxide spent fuels coming from current LWRs or future GEN IV fast reactors. The technology should be chiefly suitable for the reprocessing of advanced oxide fuels with inert matrixes of very high burn-up and short cooling time, which can be hardly reprocessed by hydrometallurgical technologies. Fluoride Volatility Method is based on direct fluorination of powdered spent fuel with fluorine gas in a flame fluorination reactor, where the volatile fluorides (mostly UF6) are separated from the non-volatile ones (trivalent minor actinides and majority of fission products). The subsequent operations necessary for partitioning of volatile fluorides are condensation and evaporation of volatile fluorides, thermal decomposition of PuF6 and finally distillation and sorption used for the purification of uranium product.