Reprocessing Of Fuels Research Articles

A practical, regional approach to nuclear waste storage

Nuclear power continues to offer the potential to be a major, worldwide, scalable, carbon-free energy source—if the challenges of safety, nonproliferation, waste management, and economic competitiveness are addressed. The international community has spent decades attempting to find a pragmatic approach to address waste-management concerns. Along the way, the advocates of open and closed nuclear fuel cycles have engaged in a running debate. For those who favor an open cycle, Sweden and Finland serve as excellent models today for successfully navigating both technical issues and public opinion to dispose of their spent nuclear fuel in a permanent repository that does not allow used fuel to be retrieved. But these successes have yet to be replicated elsewhere. For closed-cycle advocates, economically convincing technology solutions have yet to surface; as a result, leading reprocessing advocates claim that the future value of accumulated waste material can provide the economic justification for nuclear recycling. The authors discuss a middle-ground path that encourages research and development on advancements in fuel cycle technology while providing for safe waste storage on a century-long, or intermediate, timescale. Acknowledging the risks of pursuing such a venture, the authors also write on the importance of establishing performance metrics that would support nuclear energy as a sustainable, secure, and safe energy choice. The authors argue that the most important metric is the establishment of a surety index that could capture the nonproliferation and security risks of alternate fuel cycles.

Synergistic Extraction of Dysprosium and Aggregate Formation in Solvent Extraction Systems Combining TBP and HDBP

During treatment of nuclear fuel in the Plutonium/URanium EXtraction (PUREX) process, the extractant tri-n-butyl phosphate (TBP) is known to degrade to dibutylphosphoric acid (HDBP), which increases the extraction of metal ions, thereby inhibiting their stripping from the organic phase. To better understand this phenomenon, we investigated how mixtures of TBP and HDBP influenced the extraction of metal, nitric acid, and water, and correlated the results to aggregated structures in the organic phase. The mole ratios of TBP-HDBP mixtures had a non-linear effect on the extraction of Dy3+ and water from 0.2 M HNO3, indicating synergism. In 2 M HNO3, the TBP:HDBP mole ratio had a more linear relation to Dy3+ and water extraction, so the synergistic effect was less pronounced than in the low acid system. The extraction of nitric acid showed no synergistic effect and follows closely what would be expected in a system using TBP only. The small-angle X-ray scattering (SAXS) data of the 0.2 M acid system showed maximum contrast at a TBP:HDBP mole ratio of 0.25, so that the synergistic mixture is also the most aggregated at 0.2 M acid. The 2 M acid system also showed that the mixed system is more aggregated than the end members, although this does not result in peak extraction. Previous studies of synergistic extraction of metal cations explain the enhanced extraction by increased dehydration of the metal ion. Although our data do not rule out the formation of mixed complexes according to the classical mechanism of synergism, our evidence of increased water extraction and aggregate formation in systems combining TBP and HDBP are complementary to the metal-centric dehydration aspects of the process. The findings in this study give insights into the complex chemistry of solvent extraction, providing a possible link between formation of aggregates in the organic phase and synergistic extraction.

Water Content of Organic Solvents and their Relationship to Extraction of Nitric and Acetic Acid from UREX+ Streams

This study investigates the equilibrium absorption of water in various solvents and solvent-mixtures being considered for the counter-current solvent extraction of acetic acid from improved Uranium Extraction (UREX+) process solutions. It then seeks to determine if there is any correlation between the equilibrium water content of these solvents and their equilibrium extraction of 0.25 M nitric and 0.025 M acetic acid. The UREX+ process is a proliferation resistant version of the Plutonium Uranium Extraction (PUREX) process. The solvents studied were n-Dodecane (nDD), 1,2 Dichloroethane (DCE), and Phenyltrifluoromethyl Sulfone (FS-13), and mixtures of these solvents with Tributyl Phosphate (TBP). After studying both pure water and acidified aqueous systems, it seems the water absorption mechanism is independent of the diluent used and remains constant with the addition of the 0.25 M nitric and 0.025 M acetic acid.

Progress towards the Full Recovery of Neptunium in an Advanced PUREX Process

To meet the needs of future closed fuel cycles, the complete recovery of minor actinides, including neptunium, may be required. Neptunium can be fully recovered by modifications to the Plutonium URanium Extraction (PUREX) process but this requires careful control of the Np(V)-(VI) redox reaction in the first solvent extraction contactor to avoid losses to the highly active aqueous raffinate, as occurs in current reprocessing plants. As part of the on-going development of an Advanced PUREX process we report a series of solvent extraction experiments aimed at optimizing neptunium recovery in a process that is based on centrifugal contactors as the extraction equipment. Suitable experimental conditions for Np(V) oxidation were identified through simple stirred 2-phase experiments and single stage mini-centrifugal contactor experiments. A U/Np-active proof-of-principle flowsheet test in a multi-stage centrifugal contactor cascade then demonstrated > 99% extraction of neptunium, thus suggesting the aims for neptunium recovery in advanced fuel cycles can be met by an Advanced PUREX process.

Separation and recovery of ruthenium: a review

The chemistry of the noble metal fission product, ruthenium is very complex due to the existence of many oxidation states in addition to forming a large number of co-ordination complexes. In the PUREX process for the separation of U and Pu from the spent nuclear fuels from fast breeder reactors, owing to the high volatile nature of RuO4 problems arise not only during the extraction stages but also in the treatment of high active liquid waste and subsequent vitrification. As this volatile RuO4 can deposit in cooler parts, there is an increase in the radiation field due to the presence of 106Ru. The problem is very acute in the reprocessing of fast reactor fuels due to the increased concentration of ruthenium in the spent fuel. In nitric acid medium Ru can exist in various nitroso nitrate complexes and nitroso complexes are more stable than nitrates. The nitrates are non-extractable by the solvent TBP; however, they are extractable to a higher degree by DBP (the primary degradation product of TBP). The extractability of Ru nitrates into the solvent is inhibited by high acid content, temperature and prolonged hold-up time. Nevertheless, these factors promote the volatilization of Ru as RuO4. The volatilization is enhanced by the addition of phosphate ions, but is suppressed by phosphite or hypophosphite ions. Thus, it would be advantageous if ruthenium is removed so that not only the purity of the product (Pu) is improved, but also the problem related to volatilisation can be resolved. High molecular weight amines (tertiary amines) capable of forming co-ordinate bonds are reported to be ideal extractants for Ru. Gas phase separation is an effective method for the recovery of Ru from catalysts, lead button and from other platinum group metals. Separation and pre-concentration of noble metals can be accomplished from non-metals by simple sorbents like coconut shell activated carbon to complicated chelating resins, aromatic polymers and zeolites. In the electro-oxidation of active Ru from nitroso salts, Pd was found to interfere and removal of Pd prior to oxidation of Ru is recommended. Redox catalysts such as Ag2+ and Ce4+ are found to play a prominent role in the electro-oxidation of Ru. Though, various methods and extractants are reported in the literature for the separation of Ru, R&D is being pursued for the removal of Ru during aqueous reprocessing of spent fuels using extractants and methods which are conducive to plant conditions. Hence, an exhaustive survey of literature was made and the different methods reported for the removal of Ru with emphasis towards reprocessing applications are discussed in this report as a review. Attempts made by the authors in separating Ru from simulated waste solution are also included in this review.

Evaluation of fire and explosion hazards in typical nuclear recycling plants

Nuclear recycling is crucial part for the closing of nuclear fuel cycle. Reprocessing and consequent waste management allows separation and concentration of the fission products for safe disposal in the near surface disposal facilities. It facilitates recovery and recycling of the unused nuclear fuel as well as bred nuclear materials to the nuclear reactor for further production of power. Reprocessing ensures the achievement of maximum energy extraction by increasing the fissile atom burn up and maximum resource utilization. Fire safety evaluation in fuel cycle and reprocessing plants is very much important because these plants handle huge quantity of radioactive materials. In case of accidental fire in these plants, in addition to destruction and devastation at site, radioactive contamination can also spread.

Decontamination of Molten Salt Wastes for Pyrochemical Reprocessing of Nuclear Fuels

ABSTRACTPyrochemical reprocessing of nuclear fuels, in which electrochemical separation of actinides and fission products is mediated by a molten alkali chloride salt (typically a LiCl-KCl eutectic) is of interest for future nuclear energy cycles. A key challenge in the management of pyrochemical reprocessing wastes is decontamination and recycling of the molten salt medium to remove entrained actinides and radioactive lanthanide fission products. Since pyrochlore oxides are promising candidates for the immobilisation of lanthanides and actinides, we sought to use the “problematic” molten salt to our advantage as a reaction medium for low temperature synthesis of titanate pyrochlores. Through control of reaction time and temperature, we demonstrated the synthesis of lanthanide pyrochlores at temperatures as low as 700 °C in 1 h, compared to 1350 °C in 36 h for conventional solid state synthesis. The importance of this study is in demonstrating the potential feasibility for decontamination of pyrochemical reprocessing wastes by simple addition of TiO2 to form lanthanide and actinide pyrochlores by rapid molten salt assisted reaction at moderate temperature.

Extraction Studies of Phosphoric Acid Tributyl Ester from Nitric Acid Solutions Using Dodecane in the Presence of Inextractable Metal Nitrates

Sodium nitrate and calcium nitrate are not extracted by phosphoric acid tributyl ester (TBP) in the plutonium uranium extraction (PUREX) process. The effect of these inextractable metal nitrates on the solubility and distribution data of TBP has been studied. The solubility of TBP in water in the presence of different concentrations of sodium nitrate is determined, and salting coefficient of TBP is calculated. It has been observed that the solubility of TBP in nitric acid decreases in the presence of both inextractable nitrates, that is, sodium nitrate and calcium nitrate. The effect of inextractable metal nitrates on the equilibrium data for the TBP–dodecane–nitric acid system has been studied. The distribution coefficient value of TBP is greater in the presence of sodium nitrate solutions as compared with calcium nitrate solutions due to more salting out of TBP. Because sodium nitrate is a better salting agent than calcium nitrate, the variation in the concentration of nitric acid at equilibrium was more in sodium nitrate solutions in comparison with that in calcium nitrate solutions. The data obtained will be useful in separating dissolved TBP from acidic waste for safe disposal of nuclear waste.

Direct spectrophotometric determination of ruthenium in aqueous streams of nuclear reprocessing

A fiber optic aided spectrophotometric technique has been developed for determination of ruthenium in nitric acid medium. The developed method is simple, accurate and applicable to aqueous streams of nuclear reprocessing. The system obeys Lambert–Beer’s law at 468 nm in the concentration range of 30–360 μg/mL of ruthenium. The molar absorption coefficient, detection limit and Sandell’s sensitivity are 68.477 L Mol−1 cm−1, 31 μg/mL and 0.0124 μg/cm2 respectively. Relative standard deviation was less than 2 % and correlation coefficient was 0.9998. The results obtained by the developed procedure are in good agreement with those obtained by the standard ICP-OES method. Fission products like zirconium and strontium are not interfering. Uranium is interfering and needs prior separation by solvent extraction method. The developed method is adaptable for remote operation and on-line monitoring.

Modeling of Pu(VI) distribution coefficients in 30 % TBP based PUREX process

Hexavalent plutonium (Pu(VI)) is an important solute in the PUREX (plutonium uranium extraction) process. In 30 % TBP based PUREX solvent extraction system, distribution coefficient of Pu(VI) is much lower than that of Pu(IV). This lower distribution coefficient of Pu(VI) may cause unexpected Pu loss during primary HA extraction in low acid flowsheets. An empirical model for Pu(VI) distribution coefficients in 30 % TBP and its temperature dependency has been reported in this paper. Comparison with literature data revealed a reasonably good agreement between the reported experimental and model predicted values.

Elucidating the radical kinetics involved in the radiolytic destruction of lanthanide-complexed DTPA

The reprocessing of nuclear fuels to extract the remaining actinides is one of the most important strategies for viable nuclear power in the future, as geologic waste disposal of high-level radioactive waste could have considerable negative impacts on the environment. As many of these strategies are based on solvent extraction processes, our aim is to establish the radiolytic stability of the component extraction ligands. The oxidative stability of the lanthanide metal-complexed extraction ligand DTPA (diethylenetriaminepentaacetic acid) has been investigated through its reaction rate constant with the hydroxyl radical under TALSPEAK acidic pH conditions, both for the free ligand and complexed with Eu3+, Lu3+ and Gd3+. Specific rate constants were obtained over a range of pH conditions using thiocyanate competition kinetics. The rate constants for the complexed metals are seen to be greater than the corresponding values determined for only the DTPA protonated ligands at these pHs.

Simultaneous spectrophotometric determination of uranium, nitric acid and nitrous acid by least-squares method in PUREX process

Multi-wavelength linear regression spectrophotometry combined with method of least squares for simultaneous determination of uranium, nitric acid and nitrous acid in PUREX (Plutonium/URanium EXtraction) process was developed. The molar absorbance matrix was calibrated with absorbance data measured in the wavelength range of 350–500 nm for a series of standard solutions by linear least-squares regression. This method used information from the absorption spectra of U(VI)–nitrous acid–nitric acid solutions to determine U(VI), nitrous acid and nitric acid. In the range of 0.95–74.1 g/L U(VI), 5 × 10−4–2 × 10−3 mol/L nitrous acid and 3–5 mol/L aqueous nitric acid solution, the measuring precision for determination of U(VI), nitrous acid and nitric acid was 0.46, 4.09, and 0.68 % respectively. In the solution of 30 % TBP–kerosene, the measuring precision for determination of U(VI) and nitrous acid was 0.42 and 4.2 % respectively in the range of 0.95–74.1 g/L U(VI) and 5 × 10−4–2×10−3 mol/L nitrous acid. The spectrophotometric method can be valuable for monitoring and controlling of both species in PUREX process operation, thanks to its simplicity, efficiency and accuracy.

Thermal behavior pattern of tributyl phosphate under adiabatic conditions

Violent decomposition of Tributyl Phosphate (TBP), a widely employed extractant in the Plutonium Uranium Extraction process of nuclear fueling reprocessing plants in the presence of Nitric acid at temperatures in excess of 130 °C is a matter of concern in serious accidents including in the Savannah River (USA) and Tomsk (Russia). The thermal behavior of TBP under adiabatic conditions employing the world’s benchmark adiabatic calorimeter, the Accelerating Rate Calorimeter is examined. TBP shows multiple self heating exothermic activities with the onset of primary exotherm at 250 °C. The exothermic activity is accompanied by considerable pressure rise. The thermal decomposition of TBP is found to follow first order Arrhenius kinetic model. TBP loses about 70% of its chemical moieties as volatile matter during the exothermic decomposition. Spectroscopic methods are used to elucidate the degradation pathway.

Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle

Molten-Salt Reactor (MSR) is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. It is based on the experience with the development of the molten-salt reactor technology in the Oak-Ridge National Laboratory in the United States. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of radioactive wastes for final deposition and for utilization of nuclear energy for electricity and heat production as effectively as possible. There are two pyrochemical reprocessing techniques studied in NRI Rez plc which are considered to be applied both for reprocessing of already existing spent fuel and for preparation and „on-line“ reprocessing of MSR fuel: (i) the Fluoride Volatility Method (FVM), which performs chemical conversion of spent thermal oxide fuel components into fluorides and their consequent separation by means of their different volatility, thermal stability and chemical affinity to various sorbents; and (ii) electrochemical separation of the actinides (Ans) and fission products (FP), represented mainly by lanthanides (Lns), from each other by electrolytic deposition method on solid cathode in molten fluoride media.

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.

Evaluation ofN,N-dihexyloctanamide as an alternative extractant for spent fuel reprocessing: batch and mixer settler studies

ABSTRACT Solvent extraction studies carried out at BARC, India on the evaluation of N,N-dialkyl amides as alternative extractants to tri-n-butyl phosphate (TBP) for reprocessing of spent fuel have suggested that straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels. This paper presents the batch as well as mixer settler studies for uranium and plutonium extraction/stripping to evaluate DHOA vis a-vis TBP for the reprocessing of Pu rich fuels. These studies showed that uranium extraction using DHOA as extractant was comparable to that of TBP; however, it displayed better stripping behavior than TBP. Plutonium extraction behavior was better in the case of DHOA as compared to that of TBP. However, Pu stripping data indicated towards the need of reducing agent in the stripping cycle for both the extractants.

Separation, for Analytical Purpose, of Np Traces from different Solutions of Fuel Reprocessing

Four separation methods were developed for performance control of hydrometallurgical extraction processes as COEX™ or advanced PUREX. These methods used implemented the operations of radionuclides oxidation state adjustment and chromatographic separation using TEVA resin. Concerning FP raffinate, the method consisted in reducing Np traces to the valence IV by a mixture of ferrous sulfamate and ascorbic acid, to fix Np(IV) on “TEVA” resin and to eluate it by a nitrohydrofluoric acid solution. The Np recovery yield is 100%. The decontamination of Np is sufficiently high to allow its analysis by FXL (Zr/Np < 1). The study also showed that in presence of Zr and Tc, Pu behaved like Np. The mixture of ferrous sulfamate and ascorbic acid had surprisingly no action on Pu(IV). Concerning plutonium solution ([Pu] > 10g/L) and uranium solution ([U] > 100g/L), the same method used for Np recovery from FP raffinate led to an eluate containing 100% of the initial Np ([Np]: 10mg/L). The low concentration of U and Pu (< 100mg/L) allows the determination of Np by FXL. Concerning Pu(III)-U(IV) solution, the method, included 2 redox stages, the first one to oxidize all actinides to oxidation state VI et the second one to reduce Np and Pu respectively to IV and III oxidation state. Then Np(IV) was fixed on TEVA resin. The eluate contains 100% of the initial Np ([Np]: 10mg/L) and a low concentration of U and Pu ([U] < 20mg/L, [Pu] < 10mg/L). The next experiments will consist in consolidating these good results by working with real solutions of fuel reprocessing.

Fixed Contamination on Steel Surfaces: First Use of Quartz Crystal Microgravimetry to Measure Oxide Growth on Process Steels Under Conditions Typical of Nuclear Reprocessing.

ABSTRACTDuring the lifetime of a nuclear facility, radioactive material may become deposited onto process and structural material surfaces. Due to their high corrosion resistance, steels comprise the largest class of metal-based materials encountered on nuclear sites. A greater understanding of the mechanisms of how contaminant radionuclides interact with and attach to process steels in nuclear plant environments is required in order to enable informed decisions to be made about the design and effective application of decontamination techniques, reducing secondary wastes.There is limited literature relating to radionuclide sorption mechanisms on steels. Key studies have found that sorbed contamination is almost entirely located in the outermost oxide layers formed at steel surfaces. Thus, a molecular level investigation of contaminant uptake during induced oxide formation would be beneficial in developing steel decontamination strategies.Stainless steel 316L is commonly employed in the nuclear industry in process streams and pipework. Thus, we describe work carried out on electrochemically accelerated oxide growth on 316L and SS2343 (a 316L analog) in nitric acid media and its characterisation using combined voltammetric and microgravimetric measurements. These allow identification of active, passive, high voltage passive, transpassive and secondary passivation regimes in the associated current voltage curves. EQCM on SS2343 coated quartz crystal piezoelectrodes, combined with potentiodynamic polarisation data have allowed us to determine that fastest net growth of surface oxide occurs in the low voltage passive regime. Further, we have directly measured the growth of that layer by using in situ microgravimetry for the first time. We will be shortly using the methods described above and radionuclide surrogates for the study of contaminant uptake during oxide formation and uptake onto preformed oxide layers. XPS will be used to determine layer composition and mode of contaminant uptake.

A Neutronic Evaluation of Reprocess Fuel and Depletion Study of VHTR Using MCNPX and WIMSD5 Code

Among the projects of IV generation reactors available nowadays, the (High Temperature Reactors) HTR, are highlighted due to their desirable characteristics and they have been studied by the Instituto Nacional de Ciências e Tecnologia de Reatores Inovadores/CNPq(Brazil). For this work, it evaluated the neutronic behavior and fuel composition during the burnup using the codes (Winfrith Improved Multi-Group Scheme) WIMSD5 and the MCNPX2.6, inserting different percentages of reprocessed fuel in the core. The fuel type “C” coming from Angra-I nuclear power plant, in Brazil, enriched with 3.1% was burnt by three typical cycles and then reprocessed. It recovered (Pu) and minor actinides (MA)being neptunium (Np), americium (Am), curium (Cm), and processed six different fuels varying percentage insertion of reprocessed fuel and enrichment uranium. It analyzed the multiplication factor, temperatures reactivity coefficients, and the composition during the burnup. The results showed, in the analyzed conditions, only one of these fuels is possible to be used. To compare, a reference fuel using 15% enrichment (235U) was too evaluated.

Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of ±1.3M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.