Neutralizing accidental discharges of uranium hexafluoride from the isotope separation plant at the Siberian chemical works

The processes occurring during accidental discharges of liquid uranium hexafluoride in the production rooms of the Chelnok-T section in the isotope separation plant at Siberian Chemical Combine are examined. The purification system, which uses KhP-MMD and KhP-MD chemical absorbers, permits localizing very quickly any accidental discharges of uranium hexafluoride and protect the atmosphere in the plant from harmful effects of fluorine-contained gases. The absorbers effectively catch HF (99.97%) and UF₆ (>99.9%) and have a high working capacity 30.5 mass% with respect to the fluorine ion. The system for catching and dispersing harmful contaminants provides the required sanitary standard with respect to atmospheric air on the territory of the plant and at the nearest populated points.     

The Hydrothermal System Beneath Owakudani on Hakone Volcano,Japan; Traced by Stable Isotopic Ratios of H2O

In this study, we sampled natural fumarolic gases, the gas from a steam well drilled to 800 m depth and natural hot spring waters from the Owakudani geothermal area, Hakone Volcano, Japan. We then measured the chemical composition and the D/H and ¹⁸O/¹⁶O ratios of H₂O within the samples. On the basis of the analytical results, we investigated the differentiation processes of the magmatic fluid during ascent to the surface. Volcanic gas discharged from the No. 52 steam well has a CO₂/H₂O ratio that is much higher than those from fumarolic gases. The isotopic ratio of H₂O in the gas are also much higher than those in the fumarolic gases, and the gas also contains high concentrations of SO₂ and HCl. The above observations indicate a magmatic origin for the gas from the No. 52 steam well. The hydrothermal system is principally explained by the mixing of a magmatic vapor, represented by the gas from the No. 52 steam well, and cold local meteoric water. Following mixing of the fluids, a separation of primary vapor and liquid occurs. The primary liquid is discharged as a component of hot spring water, and has high isotopic ratios. The primary vapor is mixed with vapor derived from the boiling of local meteoric water. The mixed vapor is then discharged to the surface, interacting with shallow meteoric water and undergoing partial condensation of H2O vapor. Hot spring water with isotopic ratios lower than those of the primary liquid is derived from the boiling of local meteoric water.     

In situ ERDA measurements of hydrogen isotope concentrations in palladium at atmospheric pressure

In situ elastic recoil detection analysis (ERDA) measurements in gases at atmospheric pressure have been carried out using 15 MeV ⁴He ion beams. The beams are extracted through a molybdenum foil having a thickness of 5 μm. The maximum depth of analysis is about 4 μm for the palladium hydride and palladium deuteride (PdH_x and PdD_x, x = 0.7-0.8) samples. The temperature of the samples rises stepwise from room temperature to 180 °C. ERDA spectra are obtained every 2 min. Hydrogen and deuterium in the samples are discharged in the temperature range of 120-140 °C in a vacuum. Decrease in the hydrogen concentration in the PdH_x sample heated in a vacuum follows a first order in the value of x and an apparent activation energy of discharge of hydrogen is 1.05 eV. On the other hand, the hydrogen and deuterium concentrations decrease at about 80 °C in air. No isotope effects are observed in both a vacuum and air. The temperature at which the hydrogen concentration decreases in helium gas is almost the same as that in a vacuum. It indicates that hydrogen and deuterium atoms are discharged by chemical reactions with air and that there are no effects of cooling of the thermocouple by convection of air.     

In-situ Measurement of UO2 2+ Concentration in Molten NaCl-2CsCl by Differential Pulse Voltammetry

In order to assess the applicability of the Differential Pulse Voltammetry technique to the in-situ measurement of UO₂ ²⁺ concentration in the oxide electro-winning process, DPV measurements for UO₂Cl₂ in molten NaCl-2CsCl were studied. DPV measurement of UO₂ ²⁺ in NaCl-2CsCl at 923 K, with a set of optimized parameters (potential sweep rate ?0.1 V/s, pulse cycle 0.1 s, pulse width 10 ms and pulse potential height 50 mV), showed a clear current peak at ?0.9 V vs. Cl₂/Cl^?. This was attributed to the reduction of UO₂ ²⁺ to UO₂ ⁺. The relation between the current peak height and the analytical concentration of the UO₂ ²⁺ showed good proportionality in the concentration region up to 0.06 mol.l ^?1, and the applicability of UO₂ ²⁺ concentration measurement by DPV was confirmed up to 0.4 mol.l ^?1. In order to assess the interference by the coexisting fission product elements to the measurement of UO₂ ²⁺ concentration, DPV measurements of UO₂ ²⁺ concentration in molten NaCl-2CsCl containing PdCl₂, NdCl₃, SmCl₃ and CeCl₃ were also performed. Even before removing Pd, the current peak at ?0.9 V vs. Cl₂/Cl^? by the reduction of UO₂ ²⁺ to UO₂ ⁺ was found to be distinguishable from the reduction currents of Pd²⁺ to Pd at ?0.7 V vs. Cl₂/Cl^?. As a result, the application of DPV measurement technique to the in-situ monitoring of UO₂ ²⁺ concentration in the oxide electro-winning method requires the improvement of the DPV measuring condition or the electrode structure on higher UO₂ ²⁺ concentration condition.     

Escape of fission products from defective fuel rods of light water reactors

The escape behaviour of various fission product isotopes from defective fuel rods in PWRs and BWRs is analyzed.Diffusion in the UO₂ is the rate controlling step for the release of noble gases from defective fuel rods. The escape of fission iodine from defective fuel rods is controlled by a mechanism which includes migration and additional delay steps, probably in the nature of a chemical reaction.The inferred effective diffusion constants for fission gases are noticeably higher for defective fuel rods than for intact fuel rods. The difference is about two orders of magnitude. The enhancement of diffusion in defective fuel rods is believed to be due to the increase in the -ratio of the UO₂ in the defective fuel rods.     

Effect of Sodium Nitrate on the Diffusion of Cl− and I− in Compacted Bentonite

For safety assessment of TRU waste disposal, the effective diffusion coefficients (D, _e ) for Cl^? and I^? in compacted bentonite (Kunigel V1) were determined as a function of NaNO₃ concentration, ranging from 0.01 to 5 mol/dm₃. The D_e values for Cl^? and I^? increased from approximately 10^?12 to 10^?11 m²/s with increasing NaNO₃ concentration. The capacity factor α, indicative of the effective porosity, also increased with increasing NaNO₃ concentration. The maximum α values of 0.21 for Cl^? and 0.24 for I^? at 5 mol/dm³ NaNO₃ concentration were lower than the total porosity of compacted bentonite (0.40), suggesting an anion exclusion effect. Therefore, the increases in the D_e values were interpreted in terms of the decrease in anion exclusion. The D_e values for Cl^? and I^?, normalized by their diffusivities in bulk water, were found to increase in proportion to the α values in a log-log plot, while the D_e values in porous media generally increase in proportion to the total porosity. This relationship between the logD_e and the log α is considered to be derived from the lower effective porosities for Cl^? and I^? than the total porosity in compacted bentonite due to anion exclusion.     

Model of the effect of uranium hexafluoride on the human body under industrial conditions

A method for calculating certain quantities characterizing the effect of uranium hexafluoride (UF₆) on the human body under industrial conditions in uranium enrichment plants is described. It is assumed that the effect is determined by uranium and fluorine inhaled together with the products of hydrolysis of uranium hexafluoride. The proposed complex model consists of three models, the first of which describes the contamination of the industrial environment and the second and third describe inhalation and percutaneous intake. A relation is obtained between uranium and fluorine intake and the uranium hexafluoride concentration in air at the moment the compound is discharged. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 103–106, August, 2007.     

Chloride Pyrometallurgy of Uranium Ore, (II)

Selective extraction of uranium from a phosphate ore was studied by using the mixed gas of Cl₂ and O₂. On heating the ore and carbon mixture in Cl₂, the volatilized chloride of uranium is accompanied by iron, aluminum, phosphorus and silicon chlorides. The addition of O₂ gas effectively lowered the volatilization ratios of aluminum, phosphorus and silicon. The ratio decreased with increasing oxygen flow rate up to 50ml/min at 1,223 K (Cl₂: 100ml/min, O₂+N₂: 400m//min). The volatilization ratio of uranium was almost unchanged at 90% up to 50 ml/min O₂ (carbon amount: 15wt%). The effect of the other parameters, i.e. Cl₂ flow rate, carbon amount, reaction temperature and time was examined.     

High-speed processing with Cl2 cluster ion beam

Cluster ion beam processes can produce high rate sputtering with low damage compared with monomer ion beam processes. Cl₂ cluster ion beams with different size distributions were generated with controlling the ionization conditions. Size distributions were measured using the time-of-flight (TOF) method. Si substrates and SiO₂ films were irradiated with the Cl₂ cluster ions at acceleration energies of 10–30 keV and the etching ratio of Si/SiO₂ was investigated. The sputtering yield increased with acceleration energy and was a few thousand times higher than that of Ar monomer ions. The sputtering yield of Cl₂ cluster ions was about 4400 atoms/ion at 30 keV acceleration energy. The etching ratio of Si/SiO₂ was above eight at acceleration energies in the range 10–30 keV. Thus, SiO₂ can be used as a mask for irradiation with Cl₂ cluster ion beam, which is an advantage for semiconductor processing. In order to keep high sputtering yield and high etching ratio, the cluster size needs to be sufficiently large and size control is important.     

Extraction of Thorium and Uranium from Chloride Solutions by Tri-n-Butyl Phosphate and Tri-n-Octyl Phosphine Oxide

The extraction of thorium and uranium chlorides by TBP and TOPO was studied. The composition of complexes extracted from the chloride solutions of low acid concentration was established by partition study to be UO₂Cl₂ (TOPO)₂, UO₂Cl₄ (TOPO)₂, UO₂Cl₄ (TOPO)₂ and UCl₄ (TBP)₂. Composition of the thorium complex in the TBP phase free from hydrochloric acid was revealed by infrared study to be ThCl₄ (TBP)4. The extraction behavior of thorium chloride by TBP was different from that of U (N) and Pu(N) chloride, and the composition of the complex was presumed to be HThCl₅(TBP)₄ in the extraction from concentrated chloride solution containing hydrochloric acid.     

Quantum Chemical Calculation of O-H Vibration in LiOH Ab-initio Analysis of Hydrogen Isotopes in Ceramic Breeder Materials

The stretching vibrational frequency of O-H in LiOH crystal was calculated using CRYSTAL98 code. Hartree-Fock (HF) and hybrid (B3LYP) methods were used. By taking account of reduced mass, O-D or O-T frequencies could be also estimated. The obtained O-D frequencies were 2,994 cm^-1 in HF level and 2,740 cm^-1 in B3LYP. Agreement with experimental data, 2,713 cm^-1 reported by Buchanan et al., was much better in B3LYP than in HF. The result of the present calculation supported the assignment in our previous FT-IR study, where O-D vibration in the Li₂O single crystal that was treated by thermal absorption of D₂ was observed. The peak observed at 2,715 cm^-1 was attributed to O-D in LiOD phase. The present study showed that the quantum chemical analysis combined with the FT-IR was a powerful tool to study the behavior of hydrogen isotopes in the ceramic breeder materials.     

The chemistry and thermodynamics of molten salt reactor fuels

A fairly detailed knowledge of the chemistry and thermodynamics of molten salt breeder reactor fuels, typically LiF-BeF₂ThF ₄-UF₄ ($\text{72-16-12- < 1}\text{mol}\text{\%}$), has been gained from measurements of heterogeneous equilibria involving various gases (e.g. HF, H₂O) and solids (usually oxides). This information combined with the thermochemical data available for the pure compounds has given formation free energies for the fual components. Their entropies can be estimated from the known entropies of the corresponding solid oxides and the charge on the cation. Their activity coefficients depend primarily on the mol % of LiF in the salt mixture. The solubility of oxides generally decreases with the square of the cation charge divided by the radius. Thus the actinide dioxides and, especially, Pa₂O₂ (or an addition compound of it) are of low solubility. Pourbaix diagrams are presented which summarize the conditions of redox potential and oxide concentration wherein the fuel may be contained without significant corrosion or precipitation reactions, or conditions wherein selective oxide precipitation may be carried out for purposes of fuel reprocessing.     

Dissolution kinetics of Indian PHWR natural UO2 fuel pellets in nitric acid – Effect of initial acidity and temperature

Understanding the dissolution behaviour of plutonium rich MOX fuel is one of the major challenges in fast reactor spent fuel reprocessing. As an initial step, kinetics for the dissolution of Indian PHWR natural UO₂ sintered pellets in nitric acid was studied experimentally. In this paper we have reported the effect of initial nitric acid concentration and temperature on dissolution rate of sintered UO₂ pellets. The shrinking core model was used to correlate the experimental results. The apparent activation energy estimated from the temperature effect of the chemical reaction was found to be about 90.5 kJ/mol. The estimated apparent activation energy indicates that the dissolution rate of UO₂ pellets was controlled by chemical reaction under the experimental conditions.     

Investigation of the PuO2F2 - HF - H2O system (20°C isotherm)

The PuO₂F₂ — HF — H₂O system was investigated by using the method of isomermic solubility at 20 °C. It was established that there are three phases in the sytem: plutonyl fluoride dihydrate PuO₂F₂ · 2H₂O, tetrafluoroxyplutonic acid H₂PuO₂F₄ · 4H₂O, and plutonium oxyfluoride PuO₂F₂, the composition of which was determined by chemical analysis and by using the Schreinmacher method. The chemical individuality of the phases was confirmed by electron absorption spectra. The existence of the H₂PuO₂ F₄ · 4H₂O acid was proved by using the electromigration and the potentiometric titration methods.     

Measurement of Specific Alpha Radioactivity with Respect to Size for Kerala Monazite Sand Using Aerosol Centrifuge

A numerical analysis is presented on the diffusive separation behavior of a three-component gas mixture of ²³⁵UF₆, ²³⁸UF₆, and light gas in a gas centrifuge. The purpose of analysis was to examine the isotope separation performance of the centrifuge in the presence of light gases such as N2 and HF, which inevitably leak in from the atmosphere or accumulate from impurities contained in uranium hexafluoride.

An approximate basic equation is used, which is almost the same as derived by T. Kai for a ternary system from a generalized form of the Stefan-Maxwell equations, taking account of the pressure diffusion in a rapidly rotating cylinder. The method has previously been proposed for application to counter-current centrifuges, which are characterized by stable axial flow in two concentric streams, assuming that the gas mixture is in thermo-dynamical equilibrium between the light gas and UF6 mixture.

Calculations made using the above basic equation indicated that the separative power of the centrifuge is significantly lowered with increasing penetration of the light gas into the separative zone between the two concentric streams. The results of calculation are in fairly good agreement with previously reported theories over a wide range of light gas concentration.

Measurements made using an experimental centrifuge equipped with scoop substantiated the foregoing trends indicated from calculation up to about 1 mol/o average concentration of light gas in the centrifuge cylinder. At higher concentrations, a more significant lowering of separative power by the presence of light gas was indicated from measurement than from calculation.     

Effects of seawater components on radiolysis of water at elevated temperature and subsequent integrity of fuel materials

Effects of seawater components on radiolysis of water at elevated temperature have been studied with a radiolysis model and a corrosion test under gamma-ray irradiation conditions to evaluate the subsequent influence on integrity of fuel materials used in an advanced boiling water reactor. In 2011, seawater flowed into the nuclear power plant system of the Hamaoka Nuclear Power Station Reactor No. 5 during the plant shutdown operation. The reactor water temperature was 250 °C and its maximum Cl^? concentration was ca. 450 ppm when seawater was mixed with reactor water. The radiolysis model predicted that the main radiolytic species were hydrogen, oxygen and hydrogen peroxide. Concentrations of radiolytic products originating from Cl^? and other seawater components were found to be rather low. The dominant product among them was ClO₃^? and its concentration was found to be below 0.01 ppm for a 10⁵ s irradiation period. No significant corrosion of zircaloy-2 and 316L stainless steel was found in the corrosion test. These results led to the conclusion that the harmful influence of radiolytic products originating from seawater components on integrity of fuel materials must be smaller than that of Cl^? which is the main ionic species in seawater.     

A new method for detecting pressure tube failures in Indian PHWRs

For the annulus gas system (AGS) of the standardised Indian pressurised heavy water reactor, an elaborate pressure tube (PT) crack monitoring and detection system is envisaged to ensure safety through leak-before-break. The parameters that are monitored relate to the detection of D₂O moisture leaking in from the primary heat transport (PHT) system through a cracked PT. Since a slow build-up of moisture in the AGS may also occur for reasons other than PT failure, it is desirable that a diverse measurement technique should be available. This paper suggests such a technique, based on the observation that a small reference concentration of fission gases is normally present in the annulus gas. This concentration would change sharply upon PT failure, when the heavy water from the leaking PHT system releases the dissolved fission gas content into the annulus. This paper presents a theoretical study of the parameters that influence the build-up of fission product noble gases in the AGS and shows that leakage rates as low as 10 g h⁻¹ from a PT crack can be detected in a few tens of minutes by this method. This is expected to substantially increase the available time between the leak detection and the PT failure, thus serving as an important tool in meeting the leak-before-break criterion of a critical component in PHWRs.     

Pyrochemical Reprocessing of Weapons Plutonium as Nuclear Fuel for Power Reactors

Experimental work on transferring compact weapons plutonium by pyrochemical methods into its compounds was generalized. Powdered plutonium dioxide was obtained by interaction with moist air at 480–500°C. PuH₃ was synthesized by interaction of the metal with hydrogen at 280°C. PuF₃ was formed in an interaction of dioxides with HF and H₂. The reaction of plutonium hydride with HF also resulted in plutonium trifluoride. PuF₄ was synthesized by interaction of PuO₂ with HF.In summary, compact plutonium can be transfered by water-free methods into its powdered compounds, which can be used in investigations for developing new types of fuel elements for power reactors. 6 figures, 2 tables, 16 references.     

Studies of Ceramic Fuels with Use of Fission Gas Release Loop, (IV)

In-pile release of fission gas from sintered UC pellets in the presence of 8–230 ppm of water vapor in the He sweep gas was measured over the temperature range of 160°–1,000°C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established that the release of short-lived fission gases during irradiation was controlled mainly by pseudo-recoil, while chemical reaction between UC and water vapor, as well as knock-out, appeared to contribute much more significantly in the case of the longer-lived fission gases. The release of fission gas after reactor shutdown was shown to be governed by the UC-H₂O reaction. The ratio of the release due to this reaction in reference to the total release was found to be dependent not only on the concentration of the water vapor but also on the amount present of the accumulated reaction products. Also, a discussion is given on the inordinately high release of ^135mXe observed at 600°C.     

Recovery of actinides from actinide-aluminium alloys by chlorination: Part I

Pyrochemical processes in molten LiCl-KCl are being developed in ITU for recovery of actinides from spent nuclear fuel. The fuel is anodically dissolved to the molten salt electrolyte and actinides are electrochemically reduced on solid aluminium cathodes forming solid actinide-aluminium alloys. A chlorination route is being investigated for recovery of actinides from the alloys. This route consists in three steps: Vacuum distillation for removal of the salt adhered on the electrode, chlorination of the actinide-aluminium alloys by chlorine gas and sublimation of the formed AlCl₃. A thermochemical study showed thermodynamic feasibility of all three steps. On the basis of the conditions identified by the calculations, experiments using pure UAl₃ alloy were carried out to evaluate and optimise the chlorination step. The work was focused on determination of the optimal temperature and Cl₂/UAl₃ molar ratio, providing complete chlorination of the alloy without formation of volatile UCl₅ and UCl₆. The results showed high efficient chlorination at a temperature of 150 °C.