Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

Experimental study on elemental behaviors in fluorination of nuclear spent fuel with flame reactor

Our proposed spent nuclear fuel reprocessing technology named FLUOREX is a hybrid system based on reprocessing technologies of fluorination and solvent extraction for light water reactor fuel. In the current research, we experimentally clarified solid–gas transfer behaviours of the fluorides in the FLUOREX process and identified the volatile and non-volatile compounds in the fluorination. We carried out a fluorination experiment for simulated spent nuclear fuel and solid separation from the UF₆ gas stream. The distribution ratios of fission product elements in the experimental apparatus were evaluated. Molybdenum, Te, Nb, and Ru were volatilized by fluorination and they accompanied the UF₆ gas. However, 22.9% of the Ru and 3.4% of the Nb were retained as solids in the experimental apparatus, contrary to the fact that their partial pressures in the experiment were lower than their vapor pressures. Rubidium, Sr, Zr, Ce, and Nd were completely recovered as solid fluorides, and these results agreed with the prediction based on boiling points of their fluorides. Antimony was completely recovered as a solid; nevertheless, the boiling point of antimony pentafluoride was lower than the process temperature, and that was attributed to the formation of a non-volatile antimony oxyfluoride.     

Gas-Fluoride Technology for Processing Spent Oxide Fuel

The scientific-research work on reprocessing spent oxide fuel by gas-fluoride method is reviewed. The refining possibilities of the basic stages of gas-fluoride technology are studied. The possibility of separating most fission products from the ashes at the fluoridation stage is confirmed experimentally. The use of fluoride sorbents (NaF, BaF₂) permits reaching a total coefficient of removal of fission products from UF₆ at the 10⁷ level. It is shown that deep extraction of plutonium from oxide fuel is possible. The results of investigations on pyrohydrolysis of UF₆ and a mixture of UF₆ with PuF₆ with production of granulate of the oxides with the required density with fluorine content 0.005 mass % and oxygen coefficient 2–2.1 are presented.Recommendations for use of gas-fluoride technology for reprocessing spent oxide fuel from fast and light-water reactors are given taking account of the new requirements for nonproliferation of fissioning materials, and a prediction is given for a closed nuclear fuel cycle using gas-fluoride technology and separation of Np, Am, and Cm for transmutation with the aid of easily melting fluoride melts. 1 figure, 5 tables, 27 references.     

Fundamental Study of the Sulfide Reprocessing Process for Oxide Fuel (I) Study on the Pu,MA and FP Tracer-Doped U3O8

For the recovery of fuel materials from spent nuclear fuel, a novel reprocessing process based on the selective sulfurization of fission products (FP) has been proposed, where FP and minor actinides (MA) are first sulfurized by CS₂ gas, and then, dissolved by a dilute nitric acid solution. Consequently, the fuel elements are recovered as UO₂ and PuO₂. As a basic research of this new concept, the sulfurization and dissolution behaviors of U, Pu, Np, Am, Eu, Cs, and Sr were investigated by γ-ray and α spectrometries in this paper using ²³⁶Pu-, ²³⁷Np-, ²⁴¹Am-, ¹⁵²Eu-, ¹³⁷Cs-, and ⁸⁵Sr-doped U₃O₈ samples. The dependence of the dissolution ratio of each element on the sulfurization temperature was studied and reasonably explained by combining the information of the sulfide phase analysis and the chemical thermodynamics of the dissolution reaction. The sulfurization temperature ranging from 350 to 450°C seems to be promising for the separation of FP and MA from U and Pu, since a clear difference in the dissolution ratio between FP and U was derived by the sulfurization treatment in this temperature range.     

UF6 Photodissociation through Measurement of Visible and Infrared Luminescence

Infrared multiple-photon dissociation of UF₆ was induced at room temperature by irradiation simultaneously with optically pumped CF₄ laser at 615 cm^?1 and TEA CO₂ laser at 1,073 cm^?1. The dissociation of UF₆ was verified from evidence of fluorine generation, obtained by its reduction with H₂ and observing infrared emission from the resulting excited HF molecules. The rate of UF₆ dissociation was determined from the decrease of infrared absorption shown by UF₆. Quantitative indication of the dissociation yield of UF₆ is known to be provided by the more readily measurable visible luminescence from the irradiated zone. The intensity of the visible luminescence was therefore measured, to seek the dependence of dissociation yield on such factors as the fluences of the CF₄ and CO₂ lasers, and the pressures of UF₆ and of Ar buffer gas. A clear threshold proved to exist around 70 J/cm² for the fluence of CO₂ laser, but not for that of the CF₄ laser, indicating that a very slight excitation by the latter laser serves to induce UF₆ dissociation. This suggests that the dissociation energy is supplied in large part by the COZ laser. In the range of UF₆ dissociation, the visible luminescence intensity was found to rise roughly proportionally with CF₄ and CO₂ laser fluences, as well as with the pressures of UF₆ and of added Ar buffer gas.     

Gas Chromatographic Characteristics of Polytrifluoromonochloroethylene Columns as applied to Analysis of Uranium Hexafluoride and other Volatile Inorganic Fluorides

A gas chromatographic assembly for analyzing UF₆ and other volatile inorganic fluorides was constructed. Column packings were consisted of polytrifluoromonochloroethylene oil as partition liquid, and moulding powder as supporting solid, both substances being inert to UF₆ if the operation temperature is not too high.

Through examination of the curves for peak height ratio between elution and inlet, it was found that, to obtain reproducible results, it was necessary to pretreat the columns by fluorine, use purified carrier gas, and establish a definite time schedule for sample introduction.

Dependence of HETP on (1) flow rate of carrier gas, (2) operation temperature, (3) degree of polymerization of polytriflizoromonochloroethylene oils, and (4) liquid phase loading were studied, and conditions for obtaining small HETP are discussed.

Gas chromatography of TeF₆ and MoF₆ were studied, and the possibility of separating these gases from UF₆ has been demonstrated.     

核燃料元件制造厂UF_6气化工序风险分析

多种危险并存于核燃料元件制造厂,因此有必要对核燃料元件厂进行风险分析。目前有多种风险评估方法适用于核燃料元件制造厂风险评估,本文选取HAZOP和LOPA方法,对核燃料元件制造厂风险评估中的最重要工艺UF6气化工序进行了分析。HAZOP分析得到了可能产生严重后果的工艺偏离。LOPA分析得到了针对工艺偏离所采取的独立保护层措施所降低的风险和UF6气化工序的残余风险。     

Evaluation of UF6 Vapor Release in Postulated Accident

The rupture of UF₆ gas line connected to hot UF₆ cylinder, being one of various accidents in UF₆ vapor leak-out, is considered as a postulated accident for uranium enrichment plants. For this type of rupture, we will estimate the amount of UF₆ vapor release based on a simplified calculation model and then make an evaluation of UF₆ vapor release through a ventilation system of feed vaporization facility. Assuming an instantaneous steady state for the change of UF₆ states, an unsteady state thermodynamics process is solved. Numerical examples show that about 52% of the initial UF₆ quantity are vaporized at 80°C (the temperature of the liquid UF₆ in the cylinder). Furthermore, by using the amount of released UF₆ vapor and the collection capacity of HEPA filter for IiF gas, the amount of gaseous UO₂F₂, HF which may be dissipated to the environment are conservatively estimated.     

Radical Reaction Mechanisms in Infrared Multiphoton Dissociation of UF6 with Scavenger Gases

Abstract

The radical reaction mechanisms in the presence of F-atom scavenger gases were investigated in the p-H₂ Raman laser-induced infrared multiphoton dissociation (IRMPD) of gaseous ²³⁶UF₆/²³⁸UF₆ cooled to —35°C in a static gas cell. When CH₄ was added as a scavenger of F-atoms produced via IRMPD of UFC, the dissociation rate of UF6 became several tens of times larger than when no scavenger gas was added. Gas-chromatographic analysis revealed that as low as 7% of the nascent CH₃ radicals were involved in the radical reaction with UF₆. On the other hand, H₂ and C₂H₆ were found to increase both the dissociation rate of UF₆ and the contribution of this non-selective reaction. These results agreed with those obtained in the UV photolysis of UF₆ with scavengers.     

Adsorption Characteristics of Trivalent Rare Earths and Chemical Stability of a Silica-Based Macroporous TODGA Adsorbent in HNO3 Solution

In order to separate trivalent minor actinides (Am and Cm) from high-level liquid waste generated in a nuclear fuel reprocessing process, a silica-based macroporous TODGA (N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide) adsorbent (TODGA/SiO₂-P) was prepared. In this study, the adsorption characteristics of some trivalent rare earths (Y, Nd, and Eu), whose separation behaviors were similar to those of the minor actinides from HNO₃ solution with the TODGA/SiO₂-P adsorbent, and the chemical stability of the adsorbent against the HNO₃ solution were evaluated experimentally. It was found that the adsorbent exhibited a quite strong adsorption especially for Y(III). The standard enthalpy change for the Y(III) adsorption was determined to be ?2.5kJ·mol^?1 using the van't Hoff equation, which indicates that the adsorption was an exothermic reaction. The results of chemical stability experiments showed that the adsorbent had relatively excellent properties in long-time contact with the HNO₃ solution.     

Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries

The effects of evaluated nuclear data files on neutronics characteristics of a fusion–fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T = 300 K, JEFF-3.0 T = 300 K, and CENDL-2 T = 300 K evaluated nuclear data files. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li₂BeF₄) and 10% UF₄ for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe (⁶Li, ⁷Li, ¹⁹F, ⁹Be) and UF₄ (²³⁵U, ²³⁸U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m².     

Uranium Isotope Exchange between Uranium Hexafluoride and Uranium Pentafluoride

To aim at a better understanding of the uranium isotope exchange reaction between gaseous UF₆ and solid UF₅ experiments were done with natural UF₆ gas and solid UF₅ containing 3% ²³⁵U under different pressures of UF₆. The experimental results suggest a two-process reaction with an initial rapid increase of ²³⁵UF₆ in the gas phase followed by its slight and gradual increase. A rate equation based on a collision model is given for the two-process reaction which includes a primary exchange reaction on the solid surface and a secondary reaction participated by underlying UF₅ molecules. An analytical solution is provided for both of ²³⁵UF₆ concentration in the gas phase and ²³⁵UF₅ concentration on the solid surface, which is useful for determining the parameters characterizing the exchange reaction. A numerical analysis is also made to evaluate the influence of gas samplings. A remarkable agreement is found between the particle sizes of UF₅ estimated from the reaction parameter and from the direct observation with an electron microscope. The depletion of ²³⁵UF₅ concentration by the exchange reaction is very small when averaged over the whole solid UF₅, because the depletion is virtually limited to the solid surface due to the small reaction probability of underlying UF₅ molecules.     

Synthesis and thermal conductivity of Y6UO12

Y₆UO₁₂ was synthesized by solid-state reactions of Y₂O₃ and U₃O₈. The high-density pellet of Y₆UO₁₂ was prepared by the spark plasma sintering followed by heat treatment in air for oxygen supplementation. The thermal conductivity (κ) was evaluated using the laser flash method from room temperature to 1173 K. The κ of Y₆UO₁₂ decreased with increasing temperature in the whole temperature range, indicating that the phonon contribution was predominant. The room temperature κ value of Y₆UO₁₂ was 4.90 Wm^?1K^?1. The magnitude relationship of κ among Y₆UO₁₂, Y₆WO₁₂, and Yb₆WO₁₂, i.e. κ of Yb₆WO₁₂ < κ of Y₆UO₁₂ < κ of Y₆WO₁₂, was discussed based on the general lattice thermal conductivity theory.     

Studies of anions from sputtering II: 99Tc and 99Ru

A study of fluoride anion formation in PbF₂ matrices in a Cs⁺ sputter source shows that, for transition metals Groups 3, 4 and 5, distinctively large currents (similar to PbF₃^?) of their corresponding superhalogen anions [1] can be produced by Cs⁺ sputter sources. This trend however, quickly breaks down for elements from Group 6 towards 12. The phenomenon can be understood as a competitive outcome of fluorination reactions among lead, caesium and transition metals, as their fluorine-binding energies generally decrease from Group 3 to 12. Therefore, if a strong fluorine-binding element, such as Mo, coexists with Ru in a fluorinated matrix, the formation of ⁹⁹RuF₄^? can be significantly suppressed relative to MoF₄^? and ⁹⁹TcF₄^?, a possible superhalogen anion of Tc. This phenomenon is indeed observed with a neutron irradiated Mo sample that contains trace levels of both ⁹⁹Tc and Ru, leading to a possible matrix-assisted method for ⁹⁹Tc analysis using small and large accelerators.     

An Apparatus for Isotopic Ratio Measurement in UF5 Formed by Infrared Multiphoton Dissociation of UF6 in Molecular Beam

A Separation factor was measured in isotopically selective infrared multiphoton dissociation (IRMPD) of supercooled UF₆ in a supersonic expansion by multiphoton ionization (MPI) and time-of-flight mass spectrometry (TOFMS). A pulsed free-jet nozzle was used to generate a UF₆-molecular beam seeded in Ar (–10^?7 Torr in UF₆ partial pressure). Two-frequency ρ-H₂ Raman laser beams around 16μm were used for the dissociation of UF₆ under collisionless conditions in the molecular beam where the flow velocity for UF₆ is about 500m/s. The ²³⁵U/²³⁸U isotopic ratios in nascent UF₅ photoproducts were determined by selective MPI of UF₅ at 532 nm followed by TOFMS with a mass resolution as high as 1200. A separation factor of about 2 was observed under the experimental conditions chosen for the demonstration of this method.     

New Approach to the Nuclear Fuel Reprocessing in Non-acidic Aqueous Solutions

A new nuclear fuel reprocessing method based on the anodic dissolution of spent fuels in aqueous alkaline solutions (Na₂CO₃-NaHCO₃) has been proposed. Experiments of the anodic dissolution were performed by using a simulated spent fuel in a Na₂CO₃-NaHCO₃ solution. Uranyl ions produced anodically were present in the solution as stable carbonato complexes, and at the same time, most of the simulated fission products (FP) were precipitated as hydroxo or carbonate compounds. Under this condition, Cs of an alkali metal group was dissolved in the solution and precipitated by adding sodium tetraphenylborate. Uranyl ion was recovered as hydroxo compounds by adding NaOH to the solution after removing precipitates of the simulated FP. In view of waste disposal, ⁹⁹Tc having a long half-life should be removed. Precipitation behavior of Tc(VII) was examined by using Re(VII) as a simulant of Tc(VII). It was found that Re(VII) species are completely removed as a precipitate by adding tetraphenylphosphonium chloride. A large amount of Na used in the present method was recovered as NaHCO₃ by blowing CO₂ into alkaline solutions. As a result, it was clarified that the proposed method is fundamentally possible as a new reprocessing method.     

Radiation-chemical effect of fast electrons on uranium fluorides

The effect of fast electrons on UF₆, UF₅, and UF₄ is investigated. The radiation-chemical yield of the decomposition reaction of UF₆ molecules is equal to 1.1 ·9 10^–2 molecules per 100 eV. As a result.of long irradiation of UF₆, the dynamic equilibrium UF₆ UF₅ + 1/2F₂ is established, The rate of radiation--fluorination of UF₅ and UF₄ is proportional to the fluorine pressure and to the square root of the radiation intensity.Translated from Atomnaya Énergiya, Vol. 16, No. 6, pp. 510–514, June, 1964     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

Potential of a fusion–fission hybrid reactor using uranium for various coolants to breed fissile fuel for LWRs

In this study, neutronic performances of the (D,T) driven hybrid blankets, fuelled with UC₂ and UF₄, are investigated under first wall load of 5 MW/m². The fissile fuel zone is considered to be cooled with three coolants: gas (He or CO₂), flibe (Li₂BeF₄), and natural lithium. The behaviour of the UC₂ and UF₄ fuels are observed during 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. At the end of the operation time, calculations have shown that Cumulative Fissile Fuel Enrichment (CFFE) values varied between 5 and 8.5% depending on the fuel and coolant type. The best enrichment performance is obtained in UF₄ fuelled blanket with flibe coolant, followed by gas and natural lithium coolant. CFFE reaches maximum value (8.51%) in UF₄ fuelled blanket (in row #1) and flibe coolant mode after 48 months. The lowest CFFE value (4.71%) is in UC₂ fuelled blanket (in row #8) and natural lithium coolant at the end of the operation period. This enrichment would be sufficient for LWR reactor. At the beginning of the operation, tritium breeding ratio (TBR) values were 1.090, 1.3301 and 1.2489 in UC₂ fuelled blanket and 1.0772, 1.2433 and 1.1533 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. At the end of the operation, TBR reach 1.1820, 1.3983 and 1.3138 in UC₂ fuelled blanket and 1.2041,1.3266 and 1.2407 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. Nuclear quality of the plutonium increases linearly during the operation period. The isotopic percentage of ²⁴⁰Pu is higher than 5% in UF₄ and UC₂ fuel with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is very important factor for safeguarding. The isotopic percentage of ²⁴⁰Pu is lower than 5% in UC₂ fuel with gas and natural lithium coolant. In these modes, operation period must be increased to safeguarding.     

AMBIDEXTER nuclear energy complex: a practicable approach for rekindling nuclear energy application

Aiming at one of the decisive alternatives for long-term perspectives of the nuclear power, an integral and closed nuclear energy system concept is proposed; namely, the Advanced Molten-salt Break-even Inherently-safe Dual-missioning Experimental and Test Reactor (AMBIDEXTER) nuclear energy complex. This essentially comprises two mutually independent circuits of the radiation/material transport and the heat/energy conversion, centered at the integral reactor assembly, which enables one to utilize maximum benefits of nuclear energy under minimum risks of nuclear radiation. The entire reactor system resides in a thin and large Hastelloy vessel, the internal part of which is divided into a number of equipment compartments with neither connection pipings nor active valves necessary. As the reactor operates at very low FP inventory throughout its designed lifetime and there is no primary heat transport pipings outside the reactor vessel, significant release of radioactive materials due to any equipment failure should be incredible. The nuclear-thermalhydraulic characteristics of the molten ThF₄–²³³UF₄ fuel salt extend the self-sustainability of the AMBIDEXTER fuel cycle to enhance the resource security and safeguard transparency. While maintaining the break-even conversion ratio criterion, a flexible fuel management strategy using a certain choice of denaturants should improve its own proliferation-resistance characteristics. As the core technologies associated with developing the AMBIDEXTER concept are mostly available in commercialized forms at present, investigating the integral performance of the concept should be the prime research topic in ongoing 250 MW_th prototype design studies.