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.     

Adsorption of Molybdenum Hexafluoride on Magnesium Difluoride for Uranium Purification in FLUOREX Reprocessing

The nuclear fuel reprocessing method FLUOREX is a hybrid system based on fluoride volatility and using solvent extraction. Spent nuclear fuel is fluorinated, and most of the uranium is recovered as UF₆ gas. UF₆ contains some volatile fission product (FP) fluorides, so we considered their elimination from UF₆ by adsorbing them on fluoride adsorbents. We experimentally examined the adsorption of MoF₆ on MgF₂ adsorbent; MoF₆ is present as a volatile FP fluoride in UF₆ produced by the fluorination of spent nuclear fuel. The adsorption isotherm of MoF₆ adsorption on MgF₂ was obtained at MoF₆ pressures from 10^?4 to 50 kPa. The saturated adsorption amount was 1:3 ± 0:4 mg/m² at MoF₆ pressures from 10^?4 to 1 kPa. At MoF₆ pressure of about 10^?3 kPa, the saturated adsorption amount had no dependence on adsorption temperatures from 398 to 463 K. We deduced that MoF₆ was adsorbed as a monomolecule layer on the MgF₂ surface at MoF₆ pressures from 10^?4 to 1 kPa, and the MoF₆ partial pressure in UF₆ could be decreased below 1 × 10^?4 kPa, which is the specific MoF₆ concentration for the reenrichment process.     

Radiokrypton in the Atmosphere

A new reprocessing technology, FLUOREX, is proposed for the thermal reactor cycle. In the dissolution process of FLUOREX, it needs to estimate the concentration of fluoride ion in the solution to avoid formation of insoluble plutonium precipitate. We measured the solubility products of lanthanide fluorides, LaF₃, CeF₃, and NdF₃ in nitric acid solution and evaluate the concentration of fluoride ion in the dissolution process. Solubilities were measured by both of dissolution method and precipitation method. The solubility products are about 10¹ to 10² times larger than those in water. This largeness may be caused by higher ionic strength of nitric acid solution. The newly obtained solubility products for the lanthanides are useful to calculate the maximum concentration of plutonium ions in the FLUOREX dissolution process, although the further study for the solubility products for plutonium tetrafluoride is needed in the nitric acid solution.     

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.     

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.     

Fluorination of Uranium Dioxide by Bromine-Fluorine System

The addition of bromine to fluorine flow makes it possible to fluorinate UO₂ into UF₆even below 200°C, at which temperature fluorination does not proceed with fluorine alone. Presence of bromine corresponding to about 6% of the fluorine concentration is sufficient to induce the fluorination. This effect of bromine is much greater than would be expected from the presence of bromine fluorides in concentrations such as would be produced by direct combination of the fluorine with the added bromine. It appears that the fluorination is enhanced when the mixed gas is held for 3~20 sec before arriving at the sample. The main component of the reactant gas is BrF₃.     

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.     

Fluorination of Uranium Metal by Fluorine Gas

A study has been made on the fluorination of uranium metal chips to UF₆ with fluorine gas in the temperature range of 100° to 400°C. The formation of UF₆ was influenced by the concentration and the flow rate of fluorine gas and by the reaction temperature. The reaction occurred apparently above 200°C. Uranium metal was first converted to low fluorine content compound such as UF₃ or UF_4-x, then to UF₄ and finally to UF₆. The intermediate compounds were confirmed by X-ray analysis and by their color.     

Modeling and criticality evaluation of the voloxidation process

Voloxidation is a necessary process in the dry reprocessing of spent nuclear fuels. The criticality evaluation plays a considerable role in the design of voloxidation apparatus. As conservative results are always preferred in a criticality evaluation, an optimized model was built in consideration of both the geometry of voloxidation apparatus and the occurring forms of evaluation material. The criticality evaluation of fresh UO₂ fuel and PWR spent fuel were then performed by employing Monte Carlo techniques, respectively. It is demonstrated that there is no criticality risk concerning the voloxidation process dealing with fresh UO₂ or PWR spent fuel if water does not intrude into the cell. However, if water intrudes and mixes with the fuels, the subcritical mass limit is 40.1 ± 0.1 kg for fresh UO₂ and 19,155 ± 50 kg for spent fuel. The contributions of ¹H and ²³⁵U were analyzed quantitatively by the TSUNAMI code to clarify the competition between ¹H moderation effect and its dilution effect on the concentration of ²³⁵U.     

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.     

乏燃料后处理厂核材料衡算与控制国际经验及关键技术启示

本文系统地调研和分析了国内乏燃料后处理厂核材料管制现状,国外商业乏燃料后处理厂核材料衡算与控制措施的实施经验和采用的关键技术,包括典型商业乏燃料后处理厂物料平衡区和实物盘存关键测量点的设置、核材料衡算与控制措施的总体设计要求、近实时衡算的概念等。根据调研结果和分析,针对我国核材料管制的现状,提出了我国在商业乏燃料后处理厂核材料管制技术准备工作的几点初步建议。     

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.     

The criteria of flushing phenomena under microwave heating

In the microwave heating (MH) de-nitration method developed in Japan, a mixed solution of uranyl nitrate and plutonium nitrate recovered from the spent fuel in the reprocessing plant is converted directly to mixed oxide (MH-MOX) powder. This MH-MOX powder is utilized to fabricate MOX fuel with UO₂ powder for FBR. The MH method is accompanied with transient boiling phenomena such as overflow and flushing. Toward high-speed and high-capacity conversion by MH-method in the future, it is required to avoid overflow and flushing and to understand optimal conditions for design and operation. At the first step for these objectives, basic knowledge of transient boiling phenomena by the MH-method has been acquired with using distilled water. It is observed that generation of singular bubble triggers flushing and distilled water just before flushing is superheated by a temperature 10 °C higher than boiling temperature. Based on these results, it is clarified that the occurrence criteria of flushing correlate with absorbed power in the water and released power from the water surface.     

Fluorex reprocessing system for the thermal reactors cycle and future thermal/fast reactors (coexistence) cycle

A new reprocessing technology, FLUOREX was proposed for thermal reactors cycle and future thermal/fast reactors (coexistence) cycle. The proposed system is a hybrid system that combines fluoride volatility and solvent extraction methods. Spent fuel will be sheared and cladding material will be removed by dry oxidation/reduction method such as AIROX process. Fluorination and purification of most uranium can be easily achieved by fluoride volatility method with compact facility. About 10% residues including plutonium can be treated in well-established PUREX method, which means this facility load will be about 1/10 of the conventional PUREX facility with same capacity. Between fluorination process and PUREX process, there is a pyrohydrolysis process where the fluoride compounds from fluorination process are converted to the oxides. Pure mixture of Pu and U can be obtained by solvent extraction method without separating Pu and U, which is suitable for conventional MOX fuel fabrication. The system can recover pure U and MOX with the decontamination factor of over 10⁷ and can drastically reduce the cost and waste generation compared with the conventional one.

Semi engineering scale experiments for the fluorination, pyrohydrolysis, and dissolution of Pu containing materials were carried out. From those experimental results, key elemental processes were fundamentally proofed.     

压水堆核燃料循环情景模式初步研究

根据我国核电发展现状和中长期发展规划及中长期（2030、2050）发展战略研究,假设2050年前我国压水堆核电发展规模,基于压水堆乏燃料后处理,回收的钚做成MOX燃料放入压水堆中使用,MOX燃料只使用1次的循环模式,进行核能发展情景研究。基于压水堆可装载30%比例MOX燃料的已有研究结果,考虑我国主要的两种压水堆堆型M310和AP1000,进行压水堆核燃料循环分析。利用核能发展情景动态分析程序DESAE-2,给出了不同情景模式下天然铀需求量、乏燃料累计量等。结果表明：至2050年,B₁和B₂模式较A模式分别节省天然铀4.1万t和2.9万t。     

Dissociation of the UF6 molecule under the action of fragments from fission of the uranium nucleus

It was shown that when irradiated with neutrons, uranium hexafluoride decomposed to intermediate uranium fluorides (presumably UF₅ and fluorine).The rateof decomposition wasG = 0.5 mole/l00 ev or 0.21mole/hr per kw of power liberated in the gas. It was also shown that in addition to dissociation of UF₆ during irradiation there was also recombination of the decomposition products. As a result, equilibrium concentrations of fluorine and UF₆ are set up, and these depend on the strength of the radiation. When mixed with fluorine, uranium hexafluoride is a radiation-stable compound even at room temperature and may be used as a fuel in a nuclear reactor.In conclusion, the authors consider it their pleasant duty to thank I. K. Kikoin for his interest and valuable advice.     

MOX Co-deposition Tests at RIAR for SF Reprocessing Optimization

The Oxide Electrowinning method has been studied as one of the candidate dry reprocessing concepts of the future fuel cycle systems. On the MOX co-deposition process, the main process of that method, some fundamental experiments have been performed to confirm its feasibility. In the experiments, several parameters were set to study the suitable electrolysis condition to obtain desired granule of MOX. The concentrations of uranium, plutonium, fission products(FP) simulators, and corrosion products(CP) simulators were adopted as the parameters. The blowing gas composition (O₂, Cl₂, Ar) during the electrolysis was also set as the variable condition. Through these experiments, it was clarified that the partial pressure of chlorine gas during electrolysis was important to obtain MOX granule with high Pu concentration (about 30%) without generating bottom precipitation in melt. Finally, adequacy of the process control method for MOX co-electrolysis was confirmed through the test using spent fast reactor(FR) fuel.     

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.     

Behavior of Ruthenium in Fluoride-Volatility Processes, (IV)

The products of the following three reactions were studied in relation to the reprocessing of oxide fuels: (i) fluorination of Ru by F₂ (ii) fluorination of Ru by a O₂-F₂ mixture, and (iii) secondary process of RuO₂-F₂ reaction. The product of Ru-F₂ reaction was only RuF₅; the mass spectrum of RuF₅ was obtained. Fluorination of Ru by a O₂-F₂ mixture resulted in the production of RuF₅ (85～75%) and RuOF₄ (15～25%); these results are different from those reported by earlier workers. The use of radioactive Ru^*O₂ traced the behavior of RuOF₄ in the apparatus. RuOF₄ decomposes on the wall which was not preliminarily coated with ruthenium; refluorination was effective for removal of the deposit. These results suggest that the fluorination of irradiated oxide fuels volatilizes the ruthenium as a mixture of RuOF₄, RuF₅, and a small amount of RuO₄.