Treatment of molten salt wastes by phosphate precipitation: removal of fission product elements after pyrochemical reprocessing of spent nuclear fuels in chloride melts

The removal of fission product elements from molten salt wastes arising from pyrochemical reprocessing of spent nuclear fuels has been investigated. The experiments were conducted in LiCl-KCl eutectic at 550 °C and NaCl-KCl equimolar mixture at 750 °C. The behavior of the following individual elements was investigated: Cs, Mg, Sr, Ba, lanthanides (La to Dy), Zr, Cr, Mo, Mn, Re (to simulate Tc), Fe, Ru, Ni, Cd, Bi and Te. Lithium and sodium phosphates were used as precipitants. The efficiency of the process and the composition of the solid phases formed depend on the melt composition. The distribution coefficients of these elements between chloride melts and precipitates were determined. Some volatile chlorides were produced and rhenium metal was formed by disproportionation. Lithium-free melts favor formation of double phosphates. Some experiments in melts containing several added fission product elements were also conducted to study possible co-precipitation reactions. Rare earth elements and zirconium can be removed from both the systems studied, but alkaline earth metal fission product elements (Sr and Ba) form precipitates only in NaCl-KCl based melts. Essentially the reverse behavior was found with magnesium. Some metals form oxide rather than phosphate precipitates and the behavior of certain elements is solvent dependent. Caesium cannot be removed completely from chloride melts by a phosphate precipitation technique.     

Pulsed pneumatic transport of solid wastes from reprocessing of spent nuclear power plant fuel to storage

The results of experimental investigations of the effect of a gel-like residue on the transport of pieces of structural materials of fuel assemblies from nuclear power plants by a pulsed pneumatic transport system to storage are presented. The data obtained show that even increasing the viscosity of the wetting medium substantially (by a factor of 100) has little effect on the transport regime and technology. 3 figures, 2 tables, 4 references. Translated from Atomnaya énergiya, Vol. 88, No. 1, pp. 48–51, January, 2000.     

A new concept of nuclear fuel reprocessing by applying ion-exchange technology

The development of advanced technology for the spent nuclear fuel reprocessing should be achieved not only considering cost, non proliferation and reduction of radioactive wastes but also corresponding to both spent nuclear fuels of LWR and FBR.

We have proposed an ion exchange process for reprocessing using a new type ion exchanger developed to chemical method of U enrichment technology. This process possess possibility of a sharp cut in cost, since this ion exchanger is characterized by rapid adsorption-desorption rate dominating the treatment rate.

From the basic experimental results, this reprocessing process has been constructed by 3 ion exchanger columns which consist of a main separation column, the uranium-refining column and the plutonium-refining column.

Comparing ion exchange process with the conventional Purex process, this ion exchange process has many advantages such as the decrease in the number and size separation equipment, solvent-spent free and alkaline-liquid-spent free. With these advantages, this process is estimated that the construction cost of reprocessing process is greatly reduced comparing to the conventional process.     

Validation of analytical methods for nuclear spent fuel reprocessing

For more than 50 years, CETAMA, the Commission for establishment of analytical methods from the French Alternative Energies and Atomic Energy Commission, has provided Certified Reference Materials and Interlaboratory Comparisons for the development and validation of analytical methods in the nuclear field. In the future, the nuclear spent fuel reprocessing industry will require new standards and methods to comply with high content plutonium fuel and new extraction solvents. These standards and methods will have to be fully validated in order to ensure the quality of the analytical results obtained by the laboratories.In this context, a new ²⁴²Pu reference material, certified for its isotopic composition, has been recently produced. A novel statistical approach for data processing has been used and has led to a certified value of 0.985459 ± 0.000052 for the n(²⁴²Pu)/n(Pu) atomic ratio. In addition, an interlaboratory comparison has also been organized for the validation of a method for the analysis of DMDOHEMA, and its degradation products. This compound is considered as a new extractant candidate in the frame of separation processes for transmutation of long-lived radionuclides. The methodology and results obtained in both cases are presented.     

Combined reprocessing of weapons plutonium into high-purity oxide fuel

The main results of a study of regimes for dissolution of plutonium oxide and mononitride powders, obtained by the pyrochemical method from weapons plutonium, in acids for subsequent extraction removal of gallium, americium, and ballast impurities and obtaining ceramic-type plutonium oxide powder suitable for fabricating mixing oxide fuel are presented. It is established experimentally that plutonium oxide and mononitride obtained by the pyrochemical method dissolve rapidly in the acid mixture HNO₃ 12 moles/liter and HF 0.1 moles/liter. Plutonium extraction into solution reaches >99%.     

Monitoring homogeneous neutron absorbers during reprocessing of spent nuclear fuel

Bochvar All-Union Scientific-Research Institute for Standardization in Mechanical Engineering. Mayak Industrial Association. Translated from Atomnaya Énergiya, Vol. 72, No. 5, pp. 451–453, May, 1992.     

Purification of uranium products in crystallization system for nuclear fuel reprocessing

Uranium crystallization system has been developed to establish an advanced aqueous reprocessing for fast breeder reactor (FBR) fuel cycle. In the crystallization system, most part of uranium in dissolved solution of spent FBR-MOX fuels is separated as uranyl nitrate hexahydrate (UNH) crystals by a cooling operation. The targets of U yield and decontamination factor (DF) on the crystallization system are decided from FBR cycle performance and plutonium enrichment management. The DF is lowered by involving liquid and solid impurities on and in the UNH crystals during crystallization. In order to achieve the DF performance (more than 100), we discuss the purification technology of UNH crystals using a Kureha Crystal Purifier (KCP). Results show that more than 90% of uranium in the feed crystals could be recovered as the purified crystals in all test conditions, and the DFs of solid and liquid impurities on the purified UNH crystals are more than 100 under longer residence time of crystals in the column of KCP device. The purification mechanism is mainly due to the repetition of sweating and recrystallization in the column under controlled temperature.     

Environmentally acceptable nuclear fuel cycle development of a new reprocessing system

The aim of the present study is to establish a new reprocessing system for spent nuclear fuel, which would overcome the environmental problems in the nuclear fuel cycle. In order to achieve this, the following subjects have to be conquered: recoveries of high ratios of uranium and trans uranium elements from spent nuclear fuel, separations of strong radioactive elements, such as Sr and Cs, and assurance of the extreme safety during operation. The last subjects might be of particular importance in order to avoid any potential danger. Therefore, in the present system all processes were performed under mild aqueous conditions. Experiments were carried out for a simulated spent fuel solution, which was calculated from the ORIGEN CODE containing uranium and 17 major elements. The system consists of the following processes: 1. dissolution of spent UO₂ fuel involving off-gas treatment of I and Ru; 2. neutralization of the dissolved fuel solution with NaHCO₃---Na₂CO₃ mixed solution to be slightly basic at pH about 9 followed by the separation of precipitated fission products by centrifugation; 3. separation of Cs by a precipitation method using tetraphenylborate ion; 4. recovery of U, Np and Pu as precipitates of hydrolyzed compounds from alkaline solution; 5. separation of Am and Cm from lanthanide elements. The concentration of residual uranium in the final solution was measured to be ppm order, indicating that the decontamination factor of U was as large as 10⁴. Other hexa-valent actinide ions, NpO₂²⁺ and PuO₂²⁺, also have extremely large stability constants for the complex formation with carbonate ion, and are expected to behave similarly with UO₂²⁺. In conclusion, the present reprocessing system enables us to recover U, Pu and Np from spent nuclear fuel by means of a simple precipitation method in much higher ratios compared with other reprocessing methods, to separate hazardous Cs and Sr from high-level waste, and to exclude any potential danger owing to chemical processes under mild aqueous conditions.     

Equipment for vitrification of nuclear power plant wastes in a direct-heating ceramic furnace

Joint-Stock Company ‘Sverdlovsk Scientific-Research Institute of Chemical Machinery.’ N. A. Rakov Ministry of Atomic Energy of the Russian Federation. Translated from Atomnaya énergiya, Vol. 80, No. 3, pp. 219–221, March, 1996.