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.     

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.     

Development of a hydrogen mordenite sorbent for the capture of krypton from used nuclear fuel reprocessing off-gas streams

A novel new sorbent for the separation of krypton from off-gas streams resulting from the reprocessing of used nuclear fuel has been developed and evaluated. A hydrogen mordenite powder was successfully incorporated into a macroporous polymer binder and formed into spherical beads. The engineered form sorbent retained the characteristic surface area and microporosity indicative of mordenite powder. The sorbent was evaluated for krypton adsorption capacities utilizing thermal swing operations achieving capacities of 100 mmol of krypton per kilogram of sorbent at a temperature of 191 K. A krypton adsorption isotherm was also obtained at 191 K with varying krypton feed gas concentrations. Adsorption/desorption cycling effects were also evaluated with results indicating that the sorbent experienced no decrease in krypton capacity throughout testing.     

Concepts for the reduction of waste volumes from nuclear fuel reprocessing operations

It is desirable, for environmental and economic reasons, to reduce the number and volume of waste streams from nuclear fuel processing operations. Concepts for achieving this end in future plants are reviewed. These include advanced dissolution concepts, simplification of the solvent extraction flowsheet, and the use of alternative separation techniques such as crystallisation, pyrochemistry and ion exchange.     

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.     

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.     

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.     

A new concept for the nuclear fuel recycle system: Application of the fluoride volatility reprocessing

Hybrid Recycle System (HRS) is proposed as an advanced recycle system. The HRS consists of improved fluoride volatility reprocessing and vibration packing MOX fuel fabrication processing. For the former, a part of U is volatilized as hexafluoride with diluted F₂ gas, and then residual U and Pu are volatilized with concentrated F₂ gas. Plutonium content of the mixed fluoride gas can be adjusted as desired by controlling the U fluorination reaction in the first step. The U is highly decontaminated and the mixture gas of UF₆ and PuF₆ is not purified. The fluoride mixture is reacted with H₂O and H₂ and directly converted to the mixed oxide grain for the vibration packing. The HRS can reduce the costs of reprocessing and fuel fabrication, the amounts of radioactive wastes and the probability of Pu proliferation.     

Development of monitoring technique for the confirmation of spent fuel integrity during storage

A method is developed to monitor integrity of spent fuels stored in a canister that is sealed by weld. To achieve the monitoring, Kr-85 gas was newly adopted as a kind of probe. In the case of a fuel rod defect, Kr-85 gas of the fuel rod is leaked in the canister. By detection of gamma ray (514 keV) emitted from Kr-85 outside of the canister, defected rods can be detected without unsealing the canister. The monitoring technique was developed using small-scaled mock-up experiments and simulated calculation. The result showed that Kr-85 gas leakage of about 10¹¹ Bq is detectable under the noise gamma rays by using the detection system with collimator, which is about 10% of Kr-85 inventory in a fuel rod. Therefore, this monitoring technique is considered as an inspection method prior to transportation of spent fuel from an interim storage facility to a reprocessing plant.