Possibility of Various Types SNF Reprocessing at the PA Mayak exampled with AMB SNF

For the purpose of reprocessing of irradiated nuclear fuel from the water-cooled graphite-moderated pressure-tube reactor named AMB from decomissioned Russian “Atom Peaceful Big”, modernization of the process flow-sheet of the RT-1 plant is being carried out at PA Mayak with participation of FSUE KRI and VNIINM. A particular AMB SNF feature is extremely broad range of fuel compounds with the main ones being the uranium-molybdenum metal, uranium oxide and uranium carbide compositions usually dispersed in magnesium or calcium. Wide range of fuel compositions required to amend SNF dissolution, extraction processing, evaporation of high-level radioactive wastes and vitrification of high-level radioactive wastes. The above set of laboratory research was completed with dynamic tests using samples of AMB from the water-cooled graphite-moderated pressure-tube reactor. Tests have shown the possibility of processing the entire range of AMB SNF at the radiochemical plant RT-1 plant of the PA Mayak. Thus, the ability of the RT-1 plant to process different fuel compositions, including the long-term research reactor fuel have been proved experimentally.     

Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

The high level waste (HLW) generated from the reprocessing of the spent fuel of pressurized heavy water reactor has been characterized for the minor actinides. The radiation dose of the waste solution was reduced by radiochemical separation of cesium from HLW by solvent extraction with chlorinated cobalt dicarbollide dissolved in 20% nitrobenzene in xylene. Minor actinides (Np, Pu, Am, Cm) in the high level waste were assayed by alpha spectrometry following radiochemical separation. The gross alpha activity determined by liquid scintillation agrees well (within 10%) with the cumulative quantities of actinides determined by alpha spectrometry.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

Reactor physic and reprocessing scheme for innovative molten salt reactor system

The molten salt reactor is one of the six concepts retained by the Generation IV forum in 2001. Based on the MSRE and MSBR concepts developed by ORNL in the 60s which involve a liquid fuel constituted of fluorine molten salt at a temperature close to 600 °C, new developments with innovative approach and technology have been realized which contribute to strongly improve the concept. The thorium breeder potentiality is closely related to the use of a liquid fuel which is able to be periodically treated. A reprocessing scheme has been established to treat used fuel by extraction of fission products. According to the Gen IV philosophy for closed cycle nuclear reactor, the actinides are sent back in the reactor core. In this way, the wastes radiotoxicity is strongly decreased and the use of natural resource is optimized. This paper describes an innovative reactor concept, the TMSR-NM (non-moderated thorium molten salt reactor), from the nuclear physic point of view and the different steps involving in the reprocessing scheme from the chemical point of view.     

Toward mechanistic understanding of nuclear reprocessing chemistries by quantifying lanthanide solvent extraction kinetics via microfluidics with constant interfacial area and rapid mixing

The closing of the nuclear fuel cycle is an unsolved problem of great importance. Separating radionuclides produced in a nuclear reactor is useful both for the storage of nuclear waste and for recycling of nuclear fuel. These separations can be performed by designing appropriate chelation chemistries and liquid-liquid extraction schemes, such as in the TALSPEAK process (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes). However, there are no approved methods for the industrial scale reprocessing of civilian nuclear fuel in the United States. One bottleneck in the design of next-generation solvent extraction-based nuclear fuel reprocessing schemes is a lack of interfacial mass transfer rate constants obtained under well-controlled conditions for lanthanide and actinide ligand complexes; such rate constants are a prerequisite for mechanistic understanding of the extraction chemistries involved and are of great assistance in the design of new chemistries. In addition, rate constants obtained under conditions of known interfacial area have immediate, practical utility in models required for the scaling-up of laboratory-scale demonstrations to industrial-scale solutions. Existing experimental techniques for determining these rate constants suffer from two key drawbacks: either slow mixing or unknown interfacial area. The volume of waste produced by traditional methods is an additional, practical concern in experiments involving radioactive elements, both from disposal cost and experimenter safety standpoints. In this paper, we test a plug-based microfluidic system that uses flowing plugs (droplets) in microfluidic channels to determine absolute interfacial mass transfer rate constants under conditions of both rapid mixing and controlled interfacial area. We utilize this system to determine, for the first time, the rate constants for interfacial transfer of all lanthanides, minus promethium, plus yttrium, under TALSPEAK process conditions, as a first step toward testing the molecular mechanism of this separation process.     

Preliminary Design of Hybrid Energy Reactor and Related Nuclear Fuel Cycle

We propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER), based on current fusion science and technology and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The FFHER could achieve greater energy multiplication (M>10 for U-Zr alloy fuel and M>5 for UO2 fuel) and tritium sustainability (TBR>1.05). The sub-critical blanket will last 30 years without reshuffling fuel. Fission products are the only waste that needs disposal. A new dry process called Fission Product Removal (FPR) replaces conventional reprocessing. It is only necessary to remove the cladding, vent the volatiles and pulverize the solids as feedstock for EM2 fuel fabrication. The AIROX (or DUPIC) process is an example of this operation and has been well demonstrated. After removing the fission products from its 30-year discharge, the refabricated fuel is returned to the reactor for another cycle, thereby reducing the need for enrichment and the proliferation resistance would be increased.     

核化学与放射化学的研究进展

在我国核能快速发展的新形势下,新型核能资源的开发、乏燃料后处理、放射性废物处理与处置等核燃料循环化学研究日益活跃。随着科学技术的不断发展,离子加速器、反应堆、各种类型的探测器和分析设备、以及计算机技术等的发展,核化学与放射化学研究的范围和成果在不断扩展和增加,如核安全、环境放射化学、放射分析化学、放射性药物与标记化合物等,研究成果对于国防建设、核能发展、核技术应用等方面具有重要支撑作用。本文综述了近年来国内在上述领域所取得的研究进展。共引用参考文献161篇。     

离子交换净化辐照三烷基氧化膦中脂肪酸

本文研究了用大孔阴离子交换树脂D301R从乏萃取剂30%TRPO-煤油溶液中净化去除脂肪酸的方法,探讨了该法用于乏萃取剂净化的可行性。     

Development of Mathematical Model for Balance Settlement of Product and Waste Flows from SNF Reprocessing

A mathematical model has been built to estimate steady-state characteristics of flows and products in SNF processing systems. A developed calculation module will enable to change not only a type of SNF to be reprocessed and relationship among the processing units (stages) while building the model, but also will provide for entering new processing stages into the model to develop new process flow-sheets, to optimize and to compare the existing ones. Optimizing calculations have been conducted on the basis of the model regarding new promising flow-sheets of reprocessing SNF from fast reactors. The selected data structure within the model will provide for modelling of not only SNF reprocessing but also of other nuclear fuel processing stages.     

Determination of 94Nb in radioactive waste using ion exchange chromatography

This paper describes a rapid method of ⁹⁴Nb pre-concentration, separation and purification by using cation and anion exchange resins. The method is suitable for analyzing highly contaminated radioactive waste samples in a relatively short time and high decontamination factors. The use and effectiveness of the method was successfully tested by analysis of samples from nuclear reactor parts such as control rod drive shaft, shielding cassettes, neutron in-core measurement channels (KNI), pressure vessel construction material and fuel cassette construction material samples.     

MicroRaman Measurements for Nuclear Fuel Reprocessing Applications

Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices.Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size.This paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO₃, and UO₂(NO₃)₂.     

CARBEX process,a new technology of reprocessing of spent nuclear fuel

Advances in the CARBEX process, a new aqueous chemical method for reprocessing of spent nuclear fuel (SNF) in carbonate media, are considered. A review of carbonate methods for SNF reprocessing is given. The CARBEX process concept is presented and experimental data for every stage of the CARBEX process: high-temperature oxidation of spent fuel composition, its oxidative dissolution in carbonate aqueous solutions, extraction refining of U(VI) and Pu(VI), solid-phase re-extraction of carbonate complexes of U(VI) and Pu(VI), and obtaining of uranium and plutonium dioxide powders for fabrication of ceramic nuclear fuel, are discussed. It was shown that the CARBEX process can be more effective and safe than the well-known industrial PUREX process.     

Scaling-up experience of actinides partitioning from intermediate level high salted alkaline waste streams of Purex process using Versatic-10 extractant

Partitioning of minor alpha-emitting actinides, especially U, Pu and Am from medium active alkaline waste is possible from intermediate level liquid wastes (ILLW) produced during spent fuel reprocessing following Purex process. This paper deals with the efficient removal of alpha-activity from ILLW by solvent extraction process. Counter current batch extraction with O/A ratio 2:1 as well as multistage mixer settler has demonstrated that most of the alpha-activity was removed from the alkaline effluents using 20% Versatic-10 (V-10) in dodecane after giving 3 to 4 contacts, thus converting alkaline waste as non-alpha waste. Under the optimum conditions (pH 9.0-9.5 and VA-10), both Pu(IV) and Am(III) are highly extractable whereas U(VI) is relatively poorly extracted. To assess the applicability of this process for regular treatment of the waste, a feasibility study on pilot plant scale using six stage mixer settler was operated to treat the ILLW. The results indicated that almost >99.90% alpha-emitting actinides are removed. Dilute nitric acid (0.5M HNO₃) served as the most efficient strippant for all these actinides. This facilitate an easy regeneration of the extractant which can be recycled. This method is useful in obtaining alpha-free wastes and had positive impact on ease and safety aspects during subsequent waste treatment and long term storage.     

Separation of Trivalent Actinides from High-Active Waste

A typical high-active waste (HAW) arising from reprocessing of (U0.3Pu0.7)C fuel irradiated to the burn-up of 155 GWd/Te in a fast breeder test reactor (FBTR) was characterized. Partitioning of trivalent actinides from HAW was demonstrated using a solvent, 0.2 M n-octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) – 1.2 M tri-n-butylphosphate (TBP) in n-dodecane (n-DD), in a mixer settler. The results established quantitative separation of trivalents (Am(III) + Ln(III)) from HAW and recovery (> 99%) using a citric acid-nitric acid formulation. The mutual separation of lanthanides and actinides from the stripped product was studied by using bis(2-ethylhexyl)diglycolamic acid (HDEHDGA), synthesized in our laboratory.     

Quantitative Auswertung von Dünnschicht-Chromatogrammen: On line-Kopplung mit programmierbaren elektronischen Tischrechnern

Reprocessing of spherical THTR fuel elements shall be tested in the Jülich pilot plant JUPITER. This fuel type differs significantly from other fuel elements with respect to shape, composition and fissile material content. It requests special provisions for reprocessing and the necessary material balancing and safeguarding. Two material balance areas (MBA) are defined: head end and chemical extraction process. Within the 1. MBA uranium and thorium are balanced mainly by using a combination of digital counting of the fuel spheres, gammaspectrometric burn-up determination of individual spheres and X-ray fluorescence determination of uranium and thorium in nitric acid solutions which have been obtained by dissolution in Thorex reagent of the heavy metal oxides after burning of the graphite matrix. The 2. MBA begins with the solution for the chemical extraction process, collected in the so called accountability tank. After extraction according to the Thorex flowsheet the process streams are monitored in line for process control, and off line for material balancing and safeguarding. This is performed mainly by X-ray fluorescence analysis, potentiometric titrations, alpha- and mass spectrometry.     

Utilization of Slovak bentonites in deposition of high-level radioactive waste and spent nuclear fuel

The basic strategic aims in the field of managing high-level radioactive waste and liquidation of nuclear power plants are all contained in the Energy policy of the Slovak Republic. Its aim is to resolve the concept of the backside of the nuclear energetics fuel cycle??long-term deposition of high-level radioactive waste and spent nuclear fuel (SNF). The most important form of high-level radioactive waste and SNF long-term deposition is their deposition in deep geological formations created by natural as well as engineering barriers used to isolate the long-lived radionuclides from the biosphere. The basic components of these barriers are clays, of which bentonite is generally referred to as the most suitable clay material. There are a few significant bentonite deposits in the Slovak Republic: Jel?ový potok, Kopernica, Lastovce, Lieskovec, Dolná Ves. The review article summarizes the information on geotechnical properties of Slovak bentonites published up-to-date, which is inevitable to know for the intention of their use. It highlights the advantages and shows drawbacks of five Slovak deposits. It suggests further research direction, to draw a thorough hydraulical, microbial and radiation profile of Slovak bentonites.     

Spectrophotometric determination of plutonium in highly radioactive liquid waste using an internal standardization technique with neodymium(III).

A simple and rapid spectrophotometric method has been developed for the determination of Pu in highly radioactive liquid waste. This method uses Nd(III) as an internal standard, which enables us to determine the concentration of Pu and to authenticate the whole analytical scheme as well. A Nd(III) standard mixed with a sample solution and Pu was quantitatively oxidized to Pu(VI) with Ce(IV) in a nitric acid medium, having the maximum absorbance at 830 nm. A spectrophotometric measurement of Pu(VI) was subsequently performed to determine the concentration compared with the maximum absorbance of Nd(III) at 795 nm. It was estimated that the relative expanded uncertainty for a real sample is less than 10%. The limit of detection was calculated to be 1.8 mg/L (3 sigma). The proposed method was also validated through comparison experiments with isotope dilution mass spectrometry, and was successfully applied to analysis for nuclear waste management at spent nuclear fuel reprocessing plants.     

Characterization of the Insoluble Sludge from the Dissolution of Irradiated Fast Breeder Reactor Fuel

Insoluble sludge is generated in the reprocessing of spent fuel. The sludge obtained from the dissolution of irradiated fuel from the “Joyo” experimental fast reactor was analyzed to evaluate its chemical form. The sludge was collected by the filtration of the dissolved fuel solution, and then washed in nitric acid. The yields of the sludge weight were less than 1% of the total fuel weight. The chemical composition of the sludge was analyzed after decomposition by alkaline fusion. Molybdenum, technetium, ruthenium, rhodium, and palladium were found to be the main constituent elements of the sludge. X-ray diffraction patterns of the sludge were attributable to Mo₄Ru₄RhPd, regardless of the experimental conditions. The concentrations of molybdenum and zirconium in the dissolved fast reactor fuel solutions were low, indicating that zirconium molybdate hydrate is produced in negligible amounts in the process.     

Removal of radiocesium from simulated alkaline salt loaded radwaste solution using potassium cobalt(II) hexacyanoferrate(II)

The inorganic sorbent potassium cobalt(II) hexacyanoferrate(II) was tested for removal of radiocesium from alkaline salt solutions that are typical of intermediate level radioactive wastes generated at spent fuel reprocessing plants in India. Excellent results were obtained both in batch equilibration and column operation.     

Radioactive contamination of groundwater at the Andreeva Bay Shore Technical Base,Northwest Russia

The Andreeva Bay Shore Technical Base is one of the largest, most contaminated nuclear legacy sites in Northwest Russia. Radioactive contamination at the site stems from servicing and maintenance activites for Russian Northern Fleet nuclear submarine. Studies of groundwater contamination have been conducted using measurements taken at different boreholes around the site. Results indicate that groundwater contamination has occurred in some areas of the Andreeva Bay facility. These areas are primarily located near Building 5 where accidental releases occurred in 1982 during storage of spent nuclear fuel (SNF) and near the liquid and solid radioactive waste (LRW and SRW) storage facilities which have been infiltrated by precipitation waters.