Perspective of Obtaining Rare Earth Elements in Poland

Along with the increasing development of electric and electronic industries, the demand for rare earth elements is also growing due to their high position in many applications. In Poland, there are minerals containing REE; however, the concentration of these elements in raw materials is rather low, so they do not have a big impact on the national economy. The potential source of REE is secondary materials; among them are phosphogypsum, uranium tailings, and the waste electrical and electronic equipment (WEEE). Lanthanides as accompanying metals of uranium in Polish uranium ores were leached in the technology of uranium recovery from these resources. The recovery of REE from pregnant liquors was conducted by solvent extraction and ion exchange. Novel apparatus solutions like membrane contactors in extraction stage were tested. Different types of matrices (uranium ore, phosphorites, etc.) were used

Development of methods for managing radioactive waste and spent nuclear fuel - research and development works at IChTJ

Bezpieczne zagospodarowanie odpadów promieniotwórczych, a zwłaszcza wypalonego paliwa jądrowego, jest jedną z najczęściej podnoszonych kwestii przeciwników dalszego rozwoju energetyki jądrowej i stosowania radioizotopów w różnych dziedzinach życia. Prowadzenie zaawansowanych prac badawczych wspierających program jądrowy kraju i pozwalających na dalszy rozwój metod izotopowych w medycynie, przemyśle i ochronie środowiska naturalnego, jest zadaniem nauki. Jest również jednym z podstawowych celów działania instytutów związanych z atomistyką.Safe management of radioactive waste, especially spent nuclear fuel, is one of the most frequently raised issues of opponents of further development of nuclear energy and the use of radioisotopes in various areas of life. Conducting advanced research works supporting the country's nuclear program and enabling further development of isotope methods in medicine, industry and environmental protection is the responsibility of science. It is also one of the main goals of the institutes related to atomic science

Sorption-Assisted Ultrafiltration Hybrid Method for Treatment of the Radioactive Aqueous Solutions

The paper presents results of studies on the possibility of using the ultrafiltration method supported by sorption on low-cost, easily accessible aluminosilicates to purify water contaminated with radionuclides. An aqueous solution contaminated with radionuclides in the form of cations at different oxidation states—Cs(I)-137, Co(II)-60 and Am(III)-241—as well as pertechnetate anions—TcO4−-99m—was treated by the proposed hybrid method. In the presented work, the influence of the important process parameters (i.e., pH, sorbent dosage, temperature and feed flow rate) on the removal efficiency of radionuclides was studied. The obtained results showed that hazardous impurities, both in the form of cations and anions, may be effectively removed from water by the application of sorption-assisted UF (SAUF) using the clay-salt slimes as a sorbent. As a final stage of the work, we treated the simulated liquid radioactive waste using the SAUF method, also showing satisfactory results in its purification efficiency

MOF-Based Sorbents Used for the Removal of Hg²⁺ from Aqueous Solutions via a Sorption-Assisted Microfiltration

Mercury is considered to be one of the most important chemicals of public health concern. Therefore, it is necessary to develop an effective method of removing mercury ions from aqueous solutions to protect people from exposure to this element. This paper presents research on the application of a sorption-assisted microfiltration (SAMF) hybrid process for the removal of Hg2+ from aqueous solutions. As adsorbents used in the process, the metal-organic-framework-UiO-66-type materials have been considered. The methods of synthesis of two types of metal-organic-framework (MOF) sorbents were developed: UiO-66_MAA modified with mercaptoacetic acid (MAA) and a composite of UiO-66 with cellulose. The results of the experiments performed proved that the separation of Hg2+ from water solutions conducted in such a system was effective; however, a relatively long initial contact time of reagents before filtration was required. The experimental results can be used to optimize the parameters of the SAMF process in order to obtain an effective method of Hg2+ removal from aqueous solutions

Management of Radioactive Waste from HTGR Reactors including Spent TRISO Fuel—State of the Art

In light of the increasing demand for energy sources in the world and the need to meet climate goals set by countries, there is growing global interest in high temperature gas cooled reactors (HTGRs), especially as they are known to be inherently safe nuclear reactors. The safety of HTGRs results, among other, from the nature of the nuclear fuel used in them in the form of coated TRISO particles (tri-structural-isotropic) and the reduction of the total amount of radioactive waste generated. This paper reviews numerous methods used to ensure the sustainable, feasible management and long-term storage of HTGR nuclear waste for the protection of the environment and society. The types of waste generated in the HTGR cycle are presented as well as the methods of their characterization, which are important for long-time storage and final disposal. Two leading nuclear fuel cycle strategies, the once-through cycle (direct disposal or open cycle) and the twice-through cycle (recycling or partially closed cycle), are discussed also in relation to TRISO spent fuel. A short review of the possibilities of treatment of TRISO spent nuclear fuel from HTGR reactors is made

Management of Radioactive Waste from HTGR Reactors including Spent TRISO Fuel—State of the Art

In light of the increasing demand for energy sources in the world and the need to meet climate goals set by countries, there is growing global interest in high temperature gas cooled reactors (HTGRs), especially as they are known to be inherently safe nuclear reactors. The safety of HTGRs results, among other, from the nature of the nuclear fuel used in them in the form of coated TRISO particles (tri-structural-isotropic) and the reduction of the total amount of radioactive waste generated. This paper reviews numerous methods used to ensure the sustainable, feasible management and long-term storage of HTGR nuclear waste for the protection of the environment and society. The types of waste generated in the HTGR cycle are presented as well as the methods of their characterization, which are important for long-time storage and final disposal. Two leading nuclear fuel cycle strategies, the once-through cycle (direct disposal or open cycle) and the twice-through cycle (recycling or partially closed cycle), are discussed also in relation to TRISO spent fuel. A short review of the possibilities of treatment of TRISO spent nuclear fuel from HTGR reactors is made

Multibarrier system preventing migration of radionuclides from radioactive waste repository

Safety of radioactive waste repositories operation is associated with a multibarrier system designed and constructed to isolate and contain the waste from the biosphere. Each of radioactive waste repositories is equipped with system of barriers, which reduces the possibility of release of radionuclides from the storage site. Safety systems may differ from each other depending on the type of repository. They consist of the natural geological barrier provided by host rocks of the repository and its surroundings, and an engineered barrier system (EBS). The EBS may itself comprise a variety of sub-systems or components, such as waste forms, canisters, buffers, backfills, seals and plugs. The EBS plays a major role in providing the required disposal system performance. It is assumed that the metal canisters and system of barriers adequately isolate waste from the biosphere. The evaluation of the multibarrier system is carried out after detailed tests to determine its parameters, and after analysis including mathematical modeling of migration of contaminants. To provide an assurance of safety of radioactive waste repository multibarrier system, detailed long term safety assessments are developed. Usually they comprise modeling of EBS stability, corrosion rate and radionuclide migration in near field in geosphere and biosphere. The principal goal of radionuclide migration modeling is assessment of the radionuclides release paths and rate from the repository, radionuclides concentration in geosphere in time and human exposure to ionizing radiatio

Studies on uranium recovery from a U-bearing Radoniów Dump

This work reports the possibility of uranium recovery from a post-mining uranium ore dump in Poland by a bioleaching method. The studies were conducted on the dump leaching model with the mass of 570 kg of uranium bearing mineral material from Radoniów pile and in the periodic bioreactor with a work volume of 80 dm3 and with mechanical mixing and aeration of the charge. The uranium concentration in the examined material was about 800 ppm. In this process, the consortium of microorganisms isolated from former mines was used. It was composed of the following microorganisms: Bacillius, Pseudomonas, Sphingomonas, Thiobacillus, Halothiobacillus, Thiomonas, and Geothrix. The effi ciency of the uranium bioleaching process was 98% in the reactor, and a yield of 70% was obtained in the dump leaching model. The post-leaching solution contained signifi cant amounts of uranium ions that were separated in two stages: (1) by ion chromatography and then (2) by a two-step precipitation method. The resulting solution was a source of ammonium diuranate, the precursor of yellowcake (uranium oxides)

Ion exchange investigation for recovery of uranium from acidic pregnant leach solutions

The article describes studies on the separation of uranium from acid pregnant leach solutions obtained from Polish uranium ores: dictyonema shales and sandstone rocks. Ion exchange chromatography was applied for uranium sorption, using commercially available, strongly basic anion exchanger, Dowex 1. In model experiments, the influence of degree of crosslinking of Dowex 1 on the efficiency of uranium extraction was investigated. The effect of H2SO4 concentration on the breakthrough curve of uranyl ions for the Dowex 1 resins, of different crosslinking: X4, X8 and X10, was examined. Unexpectedly high increase of exchange capacity of uranium was observed in case of Dowex 1X10. This gives potential opportunity of improving the effectiveness of uranium recovery process. Applying column packed with Dowex 1X10, ‘yellow cake’ with ca. 92% yield and high purity of recovered uranium was obtained. A block diagram of the procedure for uranium and lanthanides extraction from acidic leach liquor has been proposed

Dictyonema black shale and Triassic sandstones as potential sources of uranium

The main objective of the present study was an assessment of the possibility of uranium recovery from domestic resources in Poland. In the first stage uranium was leached from the ground uranium ore by using acidic (sulfuric acid or hydrochloric acid) or alkaline (carbonate) solutions. The leaching efficiencies of uranium were dependent on the type of ore and it reached 81% for Dictyonemic shales and almost 100% for sandstones. The novel leaching routes, with the application of the helical membrane contactor equipped with rotating part were tested. The obtained postleaching solutions were concentrated and purified using solvent extraction or ion exchange chromatography. New methods of solvent extraction, as well as hybrid processes for separation and purification of the product, were studied. Extraction with the use of membrane capillary contactors that has many advantages above conventional methods was also proposed as an alternative purification method. The final product U3O8 could be obtained by the precipitation of ‘yellow cake’, followed by calcination step. The results of precipitation of ammonium diuranate and uranium peroxide from diluted uranium solution were presente