Adsorption Materials Development for the Separation of Actinides and Specific Fission Products from High Level Waste

2014 ◽  
Vol 94 ◽  
pp. 103-110 ◽  
Author(s):  
Yue Zhou Wei ◽  
Shun Yan Ning ◽  
Qi Long Wang ◽  
Zi Chen ◽  
Yan Wu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Selective Adsorption ◽  
Packed Column ◽  
Liquid Waste ◽  
Separation Method ◽  
High Level Waste ◽  
Small Scale ◽  
Fast Diffusion ◽  
High Level Liquid Waste ◽  
High Level

The long-term radiotoxicity of high level liquid waste (HLLW) generated in spent nuclear fuel reprocessing is governed by the content of several long-lived minor actinides (MA) and some specific fission product nuclides. To efficiently separate MA (Am, Cm) and some FPs such as Cs and Sr from the HLLW, we have been studying an advanced aqueous partitioning process, which uses selective adsorption as separation method. In this work, we prepared different types of porous silica-based organic/inorganic adsorbents with fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. So they are advantageously applicable to efficient separation of the MA and specific FP elements from HLLW. Adsorption and separation behaviors of the MA and some FP elements such as Cs and Sr were studied. Small scale separation tests using simulated and genuine nuclear waste solutions were carried out and the obtained results indicate that the proposed separation method based on selective adsorption is essentially feasible.

Characterization of Mechanical Strengths for Simulated High-Level Waste Forms

1989 ◽  
Vol 176 ◽  
Author(s):  
Hiroshi Igarashi ◽  
Takeshi Takahashi
Keyword(s):  
Fracture Toughness ◽  
Glass Ceramic ◽  
Bending Strength ◽  
Liquid Waste ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Waste Forms ◽  
High Level ◽  
Mechanical Strengths

ABSTRACTWaste forms have been developed and characterized at PNC (Power Reactor and Nuclear Fuel Development Corporation)to immobilize high-level liquid waste generated from the reprocessing of nuclear spent fuel.Mechanical strength tests were excecuted on simulated solidified highlevel waste forms which were borosilicate glass and diopside glass-ceramic. Commercial glass was tested for comparison. Measured strengths were three-point bending strength,uniaxial compressive strength,impact strength by falling weight method,and Vickers hardness. Fracture toughness and fracture surface energy were also measured by both notch-beam and indentation technique.The results show that mechanical strengths of waste glass form are similar and that the glass ceramic form has the higher fracture toughness.

Adsorption and separation behavior of palladium(II) from simulated high-level liquid waste using a novel silica-based adsorbents

2021 ◽  
Vol 0 (0) ◽  
Author(s):  
Hao Wu ◽  
Naoki Osawa ◽  
Masahiko Kubota ◽  
Seong-Yun Kim
Keyword(s):  
Nitric Acid ◽  
Selective Adsorption ◽  
Packed Column ◽  
Liquid Waste ◽  
Freundlich Isotherm ◽  
Effect Of Temperature ◽  
Adsorption Model ◽  
High Level Liquid Waste ◽  
Adsorption Selectivity ◽  
Successful Recovery

Abstract Aiming at selective adsorption and separation of Pd(II) in nitric acid solution, a hybrid soft N and hard O donor adsorbent (TAMIA-EH+TOA)/SiO2–P (P = Polymer) was successfully synthesized. The adsorption performances of (TAMIA-EH+TOA)/SiO2–P adsorbent towards Pd(II) were systematically investigated as a function of contact time, effect of concentration of nitric acid, effect of temperature etc. Adsorption speed of Pd(II) was fairly fast and can reach equilibrium state within only 0.5 h. The distribution coefficient of Pd(II) was more than 103 when [HNO3] = 0.1. Though it decreased gradually with an increase in the concentration of HNO3, the adsorption selectivity of (TAMIA-EH+TOA)/SiO2–P adsorbent towards Pd(II) was still significant than other co-existing metal ions in the whole HNO3 range from 0.1 to 5 M. The adsorption isotherm of Pd(II) onto (TAMIA-EH+TOA)/SiO2–P adsorbent fitted well with Langmuir adsorption model but Freundlich isotherm model. The calculated results of adsorption thermodynamic parameters indicated that the adsorption process of Pd(II) was exothermic and happened in a natural way. Furthermore, the separation chromatography experiment by utilizing (TAMIA-EH+TOA)/SiO2–P adsorbent packed column was carried out. Based on the results of plotted elution curves, it was found that the successful recovery of Pd(II) (96.27%) was achieved by eluting with thiourea solution.

French Industrial Vitrification Plant: 30 Years Old and Still Innovating

2009 ◽  
Vol 1193 ◽  
Author(s):  
E. Chauvin ◽  
C. Ladirat ◽  
R. Do Quang
Keyword(s):  
Continuous Improvement ◽  
Spent Nuclear Fuel ◽  
Atomic Energy Commission ◽  
Liquid Waste ◽  
Fission Products ◽  
High Level Liquid Waste ◽  
Final Volume ◽  
High Level ◽  
Selection Of

AbstractIn 2008, AREVA NC Industrial Vitrification of High-Level Liquid Waste blows out its 30th candle, with always two main objectives during all the time: containment of the long lived fission products and reduction of the final volume of waste. During all this time AREVA with the French Atomic Energy Commission (CEA) developed and use in their industrial installations a selection of borosilicate glass that have been demonstrated as the most suitable containment matrix for waste from spent nuclear fuel. Consistent and long-term R&D programs associated to industrial feed back from operation have enabled continuous improvement of the process: throughput and waste loading factor enhancement. The Vitrification Process used and currently implemented in the AREVA facilities will be described.

scholarly journals Micro-PIXE analysis on adsorbent of extraction chromatography for MA(III) recovery

2016 ◽  
Vol 26 (03n04) ◽  
pp. 73-83
Author(s):  
Y. Takahatake ◽  
S. Watanabe ◽  
H. Kofuji ◽  
M. Takeuchi ◽  
K. Nomura ◽  
...  
Keyword(s):  
Extraction Chromatography ◽  
Packed Column ◽  
Liquid Waste ◽  
Flow Sheet ◽  
Simple Experiment ◽  
High Level Liquid Waste ◽  
Energy Agency ◽  
Pixe Analysis ◽  
Conducting Research ◽  
High Level

Japan Atomic Energy Agency (JAEA) has been conducting research and development of MA(III) recovery from high level liquid waste (HLLW) by extraction chromatography technology for reduction in amount and environmental impact of radioactive waste. The behavior of adsorbed cations inside the adsorbent packed in a column is necessary to be evaluated for improvement of the adsorbent or flow-sheet to achieve targeted MA(III) recovery performance. In this paper, micro-PIXE analysis was carried out on the particles sampled from various positions of the column to reveal the behavior of cations inside the packed column with CMPO/SiO2-P adsorbent using RE(III) as simulated elements of MA(III). Simple experiment and data analysis were shown to be effective to reveal inside of the column, and formation and transportation of the adsorption bands were observed for some cations which are extractable by the CMPO extractant. Some part of Zr(IV) and Mo(VI) were found to remain inside the column without distinct transportation even after the elution operation. Those results will contribute to design more practical MA(III) recovery flow-sheet.

High Level Liquid Waste Solidification Using a “Cold” Crucible Induction Melter

2000 ◽  
Vol 663 ◽  
Author(s):  
Andrei V. Demine ◽  
Nina V. Krylova ◽  
Pavel P. Polyektov ◽  
Igor N. Shestoperov ◽  
Tatyana V. Smelova ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Liquid Waste ◽  
Cold Crucible ◽  
High Level Liquid Waste ◽  
Two Stage ◽  
One Stage ◽  
Waste Forms ◽  
High Level ◽  
Ceramic Melter

ABSTRACTAt the present time the primary problem in a closed nuclear fuel cycle is the management of high level liquid waste (HLLW) generated by the recovery of uranium and plutonium from spent nuclear fuel. Long-term storage of the HLLW, even in special storage facilities, poses a real threat of ecological accidents. This problem can be solved by incorporating the radioactive waste into solid fixed forms that minimize the potential for biosphere pollution by long-lived radionuclides and ensure ecologically acceptable safe storage, transportation, and disposal. In the present report, the advantages of a two-stage HLLW solidification process using a “cold” crucible induction melter (CCIM) are considered in comparison with a one-stage vitrification process in a ceramic melter.This paper describes the features of a process and equipment for a two-stage HLLW solidification technology using a “cold” crucible induction melter (CCIM) and identifies the advantages compared to a one-stage ceramic melter. A two-stage pilot facility and the technical characteristics of the equipment are described using a once-through evaporator and cold-crucible induction melter currently operational at the IA.Mayak. facility in Ozersk, Russia. The results of pilot-plant tests with simulated HLLW to produce a phosphate glass are described. Features of the new mineral-like waste form matrices synthesized by the CCIM method are also described. Subject to further development, the CCIM technology is planned to be used to solidify all accumulated HLLW at Mayak – first to produce borosilicate glass waste forms and then mineral-like waste forms.

Demonstration of a crown ether process for partitioning strontium from high level liquid waste (HLLW)

2015 ◽  
Vol 0 (0) ◽  
Author(s):  
Jianchen Wang ◽  
Shan Jing ◽  
Jing Chen
Keyword(s):  
Liquid Waste ◽  
Cold Test ◽  
Pilot Scale ◽  
Glass Block ◽  
Small Scale ◽  
High Level Liquid Waste ◽  
Uranium Recovery ◽  
Steel Tanks ◽  
Centrifugal Contactor ◽  
High Level

AbstractChinese HLLW with a higher-salt liquid that was generated via plutonium uranium recovery by extraction (PUREX) processing was temporarily stored in stainless steel tanks and is waiting for treatment. The volume and heat-loading of the glass block are reduced if the strontium, cesium, actinides and other long-life radioactive elements, such as Tc in the HLLW, are partitioned before the HLLW verification. This process is beneficial to preserve the capacity of the geological disposal repository and to minimize long-term hazards. The process of partitioning strontium from Chinese HLLW using Dicyclohexano-18Crown-6(DCH18C-6) was developed in past decades, including such fundamental studies as the small scale cold and hot test. In this work, new studies are introduced, including the cold and the long time hot cascade tests, using a miniature centrifugal contactor set and the pilot-scale cold test using pulse extraction columns. The results indicate that the crown process is promising for partitioning strontium from Chinese HLLW.

scholarly journals Development of MA separation process with TEHDGA/SiO2-P for an advanced reprocessing

2021 ◽  
Vol 330 (1) ◽  
pp. 237-244
Author(s):  
Yusuke Horiuchi ◽  
Sou Watanabe ◽  
Yuichi Sano ◽  
Masayuki Takeuchi ◽  
Fukuka Kida ◽  
...  
Keyword(s):  
Fission Product ◽  
Distribution Ratio ◽  
Extraction Chromatography ◽  
Packed Column ◽  
Liquid Waste ◽  
Separation Process ◽  
Minor Actinide ◽  
High Level Liquid Waste ◽  
Column Separation ◽  
High Level

AbstractApplicability of tetra2-ehylhexyl diglycolamide (TEHDGA) impregnated adsorbent for minor actinide (MA) recovery from high level liquid waste (HLLW) in extraction chromatography technology was investigated through batch-wise adsorption and column separation experiments. Distribution ratio of representative fission product elements were obtained by the batch-wise experiments, and TEHDGA adsorbent was shown to be preferable to TODGA adsorbent for decontamination of several species. All Ln(III) supplied into the TEHDGA adsorbent packed column was properly eluted from the column, and the applicability of the adsorbent was successfully showed by this study.

scholarly journals Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA’s Cold Crucible Melter

2003 ◽  
Author(s):  
R. Do Quang ◽  
V. Petitjean ◽  
F. Hollebecque ◽  
O. Pinet ◽  
T. Flament ◽  
...  
Keyword(s):  
Atomic Energy Commission ◽  
Liquid Waste ◽  
Glass Layer ◽  
Cold Crucible ◽  
High Level Liquid Waste ◽  
Process Waste ◽  
Glass Formulation ◽  
High Level ◽  
Fuel Reprocessing ◽  
High Frequency Induction

The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA’s R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a soldified glass layer that protects the melter’s inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities: the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybednum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs.

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