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.

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.

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.

A Comparison of the Performance of Nuclear Waste Glasses by Modeling

1987 ◽  
Vol 112 ◽  
Author(s):  
B. Grambow ◽  
D. M. Strachan
Keyword(s):  
Nuclear Waste ◽  
Nuclear Reactors ◽  
Liquid Waste ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Geologic Repository ◽  
Waste Forms ◽  
The Earth ◽  
Soluble Solid ◽  
High Level

The reprocessing of spent fuel from nuclear reactors and processing of fuels for defense purposes have generated large volumes of high-level liquid waste that need to be immobilized prior to final storage. For immobilization, the wastes must be converted to a less soluble solid, and, although other waste forms exist, glass currently appears to be the choice for the transuranic-containing portion of the reprocessed waste. Once produced, this glass will be sent in canisters to a geologic repository located some 200 to 500 m below the surface of the earth.

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.

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.

Materials Science of Radioactive Waste Forms

MRS Bulletin ◽  
1994 ◽  
Vol 19 (12) ◽  
pp. 16-19 ◽  
Author(s):  
R.C. Ewing ◽  
W. Lutze
Keyword(s):  
New York ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Nuclear Weapons ◽  
Spent Nuclear Fuel ◽  
Materials Science ◽  
Waste Forms ◽  
Mill Tailings ◽  
High Level

The materials science of radioactive waste forms and containment materials has long been a subject of interest to the Materials Research Society. One of the earliest (and continuing) MRS symposia, the Scientific Basis for Nuclear Waste Management, has been held 18 times since 1978. This symposium rotates abroad every third year: Berlin in 1982, Stockholm in 1985, Berlin in 1988, Strasbourg in 1991, and Kyoto this past October. Nearly 170 papers were presented at the Kyoto meeting.Materials science issues for nuclear waste disposal are unique in their scale and consequences. The wastes include an extremely wide variety of materials: spent nuclear fuel from commercial and research reactors; high-level liquid waste produced at West Valley, New York, during the reprocessing of commercial spent nuclear fuel; high-level waste (HLW) generated by the nuclear weapons program; nearly pure plutonium from the dismantling of nuclear weapons; highly enriched uranium from weapons; low-level, medium-level, and mixed waste from laboratories and medical facilities; and, finally, mill tailings from uranium mines and the residues from chemical processing, such as the radium-bearing filtrate presently in storage at Fernald, Ohio, and Niagara Falls, New York. Some material can be simply stabilized and monitored in situ, as is done for most uranium mill tailings and residues, but other materials require retrieval, processing, immobilization, and permanent disposal. The volumes of material that will require handling, immobilization, and disposal are enormous. In the United States, much of the weapons program waste is stored in tanks at Hanford, Washington and Savannah River, South Carolina.

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