Different methods to synthesize sodalite, as a matrix for conditioning chloride spent salts from pyroprocesses

2008 ◽  
Vol 96 (4-5) ◽  
Author(s):  
G. De Angelis ◽  
R. Nannicini ◽  
F. Martini ◽  
C. Mazzocchia ◽  
G. Modica
Keyword(s):  
Spent Nuclear Fuel ◽  
Direct Synthesis ◽  
Fission Products ◽  
Sodium Aluminate ◽  
Chloride Salt ◽  
Naturally Occurring ◽  
Zeolite 4A ◽  
Waste Salt ◽  
Pyrometallurgical Processing ◽  
Salt Waste

The pyrometallurgical processing of spent nuclear fuel generates a chloride salt waste containing alkali-metal, alkaline-earth, and some rare-earth fission products. Sodalite, a naturally occurring mineral containing chlorine, has been investigated as an immobilization matrix for this salt waste. To this end different routes for the synthesis of sodalite have been followed: on one hand, direct synthesis from kaolinite, metakaolinite, or from silica and sodium aluminate have been carried out; on the other hand, a synthesis from Zeolite 4A used for preliminary decontamination of the salt by ion-exchange has been performed. The former allows to condition the waste salt as a whole, whenever discarding an entire process salt batch becomes necessary; this is the case when an electrorefiner plant has to be decommissioned, or in the event of a severe process upset; the latter is more suitable for routine operations, which better require the clean-up of the salt and its recycle to the electrorefiner, thus avoiding the production of huge quantities of solidified wastes to be disposed of.

A New Matrix for Conditioning Chloride Salt Wastes from the Electrorefining of Spent Nuclear Fuel

2014 ◽  
Vol 94 ◽  
pp. 97-102 ◽  
Author(s):  
Giorgio de Angelis ◽  
Mauro Capone ◽  
Carlo Fedeli ◽  
Giuseppe A. Marzo ◽  
Mario Mariani ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Sol Gel ◽  
Fission Products ◽  
Nuclear Fission ◽  
Chloride Salt ◽  
The European Union ◽  
Metal Chlorides ◽  
Sol Gel Process ◽  
Waste Salt

A novel method proposed by Korea Atomic Energy Research Institute has been applied to the treatment of chloride salt wastes coming from electrorefining of spent nuclear fuel, which allows to separate uranium from fission products. It is based on a matrix, SAP (SiO2-Al2O3-P2O5), synthesized by a conventional sol-gel process, able to stabilize the volatile salt wastes due to the formation of metalaluminosilicates, metalaluminophosphates and metalphosphates. With this method a higher disposal efficiency and a lower waste volume can be obtained. Eutectic melt LiCl-KCl (59-41 mol%) has been used to simulate the waste salt. The composite SAP has been prepared by using tetraethyl ortosilicate (TEOS), aluminum chloride (AlCl3.6H2O) and phosphoric acid (H3PO4) as sources of Si, Al, and P, respectively. All reagents were dissolved in EtOH/H2O and the mixture, tightly sealed, was placed in an electric oven at 70 C. After a gelling/ageing for 3 days, the transparent hydrogels were dried at 110 C for 3 days and then thermally treated at 600 C for 2 hours. The final product (SAP) was reacted with metal chlorides at increasing temperatures for 20 hours inside an Argon-atmosphere glove-box, after mixing them at a SAP/metal chloride mixing ratio of 2. The obtained products have been characterized by means of density measurements, scanning electron microscopy, thermal analysis, as well as by XRD, FTIR and Raman spectra. Financial support from the Nuclear Fission Safety Program of the European Union (project SACSESS, contract FP7-CP-2012-323282) is gratefully acknowledged.

Pressureless Sintering of Sodalite Waste-forms for the Immobilization of Pyroprocessing Wastes

2015 ◽  
Vol 1744 ◽  
pp. 61-66 ◽  
Author(s):  
M. R. Gilbert
Keyword(s):  
Microstructural Analysis ◽  
Chloride Salt ◽  
Sintering Aid ◽  
Aluminophosphate Glass ◽  
Waste Forms ◽  
Naturally Occurring ◽  
Immobilization Matrix ◽  
Pre Treatment ◽  
Waste Salt ◽  
Pyrochemical Reprocessing

ABSTRACTSodalite (Na8[AlSiO4]6Cl2), a naturally occurring Cl-containing mineral, has long been regarded as a potential immobilization matrix for the chloride salt wastes arising from pyrochemical reprocessing operations, as it allows for the conditioning of the waste salt as a whole without the need for any pre-treatment. Here the consolidation and densification of Sm-doped sodalite (as an analogue for AnCl3) has been investigated with the aim of producing fully dense (i.e. > 95 % t.d.) ceramic monoliths via conventional cold-press-and-sinter techniques at temperatures of < 1000 °C. Microstructural analysis of pressed and sintered sodalite powders under these conditions is shown to produce poorly sintered, porous, inhomogeneous pellets. However, by the addition of a sodium aluminophosphate glass sintering aid, fully dense Sm-sodalite ceramic monoliths can successfully be produced by sintering at temperatures as low as 800 °C.

Consolidating Electrorefined Spent Nuclear Fuel Waste: Analysis and Experiment

2003 ◽  
Author(s):  
K. J. Bateman ◽  
D. D. Capson
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Processing Conditions ◽  
Production Scale ◽  
Ceramic Waste ◽  
Waste Salt ◽  
Nuclear Fuel Waste

Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in the electrometallurical treatment of spent EBR-II fuel. The ceramic waste process culminates with an elevated temperature operation. The processing conditions used by the furnace, for demonstration scale and production scale operations, are to be developed at Argonne National Laboratory. To assist in selecting the processing conditions of the furnace and to reduce the number of costly experiments, a finite difference model was developed to predict the consolidation of the ceramic waste. The model accurately predicted the heating as well as the bulk density of the ceramic waste form. The methodology used to develop the computer model and a comparison of the analysis to experimental data is presented.

scholarly journals Hot Isostatic Pressing of Ceramic Waste From Spent Nuclear Fuel

2002 ◽  
Author(s):  
Kenneth J. Bateman ◽  
Richard H. Rigg ◽  
James D. Wiest
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Hot Isostatic Pressing ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Spent Fuel ◽  
Isostatic Pressing ◽  
Ceramic Waste ◽  
Pressing Operation ◽  
Waste Salt

Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process.

Salt waste treatment from a LiCl-KCl based pyrochemical spent fuel treatment process

2008 ◽  
Vol 96 (4-5) ◽  
Author(s):  
Mike T. Harrison ◽  
Howard E. Simms ◽  
Angela Jackson ◽  
Robert G. Lewin
Keyword(s):  
Alkali Metals ◽  
Treatment Process ◽  
Spent Nuclear Fuel ◽  
Waste Treatment ◽  
Electrochemical Techniques ◽  
Fission Products ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level ◽  
Salt Waste

Spent nuclear fuel may be treated using molten salt electrochemical techniques to separate fission products and actinide metals. Salt waste arising from the electrorefining process contains alkali metals, alkaline-earth and rare earth fission products, along with residual actinides. The removal of fission product elements has been investigated using zeolite ion exchange and phosphate precipitation, which allow the salt electrolyte to be recycled back into the main electrorefining vessel. Recycling the salt minimizes the volume of high level waste (HLW) generated and yields the fission products in a form more amenable to immobilization in a final disposal matrix. Several sets of experiments have been completed, all of which have significant implications for the use of these techniques on an industrial scale, as well as their ability to clean up the salt, and potentially produce robust and durable waste forms.

Conceptual Study of Neutron Physics of Nuclear Fuel Cycle for Ceramic Fast Reactor

2021 ◽  
Author(s):  
Xuesong Yan ◽  
Yaling Zhang ◽  
Yucui Gao ◽  
Lei Yang
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Fast Reactor ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
Nuclear Fuel Cycle ◽  
Purification Method ◽  
Neutron Physics ◽  
Conceptual Study

Abstract To make the nuclear fuel cycle more economical and convenient, as well as prevent nuclear proliferation, the conceptual study of a simple high-temperature dry reprocessing of spent nuclear fuel (SNF) for a ceramic fast reactor is proposed in this paper. This simple high-temperature dry (HT-dry) reprocessing includes the Atomics International Reduction Oxidation (AIROX) process and purification method for rare-earth elements. After removing the part of fission products from SNF by a HT-dry reprocessing without fine separation, the remaining nuclides and some uranium are fabricated into fresh fuel which can be used back to the ceramic fast reactor. Based on the ceramic coolant fast reactor, we studied neutron physics of nuclear fuel cycle which consists operation of ceramic reactor, removing part of fission products from SNF and preparation of fresh fuels for many time. The parameters of the study include effective multiplication factor (Keff), beam density, and nuclide mass for different ways to remove the fission products from SNF. With the increase in burnup time, the trend of increasing 239Pu gradually slows down, and the trend of 235U gradually decreases and become balanced. For multiple removal of part of fission products in the nuclear fuel cycle, the higher the removal, the larger the initial Keff.

TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy

2000 ◽  
Vol 6 (S2) ◽  
pp. 368-369
Author(s):  
N.L. Dietz ◽  
D.D Keiser
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Metallic Sodium ◽  
Geologic Repository ◽  
Tem Analysis ◽  
Waste Forms

Argonne National Laboratory has developed an electrometallurgical treatment process for metallic spent nuclear fuel from the Experimental Breeder Reactor-II. This process stabilizes metallic sodium and separates usable uranium from fission products and transuranic elements that are contained in the fuel. The fission products and other waste constituents are placed into two waste forms: a ceramic waste form that contains the transuranic elements and active fission products such as Cs, Sr, I and the rare earth elements, and a metal alloy waste form composed primarily of stainless steel (SS), from claddings hulls and reactor hardware, and ∼15 wt.% Zr (from the U-Zr and U-Pu-Zr alloy fuels). The metal waste form (MWF) also contains noble metal fission products (Tc, Nb, Ru, Rh, Te, Ag, Pd, Mo) and minor amounts of actinides. Both waste forms are intended for eventual disposal in a geologic repository.

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