Synroc: A Suitable Waste Form for Actinides

MRS Bulletin ◽  
1994 ◽  
Vol 19 (12) ◽  
pp. 28-32 ◽  
Author(s):  
E.R. Vance
Keyword(s):  
Nuclear Power ◽  
Power Reactor ◽  
Liquid Waste ◽  
Fission Products ◽  
Nuclear Power Reactor ◽  
Process Contaminants ◽  
High Level Nuclear Waste ◽  
Reactive Mixture ◽  
High Level ◽  
Fuel Reprocessing

Synroc, a ceramic made from a reactive mixture of Al, Ba, Ca, Ti, and Zr oxides, is proving to be a suitable and effective medium for immobilizing nuclear wastes.Synroc-C, a titanate-based ceramic variant, was initially developed in 1978 by Ringwood et al. for immobilizing high-level nuclear waste (HLW) from nuclear power reactor fuel reprocessing. HLW is essentially a solution of radioactive fission products, actinides, and process contaminants in ~3 mol/L nitric acid. The developers of Synroc-C aimed to immobilize radioactive waste ions by incorporating them in a ceramic. They accomplished this by mixing the HLW solution (liquid waste) with a ceramic precursor, then forming the ceramic by drying, calcining, and hot-pressing the mixture in a metal container for two hours at 1200°C/20 MPa. The result, Synroc-C, is composed of hol-landite, zirconolite, perovskite, and rutile, together with a few percent of minor phases and metal alloys. The Synroc-C precursor has the following composition (wt%): Al203(5.4); BaO(5.6); CaO(11); TiO2(71.4); and ZrO2(6.6). Since 1984, it has been prepared by hydrolyzing a mixture of Al, Ti, and Zr alkoxides with an aqueous slurry of Ba and Ca hydroxide. The abundances of the phases, and the radionuclides contained in them in dilute solid solution, are identified in Table I.

The Application Research of Concentration & Denitration Technology in Advanced Reprocessing Process

2012 ◽  
Vol 560-561 ◽  
pp. 637-643
Author(s):  
Yong Li ◽  
Xue Gang Liu ◽  
Jin Chen
Keyword(s):  
Nitric Acid ◽  
Nuclear Power ◽  
Liquid Waste ◽  
Fission Products ◽  
Extraction Process ◽  
Purex Process ◽  
Current Status ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
High Level

The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. While conventional reprocessing process, PUREX process, was successful to recover uranium and plutonium, in recent years some countries have turned to focus on advanced reprocessing process, which features of partitioning of minor actinides (MA) and long-lived fission products(LLFP). Most advanced reprocessing processes under development involve new extractants and additional extraction cycles. In China, TRPO extraction process has been developed to partition MA/LLFP from high-level liquid waste(HLLW) since early 1980’s. In parallel to R&D work on separation technologies, studies on concentration & denitration process have been evolved to prepare feed solutions to suit qualifications of extraction. Industrially, concentration & denitration is the internationally recognized standard to treat HLLW released from PUREX before vitrification. It enables to minimize the volume of interim storage, to restrain the corrosion of storage tank, to recover nitric acid in HLLW and to reduce the required evaporation duty of the vitrification process. Generally, the constitution of concentrated HLLW has little impact on the following vitrification process. But when concentration & denitration acts as pretreatment process of partitioning, the composition of actinides, fission products, and nitric acid in concentrated HLLW solution plays significant role in extraction process. A series of technical issues relevant to the connection between concentration ﹠denitration and extractions should be solved. This paper describes current status of concentration & denitration technology utilized in industry and under reprocessing plants. The specific separation requirements in advanced reprocessing process and challenges to apply concentration & denitration process are addressed. Besides, concentration & denitration process was tested in laboratory to adjust feed solutions for TRPO and Cyanex301 partitioning. Results demonstrate its promising prospect in advanced reprocessing process.

TRLIFS study of Eu(III) spectroscopic properties to obtain structural and thermodynamic informations on lanthanide-malonamide complexes in the Eu(III)/NaNO3/TetraEthylMalonAmide system

2004 ◽  
Vol 92 (7) ◽  
Author(s):  
Laurent Couston ◽  
M. C. Charbonnel ◽  
J. L. Flandin ◽  
Christophe Moulin ◽  
F. Rancier
Keyword(s):  
Liquid Waste ◽  
Fission Products ◽  
Spectroscopic Properties ◽  
High Level Liquid Waste ◽  
Trivalent Actinides ◽  
Nuclear Fuel Reprocessing ◽  
Chemical Studies ◽  
High Level ◽  
Fuel Reprocessing ◽  
The Eu

SummaryImprovement of the nuclear fuel reprocessing involves separating the minor actinides (Am(III) and Cm(III)) from the fission products. In the French strategy, the first step consists in the separation of the trivalent actinides and lanthanides from high-level liquid waste, for which malonamides RR´NCO(CHR´´)CONRR´ are promising ligands. These molecules have been optimized for reprocessing but still require basic chemical studies to describe the complexation mechanisms at a molecular scale. This paper discusses a thermodynamic and structural study of a Ln(III)-malonamide complex formed with the hydrosoluble tetraethylmalonamide ligand (TEMA=(C

SANICRO 71

Alloy Digest ◽  
1965 ◽  
Vol 14 (12) ◽  
Keyword(s):  
Nuclear Power ◽  
Elevated Temperatures ◽  
Power Reactor ◽  
Base Alloy ◽  
Heat Treating ◽  
Good Resistance ◽  
Nuclear Power Reactor ◽  
Temperature Performance ◽  
Resistance To Stress ◽  
Nickel Base

Abstract Sanicro 71 is a nickel-base alloy having good resistance to stress-corrosion, oxidation and creep at elevated temperatures. It is recommended for nuclear power reactor heat exchanger tubes, aircraft turbojet engines and for equipment in the textile, plastic, and chemical industries. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties. It also includes information on high temperature performance and corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: Ni-108. Producer or source: Sandvik.

Radiation Effects on Hollandite Ceramics developed for Radioactive Cesium Immobilization

2003 ◽  
Vol 792 ◽  
Author(s):  
V. Aubin ◽  
D. Caurant ◽  
D. Gourier ◽  
N. Baffier ◽  
S. Esnouf ◽  
...  
Keyword(s):  
Radiation Effects ◽  
Liquid Waste ◽  
Fission Products ◽  
Radioactive Cesium ◽  
Radioactive Liquid Waste ◽  
Epr Spectrum ◽  
Paramagnetic Resonance ◽  
Γ Radiation ◽  
The Stability ◽  
High Level

ABSTRACTProgress on separating the long-lived fission products from the high level radioactive liquid waste (HLW) has led to the development of specific host matrices, notably for the immobilization of cesium. Hollandite (nominally BaAl2Ti6O16), one of the main phases constituting Synroc, receives renewed interest as specific Cs-host wasteform. The radioactive cesium isotopes consist of short-lived Cs and Cs of high activities and Cs with long lifetime, all decaying according to Cs+→Ba2++e- (β) + γ. Therefore, Cs-host forms must be both heat and (β,γ)-radiation resistant. The purpose of this study is to estimate the stability of single phase hollandite under external β and γ radiation, simulating the decay of Cs. A hollandite ceramic of simple composition (Ba1.16Al2.32Ti5.68O16) was essentially irradiated by 1 and 2.5 MeV electrons with different fluences to simulate the β particles emitted by cesium. The generation of point defects was then followed by Electron Paramagnetic Resonance (EPR). All these electron irradiations generated defects of the same nature (oxygen centers and Ti3+ ions) but in different proportions varying with electron energy and fluence. The annealing of irradiated samples lead to the disappearance of the latter defects but gave rise to two other types of defects (aggregates of light elements and titanyl ions). It is necessary to heat at relatively high temperature (T=800°C) to recover an EPR spectrum similar to that of the pristine material. The stability of hollandite phase under radioactive cesium irradiation during the waste storage is discussed.

Studies on hydrolysis and radiolysis of N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide

2002 ◽  
Vol 90 (3) ◽  
Author(s):  
Y. Sugo ◽  
Y. Sasaki ◽  
S. Tachimori
Keyword(s):  
Gamma Irradiation ◽  
Room Temperature ◽  
Degradation Products ◽  
Liquid Waste ◽  
Amide Bond ◽  
Radioactive Liquid Waste ◽  
Ether Bond ◽  
Nuclear Fuel Reprocessing ◽  
High Level ◽  
Fuel Reprocessing

SummaryHydrolytic and radiolytic stabilities of a promising extractant, N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide (TODGA), for actinides in high-level radioactive liquid waste from nuclear fuel reprocessing were investigated in air at room temperature. Hydrolysis by nitric acid was not observed, whereas radiolysis by gamma irradiation was notably observed. The radiolysis study showed that an amide-bond, an ether-bond, and a bond adjacent to the ether-bond tended to be broken by gamma irradiation, and dioctylamine and various N,N-dioctylmonoamides were identified as the main degradation products by GC/MS and NMR analyses. The

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.

Visualization of Supercritical Carbondioxide Pseudo-Boiling Inside a Rectangular Channel

2001 ◽  
Author(s):  
K. Sakurai ◽  
H. S. Ko ◽  
K. Okamoto ◽  
H. Madarame
Keyword(s):  
Forced Convection ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Power Reactor ◽  
Rectangular Channel ◽  
Visual Observation ◽  
Next Generation ◽  
Nuclear Power Reactor

Abstract The characteristics of the supercritical fluids should be precisely investigated for the next generation nuclear power reactor, i.e., Super-critical water Cooled Reactor (SCR). There are few experiments for visual observation especially in forced convection, because of the difficulty of the experiment.

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