scholarly journals Preliminary study on geo-mechanical aspects of SSiC canisters

2018 ◽  
Vol 45 ◽  
pp. 63-72 ◽  
Author(s):  
Ya-Nan Zhao ◽  
Heinz Konietzky ◽  
Jürgen Knorr ◽  
Albert Kerber
Keyword(s):  
Mechanical Damage ◽  
Fission Products ◽  
Dynamic Simulations ◽  
Package Design ◽  
Safety Requirements ◽  
First Results ◽  
High Level Nuclear Waste ◽  
Small Model ◽  
High Level ◽  
Potting Compound

Abstract. To meet safety requirements for underground storage of high-level nuclear waste, engineered barriers are an integral part of a modern defense-in-depth concept and therefore have to be considered in interaction with the host rock. This study presents preliminary results for the load behavior of a canister made of pressure-less sintered silicon carbide (SSiC), which forms the main retention barrier for the fission products in a new multi-layer waste package design denominated as TRIPLE C. This means a three-fold enclosure strategy, spreading the functionalities to three different ceramic barriers: first the porous potting compound surrounding each single fuel rod in the container, second the solid container wall of SSiC and third the over-pack of carbon concrete. Besides all the advantages a potential drawback of ceramics in general is their brittleness. Therefore, the behavior of SSiC structural components under static and dynamic loading has to be investigated. First results for a small model canister indicate that static loading will not create any relevant damage, even if stresses are extremely high and highly anisotropic on a canister all-around embedded. First dynamic simulations indicate that, under very unfavorable circumstances, the model canister can experience tensile stresses bigger than its tensile strength. Also, point loading may cause damage to the canister under certain conditions. Based on the performed calculations, the SSiC canister design will be optimized together with the carbon concrete over-pack, so that mechanical damage of main retention barrier can be excluded even under extreme static and dynamic conditions in a final repository.

Waste Package Reliability / Nalysis

1983 ◽  
Vol 26 ◽  
Author(s):  
C. Pescatore ◽  
C. Sastre
Keyword(s):  
Ordinary Sense ◽  
Regulatory Requirements ◽  
Package Design ◽  
Waste Package ◽  
Qualitative And Quantitative ◽  
Future Performance ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Package Reliability ◽  
General Method

ABSTRACTProof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt.

scholarly journals New trends in the reprocessing of spent nuclear fuel. Separation of minor actinides by solvent extraction

2016 ◽  
Vol 71 (1) ◽  
pp. 123
Author(s):  
Jerzy Narbutt
Keyword(s):  
Solvent Extraction ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Natural Uranium ◽  
Heterocyclic Ligands ◽  
High Level Nuclear Waste ◽  
Extraction Processes ◽  
High Level

<p>Recycling of actinides from spent nuclear fuel by their selective separation followed by transmutation in fast reactors will optimize the use of natural uranium resources and minimize the long-term hazard from high-level nuclear waste. This paper describes solvent extraction processes recently developed, aimed at the separation of americium from lanthanide fission products as well as from curium present in the waste. Depicted are novel poly-N-heterocyclic ligands used as selective extractants of actinide ions from nitric acid solutions or as actinide-selective hydrophilic stripping agents.</p>

Corrosion Testing of Stainless Steel-Zirconium:Metal Waste Forms

1999 ◽  
Vol 556 ◽  
Author(s):  
D. P. Abraham ◽  
L. J. Simpson ◽  
M. J. Devries ◽  
S. M. Mcdeavitt
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Waste Forms ◽  
Immersion Corrosion ◽  
Corrosion Studies ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Nuclear Waste Forms

AbstractStainless steel-zirconium (SS-Zr) alloys have been developed as waste forms for the disposal of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste forms incorporate irradiated cladding hulls, components of the alloy fuel, noble metal fission products, and actinide elements. The baseline waste form is a stainless steel- 15 wt% zirconium (SS-15Zr) alloy. This article presents microstructures and some of the corrosion studies being conducted on the waste form alloys. Electrochemical corrosio n, immersion corrosion, and vapor hydration tests have been performed on various alloy compositions to evaluate corrosion behavior and resistance to selective leaching of simulated fission products. The SS-Zr waste forms immobilize and retain fission products very effectively and show potential for acceptance as high-level nuclear waste forms.

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.

High-level nuclear waste management: a geochemical perspective

1989 ◽  
Vol 16 (4) ◽  
pp. 498-503
Author(s):  
T. T. Vandergraaf
Keyword(s):  
Contaminant Transport ◽  
Nuclear Waste ◽  
Fission Products ◽  
Crystalline Rock ◽  
Nuclear Waste Management ◽  
Geochemical Processes ◽  
Fracture Systems ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Canada Limited

Atomic Energy of Canada Limited is investigating the concept of the disposal of high-level radioactive waste in an underground vault in an intrusive crystalline rock formation. The environmental impact of such a disposal is, to a large extent, dictated by geochemical processes involving rock-forming minerals, groundwater, and fission products and actinides in the waste. These various geochemical processes impact on the transport of contaminants, including radionuclides and chemically toxic elements, from a used-fuel disposal vault towards the biosphere. The extent and importance of the geochemical processes on contaminant transport are discussed. The predominant processes controlling the velocity of contaminant transport are the various geochemical interactions of the dissolved contaminant species with the minerals lining the surfaces of conductive fractures and fracture systems. Key words: radionuclide, uranium, nuclear contaminant, transport, sorption, diffusion, geochemistry, fission products, granite.

scholarly journals First results of the 241Am(n,f) cross section measurement at the Experimental Area 2 of the n_TOF facility at CERN

2020 ◽  
Vol 239 ◽  
pp. 05014
Author(s):  
Z. Eleme ◽  
N. Patronis ◽  
A. Stamatopoulos ◽  
A. Tsinganis ◽  
M. Kokkoris ◽  
...  
Keyword(s):  
Cross Section ◽  
Nuclear Waste ◽  
Reaction Cross ◽  
Section Data ◽  
Experimental Area ◽  
First Results ◽  
Sensitivity Studies ◽  
High Level Nuclear Waste ◽  
High Level

Feasibility, design and sensitivity studies on innovative nuclear reactors that could address the issue of nuclear waste transmutation using fuels enriched in minor actinides, require high accuracy cross section data for a variety of neutron-induced reactions from thermal energies to several tens of MeV. The isotope 241Am (T1/2= 433 years) is present in high-level nuclear waste (HLW), representing about 1.8 % of the actinide mass in spent PWR UOx fuel. Its importance increases with cooling time due to additional production from the β-decay of 241Pu with a half-life of 14.3 years. The production rate of 241 Am in conventional reactors, including its further accumulation through the decay of 241Pu and its destruction through transmutation/incineration are very important parameters for the design of any recycling solution. In the present work, the 241 Am(n,f) reaction cross-section was measured using Micromegas detectors at the Experimental Area 2 of the n_TOF facility at CERN. For the measurement, the 235U(n,f) and 238U(n,f) reference reactions were used for the determination of the neutron flux. In the present work an overview of the experimental setup and the adopted data analysis techniques is given along with preliminary results.

Mcc Corrosion Tests at Reference Testing Conditions for A27 Cast steel in Hanford Ground Water

1986 ◽  
Vol 84 ◽  
Author(s):  
M.D. Merz ◽  
F. Gerber ◽  
R. Wang
Keyword(s):  
Pacific Northwest ◽  
Pressure Vessel ◽  
Static Pressure ◽  
Cast Steel ◽  
Statistical Treatment ◽  
Testing Conditions ◽  
Corrosion Tests ◽  
Corrosion Rates ◽  
High Level Nuclear Waste ◽  
High Level

AbstractThe Materials Characterization Center (MCC) at Pacific Northwest Lab- oratory is performing three kinds of corrosion tests for the Basalt Waste Isolation Project (BWIP) to establish the interlaboratory reproducibility and uncertainty of corrosion rates of container materials for high-level nuclear waste. The three types of corrosion tests were selected to address two distinct conditions that are expected in a repository constructed in basalt. An air/steam test is designed to address corrosion during the operational period and static pressure vessel and flowby tests are designed to address corrosion under conditions that bound the condi ring the post-closure period of the repository.The results of tests at reference testing conditions, which were defined to facilitate interlaboratory comparison of data, are presented. Data are reported for the BWIP/MCC-105.5 Air/Steam Test, BWIP/MCC-105.1 Static Pressure Vessel, and BWIP/MC-105.4 Flowby Test. In those cases where data are available from a second laboratory, a statistical analysis of interlaboratory results is reported and expected confidence intervals for mean corrosion rates are given. Other statistical treatment of data include analyses of the effects of vessel-to-vessel variations, test capsule variations for the flowby test, and oven-to-oven variations for air/steam tests.

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.

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