Revised Eh-pH Diagrams (25 C, One Bar) for Uranium and Transuranic Elements: Application to Radioactive Waste Studies

1988 ◽  
Vol 125 ◽  
Author(s):  
D. G. Brookins
Keyword(s):  
Radioactive Waste ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Fuel Rods ◽  
Major Phase ◽  
Geochemical Behavior ◽  
Wide Range ◽  
Carbonate Complexes ◽  
Generic System ◽  
Very High

ABSTRACTEh-pH diagrams are useful for comparing the probable geochemical behavior of many elements including U and the transuranics Np, Pu, Am. Revised Eh-pH diagrams have been constructed for U, Np, Pu and Am at 25 C, 1 bar conditions. These diagrams, based on new and revised thermodynamic data, include aqueous and solid species in the generic system M-C-O-H. The species M(OH)− is considered herein, although it's existence has not been verified.Aqueous U(VI) species are dominated by below pH 5 and by carbonate complexes at higher pH. U(IV) species include important fields of UO2, and U(OH)4. If Si is present, USiO4 may be important. U3O8, occurs between the U(IV) and U(VI) species.Np(IV) solid species are problematic. NpO2 covers a wide range of Eh-pH, but its precursors, NpO. (OH)2 and Np(OH)4 cover more restrictive Eh-pH ranges with more important. This increases potential for radionuclide migration. Np(V, VI) carbonate complexes are important only at high Eh.Pu(IV) species are dominated by PuO2, but Pu(OH)4 does not occupy as much Eh-pH space, being replaced in large part by . With carbonate present, Pu(III) species also replace part of the Pu(OH)4 field, and Pu2 (CO3)3 becomes important. Pu(VI) carbonate complexess occur only at high Eh.Am(III) species dominate Eh-pH space in the system Am-C-O-H, with Am2 (CO3)3 a major phase. Under carbonate-poor conditions, Am(OH) 3 is important. AmO2 occurs at high Eh, pH, and at high Eh and very high pH.The importance of the M(OH)4 precursors to MO2 species is the enhanced possibility of radionuclide migration from buried spent fuel rods, even under reducing condtions.

Dry Storage of Spent Research Reactor Fuel in Castor BR3® Casks at Belgoprocess in Belgium

2001 ◽  
Author(s):  
Marnix Braeckeveldt ◽  
Luc Ooms ◽  
Gustaaf Geenen
Keyword(s):  
Radioactive Waste ◽  
Nuclear Research ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage ◽  
Nuclear Site ◽  
Test Reactor ◽  
Intermediate Storage ◽  
High Level

Abstract The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd2O3) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods. Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated. Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3® cask is designed for transport and the intermediate storage of at least 50 years. Although a defect of the leaktightness of a CASTOR BR3® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.

Near field evolution of a spent fuel repository in an argillaceous rock formation and impact on radionuclide migration

2020 ◽  
Author(s):  
Vanessa Montoya ◽  
Orlando Silva ◽  
Emilie Coene ◽  
Jorge Molinero ◽  
Renchao Lu ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Spent Fuel ◽  
Exchange Reactions ◽  
Radionuclide Migration ◽  
High Level

<p>In August 2015, the German government approved the national programme for the responsible and safe management of spent nuclear fuel (SNF) and radioactive waste proposed by the Federal Ministry for the Environment, Nature Conservation, Building and Reactor Safety (BMU). The assumption is that about ~ 1 100 storage casks (10 500 tons of heavy metal) in the form of spent fuel assemblies will be generated in nuclear power plants and will have to be disposed. However, a decision on the disposal concept for high-level waste is pending and an appropriate solution has to be developed with a balance in multiple aspects. All potential types of host rocks, clay and salt stones as well as crystalline formations are under consideration. In the decision process, evaluation of the risk of different waste management options and scenarios play an enormous role in the discussion. Coupled physical and chemical processes taking place within the engineered barrier system of a repository for high-level radioactive waste will define the radionuclide mobility/retention and the possible radiological impact. The objective of this work is to assess coupled processes occurring in the near-field of a generic repository for spent nuclear fuel in a high saline clay host rock, integrating complex geochemical processes at centimetre-scale. The scenario considers that radionuclides can be released during a period of thousands of years after full saturation of the bentonite barrier and the thermal phase.</p><p>Transport parameters and the discretization of the system, are implemented in a 2D axisymmetric geometry. The multi-barrier system is emplaced in clay and a solubility limited source term for the selected radionuclides is assumed. Kinetics and chemical equilibria reactions are simulated using parameters obtained from experiments. Additionally, porosity changes due to mineral precipitation/dissolution and feedback on the effective diffusion coefficient are taken into account. Protonation/deprotonation, ion exchange reactions and radionuclide inner-sphere sorption is considered.</p><p>Numerical simulations show, that, when the canister corrosion starts, the redox potential decreases, magnetite precipitates and H<sub>2</sub> is formed. Furthermore, the aqueous concentration of Fe(II) increases due to the presence of magnetite. By considering binding to montmorillonite via ion exchange reactions, the bentonite acts as a sink for Fe(II). Additionally, magnetite forms a chemical barrier offering significant sorption capacity for many radionuclides. Finally, a decrease of porosity in the bentonite/canister interface leads to a further deceleration of radionuclide migration. Due to the complexity of reactive transport processes in saline environments, benchmarking of reactive transport models (RTM) is important also to build confidence in those modelling approaches. Development of RTM benchmark procedures is part of the iCROSS project (Integrity of nuclear waste repository systems - Cross-scale system understanding and analysis) funded by both the Helmholtz Association and the Federal Ministry of Education and Research (BMBF).</p><p> </p>

scholarly journals Structural integrity investigations of spent nuclear fuel with finite element modeling

2021 ◽  
Vol 1 ◽  
pp. 13-14
Author(s):  
Efstathios Vlassopoulos ◽  
Susanne Pudollek ◽  
Olympios Alifieris ◽  
Dimitrios Papaioannou ◽  
Ramil Nasyrow ◽  
...  
Keyword(s):  
Finite Element ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Structural Integrity ◽  
Mechanical Response ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Final Disposal ◽  
National Cooperative

Abstract. Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

scholarly journals Thermal dimensioning to determine acceptable waste package loading and spatial configurations of heat-generating waste packages

2015 ◽  
Vol 79 (6) ◽  
pp. 1625-1632 ◽  
Author(s):  
Simon Myers ◽  
David Holton ◽  
Andrew Hoch
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
High Heat ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Waste Package ◽  
Thermal Limits ◽  
Disposal Facility ◽  
Wide Range ◽  
Spatial Configurations

AbstractHeat-generating waste provides a number of additional technical challenges over and above those associated with the disposal of ILW. A priority area of work for Radioactive Waste Management (RWM) concerns the effect of heat on the engineered barrier system, and how this may be mitigated through the management of heat (thermal dimensioning) in a UK Geological Disposal Facility (GDF). The objective of thermal dimensioning is to provide a strategy to enable acceptable waste package loading and spatial configurations of the packages to be determined in order to enable high-heat generating waste to be successfully disposed in a GDF. An early focus of the work has been to develop a thermal modelling tool to support analyses of different combinations of package assumptions and other GDF factors, such as spacing of those packages, to assess the compliance with thermal limits. The approach has a capability to investigate quickly and efficiently the implications of a wide range of disposal concepts for the storage of spent fuel/HLW and the dimensions of a GDF. This study describes the approach taken to undertaking this work, which has included a robust appraisal of the key data (and the associated uncertainty); recent thermal dimensioning analysis has been performed to identify constraints on those disposal concepts.

Implementation of a Geological Disposal Facility (GDF) in the UK by the NDA Radioactive Waste Management Directorate (RWMD): The Potential for Interaction Between the Co-Located ILW/LLW and HLW/SF Components of a GDF

2009 ◽  
Author(s):  
George Towler ◽  
Tim Hicks ◽  
Sarah Watson ◽  
Simon Norris
Keyword(s):  
Radioactive Waste ◽  
White Paper ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Disposal Facility ◽  
Wide Range ◽  
The Uk ◽  
Host Rocks ◽  
System Designs

In June 2008 the UK government published a ‘White Paper’ as part of the “Managing Radioactive Waste Safety” (MRWS) programme to provide a framework for managing higher activity radioactive wastes in the long-term through geological disposal. The White Paper identifies that there are benefits to disposing all of the UK’s higher activity wastes (Low and Intermediate Level Waste (LLW and ILW), High Level Waste (HLW), Spent Fuel (SF), Uranium (U) and Plutonium (Pu)) at the same site, and this is currently the preferred option. It also notes that research will be required to support the detailed design and safety assessment in relation to any potentially detrimental interactions between the different modules. Different disposal system designs and associated Engineered Barrier Systems (EBS) will be required for these different waste types, i.e. ILW/LLW and HLW/SF. If declared as waste U would be disposed as ILW and Pu as HLW/SF. The Geological Disposal Facility (GDF) would therefore comprise two co-located modules (respectively for ILW/LLW and HLW/SF). This paper presents an overview of a study undertaken to assess the implications of co-location by identifying the key Thermo-Hydro-Mechanical-Chemical (THMC) interactions that might occur during both the operational and post-closure phases, and their consequences for GDF design, performance and safety. The MRWS programme is currently seeking expressions of interest from communities to host a GDF. Therefore, the study was required to consider a wide range of potential GDF host rocks and consistent, conceptual disposal system designs. Two example disposal concepts (i.e. combinations of host rock, GDF design including wasteform and layout, etc.) were carried forward for detailed assessment and a third for qualitative analysis. Dimensional and 1D analyses were used to identify the key interactions, and 3D models were used to investigate selected interactions in more detail. The results of this study show that it is possible for ILW/LLW and HLW/SF modules to be co-located without compromising key safety functions of different barrier components, and this reflects international precedents, e.g. the Andra and Nagra repository designs. There are two key technical issues that need to be managed in designing the geometry of the co-located GDF: (i) the heat flux from the HLW/SF module interacting with the ILW/LLW module, and (ii) the potential for development of an alkaline plume from the ILW/LLW module interacting with the HLW/SF module; particularly within fractured host rocks.

TEM study of boron additions to cobalt aluminide

1988 ◽  
Vol 46 ◽  
pp. 546-547
Author(s):  
Gerald B. Feldewerth
Keyword(s):  
High Temperature ◽  
Turbine Engines ◽  
Major Drawback ◽  
University Of Missouri ◽  
Specific Strength ◽  
Aerospace Applications ◽  
Wide Range ◽  
Ordered Alloys ◽  
The University ◽  
Very High

In recent years an increasing emphasis has been placed on the study of high temperature intermetallic compounds for possible aerospace applications. One group of interest is the B2 aiuminides. This group of intermetaliics has a very high melting temperature, good high temperature, and excellent specific strength. These qualities make it a candidate for applications such as turbine engines. The B2 aiuminides exist over a wide range of compositions and also have a large solubility for third element substitutional additions, which may allow alloying additions to overcome their major drawback, their brittle nature.One B2 aluminide currently being studied is cobalt aluminide. Optical microscopy of CoAl alloys produced at the University of Missouri-Rolla showed a dramatic decrease in the grain size which affects the yield strength and flow stress of long range ordered alloys, and a change in the grain shape with the addition of 0.5 % boron.

Features of the Christian values ​​system

2004 ◽  
pp. 21-29
Author(s):  
G.V. Pyrog
Keyword(s):  
Public Opinion ◽  
Religious Beliefs ◽  
Religious Values ◽  
Direct Influence ◽  
Moral Norms ◽  
Value System ◽  
Christian Religion ◽  
Mass Consciousness ◽  
Wide Range ◽  
Very High

In domestic scientific and public opinion, interest in religion as a new worldview paradigm is very high. Today's attention to the Christian religion in our society is connected, in our opinion, with the specificity of its value system, which distinguishes it from other forms of consciousness: the idea of ​​God, the absolute, the eternity of moral norms. That is why its historical forms do not receive accurate characteristics and do not matter in the mass consciousness. Modern religious beliefs do not always arise as a result of the direct influence of church preaching. The emerging religious values ​​are absorbed in a wide range of philosophical, artistic, ethical ideas, acting as a compensation for what is generally defined as spirituality. At the same time, the appeal to Christian values ​​became very popular.

DURCOMET 100

Alloy Digest ◽  
1993 ◽  
Vol 42 (2) ◽  
Keyword(s):  
Fracture Toughness ◽  
Wear Resistance ◽  
Corrosion Resistance ◽  
Mechanical Strength ◽  
Heat Treating ◽  
Corrosion And Wear ◽  
Annealed Condition ◽  
Excellent Corrosion Resistance ◽  
Wide Range ◽  
Very High

Abstract Durcomet 100 is an improved version of Alloy CD-4 MCu with better corrosion and wear resistance. The alloy is used in the annealed condition and possesses excellent corrosion resistance over a wide range of corrosion environments. Mechanical strength is also very high. This datasheet provides information on composition, physical properties, hardness, and tensile properties as well as fracture toughness. It also includes information on corrosion resistance as well as heat treating and joining. Filing Code: SS-540. Producer or source: Duriron Company Inc.

Influencia de las condiciones de secado solar en la coloración de plantas medicinales

2019 ◽  
pp. 28-34
Author(s):  
Margarita Castillo-Téllez ◽  
Beatriz Castillo-Téllez ◽  
Juan Carlos Ovando-Sierra ◽  
Luz María Hernández-Cruz
Keyword(s):  
Medicinal Plants ◽  
Forced Convection ◽  
Rural Areas ◽  
Color Change ◽  
Drying Process ◽  
Scientific Medicine ◽  
Drying Time ◽  
Bacterial Proliferation ◽  
Wide Range ◽  
Very High

For millennia, humans have used hundreds of medicinal plants to treat diseases. Currently, many species with important characteristics are known to alleviate a wide range of health problems, mainly in rural areas, where the use of these resources is very high, even replacing scientific medicine almost completely. This paper presents the dehydration of medicinal plants that are grown in the State of Campeche through direct and indirect solar technologies in order to evaluate the influence of air flow and temperature on the color of the final product through the L* a* scale. b*, analyzing the activity of water and humidity during the drying process. The experimental results showed that the direct solar dryer with forced convection presents a little significant color change in a drying time of 400 min on average, guaranteeing the null bacterial proliferation and reaching a final humidity between 9 % and 11 %.

scholarly journals Separate Universe calibration of the dependence of halo bias on cosmic web anisotropy

2020 ◽  
Vol 499 (3) ◽  
pp. 4418-4431 ◽  
Author(s):  
Sujatha Ramakrishnan ◽  
Aseem Paranjape
Keyword(s):  
Dark Matter ◽  
High Precision ◽  
Gaussian Distribution ◽  
Initial Perturbation ◽  
Analytical Framework ◽  
Web Environment ◽  
Wide Range ◽  
Galaxy Assembly ◽  
Very High ◽  
Good Agreement

ABSTRACT We use the Separate Universe technique to calibrate the dependence of linear and quadratic halo bias b1 and b2 on the local cosmic web environment of dark matter haloes. We do this by measuring the response of halo abundances at fixed mass and cosmic web tidal anisotropy α to an infinite wavelength initial perturbation. We augment our measurements with an analytical framework developed in earlier work that exploits the near-lognormal shape of the distribution of α and results in very high precision calibrations. We present convenient fitting functions for the dependence of b1 and b2 on α over a wide range of halo mass for redshifts 0 ≤ z ≤ 1. Our calibration of b2(α) is the first demonstration to date of the dependence of non-linear bias on the local web environment. Motivated by previous results that showed that α is the primary indicator of halo assembly bias for a number of halo properties beyond halo mass, we then extend our analytical framework to accommodate the dependence of b1 and b2 on any such secondary property that has, or can be monotonically transformed to have, a Gaussian distribution. We demonstrate this technique for the specific case of halo concentration, finding good agreement with previous results. Our calibrations will be useful for a variety of halo model analyses focusing on galaxy assembly bias, as well as analytical forecasts of the potential for using α as a segregating variable in multitracer analyses.

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