Predicting Pu Concentrations in Solutions Contacting Geologic Materials

1981 ◽  
Vol 6 ◽  
Author(s):  
Richard G. Strickert ◽  
Dhanpat Rai
Keyword(s):  
Nuclear Waste ◽  
Safety Assessment ◽  
High Level Waste ◽  
Stable Phase ◽  
Borosilicate Glasses ◽  
Geologic Materials ◽  
Solid Phases ◽  
High Level ◽  
Waste Repositories

ABSTRACTKnowledge of Pu solid phases present in nuclear wastes is important for predicting the geochemical behavior of Pu. Thermodynamic data and experimental measurements using discrete Pu compounds, Pu-doped borosilicate glasses (simulating a high-level waste form), and Pu contaminated sediments suggest that PuO2(c) is very stable and is expected to be present in the repository. The solubility of the stable phase, such as PuO2(c), can be used to predict the maximum Pu concentration in solutions for long-term safety assessment of nuclear waste repositories.

Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague

2013 ◽  
Author(s):  
A. Meleshyn ◽  
U. Noseck
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Auxiliary Information ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Stable Elements ◽  
High Level ◽  
Model Approach

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.

scholarly journals Transporting and Storing High-Level Nuclear Waste in the U.S.—Insights from a Mathematical Model

2019 ◽  
Vol 9 (12) ◽  
pp. 2437 ◽  
Author(s):  
Sebastian Wegel ◽  
Victoria Czempinski ◽  
Pao-Yu Oei ◽  
Ben Wealer
Keyword(s):  
Nuclear Waste ◽  
The United States ◽  
Nuclear Industry ◽  
High Level Waste ◽  
Geological Repository ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Interim Storage ◽  
Storage Facilities

The nuclear industry in the United States of America has accumulated about 70,000 metric tons of high-level nuclear waste over the past decades; at present, this waste is temporarily stored close to the nuclear power plants. The industry and the Department of Energy are now facing two related challenges: (i) will a permanent geological repository, e.g., Yucca Mountain, become available in the future, and if yes, when?; (ii) should the high-level waste be transported to interim storage facilities in the meantime, which may be safer and more cost economic? This paper presents a mathematical transportation model that evaluates the economic challenges and costs associated with different scenarios regarding the opening of a long-term geological repository. The model results suggest that any further delay in opening a long-term storage increases cost and consolidated interim storage facilities should be built now. We show that Yucca Mountain’s capacity is insufficient and additional storage is necessary. A sensitivity analysis for the reprocessing of high-level waste finds this uneconomic in all cases. This paper thus emphasizes the urgency of dealing with the high-level nuclear waste and informs the debate between the nuclear industry and policymakers on the basis of objective data and quantitative analysis.

The Long-Term Corrosion and Modelling of Two Simulated Belgian Reference High-Level Waste Glasses

1987 ◽  
Vol 112 ◽  
Author(s):  
P. Van Iseghem ◽  
B. Grambow
Keyword(s):  
Corrosion Behaviour ◽  
Stable Condition ◽  
High Level Waste ◽  
Forward Rate ◽  
Borosilicate Glasses ◽  
Al Content ◽  
Rate Of Dissolution ◽  
High Level ◽  
High Al Content

AbstractThe corrosion behaviour in distilled water of two simulated candidate high level waste borosilicate glasses (SAN602519L3C2 and SM58LW11) his been investigated at 90°C for different SA/V condition's 10, 100, 7800 m−1). The experimental data were modelled using the PHREEQE and GLASSOL computer codes. The model is quite successful for describing the corrosion behaviour, using experimentally derived values for the forward rate, silica saturation and the final rate. GWss SAN60 is more stable than glass SM58 at SA/V values of 10 and 100 m−1, but in the long term the relative performance is inverse. Indeed, the high Al content of SAN60 induces the creation of analcime crystals after SiO2 has reached its saturation concentration in solution, which cause an enhancement of the final rate of dissolution of the glass; for SM58 on the contrary the SiO2 solution is a stable condition.

scholarly journals Bentonite clay barriers in nuclear waste repositories

2020 ◽  
Vol 205 ◽  
pp. 01003
Author(s):  
Lyesse Laloui ◽  
Alessio Ferrari ◽  
Jose A. Bosch
Keyword(s):  
Nuclear Waste ◽  
Constitutive Models ◽  
Bentonite Clay ◽  
In Situ Experiment ◽  
Future Prediction ◽  
Bentonite Clays ◽  
High Level ◽  
Waste Repositories ◽  
Clay Barriers

Deep disposal of high-level radioactive waste is the preferred solution worldwide for the long-term disposal of nuclear waste. This concept involves a series of geological and engineered barriers that provide isolation of the waste from the biosphere. Most designs involve bentonite clays as seals in different forms. During the operation of the repository, the bentonite will be subjected to a series of complex thermo-hydro-mechanical phenomena that will interact with each other. Predicting the long-term safety of geological repositories thus involve a rigorous analysis of these multi-physical processes. This paper presents a review of recent numerical approaches and analyses that have aimed to improve the understanding of processes that will take place in clay barriers over the lifetime of nuclear waste repositories. The understanding of bentonite behavior from laboratory experiments under relevant conditions is analyzed. Constitutive models that attempt to predict such behavior are presented, focusing on the stress-strain model ACMEG-TS. These models are implemented in the finite element code Lagamine which allows for the study of real scale tests. Two application cases are presented: the performance of a clay barrier according to the Swiss design, and a model of the FEEBX in situ experiment, which was modelled after a real repository under natural conditions. Overall, the relevant processes are well captured quantitatively by the models, allowing for the establishment of sound basis for future prediction and long-term design of the final underground repositories.

Environmental Remediation of High-Level Nuclear Waste in Geological Repository: Modified Computer Code Creates Ultimate Benchmark in Natural Systems

2010 ◽  
Author(s):  
Geoffrey J. Peter
Keyword(s):  
Nuclear Waste ◽  
Environmental Remediation ◽  
Computer Code ◽  
Nuclear Waste Repository ◽  
Natural Systems ◽  
Permanent Storage ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Waste Repositories

Isolation of high-level nuclear waste in permanent geological repositories has been a major concern for over 30 years due to the migration of dissolved radio nuclides reaching the water table (10,000-year compliance period) as water moves through the repository and the surrounding area. Repositories based on mathematical models allow for long-term geological phenomena and involve many approximations; however, experimental verification of long-term processes is impossible. Countries must determine if geological disposal is adequate for permanent storage. Many countries have extensively studied different aspects of safely confining the highly radioactive waste in an underground repository based on the unique geological composition at their selected repository location. This paper discusses two computer codes developed by various countries to study the coupled thermal, mechanical, and chemical process in these environments, and the migration of radionuclide. Further, this paper presents the results of a case study of the Magma-hydrothermal (MH) computer code, modified by the author, applied to nuclear waste repository analysis. The MH code verified by simulating natural systems thus, creating the ultimate benchmark. This approach based on processes similar to those expected near waste repositories currently occurring in natural systems.

Corrosion Induced Hydrogen Evolution on High Level Waste Overpack Materials in Synthetic Groundwaters and Chloride Solutions

1988 ◽  
Vol 127 ◽  
Author(s):  
J. P. Simpson ◽  
R. Schenk
Keyword(s):  
Oxide Layer ◽  
Hydrogen Evolution ◽  
Corrosion Behaviour ◽  
Cast Steel ◽  
High Level Waste ◽  
General Corrosion ◽  
Layer Formation ◽  
High Level ◽  
Waste Repositories

ABSTRACTHydrogen evolution from anoxic corrosion of cast steel overpacks in high-level waste repositories is an important issue for design if, as has been estimated, the hydrogen is prevented from escaping by diffusion by a low permeability compacted bentonite backfill.Evaluation of the corrosion results showed three basic types of corrosion behaviour: general corrosion with oxide layer formation, unstable corrosion behaviour with pitting or macro-element formation and stable passive behaviour.Cast steel containers under Swiss repository conditions are expected to suffer general corrosion with oxide layer formation. This behaviour gives the highest long term corrosion rates (2–5 μm/a) without local attack, above the 0.03–0.8 μm/a tolerated for hydrogen escape by diffusion but below the 20 μm/a assumed for overpack design.

Development of the Code Melodie for Long Term Risk Assessment of Nuclear Waste Repositories

1985 ◽  
Vol 50 ◽  
Author(s):  
P. Goblet ◽  
P. Guetat ◽  
J. Lewi ◽  
J-P Mangin ◽  
G. De Marsily ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Nuclear Waste ◽  
Computer Code ◽  
Source Model ◽  
High Level Waste ◽  
Chemical Components ◽  
Advection Dispersion ◽  
The Individual ◽  
High Level ◽  
Waste Repositories

AbstractMELODIE is a computer code developed at the CEA/IPSN for risk assessment of nuclear waste repositories in geological formations. The interactive evolution of the source, geosphere and biosphere is studied for periods of time longer than 100 000 years. In its first version, the code can describe a repository in granite rock located at a specific site.The code is built in a modular form which allows to use different versions of the sub-system models.The basic model for radionuclide migration and hydrpgeology is a subroutine version of the METIS code developed by ENSMP. METIS is a 2D finite element code which solves the advection-dispersion equation for porous media with explicit fracture representation. Linear adsorption kinetics is included as well as matrix diffusion and radionuclide decay chains.The source model developed at CEA/DRDD**** is derived from CONDIMENT which is a 1D finite difference code describing the behaviour of high level waste packages. Four axisymetric layers are individualized: glass matrix, container, bentonite and granite. The glass leaching is modelled as a dissolution and diffusion process of the individual chemical components.The biosphere model ABRICOT developed at the CEA/DPT** is based on a detailed description of agricultural activities defined in individual systems.MELODIE is tested by participation to international exercices such as Pagis [1], Intracoin [2] and Hydrocoin [3]. Future developments will include introduction of scenarios constructed from geoprospective studies and algorithms for sensitivity studies.

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