Leaching of Spent Fuel and Simulated Fuel in the Presence of Environmental Materials: Integral Experiments

1997 ◽  
Vol 506 ◽  
Author(s):  
K. Le Lous ◽  
S. Constantin ◽  
J.L. Paul ◽  
D. Sambugaro ◽  
E. Vernaz
Keyword(s):  
Deep Geological Repository ◽  
Spent Fuel ◽  
Leach Rate ◽  
Simulated Fuel ◽  
Sandy Clay ◽  
Reducing Conditions ◽  
Geological Repository ◽  
Environmental Materials ◽  
Order Of Magnitude ◽  
Granitic Groundwater

ABSTRACTWe have designed an apparatus which simulates the conditions of a deep geological repository in order to study the behaviour of spent fuel under leaching. A spent fuel and a Simfuel have thus been leached by synthesized clayey or granitic groundwater in sandy clay (90 % sand) or granite, in reducing conditions, at 90°C and 40 bars.The apparent leach rate for spent fuel in clayey water is 3.33 µg.m−2.dr−1in the presence of clay and 3.37 µg.m-−2.d−1in the presence of granite. The apparent leach rate in granitic water is slower, being 0.37 µg.m−2d−1in the presence of clay and 0.74 µg.m−2d−1in the presence of granite.For Simfuel, the apparent leach rate in clayey water is 7.4 µg.m−2.d−1in the presence of clay and 1. 1 µg.m−2.d−1in the presence of granite, which is the same order of magnitude as that for spent fuel. In granitic water, the apparent leach rate is 20 to 40 times greater than that for spent fuel. It is 14.8 µg.m−2.d−1in the presence of clay and 18.5 µg.m−2.d−1in the presence of granite.

The Effect of Bromide on Oxygen Yields in Homogeneous α-radiolysis

MRS Advances ◽  
2017 ◽  
Vol 2 (13) ◽  
pp. 711-716 ◽  
Author(s):  
Lovisa Bauhn ◽  
Christian Ekberg ◽  
Patrik Fors ◽  
Kastriot Spahiu
Keyword(s):  
Radiation Field ◽  
Spent Nuclear Fuel ◽  
Radiation Doses ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
Mass Spectrometric Measurement ◽  
Α Decay ◽  
Geological Repository ◽  
Dissolved Ions ◽  
Dissolved Hydrogen

ABSTRACTIn a scenario where ground water enters a canister for spent nuclear fuel in a deep geological repository, the presence of dissolved ions in the water could possibly influence the fuel dissolution due to effects on radiolysis yields. One species of particular interest in this context is bromide, which has a proven ability to scavenge hydroxyl radicals much faster than molecular hydrogen does. As a result, bromide could inhibit the beneficial effect of dissolved hydrogen, which has been shown in γ-radiolysis experiments. However, already a few hundred years after repository closure, α-decay starts to dominate in the radiation field from the spent fuel. Hence, the effects of α-radiolysis are expected to govern the fuel dissolution over the geological timeframes of the repository. In the present work, α-radiolysis experiments have been performed to determine the effect of bromide ions on the yield of hydrogen peroxide by mass spectrometric measurement of its decomposition product oxygen. The use of high activity 238Pu solutions has made it possible to study this effect during pure α-radiolysis from a homogeneously distributed radiation field. To simulate deep bedrock repository conditions, and to minimize the influence of in-leaking O2 from air, the studies were performed using graphite sealed stainless steel autoclaves with an initial atmosphere of 10 bar H2. The results show that addition of 1 mM Br- to the solution gives no significant effect on the O2 yield for radiation doses up to 2 MGy. This lack of effect is most likely explained by the limited radical escape yields from radiation tracks in pure α-radiolysis.

scholarly journals Comparison of Gamma and Neutron Dose Rate Variations with Time for Cast Iron and Metal—Concrete Casks Used for RBMK-1500 Spent Fuel Storage

2021 ◽  
Vol 11 (16) ◽  
pp. 7362
Author(s):  
Arturas Smaizys ◽  
Ernestas Narkunas ◽  
Gintautas Poskas ◽  
Povilas Poskas
Keyword(s):  
Cast Iron ◽  
Dose Rate ◽  
Storage Period ◽  
Neutron Dose ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
Geological Repository ◽  
Fuel Storage ◽  
Management Concept ◽  
Neutron Dose Rate

The present SF management concept in Lithuania envisages that spent RBMK-1500 fuel will be stored in dry storage containers for 50 years, before being disposed of in a deep geological repository. However, the risk that a deep geological repository will not be constructed at the planned time should be taken into account, and the extension of SF storage over 50 years should be considered. This paper presents a comparison of gamma and neutron dose rate distributions and variations with planned and extended storage times for cast iron and metal–concrete containers loaded with RBMK-1500 SF. All calculations were performed using the SCALE computer codes system. The modeling results show that the overall shielding properties of the CONSTOR® RBMK-1500 container containing the same neutron and gamma sources are better than those of the CASTOR® RBMK-1500 container. During an extended storage period (from 50 to 300 years), the total dose rate would decrease considerably and the dose rate due to neutrons would become dominant for both containers.

scholarly journals Development of New Waste Forms to Immobilize Iodine-129 Released from a Spent Fuel Reprocessing Plant

2010 ◽  
Vol 73 ◽  
pp. 158-170 ◽  
Author(s):  
Hiromi Tanabe ◽  
Tomofumi Sakuragi ◽  
Kenji Yamaguchi ◽  
Taemi Sato ◽  
Hitoshi Owada
Keyword(s):  
Radioactive Wastes ◽  
Behavior Modeling ◽  
Spent Fuel ◽  
Waste Forms ◽  
Reducing Conditions ◽  
Geological Repository ◽  
Spent Fuel Reprocessing ◽  
Immobilization Techniques ◽  
Reprocessing Plant ◽  
Fuel Reprocessing

I-129 is a very long-lived radionuclide that is released to an off-gas stream when spent fuels are dissolved at a reprocessing plant. An iodine filter can capture I-129 in the form of AgI. However, because AgI is unstable under the reducing conditions of a geological repository and I-129 has a very long half-life, I-129 can migrate to the biosphere. These characteristics make I-129 a key radionuclide for the safety assessment of a geological disposal of radioactive wastes generated from a reprocessing plant (TRU wastes). To improve disposal safety, several new waste forms have been developed to confine I-129 for a very long period in order to reduce the leaching of I-129 from radioactive wastes. These new waste forms have technical objectives of solidifying more than 95% of I-129 into the waste form and achieving a leaching rate of less than 10-5/y. Several iodine immobilization techniques have been examined. This paper presents experimental results concerning the treatment process, leaching behavior, modeling, and related elements of these immobilization techniques.

α-Radiolysis and α-Radiation Damage Effects on uo2 Dissolution Under Spent Fuel Storage Conditions

1999 ◽  
Vol 556 ◽  
Author(s):  
V. V. Rondinella ◽  
Hj. matzke ◽  
J. Cobos ◽  
T. Wiss
Keyword(s):  
Spent Nuclear Fuel ◽  
Sol Gel ◽  
Storage Conditions ◽  
Lattice Parameter ◽  
Spent Fuel ◽  
Α Decay ◽  
Radiation Fields ◽  
Geological Repository ◽  
Fuel Storage ◽  
Order Of Magnitude

Abstractα-decay will constitute almost entirely the radiation field in and around spent nuclear fuel after a few hundred years in a geological repository. Pellets of UO 2 containing ˜0.1 and ˜10 wt. % 238Pu were fabricated using a sol-gel method and characterized, comparing their properties to those of undoped UO2. The α-radiation fields of different types of commercial LWR spent fuel are of the same order of magnitude as the fuel with the lower Pu-concentration used in this work. The results of static batch leaching tests at room temperature in demineralized water under anoxic atmosphere showed that the amounts of U released during leaching were higher in the case of UO2 containing 238pu than for undoped UO2. Relatively large amounts of Pu were released after the longest leaching times. Lattice parameter measurements using XRD and hardness measurements by Vickers indentation showed a relatively rapid build-up of α-decay damage in the material stored at ambient temperature with the higher concentration of dopant, while for the material with ˜0.1 wt. % Pu no clear variations were detected during the same time intervals.

Accelerated Dissolution Process of the Spent Fuel (UO2) under Repository Conditions

2008 ◽  
Vol 1104 ◽  
Author(s):  
Eduardo Iglesias ◽  
Javier Quiñones ◽  
Nieves Rodriguez
Keyword(s):  
Specific Surface ◽  
Surface Area ◽  
Specific Surface Area ◽  
Nuclear Energy ◽  
Fuel Pellet ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
General Law ◽  
Geological Repository ◽  
The Matrix

AbstractNowadays, nuclear energy is one of the options for developed countries in order to maintain the demand of electric energy. One of the problems of this kind of energy generation is the residual waste form after a fuel cycle (spent fuel). These kind of material is so difficult to characterize -due to their composition and the thermal treatment in the reactor- that exhaustive studies are necessaries for a complete knowledge, helping to build, with complete reliability, a very safety underground facility. In this way, the option known as Deep Geological Repository (DGR) is been developed by each country taking part in the nuclear energy industry. The unique via for the migration to the biosphere of the radionuclides -actinides and lanthanides content in the spent fuel pellet (UO2) after the closing of the deep geological repository is by the water transport phenomena. It is a fundamental question to know how much time they will spend in their trip; and the first step is the rate of liberation of these radionuclides from the spent fuel pellet. In this way, the matrix dissolution rate of the spent fuel pellet no dependent on the specific surface area after a normalization by the initial value- is a key parameter to begin the performance assessment for any deep geological repository and the specific surface value is, following the Matrix Alteration Model (MAM) sensitivity analysis, one of the most important parameters controlling the radionuclides liberation. In this way, several measurements were carried out to obtain values in different conditions for different sieves of UO2 powder, treated as fresh fuel. First of all, the specific surface area was measured with a multi-point isothermal procedure with N2 and Kr, the both. The values obtained were presented in order to obtain a general law for the evolution with the particle size. These data are part of a bigger project about the complete description of the spent fuel analogous; very useful to obtain new dissolution rates for the spent fuel under repository simulated conditions.

Modelling Radionuclide Mobility of Spent Fuel in a Deep Geological Repository Located in a Clay Bedrock

2009 ◽  
Author(s):  
Lara Duro ◽  
Abel Tamayo ◽  
Jordi Bruno ◽  
Aurora Marti´nez-Esparza
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Radioactive Decay ◽  
Secondary Phase ◽  
Diffusive Transport ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
Geological Repository ◽  
Source Term Model

Source term models are widely used to assess the behaviour of spent nuclear fuel after final disposal. However, most models do not take into account some phenomena which are expected to control the transport of radionuclides through the near field. Some uncertainties arise from this fact, thus making it difficult to obtain proper simulations of radionuclide behaviour in the near field. In this work, we have used a compartmental code to build up an integrated source term model in an attempt to overcome the abovementioned drawbacks. The model developed takes into account radiolytically-mediated matrix dissolution, radioactive decay chains, diffusive transport, and retardation by sorption and secondary phase precipitation, among other processes. In addition, this model has been used to estimate radionuclide mobility from spent fuel located in a conceptual clay geological repository.

ÚJV Řež, a.s. Research Programme Supporting Development of Deep Geological Repository in the Czech Republic

2014 ◽  
Author(s):  
Václava Havlová
Keyword(s):  
Czech Republic ◽  
Radioactive Waste ◽  
Research Programme ◽  
Deep Geological Repository ◽  
The Czech Republic ◽  
Reducing Conditions ◽  
In Situ Experiments ◽  
Geological Repository

ÚJV Řež, a.s. as a company with a long term experience in radioactive waste management (RWM) has been running a comprehensive research programme, supporting development of deep geological repository (DGR) in the Czech Republic. Recently ÚJV Řež, a.s. research has focused on the different aspects of safety functions that DGR barriers should provide. Moreover, the research has also recently paid strong attention to real conditions that can be present in DGR (anaerobic reducing conditions, increased T due to heat generation by radioactive waste, contact of different materials within repository, real scale of the rock massive etc.). Both types of experiments, laboratory and in-situ experiments in underground laboratories, were included in the research programme. The presentation gives a brief overview of experimental trends, being conducted for materials and conditions, concerned in Czech repository concept.

scholarly journals Needs of countries with longer timescale for deep geological repository implementation

2020 ◽  
Vol 6 ◽  
pp. 22
Author(s):  
Bálint Nős
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Level Waste ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
International Consensus ◽  
Geological Disposal ◽  
Geological Repository ◽  
Deep Geological Disposal ◽  
High Level

Countries operating nuclear power plants have to deal with the tasks connected to spent fuel and high-level radioactive waste management. There is international consensus that, at this time, deep geological disposal represents the safest and most sustainable option as the end point of the management of high-level waste and spent fuel considered as waste. There are countries with longer timescale for deep geological repository (DGR) implementation, meaning that the planned date of commissioning of their respective DGRs is around 2060. For these countries cooperation, knowledge transfer, participation in RD&D programmes (like EURAD) and adaptation of good international practice could help in implementing their own programmes. In the paper the challenges and needs of a country with longer implementation timescale for DGR will be introduced through the example of Hungary.

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