Retrograde Solubilities of Source Term Phases

1996 ◽  
Vol 465 ◽  
Author(s):  
William M. Murphy
Keyword(s):  
Heat Dissipation ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Radioactive Decay ◽  
Source Area ◽  
Reaction Rates ◽  
Mineral Dissolution ◽  
Spent Fuel ◽  
Natural Analog ◽  
Using Data

ABSTRACTNatural analog, experimental, and thermodynamic studies indicate that the properties of secondary uranyl minerals are likely to control the source term for U and other radioéléments incorporated in these phases in the proposed Yucca Mountain repository. Thermodynamic calculations using data from the 1992 NEA data base indicate an increase in the equilibrium constant for schoepite dissolution from 103.1 to 104.8 with decreasing temperature from 100° to 25°C, i.e., retrograde solubility. Enthalpies for mineral transformation reactions that consume protons and release cations are typically negative, suggesting that solubilities of other uranyl phases such as uranophane increase more than that of schoepite with decreasing temperature. Solubilities of mineral phases associated with spent nuclear fuel will be initially relatively low under the elevated repository temperature regime. As the temperature of the repository decreases due to radioactive decay and heat dissipation, source term mineral solubilities increase. The rate of release of U and other species is controlled by a series of processes: transport of oxidants and flux of water; oxidative dissolution of spent fuel; uranyl mineral precipitation; uranyl mineral dissolution or transformation; and radionuclide transport. Decreasing diffusion and reaction rates and increasing uranyl mineral solubilities with decreasing temperature may lead to a change with time from solubility to transport or reaction rate as a source term controlling mechanism. Preservation of large quantities of uranyl minerals formed by oxidation of uraninite and radiometrie ages of secondary uranophane at the natural analog site at Peña Blanca indicate that oxidative alteration of uraninite was fast relative to transport of U away from the deposit. The successive formation of schoepite and uranophane in natural settings where uraninite has been oxidized may represent a paragenesis reflecting increasing temperature or increasing incorporation of environmental components. In contrast, diminishing temperature conditions in a repository source area could lead to the reverse sequence of mineral formation.

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.

scholarly journals An Automated SeaFAST ICP-DRC-MS Method for the Determination of 90Sr in Spent Nuclear Fuel Leachates

Molecules ◽  
2020 ◽  
Vol 25 (6) ◽  
pp. 1429 ◽  
Author(s):  
Víctor Vicente Vilas ◽  
Sylvain Millet ◽  
Miguel Sandow ◽  
Luis Iglesias Pérez ◽  
Daniel Serrano-Purroy ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Analytical Approach ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Chemical Purification ◽  
Radiochemical Method ◽  
Good Agreement

To reduce uncertainties in determining the source term and evolving condition of spent nuclear fuel is fundamental to the safety assessment. ß-emitting nuclides pose a challenging task for reliable, quantitative determination because both radiometric and mass spectrometric methodologies require prior chemical purification for the removal of interfering activity and isobars, respectively. A method for the determination of 90Sr at trace levels in nuclear spent fuel leachate samples without sophisticated and time-consuming procedures has been established. The analytical approach uses a commercially available automated pre-concentration device (SeaFAST) coupled to an ICP-DRC-MS. The method shows good performances with regard to reproducibility, precision, and LOD reducing the total time of analysis for each sample to 12.5 min. The comparison between the developed method and the classical radiochemical method shows a good agreement when taking into account the associated uncertainties.

A Heat Dissipation Simulation Study in a CFD Way on Spent Nuclear Fuel Dry Stored in Underground Vertical Shaft

2017 ◽  
Author(s):  
Zhang Hong-jian ◽  
Yu Ren ◽  
Liu Xiao-fan
Keyword(s):  
Heat Dissipation ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Vertical Shaft ◽  
Thermal Safety ◽  
Spent Fuel ◽  
Ansys Fluent ◽  
Measuring Point ◽  
Decay Heat ◽  
Current Heat

In order to ensure the thermal safety of the spent fuel stored in an underground vertical shaft, an air current heat dissipation is simulated in CFD way using ANSYS FLUENT code. Forced convection heat dissipation is focused in the research. The air current in the shaft and the temperature distribution on the surface of the spent fuel canister are calculated. The result confirms the reliability and security of the spent fuel dry storage. Finally, based on the calculating result, a support structure is designed, and the storage position of the spent fuel canister in vertical shaft is discussed, to optimize the decay heat removal of the spent fuel, and to ensure the temperature measuring point is set in a reasonable position.

Modelling Oxidative Dissolution of Spent Fuel

1996 ◽  
Vol 465 ◽  
Author(s):  
Ivars Neretnieks
Keyword(s):  
Redox Reaction ◽  
Uranium Oxide ◽  
Reaction Rate ◽  
Spent Nuclear Fuel ◽  
Large Fraction ◽  
Reaction Rates ◽  
Spent Fuel ◽  
First Case ◽  
Rate Of Movement

ABSTRACTSpent nuclear fuel will, by the radiation, split nearby water into oxidizing and reducing compounds. The reducing compounds are mostly hydrogen that will diffuse away. The remaining oxidizing compounds can oxidize the uranium oxide of the fuel and make it more soluble. The oxidised uranium will dissolve and diffuse away. The nuclides previously incorporated in the spent fuel matrix can then be released and also migrate away from the fuel.A model is proposed where the produced oxidizing species compete for reaction with the fuel and for escaping out of the system. The chemical reaction rate of oxygen and fuel is taken from literature values based on experiments. The escape rate of oxidants to a receding redox front in the backfill is modelled assuming a redox reaction of oxidizing component and reducing component in the surrounding. The rate of movement of the redox front is determined from the rate of production of oxidants. This is estimated using a previously devised model that has been calibrated to in situ observed radiolysis.Three cases are modelled. In the first case it is assumed that the reducing compound is insoluble and that the reaction between oxygen and reducing mineral is very fast. In the second case it is assumed that the reducing component has a known solubility and that it can migrate to meet the oxygen and quickly react. In a third case a finite reaction rate is modelled between the oxygen and the reducing species.The sample calculations show that if the reducing mineral has to be supplied from the backfill a large fraction of the spent fuel could be oxidised. If the corrosion products of a degraded steel canister can supply the reducing species and the redox reaction is fast, very small amounts of the fuel could be oxidised. Literature data indicate that the redox reaction rate may not be so fast that it can be considered instantaneous and then a considerable fraction of the fuel could be oxidised. The model gives a means of exploring which mechanisms and data may be of most importance for radiolytic fuel dissolution, but the realism of the data and the model must be tested further. There is a lack of understanding and data on reaction rates, heterogeneous as well as homogeneous. This is crucial to the results.

A New Approach to the RN Source Term for Spent Nuclear Fuel under Geological Disposal Conditions

2000 ◽  
Vol 663 ◽  
Author(s):  
Christophe Poinssot ◽  
Christophe Jegou ◽  
Pierre Toulhoat ◽  
Jean-Paul Piron ◽  
Jean-Marie Gras
Keyword(s):  
Low Temperature ◽  
Closed System ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Driving Forces ◽  
Spent Fuel ◽  
Microscopic Description ◽  
New Approach ◽  
Geological Disposal ◽  
Source Term Model

ABSTRACTUnder the geological disposal conditions, spent fuel (SF) is expected to evolve during the 10,000 years while being maintained isolated from the biosphere before coming in contact with water. Under these circumstances, several driving forces would lead to the progressive intrinsic transformations within the rod which would modify the subsequent release of radionuclides. The major mechanisms are the production of a significant volume of He within the UO2 lattice, the accumulation of irradiation defects due to the low temperature which avoids any annealing, the slow migration of radionuclides (RN) within the system by (i) the α self-irradiation-induced athermal diffusion and (ii) locally the building-up of internal mechanical stresses which could turn the pellets into powder. However, the current RN source terms for SF have never accounted for this evolution as they are based on existing knowledge of the fresh SF. Two major mechanisms were considered, the leaching of the readily available fraction (one which was supposed to be instantly accessible to water), and the release of RN through alteration of the UO2 grains. We are now proposing a new RN source term model based on a microscopic description of the system in order to also take account of the early evolution of the closed system, the amplitude of which increases with the burnup and is greater for MOX fuels.

Modeling spent nuclear fuel alteration and radionuclide migration in disposal conditions

2006 ◽  
Vol 94 (9-11) ◽  
Author(s):  
Laurent de Windt ◽  
H. Schneider ◽  
C. Ferry ◽  
H. Catalette ◽  
V. Lagneau ◽  
...  
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Transport Model ◽  
Near Field ◽  
Radioactive Decay ◽  
Steel Corrosion ◽  
Spent Fuel ◽  
Fuel Pellets ◽  
Physico Chemical ◽  
Integrated Performance

A physico-chemical model developed for spent fuel alteration was integrated in a global reactive transport model of a spent fuel disposal system, considering both decaying and stable isotopes, corroded steel canisters, bentonite backfills and a clayey host-rock. Fuel evolution took into account radiolytic-enhanced corrosion and long-term solubility-controlled dissolution as well as instantaneous release fractions. The calculations show that spent-fuel dissolution has no significant alteration effect on the near-field components except an oxidizing plume in the vicinity of the waste packages. The dissolved uranyl species, partly precipitate as schoepite on the fuel pellets, and partly diffuse in the near-field where magnetite and pyrite reduce U(VI) to yield uraninite precipitation. Under disposal conditions, preliminary calculations indicate that steel corrosion may generate sufficient dissolved hydrogen as to react with radiolytic oxidants and inhibit fuel dissolution. The formation of a protective schoepite layer could also reduce the alteration of fuel pellets. Radionuclides migration (Am, Cs, I) in the near-field is discussed in a second stage discriminating between sorption, precipitation and radioactive decay processes. The migration of Cs is translated in terms of cumulative activity profiles useful for integrated performance assessment.

Release of 108Ag from Irradiated PWR Control Rod Absorbers under Deep Repository Conditions

2015 ◽  
Vol 1744 ◽  
pp. 217-222
Author(s):  
O. Roth ◽  
M. Granfors ◽  
A. Puranen ◽  
K. Spahiu
Keyword(s):  
Nuclear Fuel ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Nuclear Reactors ◽  
Reactor Irradiation ◽  
Spent Fuel ◽  
Control Rod ◽  
Potential Source ◽  
Absorber Material ◽  
Control Rods

ABSTRACTIn a future Swedish deep repository for spent nuclear fuel, irradiated control rods from PWR nuclear reactors are planned to be stored together with the spent fuel. The control rod absorber consists of an 80% Ag, 5% Cd, 15% In alloy with a steel cladding. Upon in-reactor irradiation 108Ag is produced by neutron capture. Release of 108Ag has been identified as a potential source term for release of radioactive substances from the deep repository.Under reducing deep repository conditions, the Ag corrosion rate is however expected to be low which would imply that the release rate of 108Ag should be low under these conditions. The aim of this study is to investigate the dissolution of PWR control rod absorber material under conditions relevant to a future deep repository for spent nuclear fuel. The experiments include tests using irradiated control rod absorber material from Ringhals 2, Sweden. Furthermore, un-irradiated control rod absorber alloy has been tested for comparison. The experiments indicate that the release of Ag from the alloy when exposed to water is strongly dependent on the redox conditions. Under aerated conditions Ag is released at a significant rate whereas no release could be measured after 133 days during leaching under H2.

Alternate Source Term Models for Yucca Mountain Performance Assessment Based on Natural Analog Data and Secondary Mineral Solubility

1999 ◽  
Vol 556 ◽  
Author(s):  
William M. Murphy ◽  
Richard B. Codell
Keyword(s):  
Performance Assessment ◽  
Release Rate ◽  
Source Term ◽  
Yucca Mountain ◽  
Solution Chemistry ◽  
Base Case ◽  
Total System ◽  
Spent Fuel ◽  
Natural Analog ◽  
Mineral Solubility

AbstractPerformance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Peña Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Peña Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg yr−1, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Peña Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses.

Thermal and Fluid Analysis of Test Canister for Spent Nuclear Fuel

2020 ◽  
Author(s):  
Marwan Charrouf ◽  
Allen Williams
Keyword(s):  
Nuclear Fuel ◽  
Heat Dissipation ◽  
Spent Nuclear Fuel ◽  
Flow Behavior ◽  
The United States ◽  
Time Profile ◽  
Electrical Heating ◽  
Cfd Modeling ◽  
Electric Heater ◽  
Spent Fuel

Abstract The absence of a long-term solution for the storage of spent nuclear fuel prompts utilities in the United States to establish on-site storage for used fuel. The challenges associated with placement of spent fuel in dry cask storage on the power plant’s Independent Spent Fuel Storage Installations (ISFSI’s) include aging management of the stainless steel canisters and monitoring for the possible onset of stress corrosion cracking (SCC). The San Onofre Nuclear Generating Station (SONGS) has initiated a test program to examine the effects of heat generation variations inside a test canister using an electric heater rather than spent fuel on the shell temperatures. The test helps in the evaluation of external environmental factors and shell temperature, and to monitor for SCC. This paper presents the computational fluid dynamics (CFD) modeling developed in support of the test to analyze the air natural circulation in the subgrade enclosure and within the test canister with the electrical heating. The thermal analysis is performed using ANSYS CFX and integrally simulates the flow behavior and heat transfer mechanisms both inside and outside the test canister. Comparison of results from different heat loads that represent the decay heat time-profile, sensitivity to the turbulence model, and modes of heat dissipation are discussed. The CFD results are also compared to in-situ temperature measurements to validate the analysis.

Source Trends for Performance Assessment of HLW Glass and Spent Fuel as Waste Forms

1997 ◽  
Vol 506 ◽  
Author(s):  
B. Grambow
Keyword(s):  
Mass Transfer ◽  
Performance Assessment ◽  
Source Term ◽  
Reaction Rates ◽  
Dissolution Mechanism ◽  
Experimental Measurements ◽  
Source Terms ◽  
Spent Fuel ◽  
Fuel Oxidation ◽  
Waste Forms

ABSTRACTwith respect to the state of validation for source term development. Consequences of the various mechanism on mass half lives of the waste forms are calculated with analytical equations. For glass the largest uncertainty stems from the yet unclear dissolution mechanism under silica saturated conditions. Source terms based on silica solubility coupled to Si-mass transfer are probably neither conservative nor realistic. For spent fuel the largest uncertainty is in the extrapolation of radiolytic fuel oxidation for long periods of time. Considering the uncertainties involved, reaction rates cannot yet be extrapolated reliably to values much lower than the lowest reliable experimental measurements.

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