96 research outputs found
Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction
The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%.; One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 +/- 5) degrees C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 +/- 0.62) and (3.66 +/- 0.73) for PWR fuel, and (035 +/- 0.07) and (0.51 +/- 0.10) for BWR fuel. (C) 2015 Elsevier B.V. All rights reserved.Peer ReviewedPostprint (author's final draft
Exploring alternative practices in urban water management through the lens of circular economy: a case study in the Barcelona metropolitan area
Urban water management has recently been questioned because of the fragmented nature of the urban water system and its linear model. The integration and management of water systems are currently recognized as a socio-technical challenge that must be addressed for a more sustainable urban water management. In the short term, a key factor for its transition will be integration of alternative practices that allow for experimentation, learning, and scaling up. This study aims to identify potential shifts supported by two alternative practices for water reuse: nature-based solutions and water reuse technologies, using circular economy principles as analytical categories. The research uses a case study, the Besòs river of the Barcelona metropolitan area, to show that: i) improving biodiversity and water quality helps to regenerate natural capital; ii) water reuse for streamflow augmentation keeps resources in use and promotes synergies, which benefits social livability; and iii) risk management and a potential fit-to-purpose strategy can marginally help to avoid waste externalities. This research has shown that the CE principles are applicable as a framework for identifying the interconnected shifts promoted by water systems. A reflexive understanding of the alternative practices provides deeper insight into the experiences, barriers, and shifts that allow innovative interactions in specific urban contexts and can deliver additional benefits for society. This knowledge can be useful for integrated urban management; however, further integration of cross-sectoral collaboration and flexibility are required.Peer ReviewedPostprint (published version
Immobilization of selenite onto natural siderite (FeCO3)
The sorption processes of metals and semimetals on siderite are being studied because of the potential adequacy of the mineral for removing contaminants from natural waters and because of its formation during the reduction of iron(III) (hydr)oxides under anaerobic conditions in the presence of organic matter or bacteria. This transformation could involve the release of contaminants previously sorbed on the iron(III) (hydr)oxides.
In this work, the sorption of selenite on natural siderite (Bordes de Conflent, Lleida, Spain) is experimentally studied and the results obtained are compared to the previous results on selenite sorption on magnetite, hematite and goethite.Results indicate that the process of sorption was very fast; the equilibrium was reached in less than 1 hour, which indicates that sorption will not be kinetically controlled. The experimental data were fitted with a non-competitive Langmuir isotherm, which would indicate that the main process for the decrease of selenite in solution would be its sorption on siderite via monolayer coverage of the solid surface. The values of maximum sorption capacity of siderite obtained at pH 7,1 and 8,0 were 6,9 10-7 and 6,4 10-7 mol g-1, respectively and the same value of Langmuir constant was obtained for both pH values, 4,7 104 L mol-1.Postprint (published version
UO2 surface oxidation by mixtures of water vapor and hydrogen as a function of temperature
In the present work, X-Ray Photoelectron Spectroscopy (XPS) was used to study the effect of water vapor on the UO2 surface as a function of temperature. The experiments were performed in situ inside a high pressure chamber attached to the XPS instrument. UO2 samples were put in contact with either hydrogen or argon streams, saturated with water at room temperature, and the sample surface evolution was analyzed by XPS. In the case of the water vapor/argon experiments, one experiment at 350°C was performed and, in the case of the water vapor/hydrogen experiments, the temperatures used inside the reactor were 60, 120, 200 and 350°C. On one hand, in presence of argon, the results obtained showed that the water vapor in the argon stream oxidized 93% of the U(IV) in the sample surface. On the other hand, the degree of UO2 surface oxidation showed a different dependence on the temperature in the experiments performed in the presence of hydrogen: the maximum surface oxidation occurred at 120°C, where 65.4% of U(IV) in the sample surface was oxidized, while at higher temperatures, the surface oxidation decreased. This observation is attributed to the increase of hydrogen reducing effect when temperature increases which prevents part of the oxidation of the UO2 surface by the water vapor.Peer ReviewedPostprint (author's final draft
Influence of the interpellet space to the Instant Release Fraction determination of a commercial UO2 Boiling Water Reactor Spent Nuclear Fuel
The contact of the coolant with the fuel pin during irradiation produces a gradient of temperature in the fuel pellet that segregates the radionuclides (RN) depending on its volatility and reactivity. This segregation determines the Instant Release Fraction (IRF), an important source of radiological risk in the performance assessment (PA) of a Deep Geologic Repository (DGR). RN segregation was studied radially in previous papers. In the present work, it was studied axially, taking into special consideration the cutting position of the solid sample to be studied. Iodine and caesium were the RN with the highest release, while the contribution of rubidium, strontium, molybdenum and technetium to the IRF depended on their chemical state. The interpellet presence (known also as dishing) effect was clearly observed for caesium, increasing its release by one order of magnitude. According to these results, one of the major contributions to the IRF comes from the RN trapped in the dishing and has to be considered in the sampling and data interpretation that will be performed for the PA of the DGR.Peer ReviewedPostprint (published version
Water-air volatilization factors to determine volatile organic compound (VOC) reference levels in water
The goal of this work is the modeling and calculation of
volatilization factors
(VFs) from water to air for
volatile organic compounds
(VOCs) in order to perform
human
health risk
-
base
d reference levels
(RLs) for the safe use of water. The VF models have
been developed starting from
the
overall mass
-
transfer coefficients (
K
overall
) concept from
air to water for two interaction geometries (flat surface and spherical droplets) in indoor
a
nd outdoor scenarios
.
For
a case study with
five groups of risk scenarios
and thirty VOCs
,
theoretical VFs have been calculated
by using
the developed models.
Results showed that
K
overall
values
for flat and spherical
surface
geometries
were
close to
the
m
ass transfer
coefficient for water
(
K
L
)
when
Henry
’s law
constant (
K
H
)
was high. I
n the case of
s
pherical drop geometry
,
the
fraction of volatilization
(
f
V
)
was asymptotical
when
increasing
K
H
with
f
V
values also
limited
due to
K
overall
.
VFs for flat surfa
ces were
calculated from the emission flux of VOCs
,
and results showed values close to
1000
K
H
for
the most conservative indoor scenarios and almost constant values for outdoor scenarios.
VFs for spherical geometry
in indoor scenarios
followed
also constant
VFs and
were
far
from
1000
K
H
.
The highest calculated VF values corresponded to
the
E2A, E2B, E3A and
E5A scenarios and were compared with experimental
and real
results in order to check the
goodness of
flat and sphere geometry models. Results showed an ov
erestimation of
OPEN ACCESS
Toxics
2014
,
2
277
calculated values for
the
E2A
and
E2B scenarios and an underestimation for
the
E3A and
E5A scenarios. In both cases
,
most of the calculated VFs were from 0.1
-
to 10
-
times higher
than experimental
/real
valuesPeer ReviewedPostprint (published version
Retention of cesium and strontium by uranophane, Ca(UO2)2(SiO3OH)2·5H2O
This work determines the capacity of uranophane, one of the long-term uranyl secondary solid phases formed on the spent nuclear fuel (SNF), to retain radionuclides (cesium and strontium) released during the dissolution of the SNF. Sorption was fast in both cases, and uranophane had a high sorption capacity for both radionuclides (maximum sorption capacities of 1.53·10-5¿mol¿m-2 for cesium and 3.45·10-3¿mol¿m-2 for strontium). The high sorption capacity of uranophane highlights the importance of the formation of uranyl silicates as secondary phases during the SNF dissolution, especially in retaining the release of radionuclides not retarded by other mechanisms such as precipitation.Peer ReviewedPostprint (published version
Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel
The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples.
Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel.
The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.Preprin
Oxidation by H2O(g) in the presence of H2(g) of UO2 doped with Pd nanoparticles
This is a post-peer-review, pre-copyedit version of an article published in Journal of Radioanalytical and Nuclear Chemistry. The final authenticated version is available online at: http://dx.doi.org/10.1007/s10967-018-6217-2.This work studies the influence of epsilon particles and metallic precipitates, simulated through the deposition of Pd nanoparticles on non-irradiated UO2, on the oxidation of the spent nuclear fuel (SNF). The presence of Pd nanoparticles favored the reduction by hydrogen of oxidized uranium phases located on the UO2 surface, decreased the oxidation of the UO2 by H2O in the absence of hydrogen, and, in the presence of hydrogen and water vapor, no oxidation of the UO2 by water was detected, contrasting experiments without Pd nanoparticles.Peer ReviewedPostprint (author's final draft
Contribution of phases segregated from the UO2 matrix to the release of radionuclides from spent nuclear fuel and duration of the Instant Release Fraction (IRF)
During the dissolution of the spent nuclear fuel (SNF) some radionuclides are released to the solution simultaneously from different sources in the fuel. This is of particular importance to some radionuclides that contribute to the Instant Release Fraction (IRF), which govern the initial radiation dose during the dissolution of the SNF. In this work a model that is able to discriminate between the different contributions responsible for the total concentration of a radionuclide in solution was developed. The model permits to establish that uranium and radionuclides that dissolved congruently with the UO2 matrix came from two sources as a function of time: oxidized phases on the surface of the SNF including fines and the matrix itself. Other radionuclides such as Ru and Rh were released from metallic precipitates with dissolution rates lower than the matrix dissolution rate.
In the case of radionuclides that were expected to contribute to the IRF, this work showed that Cs, Rb and Sr had initial release rates higher than uranium because a fraction of such radionuclides were segregated from the matrix during the irradiation. Actually it was calculated that the fraction of those radionuclides in the grain boundaries in a BWR SNF powder sample from the center part of a pellet (burnup 42 MWd/kgU) was 2.1%, 0.9%, and 0.6% for Cs, Rb, and Sr, respectively. In addition, the model permitted to calculate the duration of their contribution to the IRF, matrix dissolution governed the release of such radionuclides after 137 days, 75 days, and 164 days for Cs, Rb, and Sr, respectively (at these times, the contribution of the release from grain boundaries was lower than the 0.1%).Peer ReviewedPostprint (author's final draft
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