The European Journal of Physics N (EPJ-N)
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Probabilistic risk bounds for the characterization of radiological contamination
No full textThe radiological characterization of contaminated elements (walls, grounds, objects) from nuclear facilities often suffers from too few measurements. In order to determine risk prediction bounds on the level of contamination, some classic statistical methods may therefore be unsuitable, as they rely upon strong assumptions (e.g., that the underlying distribution is Gaussian) which cannot be verified. Considering that a set of measurements or their average value come from a Gaussian distribution can sometimes lead to erroneous conclusions, possibly not sufficiently conservative. This paper presents several alternative statistical approaches which are based on much weaker hypotheses than the Gaussian one, which result from general probabilistic inequalities and order-statistic based formulas. Given a data sample, these inequalities make it possible to derive prediction intervals for a random variable which can be directly interpreted as probabilistic risk bounds. For the sake of validation, they are first applied to simulated data generated from several known theoretical distributions. Then, the proposed methods are applied to two data sets obtained from real radiological contamination measurements
Formalization of the kinetics for autocatalytic dissolutions. Focus on the dissolution of uranium dioxide in nitric medium
No full textUranium dioxide dissolution in nitric acid is a complex reaction. On the one hand, the dissolution produces nitrous oxides (NOX), which makes it a triphasic reaction. On the other hand, one of the products accelerates the kinetic rate; the reaction is hence called autocatalytic. The kinetics for these kinds of reactions need to be formalized in order to optimize and design innovative dissolution reactors. In this work, the kinetics rates have been measured by optical microscopy using a single particle approach. The advantages of this analytical technique are an easier management of species transport in solution and a precise following of the dissolution rate. The global rate is well described by a mechanism considering two steps: a non-catalyzed reaction, where the catalyst concentration has no influence on the dissolution rate, and a catalyzed reaction. The mass transfer rate of the catalyst was quantified in order to discriminate when the reaction was influenced by catalyst accumulated in the boundary layer or uncatalyzed. This first approximation described well the sigmoid dissolution curve profile. Moreover, experiments showed that solutions filled with catalyst proved to lose reactivity over time. Results pointed out that the higher the liquid-gas exchanges, the faster the kinetic rate decreases with time. Thus, it was demonstrated, for the first time, that there is a link between catalyst and nitrous oxides. The outcome of this study leads to new ways for improving the design of dissolvers. Gas-liquid exchanges are indeed a lever to impact dissolution rates. Temperature and catalyst concentration can be optimized to reduce residence times in dissolvers
Independent assessment for new nuclear reactor safety
No full textA rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach
Nuclear instrumentation and measurement: a review based on the ANIMMA conferences
No full textThe ANIMMA conferences offer a unique opportunity to discover research carried out in all fields of nuclear measurements and instrumentation with applications extending from fundamental physics to fission and fusion reactors, medical imaging, environmental protection and homeland security. After four successful editions of the Conference, it was decided to prepare a review based to a large extent but not exclusively on the papers presented during the first four editions of the conference. This review is organized according to the measurement methodologies: neutronic, photonic, thermal, acoustic and optical measurements, as well as medical imaging and specific challenges linked to data acquisition and electronic hardening. The paper describes the main challenges justifying research in these different areas, and summarizes the recent progress reported. It offers researchers and engineers a way to quickly and efficiently access knowledge in highly specialized areas
The role of power sources in the European electricity mix
No full textThe ongoing debate in Europe about energy transition enhances the necessity to evaluate the performance of the envisaged mix of power sources, in terms of production cost, CO2 emissions and security of supply. In this study, we use MIXOPTIM, a Monte-Carlo simulator of the behavior of a mix of power sources on a territory, to evaluate the performance of the present EU power mix. After a validation on the French mix, we applied it to the whole EU territory and made variational calculations around the present mix to evaluate the performance impacts induced by small changes in installed renewable power and nuclear power. According to the analyzed criteria, the study shows that a plausible way to keep an affordable MWh in Europe with minimal amount of CO2 emissions and acceptable security of supply could be to extend the life of existing Gen II nuclear reactors. All other options lead to the degradation of the mix performance, on at least one of the three criteria listed above
Correlation
No full textThis paper presents a Bayesian approach based on integral experiments to create correlations which do not appear with differential data. Some quantities such as the fission cross section (σ), neutron multiplicity (ν̅p), neutron spectra (χ), etc. are usually neither modeled together nor measured in coincidence, thus there is no correlation matrices in evaluated nuclear data libraries. One can nevertheless use the information from integral experiments such as fast criticality-safety benchmarks to correlate such quantities for possible inclusion in nuclear data libraries. A simple Bayesian set of equations is presented with random nuclear data, similarly to the usual methods applied with differential data. An example for 239Pu is proposed
Methodology for the nuclear design validation of an Alternate Emergency Management Centre (CAGE)
No full textThe methodology is devised by coupling different codes. The study of weather conditions as part of the data of the site will determine the relative concentrations of radionuclides in the air using ARCON96. The activity in the air is characterized depending on the source and release sequence specified in NUREG-1465 by RADTRAD code, which provides results of the inner cloud source term contribution. Known activities and energy spectra are inferred using ORIGEN-S, which are used as input for the models of the outer cloud, filters and containment generated with MCNP5. The sum of the different contributions must meet the conditions of habitability specified by the CSN (Spanish Nuclear Regulatory Body) (TEDE <50 mSv and equivalent dose to the thyroid <500 mSv within 30 days following the accident doses) so that the dose is optimized by varying parameters including CAGE location, flow filtering need for recirculation, thicknesses and compositions of the walls, etc. The results for the most penalizing area meet the established criteria, and therefore the CAGE building design based on the methodology presented is radiologically validated
Oxygen segregation in pre-hydrided Zircaloy-4 cladding during a simulated LOCA transient
No full textOxygen and hydrogen distributions are key elements influencing the residual ductility of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). During the high temperature oxidation, a complex partitioning of the alloying elements is observed. A finite-difference code for solving the oxygen diffusion equations has been developed by Institut de Radioprotection et de Sûreté Nucléaire to predict the oxygen profile within the samples. The comparison between the calculations and the experimental results in the mixed α+β region shows that the oxygen diffusion is not accurately predicted by the existing modeling. This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200 °C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples. Experimental oxygen distribution was measured by Electron Probe Micro-Analysis (EPMA). The phase distributions within the cladding thickness, was measured using image analysis to determine the radial profile of α(O) phase fraction. It is further demonstrated and experimentally checked that the α-phase fraction in these regions follows a diffusion-like radial profile. A new phase fraction modeling is then proposed in the cladding metallic part during steam oxidation. The modeling results are compared to a large set of experiments including the influence of exposure duration and hydrogen content. Another key outcome from this modeling is that oxygen average profile is straightforward derived from the proposed modeling
Structural integrity assessment and stress measurement of chasnupp-1 fuel assembly skeleton: under tensile loading condition
No full textFuel assembly (FA) structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear power plant-1 FA skeleton at room temperature. The finite element (FE) analysis has been performed using ANSYS, in order to determine the elongation of the FA skeleton as well as the location of max. stress and stresses developed in axial direction under tensile load of 9800 N or 2 g being the FA handling or lifting load [Y. Zhang et al., Fuel Assembly Design Report, SNERDI, China, 1994]. The FE model of grids, guide thimbles with dash-pots and flow holes has been developed using Shell 181. It has been observed that FA skeleton elongation values obtained through FE analysis and experiment are comparable and show linear behaviors. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Therefore, validation of the FE methodology is confirmed. The values of stresses are less than the design limit of the materials used for the grid and the guide thimble. Therefore, the structural integrity criterion of CHASNUPP-1 FA skeleton is fulfilled safely
Dissolution of uranium dioxide in nitric acid media: what do we know?
No full textThis article draws a state of knowledge of the dissolution of uranium dioxide in nitric acid media. The chemistry of the reaction is first investigated, and two reactions appear as most suitable to describe the mechanism, leading to the formation of monoxide and dioxide nitrogen as reaction by-products, while the oxidation mechanism is shown to happen before solubilization. The solid aspect of the reaction is also investigated: manufacturing conditions have an impact on dissolution kinetics, and the non-uniform attack at the surface of the solid results in the appearing of pits and cracks. Last, the existence of an autocatalytic mechanism is questionned. The second part of this article presents a compilation of the impacts of several physico-chemical parameters on the dissolution rates. Even though these measurements have been undertaken under a broad variety of conditions, and that the rate determining step of the reaction is usually not specified, general trends are drawn from these results. Finally, it appears that several key points of knowledge still have to be clarified concerning the dissolution of uranium dioxide in nitric acid media, and that the macroscopic scale which has been used in most studies is probably not suitable