The European Journal of Physics N (EPJ-N)
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Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors
No full textDuring the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA.
The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased.
In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor
Towards spatial kinetics in a low void effect sodium fast reactor: core analysis and validation of the TFM neutronic approach
No full textThe studies presented in this paper are performed in the general framework of transient coupled calculations with accurate neutron kinetics models able to characterize spatial decoupling in the core. An innovative fission matrix interpolation model has been developed with a correlated sampling technique associated to the Transient Fission Matrix (TFM) approach. This paper presents a validation of this Monte Carlo based kinetic approach on sodium fast reactors. An application case representative of an assembly of the low void effect sodium fast reactor ASTRID is used to study the physics of this kind of system and to illustrate the capabilities provided by this approach. To validate the interpolation model developed, different comparisons have been performed with direct Monte Carlo and ERANOS deterministic S
N calculations on spatial kinetics parameters (flux redistribution, reactivity estimation, etc.) together with point kinetics feedback estimations
Local correlated sampling Monte Carlo calculations in the TFM neutronics approach for spatial and point kinetics applications
No full textThese studies are performed in the general framework of transient coupled calculations with accurate neutron kinetics models. This kind of application requires a modeling of the influence on the neutronics of the macroscopic cross-section evolution. Depending on the targeted accuracy, this feedback can be limited to the reactivity for point kinetics, or can take into account the redistribution of the power in the core for spatial kinetics. The local correlated sampling technique for Monte Carlo calculation presented in this paper has been developed for this purpose, i.e. estimating the influence on the neutron transport of a local variation of different parameters such as sodium density or fuel Doppler effect. This method is associated to an innovative spatial kinetics model named Transient Fission Matrix, which condenses the time-dependent Monte Carlo neutronic response in Green functions. Finally, an accurate estimation of the feedback effects on these Green functions provides an on-the-fly prediction of the flux redistribution in the core, whatever the actual perturbation shape is during the transient. This approach is also used to estimate local feedback effects for point kinetics resolution
Economic appraisal of deployment schedules for high-level radioactive waste repositories
No full textThe deep geological repository (DGR) is considered as the definitive management solution for high-level waste (HLW). Countries defined different DGR implementation schedules, depending on their national context and political choices. We raise the question of the economic grounds of such political decisions by providing an economic analysis of different DGR schedules. We investigate the optimal timing for DGR commissioning based on available Nuclear Energy Agency (NEA) data (2013). Two scenarios are considered: (1) rescheduling the deployment of a DGR with the same initial operational period, and (2) rescheduling the deployment of a DGR with a shorter operational period, i.e. initial closure date. Given the long timescales of such projects, we also take into account the discounting effect. The first finding is that it appears more economically favorable to extend the interim storage than to dispose of the HLW immediately. Countries which chose “immediate” disposal are willing to accept higher costs to quickly solve the problem. Another interesting result is that there is an optimal solution with respect to the length of DGR operational period and the waste flow for disposal. Based on data provided by the Organisation for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA), we find an optimal operating period of about 15 years with a flow of 2000 tHM/year
Lithium and boron analysis by LA-ICP-MS results from a bowed PWR rod with contact
No full textA previously published investigation of an irradiated fuel rod from the Ringhals 2 PWR, which was bowed to contact with an adjacent rod, identified a significant but highly localised thinning of the clad wall and increased corrosion. Rod fretting was deemed unlikely due to the adhering oxide covering the surfaces. Local overheating in itself was also deemed insufficient to account for the accelerated corrosion. Instead, an enhanced concentration of lithium due to conditions of local boiling was hypothesised to explain the accelerated corrosion. Studsvik has developed a hot cell coupled LA-ICP-MS (Laser Ablation Inductively Coupled Plasma Mass Spectrometer) equipment that enables a flexible means of isotopic analysis of irradiated fuel and other highly active surfaces. In this work, the equipment was used to investigate the distribution of lithium (7Li) and boron (11B) in the outer oxide at the bow contact area. Depth profiling in the clad oxide at the opposite side of the rod to the point of contact, which is considered to have experienced normal operating conditions and which has a typical oxide thickness, evidenced levels of ∼10–20 ppm 7Li and a 11B content reaching hundreds of ppm in the outer parts of the oxide, largely in agreement with the expected range of Li and B clad oxide concentrations from previous studies. In the contact area, the 11B content was similar to the reference condition at the opposite side. The 7Li content in the outermost oxide closest to the contact was, however, found to be strongly elevated, reaching several hundred ppm. The considerable and highly localised increase in lithium content at the area of enhanced corrosion thus offers strong evidence for a case of lithium induced breakaway corrosion during power operation, when rod-to-rod contact and high enough surface heat flux results in a very local increase in lithium concentration
Energetic and economic cost of nuclear heat − impact on the cost of desalination
No full textAn exploratory study has been carried out to evaluate the cost of heat supplied by a pressurized water reactor type of nuclear reactors to thermal desalination processes. In the context of this work, simplified models have been developed to describe the thermodynamics of power conversion, the energetics of multi-effect evaporation (MED), and the costs of electricity and heat cogenerated by the dual-purpose power plant. Application of these models show that, contrary to widespread belief, (nuclear-powered) MED and seawater reverse osmosis are comparable in terms of energy effectiveness. Process heat can be produced, in fact, by a relatively small increase in the core power. As fuel represents just a fraction of the cost of nuclear electricity, the increase in fuel-related expenses is expected to have limited impact on power generation economics
Development of the MFPR model for fission gas release in irradiated UO
No full textThe fission gas release microscopic model of the mechanistic code MFPR is further developed for modelling of enhanced release from irradiated UO2 fuel under transient conditions of the power ramp tests, along with the microstructure evolution characterised by the formation of a new population of large intragranular bubbles with a rather wide size distribution (from 30 to 500 nm), observed in transient-tested UO2 fuel samples. Implementation of the additional microscopic mechanisms results in a notable improvement of the code predictions (in comparison with the previous code version) for the fractional gas release in the Risø ramp tests with three different hold times of 3, 40 and 62 h at the terminal linear power of ≈40 kW/m
Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels
No full textFission gas behavior within the fuel structure plays a major role for the safety of nuclear fuels during operation in the nuclear power plant. Fission gas distribution and retention is determined by both, micro- and lattice-structure of the fuel matrix. The ADOPT (Advanced Doped Pellet Technology) fuel, containing chromium and aluminum additives, shows larger grain sizes than standard (undoped) UO2 fuel, enhancing the fission gas retention properties of the matrix. However, the additions of such trivalent cations shall also induce defects in the lattice. In this study, we investigated the microstructure of such doped fuels as well as a reference standard UO2 by positron annihilation spectroscopy (PAS). Although this technique is particularly sensitive to lattice point defects in materials, a wider application in the UO2 research is still missing. The PAS-lifetime components were measured in the hotlab facility of PSI using a 22Na source sandwiched between two 500-μm-thin sample discs. The values of lifetime at the center and the rim of both samples, examined to check at the radial homogeneity of the pellets, are not significantly different. The mean lifetimes were found to be longer in the ADOPT material, 220 ps, than in standard UO2, 190 ps, which indicates a larger presence of additional defects, presumably generated by the dopants. While two-component decomposition (bulk + one defect component) could be performed for the standard material, only one lifetime component was found in the doped material. The absence of the bulk component in the ADOPT sample refers to a saturated positron trapping (i.e., all positrons are trapped at defects). In order to associate a type of lattice defect to each PAS component, interpretation of the PAS experimental observations was conducted with respect to existing experimental and modeling studies. This work has shown the efficiency of PAS to detect lattice point defects in UO2 produced by Cr and Al oxides. These additives create lattice irregularities, which are acting as sinks for fission products on one hand and trapping positrons on the other hand. Fitting of the obtained experimental data with a suitable theoretical model can provide a valuable qualitative assessment of these defects. At this stage of the research, some of the existing models were used for this purpose
State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors
No full textThe paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs) and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n–γ field. A recently developed methodology based on the use of 7Li and 6Li-enriched TLDs, precalibrated both in photon and neutron fields, is a promising approach to deconvolute the two components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fibered setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1σ)
RadFET dose response in the CHARM mixed-field: FLUKA MC simulations
No full textThis paper focuses on Monte Carlo simulations aimed at calculating the dose response of the RadFET dosimeter, when exposed to the complex CHARM mixed-fields, at CERN. We study how the dose deposited in the gate oxide (SiO2) of the RadFET is affected by the energy threshold variation in the Monte Carlo simulations as well as the materials and sizes of scoring volumes. Also the characteristics of the input spectra will be taken into account and their impact on the final simulated dose will be studied. Dose variation as a function of the position of the RadFET in the test facility will be then examined and comparisons with experimental results will be shown. The contribution to the total dose due to all particles of the mixed-field, under different target-shielding configurations, is finally presented, aiming at a complete characterization of the RadFETs dose response in the CHARM mixed-fields