The European Journal of Physics N (EPJ-N)
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Comparison of SERPENT and SCALE methodology for LWRs transport calculations and additionally uncertainty analysis for cross-section perturbation with SAMPLER module
No full textIn nuclear safety research, the quality of the results of simulation codes is widely determined by the reactor design and safe operation, and the description of neutron transport in the reactor core is a feature of particular importance. Moreover, for the long effort that is made, there remain uncertainties in simulation results due to the neutronic data and input specification that need a huge effort to be eliminated. A realistic estimation of these uncertainties is required for finding out the reliability of the results. This explains the increasing demand in recent years for calculations in the nuclear fields with best-estimate codes that proved confidence bounds of simulation results. All this has lead to the Benchmark for Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of LWRs of the NEA. The UAM-Benchmark coupling multi-physics and multi-scale analysis using as a basis complete sets of input specifications of boiling water reactors (BWR) and pressurized water reactors (PWR). In this study, the results of the transport calculations carried out using the SCALE-6.2 program (TRITON/NEWT and TRITON/KENO modules) as well as Monte Carlo SERPENT code, are presented. Additionally, they have been made uncertainties calculation for a PWR 15 × 15 and a BWR 7 × 7 fuel elements, in two different configurations (with and without control rod), and two different states, Hot Full Power (HFP) and Hot Zero Power (HZP), using the TSUNAMI module, which uses the Generalized Perturbation Theory (GPT), and SAMPLER, which uses stochastic sampling techniques for cross-sections perturbations. The results obtained and validated are compared with references results and similar studies presented in the exercise I-2 (Lattice Physics) of UAM-Benchmark
Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior
No full textOne of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor) design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included
Modelling of as-fabricated porosity in UO
No full textFor consistent modelling of behaviour of as-fabricated porosity in UO2 fuel irradiated under various conditions of in-pile and out-of-pile tests as well as under normal and abnormal conditions of nuclear reactor operation, the additional analysis of experimental observations and critical assessment of available models are presented. On this base, the mechanistic MFPR code, including physically-grounded models for the fuel porosity evolution in UO2 fuel under various irradiation and thermal regimes, is refined. These modifications complete the consistent description of the fuel porosity evolution in the MFPR code and result in a notable improvement of the code predictions
Lock-in thermography for characterization of nuclear materials
No full textA simplified procedure of lock-in thermography was developed and applied for characterization of nuclear materials. The possibility of thickness and thermal diffusivity measurements with the accuracy better than 90% was demonstrated with different metals and Zircaloy-4 claddings
Dose and temperature distribution in spent fuel containing material
No full textSpent fuel containing material (SFCM) can arise during severe nuclear reactor accident by melting of a reactor core and surrounding material (corium) or during accident in spent fuel storage. It consists of nuclear fuel, fission products, activation products and materials from fuel cladding, concrete, etc. The paper deals with dose and temperature characteristics inside the SFCM after transition of the molten mixture to solid state. Calculations were made on simplified spherical models, without connection to some specific nuclear accident. The dose rate was estimated for alpha, beta and gamma radiation in times over the course of 30 years from the end of the fission chain reaction. Concentration of helium generated in the material by alpha decay was calculated. For the dose rate values estimation, computation code ORIGEN 2.2 with dosimetric library ENDF/B-IV were used. Temperature distribution inside the solid SFCM was calculated by FLUENT code. As source of heating, energy of radioactive decays was taken. Estimated dose and temperature characteristics can be used, e.g. for evaluation of radiation damage and temperature behaviour of SFCM or for radiation test design of corium simulating materials
Influence of the austenitic stainless steel microstructure on the void swelling under ion irradiation
No full textTo understand the role of different metallurgical parameters on the void formation mechanisms, various austenitic stainless steels were elaborated and irradiated with heavy ions. Two alloys, in several metallurgical conditions (15Cr/15Ni–Ti and 15Cr/25Ni–Ti), were irradiated in the JANNUS-Saclay facility at 600 °C with 2 MeV Fe2+ ions up to 150 dpa. Resulting microstructures were observed by Transmission Electron Microscopy (TEM). Different effects on void swelling are highlighted. Only the pre-aged samples, which were consequently solute and especially titanium depleted, show cavities. The nickel-enriched matrix shows more voids with a smaller size. Finally, the presence of nano-precipitates combined with a dense dislocation network decreases strongly the number of cavities
Evaluation of corrosion on the fuel performance of stainless steel cladding
No full textIn nuclear reactors, the use of stainless steel (SS) as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4) under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion
A study of different cases of VVER reactor core flooding in a large break loss of coolant accident
No full textThe paper covers the results of VVER core reflooding studies in fuel assembly (FA) mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup. The second type is bottom flooding of the FA mockup with level of boiling water. The test parameters are as follows: the range of the supplied power to the bundle is from 40 to 320 kW, the cooling water flow rate is from 0.04 to 1.1 kg/s, the maximum temperature of the fuel rod simulator is 800 °C and the linear heat flux is from 0.1 to 1.0 kW/m. The test results were used for computer code validation
Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium
No full textInternational organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III) using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV) operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR). For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU), though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment configuration, U5-FRs appear to provide the best solution for using uranium, even if the economic impact of this consumption during the first years of operation is significant
Initial economic appraisal of nuclear district heating in France
No full textAlthough cogeneration with nuclear power has been proving its feasibility for many years and in many parts of the world, the French nuclear fleet does not use this technique. Nevertheless, current developments within the energy context may offer new opportunities to review the use of nuclear cogeneration. This paper focuses on the use of cogeneration for district heating and its possible development perspectives within the French energy transition. After recapping some common assumptions about nuclear cogeneration, we will describe the techno-economic model that we built to evaluate the characteristics of introducing cogeneration into an already operating power plant. The second step consists in applying the above-described model to a use-case describing the heating of the Parisian area, which represents the largest target for this study. The last step presents the results of a simplified model derived from the first step. Summarizing the model's main input data in a few pertinent parameters gives an initial picture of the potential for developing nuclear district heating in France