The European Journal of Physics N (EPJ-N)
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CADOR “Core with Adding DOppleR effect” concept application to sodium fast reactors
No full textGeneration-IV sodium fast reactors (SFR) will only become acceptable and accepted if they can safely prevent or accommodate reactivity insertion accidents that could lead to the release of large quantities of mechanical energy, in excess of the reactor containment's capacity. The CADOR approach based on reinforced Doppler reactivity feedback is shown to be an attractive means of effectively preventing such reactivity insertion accidents. The accrued Doppler feedback is achieved by combining two effects: (i) introducing a neutron moderator material in the core so as to soften the neutron spectrum; and (ii) lowering the fuel temperature in nominal conditions so as to increase the margin to fuel melting. This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents. These preliminary results have to be confirmed and completed to meet multiple safety objectives. In particular, some margin gains have to be found to guarantee against the risk of sodium boiling during unprotected loss of supply power accidents. The main drawback of the CADOR concept is a drastically reduced core power density compared to conventional designs. This has a large impact on core size and other parameters
Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the JEFF-3.1.1 library by using integral data assimilation
No full textComparisons of calculated and experimental isotopic compositions of used nuclear fuels can provide valuable information on the quality of nuclear data involved in neutronic calculations. The experimental database used in the present study − containing more than a thousand isotopic ratio measurements for UOX and MOX fuels with burnup ranging from 10 GWd/t up to 85 GWd/t − allowed to investigate 45 isotopic ratios covering a large number of actinides (U, Np, Pu, Am and Cm) and fission products (Nd, Cs, Sm, Eu, Gd, Ru, Ce, Tc, Mo, Ag and Rh). The Integral Data Assimilation procedure implemented in the CONRAD code was used to provide nuclear data trends with realistic uncertainties for Pressurized Water Reactors (PWRs) applications. Results confirm the quality of the 235U, 239Pu and 241Pu neutron capture cross sections available in the JEFF-3.1.1 library; slight increases of +1.2 ± 2.4%, +0.5 ± 2.2% and +1.2 ± 4.2% are respectively suggested, these all being within the limits of the quoted uncertainties. Additional trends on the capture cross sections were also obtained for other actinides (236U, 238Pu, 240Pu, 242Pu, 241Am, 243Am, 245Cm) and fission products (103Rh, 153Eu, 154Eu) as well as for the 238U(n,2n) and 237Np(n,2n) reactions. Meaningful trends for the cumulative fission yields of 144Ce, 133Cs, 137Cs and 106Ru for the 235U(nth,f) and 239Pu(nth,f) reactions are also reported
Heat exchanger design studies for molten salt fast reactor
No full textIn this study, conceptual design for primary heat exchanger of the Molten Salt Fast Reactor is made. The design was carried out to remove the produced heat from the reactor developed under the SAMOFAR project. Nominal power of the reactor is 3 GWth and it has 16 heat exchangers. There are several requirements related to the heat exchanger. To sustain the steady-state conditions, heat exchangers have to transfer the heat produced in the core and it has to maintain the temperature drop as much as the temperature rise in the core due to the fission. It should do it as fast as possible. It must also ensure that the fuel temperature does not reach the freezing temperature to avoid solidification. In doing so, the fuel volume in the heat exchanger must not exceed the specified limit. Design studies were carried out taking into account all requirements and final geometric configurations were determined. Plate type heat exchanger was adopted in this study. 3D CFD analyses were performed to investigate the thermal-hydraulic behavior of the system. Analyses were made by ANSYS-Fluent commercial code. Results are in a good agreement with limitations and requirements specified for the reactor designed under the SAMOFAR project
Breed-and-burn fuel cycle in molten salt reactors
No full textThe operation of a reactor on an open but self-sustainable cycle without actinide separation is known as breed-and-burn. It has mostly been envisioned for use in solid-fueled fast-spectrum reactors such as sodium-cooled fast reactors. In this paper the applicability of breed-and-burn to molten salt reactors is investigated first on a cell level using a modified neutron excess method. Several candidate fuel salts are selected and their performance in a conceptual three-dimensional reactor is investigated. Chloride-fueled single-fluid breed-and-burn molten salt reactors using enriched chlorine are shown to be feasible from a neutronics and fuel cycle point of view at the cost of large fuel inventories
How to obtain an enhanced extended uncertainty associated with decay heat calculations of industrial PWRs using the DARWIN2.3 package
No full textDecay heat is a crucial issue for in-core safety after reactor shutdown and the back-end cycle. An accurate computation of its value has been carried out at the CEA within the framework of the DARWIN2.3 package. The DARWIN2.3 package benefits from a Verification, Validation and Uncertainty Quantification (VVUQ) process. The VVUQ ensures that the parameters of interest computed with the DARWIN2.3 package have been validated over measurements and that biases and uncertainties have been quantified for a particular domain. For the parameter “decay heat”, there are few integral experiments available to ensure the experimental validation over the whole range of parameters needed to cover the French reactor infrastructure (fissile content, burnup, fuel, cooling time). The experimental validation currently covers PWR UOX fuels for cooling times only between 45 minutes and 42 days, and between 13 and 23 years. Therefore, the uncertainty quantification step is of paramount importance in order to increase the reliability and accuracy of decay heat calculations. This paper focuses on the strategy that could be used to resolve this issue with the complement and the exploitation of the DARWIN2.3 experimental validation
Stimulating niche nurturing process for heat production with nuclear plants in France: A multi-level perspective
No full textThis paper examines how intermediaries could interact with other important actors identified by the multi-level perspective (MLP) framework, the niche actors and regime actors, to create niches for nuclear heat production in France. Whatever is the source, recovering the wasted heat is a matter of energy efficiency. Nuclear plants could remain used for several decades in France. It is thus legitimate to investigate those possible niche nurturing processes which may allow a more efficient use of this technology. Challenges are high, and our conclusions modest regarding the possible breaking through of such exploratory and collective systems. Without significant windows of opportunity, even the most willing intermediation may not be able to change the status quo. It is however important to highlight the multifarious pathways that energy transitions could follow. Drawing on lessons from the MLP, this paper proposes three key actions for intermediation willing to move beyond technology-push approaches that can lead to tension and low legitimacy. These are, sharing questions instead of knowledge; mobilise, interest, involve a legitimate place; and prevent or avoid conflicts among stakeholders. Regime changes possibly enhancing the deployment of sustainable heating systems, not only nuclear plant sourced, are also discussed
Pin to pin neutron flux reconstruction in a PWR reactor using support vector regression (SVR) technique
No full textCoarse mesh nodal methods are widely used in the analysis of nuclear reactors. However, these methods provide only average values of the neutron fluxes. From a safety point of view, it is important to have an accurate analysis of the pin to pin flux distribution that nodal methods are not able to provide. Many articles have been published that make use of mathematical techniques to determine flux distributions. Most of these techniques use expansion functions to estimate these distributions. The expansion coefficients of these works are determined by conditions that take into account the average values of certain fluxes supplied by the nodal methods. There are also methods that employ analytical solutions of the neutron diffusion equation. This article presents a different approach for calculating the pin to pin neutron flux distribution for a PWR reactor. The developed method uses support vector regression (SVR) technique to determine this pin to pin neutron flux. The SVR technique uses average data computed with the Nodal Expansion Method (NEM) for learning purposes. A total of 70% of the computed data were used for training and 30% for validation, using multifold-cross-validation. Two fuel elements were removed from the training and validation sets, to test the method. Less than 2% errors were found when compared to the values obtained by the nodal expansion method (NEM), using a fine-mesh spatial discretization. We concluded that use of SVR to reconstruct pin to pin fluxes is another option, which will be of great value in fuel reload calculations, since the same parameters will be applied to all cycles, thus expediting calculations when compared to standard procedure calculations
Influence of molten salt-(FLiNaK) thermophysical properties on a heated tube using CFD RANS turbulence modeling of an experimental testbed
No full textIn a liquid fuel molten salt reactor (MSR) a key factor to consider upon its design is the strong coupling between different physics present such as neutronics, thermo-mechanics and thermal-hydraulics. Focusing in the thermal-hydraulics aspect, it is required that the heat transfer is well characterized. For this purpose, turbulent models used for FLiNaK flow must be valid, and its thermophysical properties must be accurately described. In the literature, there are several expressions for each material property, with differences that can be significant. The goal of this study is to demonstrate and quantify the impact that the uncertainty in thermophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, computational fluid dynamics (CFD) simulations using the RANS k-ω SST model were compared to published experiment data on molten salt. Various correlations for FLiNaK’s material properties were used. It was observed that the uncertainty in FLiNaK’s thermophysical properties lead to a significant variance in the heat coefficient. Motivated by this, additional CFD simulations were done to obtain sensitivity coefficients for each thermophysical property. With this information, the effect of the variation of each one of the material properties on the heat transfer coefficient was quantified performing a one factor at a time approach (OAT). The results of this sensitivity analysis showed that the most critical thermophysical properties of FLiNaK towards the determination of the heat transfer coefficient are the viscosity and the thermal conductivity. More specifically the dimensionless sensitivity coefficient, which is defined as the percent variation of the heat transfer with respect to the percent variation of the respective property, was −0.51 and 0.67 respectively. According to the different correlations, the maximum percent variations for these properties is 18% and 26% respectively, which yields a variation in the predicted heat transfer coefficient as high as 9% and 17% for the viscosity and thermal conductivity, respectively. It was also demonstrated that the Nusselt number trends found from the simulations were captured much better using the Sieder Tate correlation than the Dittus Boelter correlation. Future work accommodating additional turbulence models and higher fidelity physics will help to determine whether the Sieder Tate expression truly captures the physics of interest or whether the agreement seen in the current work is simply reflective of the single turbulence model employed
Development of a control-oriented power plant simulator for the molten salt fast reactor
No full textIn this paper, modelling and simulation of a control-oriented plant-dynamics tool for the molten salt fast reactor (MSFR) is presented. The objective was to develop a simulation tool aimed at investigating the plant response to standard control transients, in order to support the system design finalization and the definition of control strategies. The simulator was developed employing the well tested, flexible and open-source object-oriented Modelica language. A one-dimensional modelling approach was used for thermal-hydraulics and heat transfer. Standard and validated thermal-hydraulic Modelica libraries were employed for various plant components (tubes, pumps, turbines, etc.). An effort was spent in developing a new MSR library modelling the 1D flow of a liquid nuclear fuel, including an ad-hoc neutron-kinetics model which properly takes into consideration the motion of the Delayed Neutron Precursors along the fuel circuit and the consequent reactivity insertion due to the variation of the effective delayed fractions. An analytical steady-state 2-D model of the core and the fuel circuit was developed using MATLAB in order to validate the Decay Neutron Precursors model implemented in the plant simulator. The plant simulator was then employed to investigate the plant dynamics in response to three transients (variation of fuel flow rate, intermediate flow rate and turbine gas flow rate) that are relevant to control purposes. Simulation outcomes highlight the typical reactor-follows-turbine behavior of the MSFR, and they show the small influence of fuel and intermediate flow rate on the reactor power and their strong effects on the temperatures in their respective circuits. Starting from the insights on the reactor behavior gained from the analysis of its free dynamics, the plant simulator here developed will provide a valuable tool in support to the finalization of the design phase, the definition of control strategies and the identification of controlled operational procedures for reactor startup and shutdown
Application of the lines of defence method to the molten salt fast reactor in the framework of the SAMOFAR project
No full textThe Molten Salt Fast Reactor (MSFR) with its liquid circulating fuel and its fast neutron spectrum calls for a new safety approach and adaptation of the analysis tools. In the frame of the Horizon2020 program SAMOFAR (Safety Assessment of the Molten Salt Fast Reactor), a safety approach suitable for Molten Salt Reactors has been developed and is now applied to the MSFR. For this purpose, the Lines of Defence (LoD) method is selected to drive the design consistently with the Defence in Depth principle. This paper presents the main characteristics of the method, along with some practical guidelines to apply it to the specific case of the MSFR; moreover, some initiating events are analyzed through the implementation of the LoD tool. The outcomes of this analysis drive the design evolution