Abstract DEAP-3600 is a single-phase liquid argon (LAr) direct-detection dark matter experiment, operating 2 km underground at SNOLAB (Sudbury, Canada). The detector consists of 3.3 tons of LAr ...contained in a spherical acrylic vessel. At WIMP mass of 100 GeV, DEAP-3600 has a projected sensitivity of 10 -46 cm 2 for the spin independent elastic scattering cross section of WIMPs. Radioactive sources have been used for the energy calibration and to test the detector performance. One of the most effective calibration run has been taken with a 22 Na source deployed in a tube located around the DEAP-3600 steel shell. The simultaneous emission of three γ s by the source provides an excellent tagging for the 22 Na decay. The results concerning the energy response of the detector and the agreement between data and Monte Carlo simulations in DEAP-3600 are investigated in this study.
Bridging lower length-scale calculations with the engineering-scale simulations of fuel performance codes requires the development of dedicated intermediate-scale codes. In this work, we present ...SCIANTIX, an open source 0D stand-alone computer code designed to be included/coupled as a module in existing fuel performance codes. The models currently available in SCIANTIX cover intra- and inter-granular inert gas behaviour in UO2, and high burnup structure formation as well. Showcases of validation in both constant and transient conditions are presented in this work. As for the numerical treatment of the model equations, SCIANTIX is developed with full numerical consistency and entirely verified with the method of manufactured solutions – verification of different numerical solvers is also showcased in this work.
•The characteristics of the SCIANTIX computer code are described.•The models currently available in SCIANTIX are detailed (with all the parameters).•The verification of the numerical solvers is presented.•Showcases of validation in both constant and transient conditions are presented.•Future (short and long term) development plans are outlined.
•Investigation of fuel compressibility effects on the Molten Salt Fast Reactor (MSFR) dynamics.•Modelling of the MSFR helium bubbling system.•Development of a coupled neutronics and fluid dynamics ...model for the MSFR.•Modelling of both liquid fuel and helium bubbles as compressible fluids.•Effects on compressibility due to presence and distribution of helium bubbles are investigated.
Compressible fluid dynamics is of great practical interest in many industrial applications, ranging from chemistry to aeronautical industry, and to nuclear field as well. At the same time, modelling and simulation of compressible flows is a very complex task, requiring the development of specific approaches, in order to describe the effect of pressure on the fluid velocity field. Compressibility effects become even more important in the study of two-phase flows, due to the presence of a gaseous phase. In addition, compressibility is also expected to have a significant impact on other physics, such as chemical or nuclear reactions occurring in the mixture. In this perspective, multiphysics represents a useful approach to address this complex problem, providing a way to catch all the different physics that come into play as well as the coupling between them.
In this work, a multiphysics model is developed for the analysis of the generation IV Molten Salt Fast Reactor (MSFR), with a specific focus on the compressibility effects of the fluid that acts as fuel in the reactor. The fuel mixture compressibility is expected to have an important effect on the system dynamics, especially in very rapid super-prompt-critical transients. In addition, the presence of a helium bubbling system used for online fission product removal could modify the fuel mixture compressibility, further affecting the system transient behaviour. Therefore, the MSFR represents an application of concrete interest, inherent to the analysis of compressibility effects and to the development of suitable modelling approaches. An OpenFOAM solver is developed to handle the fuel compressibility, the presence of gas bubbles in the reactor as well as the coupling between the system neutronics and fluid dynamics. The outcomes of this analysis point out that the fuel compressibility plays a crucial role in the evolution of fast transients, introducing delays in the expansion feedbacks that strongly affect the system dynamics. Moreover, it is found that the gas bubbles significantly alter the fuel compressibility, yielding even larger differences compared to the incompressible approximation usually adopted in the current MSFR solvers.
•A multiphysics model for the MSFR is extended with an SP3 neutron transport module.•The new module is successfully verified against Monte Carlo simulations.•A strong dependence of the void ...coefficient on the bubble distribution is found.•Differences are highlighted compared to a neutron diffusion approach.•The SP3 runtimes are only 17% higher compared to a diffusion module.
The aim of this paper is the extension of a multiphysics OpenFOAM solver for the analysis of the Molten Salt Fast Reactor (MSFR), developed in previous works (Cervi et al., 2017, 2018). In particular, the neutronics sub-solver is improved by implementing a new module based on the SP3 approximation of the neutron transport equation. The new module is successfully tested against a Monte Carlo model of the MSFR, in order to assess its correct implementation. Then, a neutronics analysis of the MSFR is carried out on a simplified axial-symmetric model of the reactor. Particular focus is devoted to the analysis of the MSFR helium bubbling system and its effect on reactivity. The presence of bubbles inside the reactor is handled with a two-fluid thermal-hydraulics module, previously implemented into the solver. The void reactivity coefficient is evaluated on the basis of the bubble spatial distribution calculated by the multiphysics solver. Then, the results are compared to simulations carried out with uniform bubble distributions, highlighting significant differences between the two approaches. The outcomes of this work constitute a step forward in the multiphysics analysis of the Molten Salt Fast Reactor and represent a useful starting point for the optimization of the MSFR helium bubbling system, as well as for the development of appropriate control strategies.
We propose a model describing the HBS formation and the progressive intra-granular xenon depletion in UO2. The HBS formation is modeled employing the Kolmogorov-Johnson-Mehl-Avrami (KJMA) formalism ...for phase transformations, which has been fitted to experimental data on the restructured volumetric fraction as a function of the local effective burnup. To this end, we employed available experimental data and novel data extracted in this work. The HBS formation model is coupled to a description of the intra-granular fission gas behavior, allowing to estimate the evolution of the retained xenon in order to consistently compute fission gas retention and its effect on the fuel matrix swelling. The satisfactory agreement of the model predictions to experimental data and state-of-the-art models’ results, in terms of both xenon depletion and fuel matrix swelling as a function of the local burnup, paves the way to the inclusion of the model in fuel performance codes.
The description of intra-granular fission gas behavior during irradiation is a fundamental part of models used for the calculation of fission gas release and gaseous swelling in nuclear fuel ...performance codes. The relevant phenomena include diffusion of gas atoms towards the grain boundaries coupled to the evolution of intra-granular bubbles. While intra-granular bubbles during normal operating conditions are limited to sizes of a few nanometers, experimental evidence exists for the appearance of a second population of bubbles during transients, characterized by coarsening to sizes of tens to hundreds of nanometers and that can significantly contribute to gaseous fuel swelling. In this work, we present a model of intra-granular fission gas behavior in uranium dioxide fuel that includes both nanometric fission gas bubble evolution and bubble coarsening during transients. While retaining a physical basis, the developed model is relatively simple and is intended for application in engineering fuel performance codes. We assess the model through comparisons to a substantial number of experimental data from SEM observations of intra-granular bubbles in power ramp tested uranium dioxide samples. The results demonstrate that the model reproduces the coarsening of a fraction of the intra-granular bubbles and correspondingly, predicts gaseous swelling during power ramps with a significantly higher accuracy than is allowed by traditional models limited to the evolution of nanometric intra-granular bubbles.
The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be ...particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.
•A model for transient fission gas behaviour including burst release is presented and assessed for fuel rod analysis.•Analytic sensitivity analysis is performed to evaluate the effect of model parameters on a physical figure of merit.•The same model is implemented in two fuel performance codes and assessed against 19 LWR fuel rod experiments.•Results with the transient model are more accurate than with the canonical models, for both codes.•Application in two structurally different codes isolates the effect of the specific model from the global analysis.
In this paper, the study of the heat exchange effect on the dynamic behaviour of single-phase natural circulation with internal heat generation is presented. In order to predict natural circulation ...instabilities, two different methods of analysis are developed and compared. The first approach is a linear analysis in which the governing equations are firstly linearized around a steady-state solution of the system and then treated by means of the Fourier transform. This strategy is adopted to compute, in a semi-analytical way, dimensionless stability maps for different system configurations, highlighting the heat exchange effect on the system dynamics. The second approach consists in numerically solving the nonlinear governing equations and allows investigating some transients of interest. For this purpose, an object-oriented one-dimensional model of natural circulation loops has been developed, and the corresponding results have been compared with RELAP5 and Computational Fluid-Dynamics (CFD) time-dependent simulations. The developed models have been applied to investigate the dynamic behaviour of two loop configurations characterized by large instability regions, namely the Horizontal Heater–Horizontal Cooler (HHHC) and the Vertical Heater–Horizontal Cooler (VHHC).
•1-D analytical and numerical approaches to natural circulation loop (NCL) dynamics.•Internal Heat Generation (IHG) alters the NCL stability properties.•Unstable regions of stability maps decrease when a variable heat transfer coefficient is considered.•The VHHC and HHHC loop configurations can be characterized by large instability regions.•CFD is a useful support to investigate NCL dynamics.
•Modelling of thermal conductivity and melting temperature of minor actinide-MOX fuels.•Review of state-of-the-art data and correlations available in literature or in codes.•Inclusion of the effect ...of the homogeneous minor actinides Am and Np.•Separate-effect assessment against literature data, experimental and MD-calculated.•Integral application to the performance of a fast reactor, MA-MOX fuel pin.
Recycling and burning minor actinides (MA, e.g., americium, neptunium) in mixed-oxide (MOX) nuclear fuel is a strategic option for fast reactor concepts of Generation IV, especially considering the current interest in the ultimate radioactive waste management and sustainability improvement by better use of natural resources. Among the fuel properties, thermal conductivity and melting temperature are pivotal since they determine, respectively, the fuel temperature profile and the fundamental safety limit on the margin to fuel melting, hence impacting on the overall fuel performance under irradiation and allowing the safe irradiation of the fuel pin. Nevertheless, the available literature about Am- or Np-containing MOX is currently scarce, both regarding experimental data and models. Moreover, state-of-the-art fuel performance codes (FPCs, e.g., TRANSURANUS) do not account for the effects of minor actinides on MOX fuel properties. This work presents original correlations for thermal conductivity and melting temperature of minor actinide-MOX fuels, i.e., (U, Pu, Am, Np)O2-x, derived based on the available literature and accessible data, which are herein extensively reviewed. The assessment of the novel correlations is first performed in a statistical way, evaluating the regressor p-values which indicate their significance with respect to the available experimental dataset used for the fitting procedure. Additionally, the novel correlations for MA-MOX are assessed against both measured and calculated data (from Molecular Dynamics simulations), yielding an accuracy in line with the already existing correlations and with the state-of-the-art experimental uncertainties. Finally, the potential integral impact of a homogeneous minor actinide content in the fuel is illustrated on the basis of a fuel pin fast-ramped up to fuel melting during the HEDL P-19 irradiation experiment.
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