Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Decay Heat Removal (DHR) is a fundamental safety function which is often accomplished in the advanced LWRs relying on natural phenomena. A typical passive DHR system is the two-phase flow, natural circulation, closed loop system, where heat is removed by means of a steam generator or heat exchanger, a condenser, and a pool. Different condenser tube arrangements have been developed for applications to the next generation NPPs. The two most used configurations, namely, horizontal and vertical tube condensers, are thoroughly investigated in this paper. Several thermal-hydraulic features were explored, being the analysis mainly devoted to the description of the best-estimate correlations and models for heat transfer coefficient prediction. In spite of a more critical behaviour concerning thermal expansion issues, vertical tube condensers offer remarkably better thermal-hydraulic performances. An experimental validation of the vertical tube correlations is provided by PERSEO facility (SIET labs, Piacenza), showing a fairly good agreement

Numerical Investigation of the Performance of a Solar Air Heater Equipped with a Packed Bed

In this work, a numerical study based on pore-scale CFD analysis has been carried out on a solar air heater equipped with intermittent packed beds. Hydraulic and thermal characteristics of the air were determined by testing a range of mass flow rates and various collector configurations, including one (SAH-M1), two (SAH-M2), and three (SAH-M3) packed beds. After the validation of the numerical simulations with general correlations, results showed that as the number of packed zones increases, the total pressure drop grows. As a result, the maximum pressure drop was obtained as 115 kPa for the collector with three packed zones, operating at Re ~ 1.3 × 105. Detailed thermal analysis proved that the intermittent integration of packed beds has the potential to also increase the heat transfer exchange between each packed zone, improving the convection by 25% in the second zone to the first zone and 35% in the third zone to the second zone at Re ~ 1.3 × 105. This enhancement effect can be attributed to the increase in the turbulence conditions of the fluid regime from one bed to the other. Evaluating the effects of design parameters showed that the bed spacing could not change the thermo-hydraulic characteristics significantly, while particles’ randomness may only affect the pressure drop by 13%

Thermo-Hydrodynamics of Internally Heated Molten Salts for Innovative Nuclear Reactors

The problem of heat transfer in pipe flow has been extensively investigated in the past. Many different models have been proposed and adopted to predict the velocity profile, the eddy diffusivity, the temperature distributions, the friction factor and the heat transfer coefficient (Kays et al., 2004; Schlichting & Gersten, 2000). However, the majority of such studies give a description of the problem for non-internally heated fluids. Models regarding fluids with internal heat generation have been developed more than 50 years ago (Kinney & Sparrow, 1966; Poppendiek, 1954; Siegel & Sparrow, 1959), giving in most cases a partial treatment of the problem in terms of boundary conditions and heat source distribution, and relying on a turbulent flow treatment that does not seem fully satisfactory in the light of more recent investigations (Churchill, 1997; 2002; Kays, 1994; Zagarola & Smits, 1997). Internally heated fluids are of great interest in the current development of Molten Salt Reactors (MSR) (LeBlanc, 2010), included as one of the six innovative nuclear reactors selected by the Generation IV International Forum (GIF-IV, 2002) for a more sustainable version of nuclear power. MSRs are circulating fuel reactors (Nicolino et al., 2008), which employ a non-classical (fluid-type) fuel constituted by a molten halide (fluoride or chloride) salt mixture playing the distinctive role of both heat source and coolant. By adopting classical correlations for the Nusselt number (e.g., Dittus-Boelter), the heat transfer coefficient of the MSR fuel can be overestimated by a non-negligible amount (Di Marcello et al., 2008). In the case of thermal-neutron-spectrum (graphite-moderated) MSRs (LeBlanc, 2010), this has significant consequences on the core temperature predictions and on the reactor dynamic behaviour (Luzzi et al., 2011). Such influence of the heat source within the fluid cannot be neglected, and thus required proper investigation. The present chapter deals with this critical issue, first summarizing the main modelling efforts carried out by the authors (Di Marcello et al., 2010; Luzzi et al., 2010) to investigate the thermo-hydrodynamics of internally heated fluids, and then focusing on the heat transfer coefficient prediction that is relevant for analysing the molten salt behaviour encountered in MSRs. The chapter is organized as follows. Section 2 provides a brief description of the Molten Salt Reactors, focusing on their distinctive features, in terms of both sustainability (i.e., reduced radioactive waste generation, effective use of natural resources) and safety, with respect to the traditional configuration of nuclear reactors. Section 3 deals with the study of molten salt heat transfer characteristics, which represent a key issue in the current development of MSRs. In particular, a "generalized approach" to evaluate the steady-state temperature distribution in a representative power channel of the reactor core is presented. This approach incorporates recent formulations of turbulent flow and convection (Churchill, 1997; 2002), and is built in order to carefully take into account the molten salt mixture specificities, the reactor core power conditions and the heat transfer in the graphite core structure. In Section 4, a preliminary correlation for the Nusselt number prediction is advanced in the case of simultaneous uniform wall heat flux and internal heat generation, on the basis of the results achieved by means of the presented generalized approach. In Section 5, the main conclusions of the present study are summarized

Analysis of Different Containment Models for IRIS Small Break LOCA, using GOTHIC and RELAP5 Codes

Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work IRIS reactor was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a Small Break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. IRIS containment drywell was modelled with RELAP according to a sliced approach, based on the two-pipe-with-junction concept, while it was simulated with GOTHIC testing several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, Heat Transfer Coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flowrate. The objective of the paper is to compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix. The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due e.g. to a marked overestimation of internal natural recirculation, RELAP confirmed its capability to satisfactorily model the IRIS containment

On Density Wave Instability Phenomena – Modelling and Experimental Investigation

Density Wave Oscillations (DWOs) are dealt with in this work as the most representative instabilities frequently encountered in the boiling systems. This dynamic type instability mode – resulting from multiple feedback effects between the flow rate, the vapour generation rate and the pressure drops in the boiling channel – constitutes an issue of special interest for the design of industrial systems and equipments involving vapour generation. The chapter is structured as follows. Physical insight into the distinctive features leading to DWO mechanism is provided in Section 2. Modelling and experimental investigations on instability phenomena available from the open literature are described in Section 3. Section 4 and 5 present the analytical modelling developed in this work for DWO theoretical predictions, whereas numerical modelling (using RELAP5 and COMSOL codes) is briefly discussed in Section 6. Modelling efforts start necessarily from the simplifying and sound case of straight vertical tube geometry, which is referenced for validating the whole modelling tools. Description of the experimental campaign for DWO characterization in helical coil tubes is shortly presented in Section 7. The peculiar influence of the helical shape on the instability occurrence is examined in Section 8. Suited modifications of the models are introduced in order to simulate the experimental results

Erratum to: A Serpent/OpenFOAM coupling for 3D burnup analysis

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Development of a Reduced Order Model for Fuel Burnup Analysis

Fuel burnup analysis requires a high computational cost for full core calculations, due to the amount of the information processed for the total reaction rates in many burnup regions. Indeed, they reach the order of millions or more by a subdivision into radial and axial regions in a pin-by-pin description. In addition, if multi-physics approaches are adopted to consider the effects of temperature and density fields on fuel consumption, the computational load grows further. In this way, the need to find a compromise between computational cost and solution accuracy is a crucial issue in burnup analysis. To overcome this problem, the present work aims to develop a methodological approach to implement a Reduced Order Model (ROM), based on Proper Orthogonal Decomposition (POD), in fuel burnup analysis. We verify the approach on 4 years of burnup of the TMI-1 unit cell benchmark, by reconstructing fuel materials and burnup matrices over time with different levels of approximation. The results show that the modeling approach is able to reproduce reactivity and nuclide densities over time, where the accuracy increases with the number of basis functions employed

A new model with Serpent for the first criticality benchmarks of the TRIGA Mark II reactor

We present a new model, developed with the Serpent Monte Carlo code, for neutronics simulation of the TRIGA Mark II reactor of Pavia (Italy). The complete 3D geometry of the reactor core is implemented with high accuracy and detail, exploiting all the available information about geometry and materials. The Serpent model of the reactor is validated in the fresh fuel configuration, through a benchmark analysis of the first criticality experiments and control rods calibrations. The accuracy of simulations in reproducing the reactivity difference between the low power (10 W) and full power (250 kW) reactor condition is also tested. Finally, a direct comparison between Serpent and MCNP simulations of the same reactor configurations is presented