• Energy Research

Abstract Heavy-liquid metals, such as lead and lead–bismuth eutectic, are promising candidates as coolant for advanced GEN-IV fast reactors as well as for Accelerator-Driven Systems. The advancing knowledge of the thermal-hydraulic behavior of these fluids leads to explore new geometries and new concepts aimed at optimizing the key components of a GEN-IV reactor for these fluids. In this paper, a theoretical and computational analysis is presented of a jet pump evolving liquid lead as primary pump for ALFRED (Advanced Lead Fast Reactor European Demonstrator). The jet pump is modeled with a 3D CFD code (FLUENT) and at design operating conditions. The analysis shows that a jet pump could be a viable solution for ALFRED (at least from a thermal-hydraulic point of view), albeit some technological issues remain to be fully addressed.

Abstract This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MWth reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

Abstract Lead-cooled fast reactor (LFR) has both a long history and a penchant of innovation. With early work related to its use for submarine propulsion dating to the 1950s, Russian scientists pioneered the development of reactors cooled by heavy liquid metals (HLM). More recently, there has been substantial interest in both critical and subcritical reactors cooled by lead (Pb) or lead–bismuth eutectic (LBE), not only in Russia, but also in Europe, Asia, and the USA. The growing knowledge of the thermal-fluid-dynamic properties of these fluids and the choice of the LFR as one of the six reactor types selected by Generation IV International Forum (GIF) for further research and development has fostered the exploration of new geometries and new concepts aimed at optimizing the key components that will be adopted in the Advanced Lead Fast Reactor European Demonstrator (ALFRED), the 300 MWt pool-type reactor aimed at proving the feasibility of the design concept adopted for the European Lead-cooled Fast Reactor (ELFR). In this paper, a theoretical and computational analysis is presented of a multi-blade screw pump evolving liquid Lead as primary pump for the adopted reference conceptual design of ALFRED. The pump is at first analyzed at design operating conditions from the theoretical point of view to determine the optimal geometry according to the velocity triangles and then modeled with a 3D CFD code (ANSYS CFX). The choice of a 3D simulation is dictated by the need to perform a detailed spatial simulation taking into account the peculiar geometry of the pump as well as the boundary layers and turbulence effects of the flow, which are typically tri-dimensional. The use of liquid Lead impacts significantly the fluid dynamic design of the pump because of the key requirement to avoid any erosion affects. These effects have a major impact on the performance, reliability and lifespan of the pump. Albeit some erosion-related issues remain to be fully addressed, the results of this analysis show that a multi-blade screw pump could be a viable option for ALFRED from a thermo-fluid-dynamic point of view.

Nuclear engineering requires computationally efficient methods to simulate different components and systems of plants. The Lattice Boltzmann Method (LBM), a numerical method with a mesoscopic approach to Computational Fluid Dynamic (CFD) derived from the Boltzmann equation and the Maxwell–Boltzmann distribution, can be an adequate option. The purpose of this paper is to present a review of the recent applications of the Lattice Boltzmann Method in nuclear engineering research. A systematic literature review using three databases (Web of Science, Scopus, and ScienceDirect) was done, and the items found were categorized by the main research topics into computational fluid dynamics and neutronic applications. The features of the problem addressed, the characteristics of the numerical method, and some relevant conclusions of each study are resumed and presented. A total of 45 items (25 for computational fluid dynamics applications and 20 for neutronics) was found on a wide range of nuclear engineering problems, including thermal flow, turbulence mixing of coolant, sedimentation of impurities, neutron transport, criticality problem, and other relevant issues. The LBM results in being a flexible numerical method capable of integrating multiphysics and hybrid schemes, and is efficient for the inner parallelization of the algorithm that brings a widely applicable tool in nuclear engineering problems. Interest in the LBM applications in this field has been increasing and evolving from early stages to a mature form, as this review shows.

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.

Abstract This review article is an overview on some recent developments of Generation IV (GEN-IV) gas cooled nuclear reactor performing in the frame of innovative symbiotic fuel cycle mainly by our research group. In fact such reactors, in addition to energy generation, can also improve nuclear fuel exploitation and decrease nuclear waste radiotoxicity mainly by transmutation processes. Transmutation of Trans-Uranic nuclides (TRU) can be carried on in GEN-IV reactors mainly due to their neutronic behaviour and design features. An optimization of the burning process, by using different reactors chains (symbiotic cycles), has been analyzed in order to allow an efficient TRU transmutation with the aims to reduce waste radiotoxicity, optimize raw material exploitation and close nuclear fuel cycle.

Accelerator Driven Systems (ADS) seem to be a good solution for safe nuclear waste transmutation. One of the most important challenges for this kind of machine is the target design, particularly for what concerning the target cooling system. In order to optimize this component a CFD-based approach has been chosen. After the definition of a reference design (Be target cooled by He), some parameters have been varied in order to optimize the thermal-fluid-dynamic features. The final optimized target design has an increased security margin for what regarding Be melting and reduces the maximum coolant velocity (and consequently even more the pressure drops).

Abstract The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, started at SCK·CEN since 1999, aims at the construction of a pool-type sub-critical Accelerator-Driven System (ADS) which could also operate as a critical reactor. The primary system, enclosed in the primary vessel, is filled with Lead Bismuth Eutectic (LBE) which acts as primary coolant. The power is then delivered through four heat exchangers to four secondary loops. The secondary cooling fluid is two-phase water operating at relatively low pressure (16 bar). Four aero-condensers act as heat sinks, since MYRRHA design does not foresee any electricity generation. The MYRRHA Primary Heat eXchangers (PHXs) cover a role of fundamental importance in normal operation and accidental conditions, being part of the primary and secondary cooling system and of the Decay Heat Removal (DHR) system. It is thus highly relevant to understand the PHXs behavior under all the potential working conditions. In particular, the stability of the PHXs must be guaranteed under all operating conditions. System code models play an important role in understanding and predicting the behavior of the reactor in all conditions, from steady state to operational and accidental transients, and simulating all the postulated scenarios. A solid PHX design requires a complete assessment of two-phase flow instabilities in the secondary system water tube bundle and the potential implementation of a suitable stabilizing device (orifice) to reduce the impact of the perturbations along the channel. The stability assessment should take in consideration all the possible reactor operational power levels in order to prove the stable behavior under all operational conditions. The tube bundle stability assessment has been carried out by following a similar procedure used for BWR fuel channels, through a specific RELAP5-3D model representing the PHX and able to evaluate the propagation of a density wave in the tube length. A series of suitable boundary conditions, on both primary and secondary side, and perturbation triggers have been foreseen into the model, so to discover all kind of unstable behavior and to dimension the needed orifice to guarantee the flow stability in all operating conditions. The PHX stability analysis is initially performed on the original tube bundle without the adoption of any stabilizing devices, in order to check the natural behavior of the system. The possible adoption and design of an orifice is then conducted on the basis of this preliminary study. The system response against the various types of instabilities, before the introduction of an orifice, is not completely satisfactory: a stable flow is found within certain specific system parameters ranges. After the introduction of a suitable orifice, the system behavior becomes stable under all operating conditions against all types of two-phase flow instabilities.

The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.