Post-test simulations for the NACIE-UP benchmark by STH codes

This paper illustrates the results obtained in the last phase of the NACIE-UP benchmark activity foreseen inside the EU SESAME Project. The purpose of this research activity, performed by system thermal–hydraulic (STH) codes, is finalized to the improvement, development and validation of existing STH codes for Heavy Liquid Metal (HLM) systems. All the participants improved their modelling of the NACIE-UP facility, respect to the initial blind simulation phase, adopting the actual experimental boundary conditions and reducing as much as possible sources of uncertainty in their numerical model. Four different STH codes were employed by the participants to the benchmark to model the NACIE-UP facility, namely: CATHARE for ENEA, ATHLET for GRS, RELAP5-3D© for the “Sapienza” University of Rome and RELAP5/Mod3.3(modified) for the University of Pisa. Three reference tests foreseen in the NACIE-UP benchmark and carried out at ENEA Brasimone Research Centre were analysed from four participants. The data from the post-test analyses, performed independently by the participant using different STH codes, were compared together and with the available experimental results and critically discussed

Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project

Since some years, there is a worldwide trend to move towards “higher-fidelity” simulation techniques in reactor analysis. One of the main objectives of the research in this area is to enhance the prediction capability of the computations used for safety demonstration of the current LWR nuclear power plants through the dynamic 3D coupling of the codes simulating the different physics of the problem into a common multi-physics simulation scheme. In this context, the NURESAFE European project aims at delivering to the European stakeholders an advanced and reliable software capacity usable for safety analysis needs of present and future LWR reactors and developing a high level of expertise in Europe in the proper use of the most recent simulation tools including uncertainty assessment to quantify the margins toward feared phenomena occurring during an accident. This software capacity is based on the NURESIM European simulation platform created during FP6 NURESIM project which includes advanced core physics, two-phase thermal–hydraulics, fuel modeling and multi-scale and multi-physics features together with sensitivity and uncertainty tools. These physics are fully integrated into the platform in order to provide a standardized state-of-the-art code system to support safety analysis of current and evolving LWRs

Multi-Scale Thermal-hydraulic Developments for the Detailed Analysis of the Flow Conditions within the Reactor Pressure Vessel of Pressurized Water Reactors

Die mehrskalige thermohydraulische Analyse und die Entwicklung von mehrskaligen thermohydraulisch gekoppelten Programmen haben sich in den letzten Jahren zu einem vielversprechenden Gebiet im Bereich der Reaktortechnik entwickelt. Sie zielen darauf ab, die Fähigkeiten der thermohydraulischen Simulationswerkzeuge zu verbessern und die thermohydraulischen Phänomene in den Kernkraftsystemen umfassender zu beschreiben. Die mehrskalige thermohydraulische Simulation eines Druckwasserreaktors (PWR) bildet den Schwerpunkt dieser Arbeit. Eine generische Klassifizierung der verschiedenen Multi-Skalen-Kopplungsansätze wird vorgeschlagen und die gekoppelten Programme und Methoden werden verglichen. Die Dissertation entwickelt zwei mehrskalige thermohydraulische Kopplungssysteme: 1) die Verbindung des Systemcodes TRACE der US Nuclear Regulatory Commission (NRC) mit dem Unterkanalprogramm SubChanFlow (SCF) des Karlsruher Instituts für Technologie (KIT) unter Verwendung einer externen Kommunikationsschnittstelle (ECI); 2) die Verbindung des US-amerikanischen NRC-Systemcodes TRACE mit dem französischen Open-Source-CFD-Code TrioCFD unter Verwendung der ICoCo-Methode (Interface for Code Coupling). Die Kopplung TRACE / SCF-ECI wurde als serverloses, paralleles und explizites Kopplungssystem entwickelt, das die Methode der Domänenzerlegung anwendet. Ein neu entwickeltes Toolkit löst dabei die Feldzuordnungsprobleme. Dieses System wurde anhand eines akademischen Kühlmittelmischproblems für einen VVER-1000 als Referenz verifiziert und validiert. Es zeigt sich, dass der gekoppelte Code die Vermischung des Kühlmittels im Reaktordruckbehälter genauer vorhersagen kann. Darüber hinaus wurden die gekoppelten Codes optimiert, um effizienter zu arbeiten. Die Kopplung von TRACE / TrioCFD mithilfe von ICoCo wurde als explizites Server-Client-Kopplungssystem entwickelt, das die domänenübergreifende Methode anwendet und die SALOME MED-Kopplungs-Bibliothek verwendet, um die Feldzuordnung und den Datentransfer zwischen verschiedenen Volumenzellen zu verwalten. Eine neuartige DIAS-Methode (Dynamic-Implicit-Additional-Source) wurde implementiert. Dabei werden die von der MED-Bibliothek aus TrioCFD übersetzten Felder für die Kühlmittelgeschwindigkeit, den Druck, die Kühlmitteltemperatur und die Borkonzentration verwendet, um die vier entsprechenden Felder in TRACE abzubilden. Die Übertragung findet in der gesamten überlappenden Domäne statt. Das Ergebnis wird mit dem VVER-1000-Referenzwert für die Kühlmittelvermischung verglichen, und die verbesserte Fähigkeit der Codes, das Mischen von Kühlmittel im Reaktordruckbehälter vorherzusagen, wird aufgezeigt

Current State of Research on Pressurized Water Reactor Safety

For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study – loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. –, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today’s reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

Reactor antineutrino fluxes - status and challenges

In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.Comment: 10 pages, 2 figures, 3 tables, prepared for the Nuclear Physics B special issue on Neutrino Oscillations celebrating the Nobel Prize in Physics 201

Plenum-to-plenum heat transfer characteristics under natural circulation in a scaled-down prismatic modular reactor

“Gas-cooled reactor (GCR) is being developed under the Next Generation Nuclear Plant Program (NGNP) in nuclear engineering studies. As the world searches for an energy source with high energy density, clean, abundant, and storable nature to avoid global warming issues, GCR seems to be a promising solution, particularly the possibility of producing hydrogen. Studying and developing the safety analysis and GCR technologies are required for the optimum design and safety of GCR system. Multiphase Reactors Engineering and Applications Laboratory (mReal) at Missouri University of Science and Technology (S&T) has developed a natural convection heat transfer test facility with one riser and one downcomer between two plena to investigate loss of flow accident scenario (LOFA) for a prismatic very high temperature reactor (VHTR). Using advanced heat transfer coefficient probe and T-thermocouples. The facility represents a scaled down prismatic modular reactor with a reference to High-Temperature Test Facility at Oregon State University (OSU-HTTF). The natural circulation heat transfer in terms of temperature fields and heat transfer coefficients across the core of current facility (i.e., channels) has been investigated at different conditions to understand the passive safety features of prismatic modular reactors (PMR). This dissertation advance knowledge and understanding of the intra core natural circulation phenomenon in the PMRs. Moreover, fill the gaps in the open literature and understanding of the natural circulation heat transfer and gaseous dynamics in the prismatic VHTR and provide experimental benchmark data, which is much needed and is missing in the literature for verification and validation”--Abstract, page iii