Analysis of unmitigated large break loss of coolant accidents using MELCOR code

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation

Application of Artificial Intelligence in Detection and Mitigation of Human Factor Errors in Nuclear Power Plants: A Review

Human factors and ergonomics have played an essential role in increasing the safety and performance of operators in the nuclear energy industry. In this critical review, we examine how artificial intelligence (AI) technologies can be leveraged to mitigate human errors, thereby improving the safety and performance of operators in nuclear power plants (NPPs). First, we discuss the various causes of human errors in NPPs. Next, we examine the ways in which AI has been introduced to and incorporated into different types of operator support systems to mitigate these human errors. We specifically examine (1) operator support systems, including decision support systems, (2) sensor fault detection systems, (3) operation validation systems, (4) operator monitoring systems, (5) autonomous control systems, (6) predictive maintenance systems, (7) automated text analysis systems, and (8) safety assessment systems. Finally, we provide some of the shortcomings of the existing AI technologies and discuss the challenges still ahead for their further adoption and implementation to provide future research directions

Thermo-structural analysis of a reactor pressure vessel lower head during core-melt severe accidents

Schwere Störfälle in Kernkraftwerke haben in Anbetracht der implementierten Sicherheitsmaßnahmen sehr geringe Eintrittswahrscheinlichkeit. Sie sind jedoch möglich, wie im Jahr 2011 am Kernkraftwerk Fukushima Daiichi eingetreten. Die Freilassung radioaktiver Stoffe vom versagten Reaktordruckbehälter (RDB) hatte dort massiv Konsequenzen auf die Umgebung. Um radioaktives Material im RDB einzuschließen, wird eine Rückhaltung der Kernschmelze im Behälter (IVR: in-vessel melt retention) durch externe Reaktorbehälterkühlung (ERVC: external reactor vessel cooling) betrachtet als eine vielversprechende Sicherheitsmaßnahme um schwere Störfälle zu begrenzen. Gemäß der derzeitigen Forschungsprioritäten für schwere Störfälle werden folgende Phänomene im unteren Plenum des RDB als wichtig betrachtet, um den Unfallablauf zu erklären: (a) Schmelzeverhalten in unteren Plenum, (b) Integrität des RDBs durch externe Behälterkühlung, und (c) RDB Versagensmodus. Es wird weiterhin ein tieferes Verständnis dieser Phänomene und des Unfallverlaufs gefordert. Die Kosten großer Experimente sind sehr hoch. Sie sind daher nicht wiederholbar. Daher sollen numerische Analysemodelle weiterentwickelt werden, um schwere Unfälle zu analysieren und um Gegenmaßnahmen schwerer Störfälle zu verbessern. Das Ziel dieser Arbeit ist die Bewertung und Verbesserung des Analysemodells für das untere Plenum. RELAP/SCDAPSIM ist ein Reaktorberechnungsprogramm, das weiterhin für Analysen schwerer Unfälle verwendet wird. Es beinhaltet das COUPLE-Modul, das den Wärmeübergang des Schmelzpools im unteren Plenum sowie die Kriechschädigung simuliert. Bisher war die wichtigste Anwendung dieses Moduls die Analyse der Aufheizung des Druckbehälters, so dass Versagens- sowie Schmelzzeit abgeschätzt werden können. In letzter Zeit gewinnt jedoch die Anwendbarkeit des vorliegenden Reaktorberechnungsprogramms auf die Reaktorbehälterkühlung immer größeres Interesse. Dennoch wurden bis jetzt nur wenige Evaluierungen mit Schmelzpoolexperimenten durchgeführt. Die Modifikation des Programms in dieser Arbeit auf die Anwendung auf externe Kühlung ermöglichte die Evaluierung des Codes anhand der LIVE-Experimente des KIT. Das COUPLE-Modul ist dadurch eingeschränkt, dass ein homogener Schmelzepool angenommen wird und dass der Einfluss eines geschichteten Pools nicht erfasst werden kann. Ferner wird nur Kriechschädigung betrachtet und detaillierte mechanische Analysen sind nicht möglich. Das Phase-Change Effective Convectivity Model (PECM) ist ein Spezialmodel für detaillierten Wärmetransport im Schmelzpool, das basierend auf CFD-Untersuchungen entwickelt wurde. Dieses Model verwendet empirische Korrelationen um den konvektiven Wärmetransport in der Energieerhaltungsgleichung abzuschätzen. Eine Simulation des Wärmetransports in einem geschichteten Pool ist ebenfalls möglich. Das Model wurde in OpenFOAM integriert und um mechanische Analysemodelle erweitert, um Wärmeausdehnung, Plastizität, Kriechen sowie Materialschädigung berücksichtigen zu können. Der erweiterte Löser (PECM/S) wurde mit LIVE und FOREVER Experimenten validiert. Mit PECM/S ist es jedoch nicht möglich ein ganzes Unfallszenario aufzulösen. Daher wurden RELAP/SCDAPSIM und PECM/S mit OpenMPI gekoppelt, ein Message Passing Interface, um die Vorteile beider Codes zu nutzen. Das gekoppelte System wurde mit LIVE-Experimenten, die je nach Versuch verschiedenen Heiz- und Kühlungsbedingungen hatten, validiert. Die Ergebnisse wurden ferner mit RELAP/SCDAPSIM Einzelanalysen verglichen und zeigte eine detailliertere und bessere Übereinstimmung mit den Experimenten. Die Anwendung des gekoppelten Systems auf die Simulation eines schweren Störfalls zeigte die Stärken der Kopplung und lässt erwarten, dass die Kopplungsmethode bei Analysen des unteren Plenums des RDBs für schwere Störfällen mit Kernschmelze verwendet werden kann

Fundamentals of 3-D Neutron Kinetics and Current Status

This lecture includes the following topics: 1) A summary of the cell and lattice calculations used to generate the neutron reaction data for neutron kinetics, including the spectral and burn up calculations of LWR cells and fuel assembly lattices, and the main nodal kinetics parameters: mean neutron generation time and delayed neutron fraction; 2) the features of the advanced nodal methods for 3-D LWR core physics, including the treatment of partially inserted control rods, fuel assembly grids, fuel burn up and xenon and samarium transients, and ex core detector responses, that are essential for core surveillance, axial offset control and operating transient analysis; 3) the advanced nodal methods for 3-D LWR core neutron kinetics (best estimate safety analysis, real time simulation); and 4) example applications to 3-D neutron kinetics problems in transient analysis of PWR cores, including model, benchmark and operational transients without, or with simple, thermal-hydraulics feedback

Analysis of Primary/Containment Coupling Phenomena Characterizing the MASLWR Design During a SBLOCA Scenario

Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the “Power Reactor Information System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction (IAEA PRIS, 2011). In the last 20 years, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs, to satisfy the demands of the people to improve the safety of nuclear power plants and the demands of the utilities to improve the economic efficiency and reduce the capital costs (D'Auria et al., 1993; Mascari et al., 2011c). Design simplifications and increased design margins are included in the advanced Light Water Reactors (LWR) (Aksan, 2005). In this framework, the project of some advanced reactors considers the use of emergency systems based entirely on natural circulation for the removal of the decay power in transient condition and in some reactors for the removal of core power during normal operating conditions (IAEA-TECDOC-1624, 2009; Mascari et al., 2010a; Mascari et al., 2011d). For example, if the normal heat sink is not available, the decay heat can be removed by using a passive connection between the primary system and heat exchangers (Aksan, 2005; Mascari et al., 2010a, Mascari, 2010b). The AP600/1000 (Advanced Plant 600/1000 MWe) design, for example, includes a Passive Residual Heat Removal (PRHR) system consisting of a C-Tube type heat exchanger immersed in the In-containment Refueling Water Storage Tank (IRWST) and connected to one of the Hot Legs (HL) (IAEA-TECDOC-1391, 2004; Reyes, 2005c; Gou et al., 2009; Mascari et al., 2010a). A PRHR from the core via Steam Generators (SG) to the atmosphere, considered in the WWER-1000/V-392 (Water Moderated, Water Cooled Energy Reactor) design, consists of heat exchangers cooled by atmospheric air, while the PRHR via SGs, considered in the WWER-640/V-407 design, consists of heat exchangers immersed in emergency heat removal tanks installed outside the containment (Kurakov et al., 2002; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a). In the AC-600 (Advanced Chinese PWR) the PRHR heat exchangers are cooled by atmospheric air (IAEATECDOC 1281, 2002; Zejun et al., 2003; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a) and in the System Integrated Modular Advanced Reactor (SMART) the PRHR heat exchangers are submerged in an in-containment refuelling water tank (IAEA-TECDOC- 1391, 2004; Lee & Kim, 2008; Gou et al., 2009; Mascari et al., 2010a). The International Reactor Innovative and Secure (IRIS) design includes a passive Emergency Heat Removal System (EHRS) consisting of an heat exchanger immersed in the Refueling Water Storage Tank (RWST). The EHRS is connected to a separate SG feed and steam line and the RWST is installed outside the containment structure (Carelli et al., 2004; Carelli et al., 2009; Mascari, 2010b; Chiovaro et al., 2011). In the advanced BWR designs the core water evaporates, removing the core decay heat, and condenses in a heat exchanger placed in a pool. Then the condensate comes back to the core (Hicken & Jaegers, 2002; Mascari et al., 2010a). For example, the SWR-1000 (Siede Wasser Reaktor, 1000 MWe) design has emergency condensers immersed in a core flooding pool and connected to the core, while the ESBWR (Economic Simplified Boiling Water Reactor) design uses isolation condensers connected to the Reactor Pressure Vessel (RPV) and immersed in external pools (IAEA-TECDOC-1391, 2004; Aksan, 2005; Mascari et al., 2010a). The designs of some advanced reactors rely on natural circulation for the removing of the core power during normal operation. Examples of these reactors are the MASLWR (Multi- Application Small Light Water Reactor), the ESBWR, the SMART and the Natural Circulation based PWR being developed in Argentina (CAREM)(IAEA-TECDOC-1391, 2004; IAEA -TECDOC-1474, 2005; Mascari et al., 2010a). In particular the MASLWR (Modro et al., 2003), figure 1, is a small modular integral Pressurized Water Reactor (PWR) relying on natural circulation during both steady-state and transient operation. In the development process of these advanced nuclear reactors, the analysis of single and two-phase fluid natural circulation in complex systems (Zuber, 1991; Levy, 1999; Reyes & King, 2003; IAEA-TECDOC-1474, 2005; Mascari et al., 2011e), under steady state and transient conditions, is crucial for the understanding of the physical and operational phenomena typical of these advanced designs. The use of experimental facilities is fundamental in order to characterize the thermal hydraulics of these phenomena and to develop an experimental database useful for the validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. In general it is expensive to design a test facility to develop experimental data useful for the analyses of complex system, therefore reduced scaled test facilities are, in general, used to characterize them. Since the experimental data produced have to be applicable to the full-scale prototype, the geometrical characteristics of the facility and the initial and boundary conditions of the selected tests have to be correctly scaled. Since possible scaling distortions are present in the experimental facility design, the similitude of the main thermal hydraulic phenomena of interest has to be assured permitting their accurate experimental simulation (Zuber, 1991; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). Fig. 1. MASLWR conceptual design layout (Modro et al, 2003; Reyes et al., 2007; Mascari et al., 2011a). Different computer codes have been developed to characterize two-phase flow systems, from a system and a local point of view. Accurate simulation of transient system behavior of a nuclear power plant or of an experimental test facility is the goal of the best estimate thermal hydraulic system code. The evaluation of a thermal hydraulic system code’s calculation accuracy is accomplished by assessment and validation against appropriate system thermal hydraulic data, developed either from a running system prototype or from a scaled model test facility, and characterizing the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs (Mascari et al., 2011a; Mascari et al., 2011c). In this international framework, Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a system level test facility to examine natural circulation phenomena of importance to the MASLWR design. The scaling analysis of the OSUMASLWR experimental facility was performed in order to have an adequately simulation of the single and two-phase natural circulation, reactor system depressurization during a blowdown and the containment pressure response typical of the MASLWR prototype (Zuber, 1991; Reyes & King, 2003; Reyes, 2005b). A previous testing program has been conducted in order to assess the operation of the prototypical MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems under transient conditions (Modro et al. 2003; Reyes & King, 2003; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). The experimental data developed are useful also for the assessment and validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. For many years, in order to analyze the LWR reactors, the USNRC has maintained four thermal-hydraulic codes of similar, but not identical, capabilities, the RAMONA, RELAP5, TRAC-B and TRAC-P. In the last years, the USNRC is developing an advanced best estimate thermal hydraulic system code called TRAC/RELAP Advanced Computational Engine or TRACE, by merging the capabilities of these previous codes, into a single code (Boyac & Ward, 2000; TRACE V5.0, 2010; Reyes, 2005a; Mascari et al., 2011a). The validation and assessment of the TRACE code against the MASLWR natural circulation database, developed in the OSU-MASLWR test facility, is a novel effort. This chapter illustrates an analysis of the primary/containment coupling phenomena characterizing the MASLWR design mitigation strategy during a SBLOCA scenario and, in the framework of the performance assessment and validation of thermal hydraulic system codes, a qualitative analysis of the TRACE V5 code capability in reproducing it