Nuclear safety has been one of the major issues studied since the inception of the nuclear industry. Establishing and maintaining core cooling and ensuring containment integrity are two main goals that nuclear safety must guarantee. Improvement in these safety systems has generally involved the development of suitable Passive Containment Cooling Systems (PCCSs). This kind of safety approach poses significant issues for computational and analysis methods since the vessel and containment are strongly coupled and the system response is based on the interaction between the two. This is the case of Small Modular Reactors (SMRs), which adopt a completely passive safety approach, and the integral design eliminates the large coolant loop piping, which in turn eliminates large Loss-Of-Coolant-Accidents (LOCAs) as well as the individual component pressure vessels and supports. For these reasons, no severe accident analyses have yet been conducted on this type of plant. Nevertheless, it is useful to investigate the possible consequences of a multiple failure scenario in these advanced systems. In order to perform these analyses, starting from the available data of the SPES3 experimental facility, a SMR model was developed for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. The facility based on the IRIS reactor (International Reactor Innovative and Secure) design reproduces the primary, secondary and containment systems with a 1:100 volume scale, full elevation and prototypic fluid and thermal hydraulic conditions. The IRIS reactor is a SMR developed by an international consortium led by Westinghouse/BNFL, which includes universities, national laboratories, commercial companies and utilities. The Design Basis Accident (DBA) Direct Vessel Injection (DVI) line double-ended guillotine break was the reference transient that allowed matching the trend of the main physical parameters predicted by the ASTEC code model with those computed by the well-established best estimate coupled codes RELAP-GOTHIC. In this paper, a multiple system failure scenario was reproduced, to investigate and evaluate the in-vessel phase phenomena and the effectiveness of the passive mitigation measures. The results of the calculations confirmed the good performance of the IRIS system during the DBA accident, and showed for the first time how the ASTEC code can reproduce well the behaviour of this non- prototypic system.
Exploratory Studies of Small Modular Reactors Using the ASTEC Code
DI GIULI, MIRCO;SUMINI, MARCO;
2015
Abstract
Nuclear safety has been one of the major issues studied since the inception of the nuclear industry. Establishing and maintaining core cooling and ensuring containment integrity are two main goals that nuclear safety must guarantee. Improvement in these safety systems has generally involved the development of suitable Passive Containment Cooling Systems (PCCSs). This kind of safety approach poses significant issues for computational and analysis methods since the vessel and containment are strongly coupled and the system response is based on the interaction between the two. This is the case of Small Modular Reactors (SMRs), which adopt a completely passive safety approach, and the integral design eliminates the large coolant loop piping, which in turn eliminates large Loss-Of-Coolant-Accidents (LOCAs) as well as the individual component pressure vessels and supports. For these reasons, no severe accident analyses have yet been conducted on this type of plant. Nevertheless, it is useful to investigate the possible consequences of a multiple failure scenario in these advanced systems. In order to perform these analyses, starting from the available data of the SPES3 experimental facility, a SMR model was developed for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. The facility based on the IRIS reactor (International Reactor Innovative and Secure) design reproduces the primary, secondary and containment systems with a 1:100 volume scale, full elevation and prototypic fluid and thermal hydraulic conditions. The IRIS reactor is a SMR developed by an international consortium led by Westinghouse/BNFL, which includes universities, national laboratories, commercial companies and utilities. The Design Basis Accident (DBA) Direct Vessel Injection (DVI) line double-ended guillotine break was the reference transient that allowed matching the trend of the main physical parameters predicted by the ASTEC code model with those computed by the well-established best estimate coupled codes RELAP-GOTHIC. In this paper, a multiple system failure scenario was reproduced, to investigate and evaluate the in-vessel phase phenomena and the effectiveness of the passive mitigation measures. The results of the calculations confirmed the good performance of the IRIS system during the DBA accident, and showed for the first time how the ASTEC code can reproduce well the behaviour of this non- prototypic system.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
