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Science and Technology of Nuclear Installations
Volume 2012, Article ID 430471, 2 pages

Severe Accident Analysis in Nuclear Power Plants

1Departamento de Ingeniería de Procesos e Hidráulica, Universidad Autónoma Metropolitana-Iztapalapa, Avenida San Rafael Atlixco 186 Col. Vicentina, 09340 México, DF, Mexico
2Department of Physics and Nuclear Engineering, Universitat Politècnica de Catalunya (BarcelonaTECH), Av. Diagonal 647, 08028 Barcelona, Spain
3Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México City, DF, Mexico
4Department of Nuclear Engineering, Graduate School of Engineering, Kyoto University, Yoshida, Sakyo, Kyoto 606-8501, Japan

Received 9 September 2012; Accepted 9 September 2012

Copyright © 2012 Gilberto Espinosa-Paredes et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Safety of nuclear power plants is essential and safety standards are continuously reviewed and upgraded as new developments and research are performed. Continuous research regarding this subject is fundamental for the nuclear industry. Although severe accident analysis and research have been performed throughout the evolution of nuclear industry, it has not yet considered all plausible scenarios. Adequate analyses are needed for all phases of severe accidents in order to maintain or improve safety margins. Hence, it is essential to encourage researchers to keep performing and developing research, codes, and simulations of these potentially hazardous events. The original works published in this special issue can help to improve safety and understand the phenomena involved in severe accidents and their consequences in existing generation II nuclear power plants (NPP) as well as in generation III NPP being built and in generation III+ and IV NPP being developed.

In this special issue ten research articles were published. J. Arndt et al. describe a method to perform simplified analyses concerning integrity of the components of the primary cooling circuit during a severe accident. A second method, using complex calculation models, was used to analyze a postulated high-pressure core melt accident scenario in a PWR caused by a station blackout. Authors found that temperature values of more than 800°C can be reached in the reactor coolant line and the surge line before the bottom of the reactor pressure vessel experience a significant temperature increase due to core melting.

K.-M. Koo et al. presented a response analysis on electrical pulses under severe accident temperature conditions using an abnormal signal simulation analysis. These authors obtained a special function for abnormal pulse signal patterns through a characteristic response under severe accident temperature conditions, which in turn makes it possible to analyze the abnormal output pulse signals through a characteristic response of a 4~20 mA circuit model and a range of the elements changing with temperature under an accident condition.

G. Espinosa-Paredes et al. presented the simulation and analysis of the loss-of-coolant accident (LOCA) in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. The LVNPP model was developed using the RELAP/SCDAPSIM code. The lack of cooling water after the LOCA leads the LVNPP to core melting that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt it is necessary to fully understand the progression of core damage, since such action has effects that may be decisive in accident progression. During the progression of core damage, these authors analyzed the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

A description of the results for a Station Blackout analysis for Atucha 2 Nuclear Power Plant is presented by A. Bonelli et al. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the station blackout sequence the first pressurizer safety valve fails stuck to open after 3 cycles of water release. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. These authors found that this feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than those of German PWRs.

R. P. Martin presented a general evaluation methodology development and application process (EMDAP) paradigm for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management guidelines. The basic framework described in this paper extends the top-down, bottom-up strategy described in the US Nuclear Regulatory Commission Regulatory Guide 1.203 to severe accident evaluations addressing US NRC expectation for plant design certification applications.

The numerical analysis of heat and mass transfer during hydrogen generation in an array of fuel cylinder rods, each coated with a cladding and a steam current flowing outside the cylinders, is presented by H. Romero-Paredes et al. The analysis considers the fuel element without mitigation effects. The system consists of a representative periodic unit cell where the initial and boundary value problems for heat and mass transfer were solved. In this unit cell, it is considered that a fuel element is coated by a cladding with steam surrounding it as a coolant. The numerical simulations allow describing the evolution of temperature and concentration profiles inside the nuclear reactor and could be used as a basis for hybrid upscaling simulations.

R. Kapulla et al. presented validation experiments, conducted in the frame of the OECD/SETH-2 Project. These experiments address the combined effects of mass sources and heat sinks related to gas mixing and hydrogen transport within containment compartments. A wall jet interacts with an operating containment cooler located in the middle (M-configuration) and the top (T-configuration) of the containment vessel. The experiments are characterized by a 3-phase injection scenario. In Phase I, pure steam is injected, while in Phase II, a helium-steam mixture is injected. Finally, in Phase III, pure steam is injected again. For the M-configuration, a strong degradation of the cooler performance was observed for these authors during the injection of the helium/steam mixture (Phase II). For the T-configuration, we observe a mainly downwards acting cooler resulting in a combination of forced and natural convection flow patterns. The cooler performance degradation was much weaker compared with the M-configuration and a good mixing was ensured by the operation of the cooler.

Forty-three organizations from 22 countries networking their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP) are discussed by J. P. Van Dorsselaere et al. According to these authors, the first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2.5 years, some main outcomes of joint research (modeling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete interaction, containment phenomena (water spray, hydrogen combustion), and source-term issues (mainly iodine behavior). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behavior, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented.

C. M. Allison et al. focused on the Fukushima Daiichi accident; they present an assessment which includes a brief review of relevant severe accident experiments and a series of detailed calculations using the RELAP/SCDAPSIM model which were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The authors concluded that detailed analysis for realistic bounding scenarios can provide general guidance of timing of important events and that the responses to the accident once the accident is underway can make a significant difference in the consequences of the accident.

Finally, A. Núñez-Carrera et al. presented the analysis of the Boiling Water Reactor (BWR) lower head during a severe accident using SCDAPSIM/RELAP5 3.2. The computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel due to a loss-of-coolant accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water. The authors conclude that SCDAPSIM/RELAP5 has the capability to predict the melting of the core, control rod, and some structures, with an estimation of the main parameter of the molten pool until the failure of the crust.

Gilberto Espinosa-Paredes
Lluís Batet
Alejandro Nuñez-Carrera
Jun Sugimoto