Last update on september 2015
The international BSAF project (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) was launched in 2012 under the direction of the OECD Nuclear Energy Agency (NEA). It is coordinated by the Japanese Atomic Energy Agency (JAEA) and involves 16 partners from eight countries. It primarily consists of carrying out and analyzing numerical simulations of the 2011 Fukushima Daiichi Power Plant accident in Japan. The project has a two-fold objective of improving the scientific software used for the numerical simulation of nuclear reactor core melt accidents, and providing indications on the condition of each of the power plant’s reactors, particularly the location and composition of the melted fuel, in order to help prepare for the decommissioning of the damaged reactors.
Context and objectives
During the March 2011 accident at Fukushima Daiichi power plant, three reactors (units 1, 2 and 3) suffered at least partial core melt. Four years later, one of the difficulties is locating the fuel, which flowed out of some of the reactor vessels, in order to prepare decommissioning operations, particularly fuel extraction in the presence of high ambient radioactivity.
The international BSAF project was initiated in 2012 by the Nuclear Energy Agency (OECD/NEA) at the proposal of the Japanese Ministry of Economy, Trade and Industry (METI) to compare numerical simulations of the Fukushima accident using different severe accident simulation software. IRSN is involved in the project with the ASTEC severe accident simulation tool, developed in collaboration with its German counterpart, GRS.
The project involves simulating the progression of the accident in the vessel and containment building of units 1, 2 and 3, in order to assess the current condition of the degraded fuel (location, composition, quantity melted, etc.). There are two major objectives:
- to provide Japan with information and support for its decommissioning and decontamination operations at Fukushima nuclear power plant, which should be carried out with as few risks as possible for workers’ health and the environment;
- to test and improve severe accident simulation software for NEA member countries by confronting them with a real situation, and thereby reduce uncertainties in the analysis of severe accidents.
During the first phase of the project, which was completed in November 2014, studies focused on the accident progression during the first six days of the accident. The second phase of the project began in April 2015 and will continue until 2018. It will involve extending numerical simulations over the three weeks following the earthquake, focusing especially on the behavior of fission products in the containment building and the release of radioactive products into the environment.
Main results from the first phase
The first phase of the BSAF project simulated the physical and chemical phenomena that occurred in units 1 to 3, particularly in the vessel (core damage and melt, vessel breach) and the reactor containment building (progression of melted fuel). Calculation is difficult because of the high level of uncertainty for some data and the limited amount of information, resulting in a large number of hypotheses.
For unit 1, the various numerical simulations are consistent as to total meltdown of the core and internal structures, which led to vessel breach. However, because of differences between the models of various simulation tools, there are disparities in the estimated quantities of hydrogen produced during core meltdown, which varies from 350 to 1000 kg.
The calculated extent of core degradation differs significantly between the various numerical simulations for units 2 and 3, in part due to differences between models of simulation tools, but also to uncertainties concerning the quantities of coolant injected into the reactors in the hours following the start of the accident. The exact quantities are unknown, which makes it difficult to determine the specific chronology of events. The most important difference relate to the final distribution of the fuel debris (quantity and location) and whether or not there has been a breach in the reactor vessel.