Radioactive Release from the Prototype Fast Breeder Reactor under Severe Accident Conditions

Closed for proposals

Project Type

Coordinated Research Project

Project Code




Approved Date

23 February 2015


3 - Active - Ongoing

Start Date

13 November 2015

Expected End Date

31 May 2020

Participating Countries

Republic of Korea
Russian Federation
United States of America


In a sodium cooled fast reactor (SFR), a severe transient condition (i.e. hypothetical core disruptive accident (CDA)) is the design extension condition resulting from the mismatch of power produced and power removed from the reactor, as well as from the shutdown system not responding on demand, typically under conditions of either unprotected loss of flow or unprotected transient overpower events. The assessment of the consequences of a CDA in terms of radioactivity release outside the containment system, which may affect the environment and the public, is of paramount importance from the point of view of public acceptance of nuclear power, especially after the Fukushima Daiichi accident. The objective of this CRP is to make realistic estimates, through numerical simulations, of the fission product transport mechanisms in typical pool type SFRs and to determine the fission products retained within the reactor primary vessel and ejected into the reactor containment building. The exercise would be carried out for a reference pool type SFR of 500 MW(e) capacity fuelled with mixed oxide fuel, i.e. the Prototype Fast Breeder Reactor in Kalpakkam (India).


The scope of the CRP is to perform realistic estimations of fission product and fuel particle inventory inside some specific PFBR reactor areas (i.e. in-primary vessel, cover gas system and in-containment building) at different time scales (few seconds for the instantaneous source terms and several days for the long term source term), under severe accident conditions. The objective is to improve the understanding of the key phenomena involving fission products and fuel particle transport inside the reactor and the containment compartments, in order to reduce uncertainties in estimation of the releases to the environment under severe accident condition in the PFBR. Therefore, this CRP will contribute to extend the predictive capabilities of existing simulation tools devoted to SFR design and safety analysis.The CRP will be implemented as a specific task of the IAEA Project 1000154 “Advanced Technology for Fast and Gas-cooled Reactors”, starting with the IAEA Programme & Budget Cycle 2014-2015. Among others, the project 1000154 has the objective to enable Member States to take informed decisions on the development of advanced fast reactor designs, and to increase cooperation between Member States in achieving advances in fast reactor development through international collaborative R&D, in particular in the area of verification, validation and qualification of advanced simulation tools and data for the design and the safety analysis of fast reactors. In this respect the CRP fully responds to the objectives of the IAEA project 1000154.

Specific objectives

Primary system/containment system interface source term estimation

In-containment phenomenology analysis

In-vessel source term estimation

WP1. In-vessel source term estimation

WP2. Primary system/containment system interface source term estimation

WP3. In-containment phenomenology analysis

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