Last update on August 2019
The PERFROI experimental research project was launched in January 2014 for a six years long period. An extension of the project until the end of 2021 has been accepted by the ANR with additional funding. It aims to better understand the cooling behavior of a nuclear reactor core in case of a loss of coolant accident (LOCA). It is one of seven IRSN projects selected by the National Research Agency (ANR) in response to the RFP for "Nuclear safety and radiation protection-related projects" issued in 2012.
This multidisciplinary research project includes thermal mechanical and thermal hydraulic aspects. The results obtained will be used for validation of the DRACCAR simulation tool developed by IRSN to simulate nuclear fuel behavior under LOCA conditions.
Context and objectives
In case of a leakage of the coolant circuit in a pressurized water reactor (PWR), the fuel rods of the fuel assemblies, if not sufficiently cooled, could undergo a strong temperature increase and significant cladding deformation. On the one hand, the fuel rod cladding may swell and burst due to internal pressure. This can result in radioactive fission products release into the coolant system
and may also limit the water circulation through the core (blockage phenomenon) and compromise the possibility of subsequent cooling using water from safety injection systems (reflood phenomenon). On the other hand, water vapor may cause oxidation and embrittlement of the cladding, with a significant risk of loss of mechanical resistance and subsequent collapse to form a debris bed that can no longer be properly cooled.
IRSN has prepared a synthesis of the research conducted over the past thirty years on potential in-core phenomena during and after a LOCA transient. This state of the art addressing cladding oxidation, hydriding, swelling and bursting phenomena, as well as the potential impact of these phenomena as a whole, was used to identify three main questions requiring additional answers:
- What is the maximum admissible blockage rate beyond which core coolability can no longer be guaranteed?
- Can the fragmentation of fuel pellets within cladding subject to local or balloon-like swelling cause them to migrate from the upper region toward the balloon-like swell region, or even to exit the cladding and disperse into the coolant system? And if so, what are the potential consequences?
- Does the degradation of thermal mechanical properties of the cladding (as a result of oxidation and hydriding due to chemical reactions with water or steam), under both normal and accident conditions, compromise the mechanical resistance of the fuel rods and the long-term coolability of the fuel rod assemblies?
The PERFROI project was launched in order to answer, at least partially, to these pending questions.
The PERFROI project is structured into two research items, i.e.
Thermal-mechanical deformation and rupture of fuel rods under LOCA conditions: Experiments (ELFE and COCAGNE tests) conducted using IRSN's MAESTRO experimental platform will allow characterizing the thermal mechanical properties of the cladding and developing the numerical simulation models needed to simulate cladding deformation and bursts. These models will be integrated into the DRACCAR simulation tool in order to evaluate potential blockage rates in case of a LOCA transient. In particular, these models will take into account the physico-chemical behavior of hydrogen within the cladding and its potential impact on the mechanical behavior of the fuel rods. ELFE tests has been conducted between 2014 and 2017 and those regarding
the cladding material Zy-4 have been the subject of a publication. The COCAGNE
installation was received in December 2016 and tests on it are in progress with
a first test carried out in December 2017.
- Reflooding of a partially blocked fuel assembly: Experiments (COAL tests) will be conducted with a test device containing a 7x7 assembly (46 fuel rods and three
guide tubes). This device will be installed in a thermal-hydraulic test loop
made available in STERN Lab in Canada (2020-2021). Results will be used to
fine-tune and validate the thermal hydraulic models used in DRACCAR.
In particular, the results of this experimental research project,
co-funded by EDF, US-NRC and more recently KAERI (COCAGNE tests) will
be used by IRSN to evaluate the relevance of assumptions considered by operators in order to demonstrate the reactor core coolability under LOCA conditions.
 Fuel rod cladding constitutes the first of the three containment barriers between radioactive products and the environment (the other two are the reactor coolant pressure boundary and the reactor containment).