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Enhancing Nuclear Safety


Research

Research programs

PROGRES program

Last update in October 2014

 


 

 

Launched in 2007, the PROGRES (ex-PEARL) experimental research program aims to better understand the behavior of steam and water flow in a porous medium composed of solid particles at very high temperature under conditions representative of a core melt accident (or severe accident). Experiments are conducted using the PRELUDE and PEARL test facilities of the IRSN's THEMA platform.


 

Context and objectives

 

 

 

In case of prolonged loss of cooling accident, the fuel rods of the core of a pressurized water reactor (PWR) will be damaged, and will collapse to form what is called a "debris bed", i.e. an agglomeration of fragments of zircaloy cladding and UO2 pellets (or UO2 and PuO2 pellets in the case of MOX fuel rods) which, if not rapidly cooled, will melt and become increasingly difficult to cool. This problem was identified through analysis of the Three Mile Island accident (TMI-2) which occurred in the United States in 1979.


One of the recommended actions to mitigate such accident sequences consists of reinjecting cooling water into the core, an action so-called "reflooding". Although essential for cooling the fuel assemblies, this action may nevertheless compromise the integrity of the reactor containment building. Indeed, reflooding a melting core at very high temperature may cause an explosive thermal reaction, so-called "steam explosion", between the cooling water and the molten corium. Such an explosion can generate projectiles which could damage the containment building. Furthermore, the water vapor resulting from the vaporization of the injected water will oxidize the metallic compounds of the core (zircaloy cladding, steel structures) and generate hydrogen with the potential to undergo a combustion inside the containment, as it was observed during the Fukushima accident.


The PROGRES experimental research program was launched in order to better understand and model these phenomena, the final objective being to determine the conditions under which cooling water can be injected so as to cool the core in an efficient manner with an acceptable risk for the containment. This additional knowledge will be subsequently used to clarify the choice of emergency operating procedures for severe accident conditions and to support the assessment of the relevance of EDF's Severe Accident Operating Guidelines.


 

 

 

 

 

 

 Simplified diagram of a highly degraded PWR core. © IRSN

 

 

 

 

 

Basis of the program

 

 

 

When "reflooding" a degraded core, the injected water flows through an agglomeration of particles (or "debris bed") at very high temperature with locally variable geometry and porosity. This debris bed can be difficult to cool as the cooling water may flow around the less porous regions of the debris (i.e. the hottest regions) and directly into the more porous regions. In order to predict the flow of water within the debris bed and the resulting production of steam, detailed thermal-hydraulic models have been developed describing the damaged core as a porous medium with variable geometry and characteristics. These models have been developed using the ICARE-CATHARE core degradation simulation tool. Once validated with experimental data acquired from the PRELUDE and PEARL test facilities, these models will be simplified and implemented in the ASTEC severe accident simulation tool.

 

 

 

 

 
 
 
 
 
 
 
  
 

Top view of the bed of particles surrounded by a “by-pass”, which simulates less degraded peripheral areas through which the water can flow around the heated beads if the conditions are unfavourable to reflooding. © IRSN

 

View of the bed of particles inside the inductor used for heating. © IRSN

 

 

 

 
 

 

The PROGRES experimental research program consists of reflooding an agglomeration of metal particles heated to high temperature (1000°C), whose size and spatial distribution are representative of those to be expected in a pressurized water reactor. Spherical steel particles with diameters ranging from 1 to 8 mm are placed inside a quartz cylinder surrounded by an induction solenoid used to heat the particles so as to simulate the residual heat from the core debris. Once the setpoint temperature has been reached, "reflooding" cooling water is injected through an injection nozzle at the base of the cylinder.


 

Instrumentation placed inside the debris bed is used to characterize water flow within the porous medium, steam production and the resulting pressure rise as a function of parameters such as water injection rate, particle size, pressure, and power deposition in the debris bed.


 

 

 

 

Program outline

 

 

 
 
    The experimental research program comprises two phases:
  • The first phase of experiments was conducted using the PRELUDE test facility, a small-scale mockup of the PEARL test facility with test section diameters ranging from 110 to 290 mm and containing approximately 5 to 60 kg of particles.

    This first phase was devoted to the validation of technological options for the PEARL test facility, particularly through induction heater feasibility studies and instrumentation qualification tests. Experiments were conducted from 2010 to 2012, leading to the improvement and validation of simulation tools.
 
 

 

 

Video of PRELUDE tests made in 2009.

 

 


Water flow: 40g/sec

Tube diameter: 180 mm

Beads diameter: 4 mm

 

 
     
  • Scheduled to begin in late 2014, the second phase of experiments will be conducted using the PEARL test facility, consisting of a quartz cylinder 540 mm in diameter and containing approximately 500 kg of steel particles.

 

 

The results obtained between 2010 and 2012

 

Experiments conducted from 2010 to 2012 using the PRELUDE test rig successfully demonstrated the feasibility of reflooding a particle bed heated to 900 °C while maintaining 200 W/kg of power deposition in the particle bed during water injection. Thermocouples placed at different levels within the particle bed (see figure below) were used to monitor the progression of the quench front, i.e. the boundary separating the region where particles have been cooled (bottom region) from the region where particles are still hot and intense boiling occurs. The evolution of pressure within the particle bed and the steam flow rate were also measured.

 

 

 Temperature inside the debris bed during reflooding © IRSN

 

 

 

 

The PRELUDE test facility was used not only for qualification testing of techniques to be implemented in the PEARL test facility, but also for conducting several test campaigns comprising over one hundred tests designed to investigate physical phenomena observed during reflooding and improve correlations currently used in the ASTEC simulation tool.


The results obtained were used to estimate the maximum heat flux extracted from the particles by the cooling water as a function of the particle temperature and to determine the rate of quench front progression as a function of injection rate, initial temperature and particle size. Steam flow rate measurements were used to evaluate the cooling efficiency under different reflooding conditions. Furthermore, it was observed that under certain conditions, some particles may be lifted by steam passing through the upper part of the particle bed at very high velocity. This phenomenon, referred to as "fluidization", contributes to enhance the efficiency of the cooling.

 

 

 

 

 

 

 

 

 

 

 Example of calculation of water progression in the bed of particles (the quench front is situated at the limit of the blue area). © IRSN

Comparison of measured and calculated quench progression rates at different radial positions of the particle bed. © IRSN

 

 

 

 

Tests conducted in 2012 using a debris bed surrounded by a more porous region (so-called bypass) where water can flow more easily. Such conditions are far more representative of reactor conditions, where a damaged core exhibits significant heterogeneities and water initially progresses through pathways offering the least resistance. The results obtained were used to assess the impact of this bypass region on particle cooling.


 


 

Test campaigns to be conducted using the PEARL test facility


Its dimensions and parameter variation capabilities make the PEARL test facility unique in the world. The facility is designed to simulate roughly a reactor core debris at 1:4 scale.


The test section diameter (5 times greater than that of the PRELUDE test facility) will allow for improved analysis of two-dimensional effects on water and steam flow with a bypass present at the debris bed boundary. In addition, tests will be conducted at pressures up to 10 bar, which was not possible with the PRELUDE test facility.


The PEARL test facility is designed to study the effects on debris bed reflooding associated with the water injection rate (2 to 50 m3/h), system pressure (1 to 10 bar), initial temperature of the debris bed (400 to 900 °C), heat input (0 to 300 W/kg, simulating the residual heat content of the core) and subsaturation of injected water. The debris bed may consist of particles of identical size (homogeneous), particles of variable size (heterogeneous), heated metal particles or unheated quartz particles.


 

Installation Pearl.jpg
 

PEARL facility © IRSN


 

Four test campaigns are currently planned, some of which will involve use of the PRELUDE test facility:


 

  • Test campaign #1 (2014-2015): Reflooding of a homogeneous debris bed (same size particles) – Investigated parameters:  water flow rate, pressure and temperature
  • Test campaign #2 (2016): Effects of debris bed geometry – Investigated parameters: particle size distribution and bypass dimensions
  • Test campaign #3 (2017): Analysis of a compact region within a debris bed
  • Test campaign #4 (2018): Effects due to oxidation of metallic materials and hydrogen production


 

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Characteristics

Dates: 2007-2019

Coordinator: IRSN
Funding: IRSN, EDF, European Commission

Partner: EDF

Involved IRSN laboratories

Mechanical and Material Experiment Laboratory (LE2M)

Corium Physics Study Laboratory (LEPC)

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