Degradation and oxidation of B4C control rod segments at high temperatures. A review and code interpretation of the BECARRE program

The paper gives a code-based interpretation of the experimental BECARRE program carried out at the "Institut de Radioprotection et de Sureté Nucléaire" (IRSN) between 2005 and 2010, using the severe accident (SA) code ASTEC/ICARE. As part of the International Source Term Program (ISTP), the BECARRE program focuses on boron carbide (B4C) effects during SA conditions, when B4C is used as an absorber material in French PWRs. Steam oxidation of solid B4C pellets (at 1200-1800 °C), as well as oxidation of molten B4C bearing mixtures up to 9 wt% of B4C dissolution (at 1289-1527 °C) are studied, as well as degradation of 60 cm-long control rod (CR) segments representative of a PWR geometry (up to ∼2000 °C). Temperature and outlet gas releases (steam, H2, CO, CO2) measured in line, and post-test examinations (radiography, tomography, microscopic examinations) give available data to code validation and interpretation. The oxidation rates of the B4C bearing melts have been found always lower than the rates of the solid pellet oxidation in similar conditions, as modeled in ASTEC. For the degradation of the CRs, it is shown that for temperature above 1600 °C, the main effect of the B4C is more toward a mitigation of the hydrogen production rather than increasing it by additional oxidation of boron compounds. No large increases of the hydrogen release after the failure of the guide tube (GT) have been measured, due to downward relocation of the low viscosity B4C bearing melts inside the 60 cm-height CR segments. The ZrO2 oxide layer formed on the outer surface of the GT has been found very protective, leading to failure only above 1650 °C. Both isothermal and runaway thermal conditions have been used to bring about limited GT failure showing that the main mechanisms leading to failure are linked to deformation of the initial geometry, close contacts between materials, and subsequent eutectic material formations. Such a limited degradation for the GT is not modeled in the code. © 2013 Elsevier B.V

Simulation of the OECD/NEA Sandia Fuel Project Phases I and II ignition tests with DRACCAR

International audienceThis paper describes simulations of two ignition tests performed at full power in the frame of the Sandia Fuel Project (SFP) with the thermo-mechanical code DRACCAR v2.3. The OECD/NEA Sandia Fuel Project was built on an agreement between 12 countries from OECD, the Nuclear Energy Agency (NEA) and the US-NRC for the characterization of thermal-hydraulic and zirconium fire phenomena in pressurized-water reactor (PWR). The experimental program was split in two phases to focus at first on axial heating and burn propagation in one prototypic fuel assembly (Phase I), and then on axial and radial heating and burn propagation in 1 × 4 fuel assemblies (Phase II). DRACCAR is a simulation tool, developed at IRSN, for fuel assembly mechanical behavior and coolability assessment during a LOCA transient. The flexibility of DRACCAR allows the modeling of many kinds of geometries. Because the code is based on a 3D non-structured meshing, it can be used to model any non-axisymmetric geometry, like the 1 × 4 fuel assemblies geometry of the Phase II of the SFP program. In order to check the consistency of the modeling, we have optimized the code options to get best results for the Phase I, and applied the same options to the Phase II. Most of the DRACCAR results have been successfully checked against experimental ones, using additional code improvements. Air oxidation and breakaway modeling of the zircaloy claddings were successfully tested against the experimental results. Nevertheless parts of the experimental results of Phase II have been difficult to reproduce. As many causes could be involved in these difficulties, such as detailed evolution of the air convective loop, radiative heat transfers in the bundles, and the modeling of additional reactions of zirconium with nitrogen in places where oxygen is lacking, there is still room for improvement in the work of interpretation and modeling of the SFP tests. © 2018 Elsevier B.V

Development and validation of the multi-physics DRACCAR code

International audienceTo meet the simulation needs of its LOCA RandD program, the IRSN is developing a multi-rod computational tool named DRACCAR. In order to realistically describe the behavior of the reactor core during a Loss Of Coolant Accident (LOCA), modeling has to take into account many coupled phenomena such as thermics (heat generation, radiation, convection and conduction), hydraulics (multi dimensional 1-3 phase flow, shrinkage), mechanics (thermal dilatation, creep, embrittlement) and chemistry (oxidation, oxygen diffusion, hydriding,.). This paper presents several aspects of the DRACCAR code abilities first to handle thermal-hydraulics during reflooding of an intact and of a partially ballooned bundle and secondly the simulation of the OECD SFP phase II experiment dealing with the instantaneous draining of a spent fuel pool. © 2014 Elsevier Ltd All rights reserved © 2014 Elsevier Ltd. All rights reserved

Early phase fuel degradation in Phébus FP: Initiating phenomena of degradation in fuel bundle tests

The international Phébus Fission Product Programme investigated key phenomena occurring in light water reactor core meltdown accidents in a series of five in-pile experiments. Four of these tests focused on the degradation of fuel rod bundles, containing a central control rod, and on the resulting release of fission products, structural materials and actinides from the fuel rod bundle, their transport in the reactor coolant system (RCS) and their subsequent behaviour in the containment vessel. Various steam contents were used in the RCS, from highly oxidising conditions (in FPT0 and FPT1) to more reducing ones (in FPT2 and FPT3). The "early degradation phase" took place at the beginning of the Phébus driver core and fuel bundle heat-up phase, with a quasi-intact fuel bundle geometry. During this phase, the degradation of the control rod and the oxidation runaway due to the fast oxidation of the Zircaloy claddings of the fuel rods, were two major events which took place. The oxidation runaway locally increased the temperatures much above the temperatures resulting from the Phébus driver core heat transfer to the bundle and yielded a large hydrogen release, which amounted to 70-80% of the whole hydrogen production during the tests. The maximum hydrogen flow rates increased with increasing steam flow rates injected at the fuel bundle inlet. The failure mechanisms of silver-indium-cadmium (used in three tests) and boron carbide (used in one test) control rods involve eutectic interactions amongst the components of these control rods. Mechanical deformations of the control rod stainless steel cladding against the Zircaloy control rod guide tube are the main presumed mechanisms for the beginning of these eutectic formations. However, different post-failure scenarios can be postulated for the effect of control rod degradation on fuel bundle degradation for both types of control rods. The exothermic oxidation of the exposed boron carbide pellets led to the release of carbonaceous species (CO, CO2) as well as of additional hydrogen, but no significant methane release could be detected above the limits of detection. Overall, the results confirmed existing knowledge concerning early phase degradation phenomenology found in previous integral experiments such as CORA and QUENCH (Karlsruhe Institute of Technology) and Phébus SFD (IRSN Cadarache), and formed a sound basis for analysis of the late phase degradation subsequently observed. Quantitative analysis of boron carbide control rod degradation in FPT3 pointed to a need for improved modelling of chemical reactions involving this material, particularly its oxidation in steam; this has been studied in the BECARRE experiments conducted by IRSN in the International Source Term Programme, leading to better quantitative understanding and improved modelling in codes such as the European reference severe accident analysis code ASTEC. © 2012 Elsevier Ltd. All rights reserved

Late phase fuel degradation in the Phébus FP tests

The aim of the experimental Phébus FP program was to study the degradation phenomena and the behaviour of the fission products during the progression of a severe accident. The present paper focuses on the late phase fuel degradation in these tests and more particularly on some phenomena that happened during this phase, which could explain the transition processes from the fuel bundle geometry to a debris bed or molten pool. One of the main results produced by the program is the evidence of the loss of the fuel rod geometry systematically in the same range of temperature, (2200 ± 200 C), in spite of the different test conditions. The severe degradations at these temperatures appear linked to important chemical interactions between the fuel and structural materials, principally the Zircaloy cladding of the fuel rods and possibly with stainless steel oxides from the control rod guide tube (and with boron oxide in FPT3). The oxidation of fuel itself could lead to a lowering of the fuel rod relocation temperature. The irradiation effect was not clearly identified as important in these degradations. © 2013 Elsevier Ltd. All rights reserved

DRACCAR A multi-physics code for computational analysis of multi-rod ballooning and fuel relocation during LOCA transients Part one General modeling description

International audienceComputational predictions concerning ballooning of multiple fuel pin bundles during a loss of coolant accident with a final reflooding phase are now more than ever of interest in the framework of light water reactor nuclear safety. To carry out these studies, two difficulties have to be overcome. First, the modeling has to take into account many coupled phenomena such as heat transfer (heat generation, radiation, convection and conduction), hydraulics (multidimensional 2-phase flow, blockage), mechanics (thermal expansion, creep, embrittlement) and chemistry (oxidation, hydriding). Secondly, there are only a few experimental investigations that can help to validate such complex coupled modeling. Over several years, IRSN has developed the 3D computational tool DRACCAR to investigate rod bundle strain during LOCA transients including prediction of the reflooding phase. DRACCAR code is dedicated to study complex configurations such as the deformation and possible contact between neighboring rods and the associated blockage of thermalhydraulic channels in the ballooned zone of the fuel assembly. Modeling efforts have been devoted to the assessment of the coolability of deformed geometries by coupling the thermo-mechanical behavior of the fuel assembly to the thermalhydraulics. The physical modeling available in the current version of DRACCAR V2.3.1 as well as its flexibility are depicted. As a conclusion, some prospects regarding the development of the future version DRACCAR V3 are provided, in particular accounting for the knowledge acquired through IRSN RandD project PERFROI

A comparison of core degradation phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP experiments

International audienceOver the past 20 years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP - fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH - electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that occur under severe accident conditions, relevant to such safety issues as in-vessel retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. These data form an important basis for development and validation of severe accident analysis codes such as ASTEC (IRSN/GRS, EC) and MELCOR (USNRC/SNL, USA) that are used to assess the safety of current and future reactor designs, so helping to reduce the uncertainty associated with such code predictions. Following the recent end of the Phébus FP project, it is appropriate now to compare the core degradation phenomena observed in these four major experimental series, indicating the main conclusions that have been drawn. This covers subjects such as early phase degradation up to loss of rod-like geometry (all the series), late phase degradation and the link between fission product release and core degradation (Phébus FP), oxidation phenomena (all the series), reflood behaviour (CORA and QUENCH), as well as particular topics such as the effects of control rod material and fuel burn-up on core degradation. It also outlines the separate-effects experiments performed to elucidate specific phenomena such as the impact of chemical reactions involving boron carbide absorber material. Finally, it indicates the remaining topics for which further investigation is still required and/or is under way. © 2014 Elsevier B.V. All rights reserved

DRACCAR A multi-physics code for computational analysis of multi-rod ballooning and fuel relocation during LOCA transients. Part Two Overview of modeling capabilities for LOCA

Article paru sous le titre final "DRACCAR: A multi-physics code for computational analysis of multi-rod ballooning, coolability and fuel relocation during LOCA transients. Part Two: Overview of modeling capabilities for LOCA"International audienceComputational predictions concerning ballooning of multiple fuel pin bundles during a loss-of-coolant accident with a final reflooding phase are now more than ever of interest in the framework of light water reactor nuclear safety. To carry out these studies, two difficulties have to be overcome. First, the modeling has to take into account many coupled phenomena such as heat transfer (heat generation, radiation, convection and conduction), hydraulics (multidimensional 2-phase flow, blockage), mechanics (thermal expansion, creep, embrittlement) and chemistry (oxidation, hydriding). Secondly, there are only a few experimental investigations that can help to validate such complex coupled modeling. Over several years, IRSN has developed the 3D computational tool DRACCAR to investigate rod bundle strain during LOCA transients including prediction of the reflooding phase. The DRACCAR code is dedicated to study complex configurations such as the deformation and possible contact between neighboring rods and the associated blockage of thermalhydraulic channels in the ballooned zone of the fuel assembly. To accompany the development of DRACCAR, efforts have been devoted to the validation of the coupling between the thermo-mechanics and thermalhydraulic models – including reflooding – through a comparison to integral experiments dedicated to LOCA. The DRACCAR capabilities and validation status are depicted for the version DRACCAR V2.3.1. DRACCAR provides an interesting insight on LOCA by simulating multi-rod and fluid interaction which cannot be investigated with a classical single rod approach. As a conclusion, some prospects regarding the development and validation of the future version DRACCAR V3 are mentioned. In particular significant evolutions are expected regarding the cladding rupture prediction, the contact simulation and the assessment of the coolability of deformed geometries. These evolutions will be based on the knowledge acquired through the RandD project PERFROI, a project dedicated to LOCA, launched by IRSN in association to other partners and supported by the French National Research Agency (ANR). © 2018 Elsevier B.V