Third Yearly Activity Report

The calculation work performed during the 3rd project year in WP2 as well as the R&D activities carried out in WP3, WP4 and WP5 are described in this report. In addition, the work dedicated to the project management (WP1) as well as to WP6 regarding the dissemination/communication activities and the education/training program (e.g. the follow-up of the mobility program between different organizations in the consortium, training on simulation tools and activities accomplished by PhD/post-doctoral students) is also reported

Behavior of chromium coated M5 claddings upon thermal ramp tests under internal pressure (loss-of-coolant accident conditions)

International audienceIt has been previously shown that Cr-coatings, even with a thickness limited to 10-15µm, can have a positive impact on the high temperature isothermal creep and burst behavior of the cladding under internal pressure, typical of hypothetical LOss-of-Coolant Accident (LOCA) situations. Then, in view of being more representative of LOCA conditions, thermal ramp tests were recently performed at CEA thanks to the "EDGAR" facilities on 50 cm long M5 clad segments with 12-15 $\mu$m thick Cr-coating. Heating rates ranging from 0.1 up to 25°C/s and internal pressures from 10 to 100 bars have been applied up to the rupture temperature. It was observed that, compared to the uncoated reference material, failure of the Cr-coated cladding segments occurred at comparable or higher temperatures with significantly smaller balloon and/or rupture opening. Very limited oxidation and excellent adhesion of the Cr-coating were also confirmed, even at the balloon location where the cladding was highly deformed. M5 is a trademark or a registered trademark of Framatome or its affiliates in the USA or other countries

Quench Behavior Of Sic/Sic Cladding After High Temperature Ramp Under Steam Conditions

International audienceSilicon carbide based continuous fiber ceramic matrix composite materials (SiCf/SiC) are considered by the French Nuclear Institute as a long-term option for Gen III/III+ light water reactor cladding to improve the accident tolerance of the fuel (ATF). Consistent with this ambition, the extensive RetD activities over the post-Fukushima period has resulted in significant progress in the fabrication of representative and functional specimen, removing some of technological barriers that prevent such advanced ceramics materials from using in a nuclear environment. In addition to on-going basic research, a collaborative program was launched to assess the thermo-mechanical performances of SiC/SiC composites produced at CEA and to collect the required data for a preliminary conceptual design.The present work reports the results from experiments on SiC/SiC clad segments demonstrating their outstanding ability to preserve a coolable geometry during the reflood phase of a postulated design basis Loss of Coolant Accident (LOCA) and beyond. Experimental assessments of thermal shock performance were investigated for this purpose by quenching specimens from 1200DC to 1500DC under steam environment into room temperature water. SiC/SiC composites remain structurally sound and exhibit a slightly affected residual mechanical behavior depending on the oxidation time and the quenching temperature. Post-quenched SiC/SiC tubes were subjected to internal helium gas pressure to test the permeation of the structures. These tests clearly established the link between the leak-tightness retention and damages in the microstructure through micro crack initiation. Close examinations of the oxidized surfaces evidence the efficiency of the passivation layer in protecting substrates. Finally, a correlation of the observations with the oxide growth rate is developed to support a better understanding of the mechanisms as well as the prediction of the acceptable limits of application. These positive results reinforce the interest of a SiC-based fuel cladding concept to enhance the accident tolerance of fuel for future reactors

Advances in the RetD at CEA on ATF claddings

International audienceDescription du programme de la DEN sur les AT

Out-of-pile RandD on chromium coated nuclear fuel zirconium based claddings for enhanced accident tolerance in LWRs

International audienceIn the framework of the CEA/AREVA/EDF French nuclear institute, coatings (with thicknesses of a few microns up to $\sim$20$\mu$m) on zirconium alloy cladding have been studied with the objective to provide a significant reduction in the embrittlement of the cladding in accidental conditions, such as the lossof-coolant accident (LOCA). Early studies performed at CEA on several types of coatings obtained by a physical vapor deposition (PVD) process, including ceramic nitride and metallic multi-layered coatings, are discussed in a first part of this paper. The results of this screening analysis showed that chromium coatings have exhibited the best compromise between oxidation resistance and adhesion of the coating. In a second part, the present paper gives an overview of out-of-pile results obtained so far on Cr-coated Zircaloy 4 cladding in both nominal and accidental (LOCA) condition

Chromium hardening and Zr-Cr interface stability of irradiated chromium-coated Zircaloy-4 alloy

International audienceIn Light Water Reactors, under hypothetical LOss of Coolant Accident (LOCA) scenario, zirconium alloy nuclear fuel claddings undergo significant steam oxidation at high temperatures. In such accidental situation, decreasing the oxidation rate of nuclear fuel claddings is a key issue to improve the accident tolerance of the nuclear fuel sub-assembly. Hence, CEA has engaged specific studies on chromium-coated zirconium alloys in the framework of CEA-FRAMATOME-EDF collaborative program on Enhanced Accident Tolerant Fuel (E-ATF) cladding materials. Beyond the LOCA issues, the adhesion strength of the Cr coating has to be assessed under irradiation. Thus, the purpose of the study is to determine the Cr hardening and the stability of the Zr/Cr interface of a Cr-coated Zircaloy-4 alloy after irradiations by means of Transmission and Scanning Electron Microscopy.On a first type of Zr/Cr interface, it is reported the formation of a few tenths nanometers thick Zr(Fe,Cr)2 polytype structured Laves phase displaying both C14 and C15 lattice symmetry. After ion irradiation up to 10 dpa at 400°C, only the C14 phase is observed while the destabilization of the C15 phase could be attributed to some segregation of iron at the interface. For a second interface, obtained under different deposition conditions, only the C15 phase is observed at the interface. The in-situ ion irradiation at 400°C up to 20 dpa showed a partial dissolution of the C15 phase accompanied with a sharpened distribution of the Zr and Cr atoms. In all cases, the semi-coherent atomic structure of the interface observed before and after irradiation may explain the good residual adhesion properties of the coating to the substrate.Additionally, some tensile tests were performed at 350°C on neutron irradiated chromium-coated Zircaloy-4 alloys. Fractograph analysis revealed no peeling of the chromium coating irradiated up to ~2 dpa and confirms the good residual adhesion properties

Behavior under LOCA conditions of Enhanced Accident Tolerant Chromium Coated Zircaloy-4 Claddings

International audienceFor enhanced accident tolerant fuels for light water reactors application, chromium coatings on zirconium based nuclear fuel claddings are developed and studied at CEA in the framework of the French CEA-EDF-AREVA collaborative program. The results obtained so far, mainly on Zircaloy-4 substrate, show very good corrosion resistance in nominal conditions and significant enhancement of the resistance of the material to oxidation in steam at high temperature (HT), up to 1300C, with a drastic decrease of hydrogen release and/or pick-up. The present paper reports some new results obtained on chromium coated Zircaloy-4 claddings tested in loss-of-coolant accident (LOCA) conditions. In order to investigate the potential effect of the coating on the cladding mechanical behavior at HT and the capacity of the coating to sustain significant substrate deformation (i.e., during ballooning until burst occurrence) without generalized cracking/peeling, a preliminary limited set of internal pressure creep and temperature ramp tests have been performed in steam environment thanks to the EDGAR facility. The thermalmechanical tests were done for testing/burst temperatures ranging from 600C ($\alpha_{Zr}$ phase domain) up to 1000C (($\beta_{Zr}$ phase domain) on 50 cm long low-tin Zircaloy-4 cladding samples with a 15$\mu$m thick outer chromium coating

Behavior of cr-coated m5 claddings during and after high temperature steam oxidationfrom 800c up to 1500c

International audienceOne-sided steam oxidation tests have been conducted at various temperatures between 1000 and 1500°C on several facilities on 12-15µm thick Cr-coated M5 claddings. Microstructural observations and micro-chemical analysis have been performed after oxidation and quenching. Some post-quenching ring compression tests have been also carried out to assess the residual strength/ductility of the Cr-coated materials oxidized at High Temperature (HT). It was confirmed that in the 1000-1300°C oxidation temperature range, the oxidation resistance of the coated materials was enhanced, with a significant additional "coping period" at HT before the material became macroscopically brittle during and/or after quenching. The study was then extended to steam oxidation temperatures higher than 1300°C (Design Extension Conditions). Those tests have confirmed that a eutectic reaction occurred between the zirconium-based substrate and the residual metallic Cr coating above 1300°C. For the DEC-type conditions applied, the tested Cr-coated clad segments did not fail upon the final water quenching while some uncoated reference segments did

High temperature steam oxidation of chromium-coated zirconium-based alloys: Kinetics and process

International audienceThe oxidation of chromium-coated zirconium-based alloys is studied under steam at temperatures ranging from 800 °C up to 1500 °C and for oxidation times ranging from a few minutes up to a few hours. For oxidation temperatures up to 1300 °C, the overall oxidation kinetics is nearly parabolic at the beginning of oxidation, when the Cr outer layer is protective. Finally, it significantly accelerates and hydrogen is absorbed during a short period. These steps correspond to different oxidation and diffusion mechanisms, involving: growth of outer chromia scale; Zr-Cr interdiffusion, inducing Zr(Cr,Fe)2 intermetallic layer thickening then disappearance due to transformation into metallic chromium and zirconia; transport of oxygen through residual chromium (in particular along grain boundaries) and into the zirconium substrate, and finally growth of a sub-coating zirconia. The additional effect of the Zr-Cr eutectic reaction occurring when the oxidation temperature is increased beyond 1300 °C is also studied and briefly discussed. © 2020 Elsevier Lt

behavior of chromium coated m5tm claddings under loca conditions

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