AUSTENITIC STAINLESS STEELS FOR FUTURE NUCLEAR FUEL CLADDINGS

Nuclear power systems have been under continuous development since the first nuclear power plant started operation in 1954. They are categorized into different generations, with each new generation having significant technological advances over the previous one. The worldwide effort to develop the next generation of nuclear reactors was defined at the Generation IV International Forum (GIF) in 2000. Six types of design were proposed, including supercritical water cooled reactor (SCWR). Materials in this reactor will be exposed to more severe environments than the current generation of reactors to assure higher efficiency in energy production and the current materials used for fuel cladding need to be improved or new materials should be developed. In this thesis, the behavior of two existing nuclear materials, stainless steels 310S and 316L was investigated, under conditions approximating the nuclear reactor environment. An environment with dynamic loop of supercritical water (SCW) was used to test the performance of the alloys and the oxides formed were analyzed. Oxidation of the alloys in air was also performed for comparison. It was found that although both alloys showed good oxidation resistance in air at 600ºC, stainless steel 310S has better resistance in SCW environment compared to stainless steel 316L. A thin protective oxide layer of Mn2CrO4 spinel delays oxidation in alloy 310S. In order to improve the oxidation resistance of 310S and 316L stainless steels, thermo-mechanical processing (TMP) was applied to modify their microstructures. The deformation and annealing texture of the as-received and processed samples were investigated by means of X-ray diffraction (XRD) and orientation imaging microscopy (OIM). Different rolling paths and different deformation levels before annealing were used to produce samples of different grain size with similar texture and samples of similar grain size with different textures. Subsequently, the oxidation resistance of thermo-mechanically processed 316L and 310S samples in SCW was studied. It was found that the oxidation resistance of stainless steels 316L and 310S can be improved up to four and five times, respectively, by decreasing the grain size below a critical value of 3 µm. It was demonstrated that samples with smaller grain size provided higher fraction of grain boundaries for fast diffusion of chromium to reach the surface and compensate losses due to dissolution of chromium in the oxidation media. External oxide layers formed on as-received and thermo-mechanically processed stainless steel 316L samples was characterized to establish possible correlation between orientation of the substrate and oxide grains. Micro and macro textures of the substrate and the oxide layers were examined and the results showed that the texture of substrate did not affect the texture of magnetite (Fe3O4) in the upper oxide layer. In addition, the texture of magnetite did not affect the texture of hematite (Fe2O3) on samples where hematite was an additional oxide phase. The strong texture of both oxides was explained with surface free energy minimization and strain energy minimization theory. This means that the texture of both oxides is dictated by a competition between their surface and strain energies

AUSTENITIC STAINLESS STEELS FOR FUTURE NUCLEAR FUEL CLADDINGS

Nuclear power systems have been under continuous development since the first nuclear power plant started operation in 1954. They are categorized into different generations, with each new generation having significant technological advances over the previous one. The worldwide effort to develop the next generation of nuclear reactors was defined at the Generation IV International Forum (GIF) in 2000. Six types of design were proposed, including supercritical water cooled reactor (SCWR). Materials in this reactor will be exposed to more severe environments than the current generation of reactors to assure higher efficiency in energy production and the current materials used for fuel cladding need to be improved or new materials should be developed. In this thesis, the behavior of two existing nuclear materials, stainless steels 310S and 316L was investigated, under conditions approximating the nuclear reactor environment. An environment with dynamic loop of supercritical water (SCW) was used to test the performance of the alloys and the oxides formed were analyzed. Oxidation of the alloys in air was also performed for comparison. It was found that although both alloys showed good oxidation resistance in air at 600ºC, stainless steel 310S has better resistance in SCW environment compared to stainless steel 316L. A thin protective oxide layer of Mn2CrO4 spinel delays oxidation in alloy 310S. In order to improve the oxidation resistance of 310S and 316L stainless steels, thermo-mechanical processing (TMP) was applied to modify their microstructures. The deformation and annealing texture of the as-received and processed samples were investigated by means of X-ray diffraction (XRD) and orientation imaging microscopy (OIM). Different rolling paths and different deformation levels before annealing were used to produce samples of different grain size with similar texture and samples of similar grain size with different textures. Subsequently, the oxidation resistance of thermo-mechanically processed 316L and 310S samples in SCW was studied. It was found that the oxidation resistance of stainless steels 316L and 310S can be improved up to four and five times, respectively, by decreasing the grain size below a critical value of 3 µm. It was demonstrated that samples with smaller grain size provided higher fraction of grain boundaries for fast diffusion of chromium to reach the surface and compensate losses due to dissolution of chromium in the oxidation media. External oxide layers formed on as-received and thermo-mechanically processed stainless steel 316L samples was characterized to establish possible correlation between orientation of the substrate and oxide grains. Micro and macro textures of the substrate and the oxide layers were examined and the results showed that the texture of substrate did not affect the texture of magnetite (Fe3O4) in the upper oxide layer. In addition, the texture of magnetite did not affect the texture of hematite (Fe2O3) on samples where hematite was an additional oxide phase. The strong texture of both oxides was explained with surface free energy minimization and strain energy minimization theory. This means that the texture of both oxides is dictated by a competition between their surface and strain energies

NANOINDENTATION OF HYDROGEN ENRICHED Zr-1Nb ZIRCONIUM ALLOY NUCLEAR FUEL CLADDINGS

Zirconium alloys are being commonly used as a material of choice for nuclear fuel claddings in water cooled nuclear reactors for decades due to their good corrosion resistance and low neutron absorption. However, the increasing operation conditions of next generation nuclear reactors (Gen-V) in terms of higher temperatures, pressures and higher neutron flux requires evaluation of further Zr cladding usability. The embrittlement of Zr claddings due to hydrogen pickup from reactor coolant is one of the issues for its potential use in Gen-IV reactors. Nanoindentation is an effective tool for analysis of the change of mechanical properties of hydrogen enriched Zr claddings from localised material volume. Zirconium alloy Zr-1Nb (E110) with experimentally induced hydrides was analysed by the means of nanoindentation. Zirconium hydrides were formed in the material after exposure in high temperature water autoclave. The optimized methodology of surface preparation suitable for nanoindentation is described and the resulting surface quality is discussed. The nanoindentation measurements were performed as an array of 10x10 indents across areas with hydrides. Depth dependent hardness and reduced modulus values measured by nanoindentation were compared between the material with no hydrogen content, low hydrogen content (127 ppm H) and high hydrogen content (397 ppm H). Complementary microhardness measurements at HV 0.1 were performed on all materials for bulk material hardness comparison

In Situ Neutron Radiography Investigations of Hydrogen Related Processes in Zirconium Alloys

In situ neutron radiography experiments can provide information about diffusive processes and the kinetics of chemical reactions. The paper discusses requirements for such investigations. As examples of the zirconium alloy Zircaloy-4, the hydrogen diffusion, the hydrogen uptake during high-temperature oxidation in steam, and the reaction in nitrogen/steam and air/steam atmospheres, results of in situ neutron radiography investigations are reviewed, and their benefit is discussed

Study of the thermo-mechanical behaviour of an innovative multi-layered ceramic-based nuclear fuel cladding

The CEA is the French Alternative Energies and Atomic Energy Commission (Commissariat à l'énergie atomique et aux énergies alternatives). It is a public body established in October 1945 by General de Gaulle. A leader in research, development and innovation, the CEA mission statement has two main objectives: To become the leading technological research organization in Europe and to ensure that the nuclear deterrent remains effective in the future. The CEA is based in ten research centres in France, each specializing in specific fields. The laboratories are located in the Paris region, the Rhône-Alpes, the Rhône valley, the Provence-Alpes-Côte d'Azur region, Aquitaine, Central France and Burgundy. The Cadarache facility at the Provence-Alpes-Côte d'Azur region is one of the largest nuclear research sites in Europe, hosting 21 fixed nuclear installations, including reactors, waste stockpiling and recycling facilities and research centres. It employs over 4,500 people, and approximately 350 students and foreign collaborators carry out research in the facility‟s laboratories. CEA-Cadarache„s host laboratory is the LC2I (Conception and Irradiation Laboratory for Innovative Nuclear Fuels). This laboratory, directed by Mme Sylvie Pillon, is dependant of the SESC (Fuel‟s Behavior Study and Simulation Service), itself included in the DEC (Fuel‟s Study Department). DEC belongs do the DEN (Energy Nuclear Direction) LC2I mission is to conceive, to dimension and to qualify fuel assemblies for fast neutron nuclear reactors and to design and perform in-core radiation experiences. The team is composed by 14 engineers, one technician and one secretary. The specialities are mainly thermo-mechanics and thermo-hydraulics but with extended knowledge in various fields from materials to computer assisted conception, all merging in the nuclear engineering field.Outgoin

Zr alloy protection against high-temperature oxidation: Coating by a double-layered structure with active and passive functional properties

In this work, a new concept of metal surface protection against degradation caused by high-temperature oxidation in water environment is presented. We were the first to create a double-layered coating consisting of an active and passive part to protect Zr alloy surface against high-temperature oxidation in a hot water environment. We investigated the hot steam corrosion of ZIRLO fuel cladding coated with a double layer consisting of 500 nm nanocrystalline diamond (NCD) as the bottom layer and 2 m chromium-aluminum-silicon nitride (CrAlSiN) as the upper layer. Coated and noncoated ZIRLO samples were exposed for 4 days at 400 °C in an autoclave (working water-cooled nuclear reactor temperature) and for 60 minutes at 1000 °C (nuclear reactor accident temperature) in a hot steam furnace. We have shown that the NCD coating protects the Zr alloy surface against oxidation in an active way: carbon from NCD layer enters the Zr alloy surface and, by changing the physical and chemical properties of the Zr cladding tube surface, limits the Zr oxidation process. In contrast, the passive CrAlSiN coating prevents the Zr cladding tube surface from coming into physical contact with the hot steam. The advantages of the double layer were demonstrated, particularly in terms of hot (accident-temperature) oxidation kinetics: in the initial stage, CrAlSiN layer with low number of defects acts as an impermeable barrier. But after a longer time (more than 20 minutes) the protection by more cracked CrAlSiN decreases. At the same time, the carbon from NCD strongly penetrates the Zr cladding surface and worsen conditions for Zr oxidation. For the double-layer coating, the underlying NCD layer mitigates thermal expansion, reducing cracks and defects in upper layer CrAlSiN

Test methodologies for determining high temperature material properties of thin walled tubes

This report presents briefly the test methods used, within the in the EERA JPNM Pilot Project TASTE, for defining the tensile and creep material properties relevant to the integrity of nuclear fuel claddings. These properties are challenging to extract from thin walled tubes since the standard test methods use test specimen that require minimum material thicknesses in the order of 10 mm or more. In consequence the thin walled material properties are acquired through a number of testing techniques and evaluation methodologies suitable for the thin walled product form. In this report the different test methods and their data assessment requirements are briefly described. The test methods evaluated here comprises of sub-size (curved specimen) tensile testing (ST) of the cladding tube, micro specimen (dog-bone) tensile testing (MT), Small Punch testing (SP), Segmented Expanding Cone Mandrel tests (SCM), the ring tension (RT) and ring compression (RC) tests and internal pressure testing (IP).JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy

The resistance upset welds (RUW) made from E110 alloy without and with Cr coatings were oxidized in air atmosphere at 1100 °C for 2, 10 and 30 min. The cross-section microstructure, elemental composition and hardness were studied before and after oxidation using optical and scanning electron microscopy, and indentations in welding region. The RUW welding does not noticeably change oxidation kinetics of E110 alloy. The most crucial effect has surface non-regularities formed after welding, which prevent uniform coating deposition on full surface of welded cladding tube and end plug. Cr coating deposition can strongly reduce oxidation of welded E110 alloy, while additional post-processing treatment should be applied to improve surface morphology after RUW welding. Several suggestions favorable to development of ATF Zr-based claddings using Cr coating deposition on welded nuclear rods were discussed