TENSILE BEHAVIOUR OF T91 STEEL OVER A WIDE RANGE OF TEMPERATURES AND STRAIN-RATE UP TO 10^4 s^-1

High cromium ferritic/martensitic steel T91 (9% Cr, 1% Mo), on account of its radiation resistance, is a candidate material for nuclear reactor applications. Its joining by an impact method to create a cold joint is tested in the realm of scoping tests towards the safe operation of nuclear fuels, encapsulated in representative T91 materials. Hitherto, T91 mechanical characterisation at high strain rates is relatively unknown, particularly, in relation to impact joining and also to nuclear accidents. In this study, the mechanical characterization of T91 steel was performed in tension by varying the strain-rate (10-3 up to 104 s-1) and temperature (20-800°C) on dog-bone specimens, using standard testing machines or Hopkinson Bar apparati. As expected, the material is both temperature and strain-rate sensitive and different sets of parameters for the Johnson-Cook strength model were extracted via a numerical inverse procedure, in order to obtain the most suitable set to be used in this field of applications

Tensile Behavior of T91 Steel Over a Wide Range of Temperatures and Strain-Rate Up To 104 s1

High chromium ferritic/martensitic steel T91 (9% Cr, 1% Mo), on account of its radiation resistance, is a candidate material for nuclear reactor applications. Its joining by an impact method to create a cold joint is tested in the realm of scoping tests toward the safe operation of nuclear fuels, encapsulated in representative T91 materials. Hitherto, T91 mechanical characterization at high strain rates is relatively unknown, particularly, in relation to impact joining and also to nuclear accidents. In this study, the mechanical characterization of T91 steel was performed in tension by varying the strain-rate (1023 up to 104 s21) and temperature (20-800C) on dog-bone specimens, using standard testing machines or Hopkinson Bar apparati. As expected, the material is both temperature and strain-rate sensitive and different sets of parameters for the Johnson-Cook strength model were extracted via a numerical inverse procedure, in order to obtain the most suitable set to be used in this field of applications.JRC.E.4-Nuclear Fuel Safet

Zirconium carbonitride pellets by internal sol gel and spark plasma sintering as inert matrix fuel material

Inert matrix fuel is a fuel type where the fissile material is blended with a solid diluent material. In this work zirconium carbonitride microspheres have been produced by internal sol gel technique, followed by carbothermal reduction. Material nitride purities in the produced materials ranged from Zr(N0.45C0.55) to Zr(N0.74C0.26) as determined by X-ray diffraction and application of Vegard's law. The zirconium carbonitride microspheres have been pelletized by spark plasma sintering (SPS) and by conventional cold pressing and sintering. In all SPS experiments cohesive pellets were formed. Maximum final density reached by SPS at 1700 C was 87% theoretical density (TD) compared to 53% TD in conventional sintering at 1700 C. Pore sizes in all the produced pellets were in the mm scale and no density gradients could be observed by computer tomography

Structural investigation of self-irradiation damaged AmO2

Studying self-irradiated materials is an ideal means to investigate the effect of the damages on material structure and to better understand the behaviour of irradiated nuclear fuels. In this context, X-ray diffraction, X-ray absorption spectroscopy and transmission electron microscopy have been used to investigate self-irradiation damaged AmO2. Combining these techniques allows studying the microstructure and the variation of the fluorite structure at both short-range and long-range order. Thus, the increase of both interatomic distances and lattice parameter was shown, as well as the presence of nanometer sized He bubbles. As confirmed by the observed high-level of crytallinity, the fluorite structure exhibits a high radiation tolerance, which is confirmed by the low increase of the lattice parameter. This could be explained by a self-annealing mechanism of the created defects at room temperature.JRC.E.4-Nuclear fuel

The effect of high dose rate gamma irradiation on the curing of CaO-FexOy-SiO2 slag based inorganic polymers: Mechanical and microstructural analysis

In search for alternative cementitious materials for radioactive waste encapsulation, geopolymers and inorganic polymers (IPs) have received wide attention. Moreover, Fe-rich IPs offer an interesting alternative to high density concretes for use in radiation shielding applications. Materials can however be altered when subjected to ionizing radiation, creating the necessity to evaluate the material’s behaviour under irradiation conditions. In this study the effect of high dose rate (8.85 kGy/h) gamma irradiation is investigated on CaO-FexOy-SiO2 slag-based IPs. Samples with different curing times (1 h, 24 h and 28 days) prior to the irradiation were irradiated to a dose of 200 kGy using a60Co source. The effect of gamma radiation is observed to be highly dependent on the curing time prior to irradiation. 28 days cured samples are found to be resistant to the irradiation for the dose (rate) and properties tested without any significant change in strength, indentation characteristics, porosity and Fe3+ content. The IPs studied show a different behaviour when irradiated immediately after casting or after 24 h of curing. It is therefore thought that the mechanism behind the effect of irradiation is different for the non-hardened samples compared to hardened samples. For the 1 h cured samples prior to irradiation multiple effects were observed: an increase of the compressive strength by a factor 2.20, a decrease in hardness of the binder by a factor of 0.73, a lower Young’s-modulus of the binder by a factor of 0.67, a decrease of creep in time for the binder by a factor of 0.72, a decrease in porosity by a factor of 0.92 and an increase of the Fe3+/ΣFe ratio by a factor of 1.95.JRC.G.I.3-Nuclear Fuel Safet

Preparation of bulk nanostructured UO2 pellets using high-pressure spark plasma sintering for LWR fuel safety assessment

Nuclear fuel undergoes a significant structural restructuration during its lifetime in the nuclear reactor. Especially at the rim of the pellet large UO2 grains disintegrate into a nanosized material. In this paper we focus on the preparation of bulk UO2 with grain sizes below 100 nm to investigate physico-chemical properties of this so-called "High Burn up Structure". Preparation of bulk nanocrystalline materials is a challenge that can be overcome using the high pressure spark plasma sintering (HP SPS) technique. In house developed HP SPS with 500 MPa applied pressure was used for compaction of 11 nm UO2 powder obtained by oxalate conversion. The procedure yielded dense (>90%) compacts with grain size as low as 34 nm for samples sintered at 800°C.JRC.G.I.3-Nuclear Fuel Safet

Spark plasma sintering and mechanical properties of tantalum carbonitrides Ta2CxNy

Dense tantalum-based carbonitride materials featuring >95 % relative density, TaC, Ta2C1.5N0.5, Ta2CN, Ta2C0.5N1.5 and TaN, were prepared by Spark Plasma Sintering (SPS) at 1873 K with a dwell time of 10 min and a pressure of 50 MPa. Despite the presence of an oxide phase (i.e. 5 vol%) and some W inclusions (i.e. 1 vol%), the mechanical properties of such ultra-high temperature ceramics (UHTC) show promising values. Indeed, the Young's moduli measured by nano-indentation were approx. 600–700 GPa, which is higher than literature values reported for similar UHTC. The hardness values increased from 16.4 ± 0.8 GPa for TaC to 27.9 ± 1.3 GPa for TaN, with an approximately linear trend for the carbonitride samples while the ratio of plastic deformation work over the total indentation work followed the opposite trend