The understanding of radiation-induced grain boundary segregation (RIS) has considerably improved over the past decade. New models have been introduced and much effort has been devoted to obtaining comprehensive information on segregation from the literature. Analytical techniques have also improved so that chemical analysis of layers 1 nm thick is almost routine. This invited paper will review the major methods used currently for RIS prediction: namely, Rate Theory, Inverse Kirkendall, and Solute Drag approaches. A summary is made of the available data on phosphorus RIS in reactor pressure vessel (RPV) steels. This will be discussed in the light of the predictions of the various models in an effort to show which models are the most reliable and easy to use for forecasting P segregation behaviour in steels. A consequence of RIS in RPV steels is a radiation induced shift in the ductile to brittle transition temperature (DBTT). It will be shown how it is possible to relate radiation-induced P segregation levels to DBTT shift. Examples of this exercise will be given for RPV steels and for ferritic steels being considered for first wall fusion applications. Cr RIS in high alloy stainless steels and associated irradiation-assisted stress corrosion cracking (IASCC) will be briefly discussed.

From the irradiation resistance and high-temperature strength, oxide dispersion strengthened (ODS) ferritic steels are candidate materials for advanced and fusion reactors. For the development of advanced steels the key issue is to homogenize nano-particles into matrix. Recent studies have indicated that Ti addition can homogenize Y-Ti complex particles into ferrite matrix, but the reason of the effect of additional elements has not been clarified. In this model study, we focus on the effect of additional elements, such as IV and V families and other oxide formers, which can control potentially the distribution of the oxide particles. The materials used in this study were based on Fe-9Cr-Y2O3 alloys which were mechanical alloyed (MA) from the powder of Fe, Cr and Y2O3, which was added systematically with the element of Ti, Zr, Ta, V, Nb, Hf, Al, Si and others. Usually ODS fabrication process is required for hot extrusion, but we annealed up to 1150 C for simplify the microstructure. To evaluate the distribution of ODS particles; we used TEM equipped with EDS after electro-polishing or FIB techniques. (1) In the case of Si or Al addition, oxides were disappeared after MA process, which means Y2O3 and other elements should be in solution at non-equilibrium condition. Two types of oxides of Y2O3 and Al2O3 or SiO2 developed after the annealing at 850 C, but only complex oxides were developed after the annealing at 1150 C. This result suggests that the oxide formation is independent process for Y and Si or Al. (2) In the case of Ti addition, oxides also were disappeared after MA process, but developed after annealing at 1150 C. This means that Ti can stabilize complex oxides of Y and Ti, and enhance the fine distribution of the oxides comparing with simple Fe-9Cr-Y2O3 alloy.

Large scale classical molecular dynamics simulations have been carried out to study the primary damage state due to a-decay self irradiation in UO2 matrix. Simulations of energetic displacement cascades up to the realistic energy of the recoil nucleus at 80 keV provide new informations on defect production, their spatial distribution and their clustering. The discrepancy with the classical linear theory NRT (Norton-Robinson-Torrens) law on the creation of the number of point defects is discussed. Study of cascade overlap sequence shows a saturation of the number of point defects created as the dose increases. Toward the end of the overlap sequence, large stable clusters of vacancies are observed. The values of athermal diffusion coefficients coming from the ballistic collisions and the additional point defects created during the cascades are estimated from these simulations to be, in all the cases, less than 10-26 m2/s. Finally, the influence of a grain boundary of type Sigma5 is analysed. It has been found that the energy of the cascades are dissipated along the interface and that most of the point defects are created at the grain boundary.

High chromium ( 9-12 wt %) ferritic/martensitic steels are candidate structural materials for future fusion reactors and other advanced systems such as accelerator driven systems ADS). Their use for these applications requires a careful assessment of their mechanical stability under high energy neutron irradiation and in aggressive environments. In particular, the Cr concentration has been shown to be a key parameter to be optimized in order to guarantee the best corrosion and swelling resistance, together with the least embrittlement.

In this work, the characterization of the neutron irradiated Fe-Cr model alloys with different Cr % with respect to microstructure and mechanical tests will be presented. The behavior of Fe-Cr alloys have been studied using tensile tests at different temperature range ( from -160°C to 300°C). Irradiation-induced microstructure changes have been studied by TEM for two different irradiation doses at 300°. The density and the size distribution of the defects induced have been determined. The tensile test results indicate that Cr content affects the hardening behavior of Fe-Cr binary alloys. Hardening mechanisms are discussed in terms of Orowan type of approach by correlating TEM data to the measured irradiation hardening.

Boron carbide diode detectors, fabricated from two different polytypes of semiconducting boron carbide, will detect neutrons in reasonable agreement with theoretical expectations. The performance of the all boron carbide neutron detector differs, as expected, from devices where a boron rich neutron capture layer is distinct from the diode charge collection region (i.e. a conversion layer solid state detector).

Diodes were fabricated from natural abundance boron (20% 10B and 80% 11B.) directly on the metal substrates and metal contacts applied to the films as grown. The total boron depth was on the order of 2 microns. This is clearly not a conversion-layer configuration. The diodes were exposed to thermal neutrons generated from a paraffin moderated plutonium-beryllium source in moderated and unmoderated, as well as shielded and unshielded experimental configurations, where the expected energy peaks at at 2.31 MeV and 2.8 MeV were clearly observed, albeit with some incomplete charge collection typical of thinner diode structures. The results are compared with other boron based thin film detectors and literature models.

The susceptibility of liquid metal embrittlement of the T91 martensitic steel is investigated from slow deformation rate (6.67x10-8 m.s-1) to high deformation rate (6.67x10-3 m.s-1), using a Center Cracked in Tension geometry. Brittle fracture, characterized by elongated river cracks on all the fracture surfaces, indicates that T91 is sensitive to the Embrittlement by LBE. This embrittlement effect is very marked at the low deformation rate (~10-5 mm.s-1). A ductile-brittle transition is observed in the high strain rate range investigated. In this transition regime, there is a competition between the growth of dimples and the cracking induced by the liquid metal. Ductility recovery is complete at the highest investigated displacement rate. This dependence of T91/LBE susceptibility to LME on the displacement rate and the brittle to ductile transition at high strain rate is confirmed by a fracture mechanics analysis.

Future designs of nuclear fuels require an increased understanding of fission gas bubble density evolution. Derivations will be provided for a generic Boltzmann bubble density evolution equation, a bubble density deformation field equation, and a Cauchy stress/bubble-pressure equilibrium equation.

In the context of research on new materials for next generation nuclear reactors, it becomes more and more interesting to know what can be the advantages of nanostructured materials for such applications. In this study, we performed irradiation experiments on microstructured and nanostructured â-SiC samples, with 95 MeV Xe and 4 MeV Au ions. The structure of the samples was characterized before and after irradiation by grazing incidence X ray diffraction and Raman spectroscopy. The results showed the occurrence of a synergy between electronic and nuclear energy loss in both samples with 95 MeV Xe ions, while the nanostructured pellet was found to have a better resistance to the irradiation with 4 MeV Au ions.

In this investigation, a thermo-mechanical model of pebble beds is adopted and developed based on experiments by Dr. Reimann at Forschungszentrum Karlsruhe (FZK). The framework of the present material model is composed of a non-linear elastic law, the Drucker-Prager-Cap theory, and a modified creep law. Furthermore, the volumetric inelastic strain dependent thermal conductivity of beryllium pebble beds is taken into account and full thermo-mechanical coupling is considered.

Investigation showed that the Drucker-Prager-Cap model implemented in ABAQUS can not fulfill the requirements of both the prediction of large creep strains and the hardening behaviour caused by creep, which are of importance with respect to the application of pebble beds in fusion blankets [1]. Therefore, UMAT (user defined material's mechanical behaviour) and UMATHT (user defined material's thermal behaviour) routines are used to re-implement the present thermo-mechanical model in ABAQUS. An elastic predictor radial return mapping algorithm is used to solve the non-associated plasticity iteratively, and a proper tangent stiffness matrix is obtained for cost-efficiency in the calculation. An explicit creep mechanism is adopted for the prediction of time-dependent behaviour in order to represent large creep strains in high temperature. Finally, the thermo-mechanical interactions are implemented in a UMATHT routine for the coupled analysis.

The oedometric compression tests and creep tests of pebble beds at different temperatures are simulated with the help of the present UMAT and UMATHT routines, and the comparison between the simulation and the experiments is made.

Higher conversion efficiencies require high operation temperatures that are difficult to obtain due to the actual thermo-physical properties of the nuclear fuels. The initial intrinsic thermal conductivity of the actual fuel pellets, mainly ceramics like structures (oxides, nitrides, carbides, MOX, beads) is low. The center of the pellet is near melting temperature while the cladding operation temperature has to be low. The fission products deposition and burnup effects are further dimming the thermal conductivity. More the cooling agent's chemical reactivity increases with temperature is another main reasons of keeping the operation temperature low. The usage of a hetero-structure of solid fuel soaked into a drain fluid is increasing the thermal conductivity. Properly shaped beads structure drives to the possibility of preventing most of the fission products of being stored inside the fuel lattice deteriorating its properties, being drained outside the nuclear reactor. This changes inspired from the nature, makes the nuclear reactor resembling with a plant having self cleaning and curing properties while operating at higher temperatures. Immersing the fuel into liquid metal higher operation temperatures are allowed due to increase in thermal conductivity by 3 to 20 times. Making the fuel beads shorter than the range of fission products, their trajectories end in the drain liquid that is tolerant to nuclear recoils damage. Due to better thermal conductivity the temperature field differences inside the nuclear reactor becomes smaller, allowing the operating temperature to rise significantly without safety concerns. There is possible to continuously remove the fission products by smoothly circulating the drain liquid. The low flow is needed to give time to short lives fission products to decay inside the reactor highly shielded volume. There are several fissionable materials and drain liquids matching which assure high operation temperatures, and allow He cooling, and high temperatures gas turbines cycles. The conversion efficiency might be higher than 70% depending on the chosen actinide / drain-liquid / cooling-liquid combination. The new concept on fission products continuous release and separation minimizes the waste and the total radioactivity stored inside the reactor to few weeks integrated operation amount remaining constant over the time. That makes the fuel's remnant radioactivity lower by a factor >100 than the actual reactors level. The fuel reactivity might be controlled by poisoning and transmutation or by assuring specific reactive geometries which to allow a ultrahigh burnup without the need of over-criticality loads. The fuel's deformability opens the way for interesting applications. The advantages of micro structured nuclear fuel are higher thermal conductivity, fission products removal, appropriate reactivity, higher efficiencies and longer fuel life.

Refractory carbide nanostructured ceramics as SiC constitute interesting materials for high temperature applications and particularly for fourth generation nuclear plants. To elaborate such nanomaterials, weighable amounts of SiC nanopowders have to be synthesized first with an accurate control of the grain size and stoichiometry. The inductively coupled plasma and the laser pyrolysis techniques, respectively developed at EMPA Thun and CEA Saclay, allow meeting these requirements. Both techniques are able to produce dozens of grams per hour of silicon carbide nanopowders. The particle size can be adjusted down to around 20 nm for the plasma synthesis and even down to 5-10 nm for the laser pyrolysis. The stoichiometry Si/C can be tuned by the addition of methane into the plasma and acetylene for the laser process.

Density functional theory calculations of the solution energies of helium in substitutional, tetrahedral and octahedral sites have been performed for all BCC transition metals: V, Nb, Ta, Cr, Mo, W and Fe. The effects of exchange correlation functional and of pseudopotential have been investigated in Fe; they are relatively small. The solution energies are found to be weakly dependent on the element for the substitutional site whereas for the interstitial sites they are much smaller in group V than in group VI and they decrease from 3d to 4d and 5d metals. As a result an inversion is observed from V, Nb and Ta - which tend to favor the interstitial site - to Mo and W, which favor the substitutional one, with an intermediate behavior for Cr and Fe. Finally, the results indicate that the tetrahedral site is always energetically more favorable than the octahedral one by 0.2 to 0.3 eV.

Amorphous Silica is one of candidate materials for both final focusing optics of lasers for NIF and future inertial fusion reactors and diagnostics of the Safety and Control Systems of the ITER machine as well as DEMO magnetic fusion reactors. In operation, these materials will be exposed to high neutron irradiation fluxes and it can result in point defect and vary the optical absorption, that is, degradation of the optical properties. In this paper we present molecular dynamic simulation of displacement cascade due to energetic recoils in amorphous silica without hydrogen atoms and with 1% of hydrogen atoms trying to identify defects formation. We have made a statistics of the different kind of defects at different energy of primary knock-on atoms (PKA). The range of studied PKA energies are from 400 eV to 3.5 keV and it is made to both component of this material Silicon and Oxygen.

The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 ¡ãC). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with subsize specimens.

Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C=DDBTT/Ds indicates hardening-dominated embrittlement at irradiation temperatures below 350 ¡ãC with 0.17 ¡Ü C ¡Ü 0.53 ¡ãC/MPa. Scattering of C at irradiation temperatures above 400 ¡ãC indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated 10B-isotope (0.008-0.112 wt.%).

Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns.

Radiation damage has traditionally been modeled using cascade simulations however such simulations generally neglect the effects of electron-ion interactions, which may be significant in high energy cascades. A model has been developed which includes the effects of electronic stopping and electron-phonon coupling in Molecular Dynamics simulations by means of an inhomogeneous Langevin thermostat. The energy lost by the atoms to electronic excitations is gained by the electronic system and the energy evolution of the electronic system is modeled by the heat diffusion equation. Energy is exchanged between the electronic system and the atoms in the Molecular Dynamics simulation by means of a Langevin thermostat, the temperature of which is the local electronic temperature. The model is applied to a 10 keV cascade simulation for Fe.