The objective of this study was to determine the mechanism of decarburization of alloy 617 by investigating the surface and bulk stabilities of the alloy in helium gas containing only CO and CO2 impurities over the temperature range 900-1000°C. For this purpose, the alloy was pre-oxidized at 900°C for 150h in He + 15 ppm CO + 1.6 ppm CO2 gas mixture and subsequently decarburized at 1000°C for additional 100h. The reaction mechanism was corroborated through gas phase analysis, microstructural examination and micro-probe measurements. It was determined that decarburization of the alloy occurred via a reaction between the porous Cr2O3 scale (formed during pre-oxidation) and carbon in the alloy, resulting in chromium metal and carbon monoxide. The chromium diffused back into the matrix, whereas the carbon monoxide escaped the sample through existing cracks and pores in the oxide. Therefore, this study shows that decarburization of alloy 617 occurs in a helium gas containing only CO and CO2 impurities.

Results of the microstructural characterization of four different RAFM ODS Eurofer 97 batches are presented and discussed. Analyses and observations were performed by nuclear microprobe and scanning and transmission electron microscopy. X-ray elemental distribution maps obtained with proton beam scans showed homogeneous composition within the proton beam spatial resolution and, in particular, pointed to a uniform distribution of ODS (yttria) nanoparticles in the Eurofer 97 matrix. This was confirmed by transmission electron microscopy. Scanning electron microscopy coupled with energy dispersive spectroscopy made evident the presence of chromium carbide precipitation. Precipitates occurred preferentially along grain boundaries (GB) in three of the batches and presented a discrete distribution in the other, as a result of different thermo-mechanical routes. Additional electron backscattered diffraction experiments revealed the crystalline textures in the ferritic polycrystalline structure of the ODS steel samples.

To investigate fuel cladding chemical interaction in irradiated metallic nuclear fuels, diffusion couple experiments have been performed using prototypic metallic fuel alloys with additions of noble metal and lanthanide fission product components mated against stainless steel claddings. The developed interdiffusion zones have been characterized using SEM/EDS/WDS to determine the development of phases and the interdiffusion behavior of specific fuel, cladding, and fission product components. The formed diffusion structures have been compared to actual interaction zones that form in irradiated metallic SFR fuels. This paper discusses how the structures compare between the diffusion couple-generated interdiffusion zones and those that develop in irradiated metallic nuclear fuels. It was found that similarities exist between the phase development and interdiffusion behavior in the annealed diffusion couples and the irradiated fuels. Nd, Mo, and Ru, which were added to a fuel alloy to represent fission products that are present in irradiated metallic nuclear fuels, were found to exhibit interdiffusion behavior in annealed diffusion couples that was similar to what has been observed in actual irradiated metallic fuels. This was also true for the original fuel components U, Pu, and Zr, along with the cladding constituent Fe, Ni, and Cr.

The materials used for the SO3 decomposer in Iodine-Sulfur (IS) cycle for Nuclear Hydrogen Production System require excellent mechanical properties as well as a high corrosion resistance in SO2/SO3 environment at an elevated temperature up to 950°C. So far, no metallic materials have been suggested to be useful in such an environment. A surface modification of Hastelloy X by a SiC coating processed by an electron beam evaporative deposition has been studied in combination with an ion beam mixing (IBM) and an ion beam hammering (IBH). The simply deposited SiC film on the Hastelloy X substrate is easily peeled-off during an annealing at a high temperature due to a huge difference in their coefficients of thermal expansion (CTE), however the SiC coating on Hastelloy X prepared with IBM is sustained above 900 °C when the heating rate is less than 10 °C/min. The process of coating and IBM consists of a thin SiC film deposition, a subsequent N ion beam bombardment, and then an additional deposition of the film to the designed thickness. IBM plays a role of fastening the SiC film on the Hastelloy X substrate until the interfacial reaction takes place. Once the reaction takes place, new phases are developed at the interface under the consumption of the film and the substrate materials, producing a buffer layer. Without IBM, the SiC film tends to be easily detached during an annealing before the interfacial reaction initiates. The SiC film prepared with IBM requires a post-deposition annealing in vacuum for the interfacial reaction. However, the sublimation of SiC film prepared by an electron beam evaporative deposition occurs at the temperature above 900 °Cdecreasing the thickness of the deposited film. The sublimation of the SiC film can be prevented by IBH in which ion beams are bombarded onto the deposited film. This may be attributed to an ion beam bombardment induced densification of the deposited film. The resultant SiC coated Hastelloy X prepared by IBM and IBH exhibits a high corrosion resistance in a sulfuric acid at 300 °C, suggesting a possible application for the IS cycle.

Efforts of many scientists for more than a half of a century have resulted in substantial understanding of the response of various materials to irradiation. The theory has contributed significantly to this process but has not acquired a status allowing it to play a decisive role in creating radiation-resistant materials. Moreover, some theoretical predictions are in contradiction with observations, which indicates that something important has escaped attention. In the present paper, the current theoretical framework and experimental data are analyzed to elucidate the reasons for such a situation. A way of developing a predictive theory is proposed.

The structure and role of grain boundaries is investigated using an atomic analysis of the grain boundary movement during Molecular Dynamics displacement cascade simulations of bcc Fe. The results show the grain boundary to be a flexible entity. Local restructuring of the GB accommodates the incoming self interstitial atoms with local kinks, or small movements of a few atomic spacings occurring when the grain boundary is engulfed in the displacement cascade. The damage created is investigated using two potentials: the Ackland (non-magnetic) and the Dudarev- Derlet (magnetic) to study the role and influence of magnetism on the results obtained.

Molecular dynamics simulations provide an atomistic description of the processes that control primary radiation damage formation in atomic displacement cascades. An extensive database of simulations describing cascade damage production in single crystal iron has been compiled using a modified version of the interatomic potential developed by Finnis and Sinclair. This same potential has been used to investigate primary damage formation in nanocrystalline iron in order to have a direct comparison with the single crystal results. A statistically significant number of simulations were carried out at cascade energies of 10 keV and 20 keV and temperatures of 100 and 600K to make this comparison. The results demonstrate a significant influence of nearby grain boundaries as a sink for mobile defects during the cascade cooling phase. This alters the residual primary damage that survives the cascade event. Compared to single crystal, substantially fewer interstitials survive in the nanograined iron, while the number of surviving vacancies is similar or slightly greater than the single crystal result. The fraction of the surviving interstitials contained in clusters is also reduced. The asymmetry in the survival of the two types of point defects is likely to alter damage accumulation at longer times.

This study describes the microstructures of an ODS/Fe12Cr alloy before and after Fe+ irradiation at 500°C to a dose of 2.5 dpa and compares these microstructures with those of a reference Fe12Cr alloy produced following the same pulvimetallurgical route. The grain and dislocation structures of the alloys did not seem to change after irradiation. The main difference after irradiation was the appearance of small loops, which had similar sizes, distribution and densities in the ODS and reference alloys. The size distributions and chemical compositions of the ODS particles were similar to those found before irradiation.

The interaction between dislocation sliding and damage structure in ion-irradiated austenitic stainless steels was investigated. Solution annealed type 316 and 304 stainless steels (316SS and 304SS) were irradiated with 2.8 MeV Fe2+ ions at 300 °C up to 10 dpa and tensiled to 2% plastic strain at 300 °C. Dislocations moving from unirradiated matrix were prevented due to the interactions with the damage structures consisted of dislocation loops and voids in the damage region. The prevention of dislocation movements by the damage structures became strong in 304SS compared in 316SS; probably due to lower stacking fault energy in 304SS. The prevention of dislocation movements was weak for Fe ion-irradiated specimens in which the increase in shear strength calculated from the size and number density of the defects was small compared to He ion-irradiated specimens.

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances under creep-fatigue hold the key to success. This paper presents extended experimental results obtained from creep, fatigue and creep-fatigue tests on the main structural materials retained for these concepts, namely: stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and its low activation derivatives such as Eurofer steel, and their more advanced grades strengthened by oxide dispersion. It shows that the existing recommendations made in design codes adequately cover individual damage due to creep or fatigue but often fall short under combined creep-fatigue interaction. This is partly due to the difficulties of reproducing service conditions in laboratory. In this paper, results from tests performed on components removed from reactor, after long service, are used to refine code recommendations.

Using the above combined assessment, it is concluded that there is good confidence in predicting creep-fatigue damage for austenitic stainless steels. For the martensitic steels the effects of cyclic softening and microstructure coarsening throughout the fatigue life need more consideration in creep-fatigue recommendation. In the long-term development of ferritic/martensitic oxide dispersion strengthened grades with stable microstructure and no cyclic softening, appears promising provided problems associated with their fabrication and embrittlement are resolved.

The Metropolis Monte Carlo (MMC) algorithm is a computational method to study equilibrium thermodynamic properties of a system at the atomic level. The algorithm accounts for all terms that contribute to defining the free energy difference between states: not only chemical, configurational and interfacial, but also due to strain fields and thermal vibrations. In this work, the MMC method with a two bands empirical many-body potential is used to predict the ordering properties of Fe1-xCrx alloys at various compositions and temperatures in the absence of defects. The particular goal of the work was to reveal the effect of atomic relaxations and vibrations on the phase diagram. It is found that vibrations and local relaxation effects contribute to lowering the order-disorder transition temperature by about 25 percent as compared to MMC predictions with a rigid lattice.

HCM12A is a ferritic-martensitic steel alloy envisioned for cladding and structural material in the Generation IV Supercritical Water Reactor (SCWR). This alloy was oxidized in 600°C supercritical water for 2, 4 and 6 weeks, and the oxide layers formed were analyzed using microbeam synchrotron radiation and electron microscopy. The oxide layers show a three-layer structure with an Fe3O4 outer layer, an inner layer containing a mixture of Fe3O4 and FeCr2O4 and a diffusion layer containing FeCr2O4 and Cr2O3 precipitates along ferrite lath boundaries. The base metal microstructure has a strong influence on the advancement of the oxide layers, due to the segregation at the lath boundaries of chromium rich particles, which are oxidized preferentially.

A novel material design for nuclear fusion reactors is proposed based on Cu-Diamond and W-Diamond nanocomposites. The proposed design involves the production of W/W-Diamond/CuDiamond/Cu functionally graded material. W, W-Diamond, Cu-Diamond and Cu nanostructured composite powders were produced independently by mechanical alloying and subsequently consolidated/welded through spark plasma sintering. Modulation of processing parameters allowed controlling the extent of unfavourable conversion of Diamond into tungsten carbide, as well as to overcome the Diamond intrinsically difficult bonding to copper. Microstructural features and microhardness of the as-produced materials are presented.

We have evaluated phosphorus (P) segregation in ion-irradiated nickel (Ni) by the rate theory model incorporating the results of first principles calculations. We find from our first principles calculation that the transport of P via the rotation mode of a mixed-dumbbell is unlikely to occur, and the transport coefficient of phosphorus by the vacancy mechanism is much larger than that reported previously. On the basis of our first principles results, we have also proposed to include the effect of free migration of P via the octahedral interstitial site of FCC Ni crystal in the rate theory model. With all these renewed parameters, we have successfully obtained the P distribution in irradiated Ni, which is very close to experiment, by adjusting the effect of P transport by the vacancy mechanism.

The microstructures of tungsten, molybdenum and copper wires used as Langmuir probes at ISTTOK edge plasma have been investigated. The probes cross-sections evidenced extensive grain growth, intergranular bubbles and increased hardness at the plasma exposed regions. Internal surfaces of large bubbles exhibited slip bands resulting from plastic deformation induced by high H2 pressure. Elastic recoil detection analysis was used to measure H concentration profiles. The present results suggest that H2 bubble formation in first wall components under long-term high thermal loads should be closely monitored in nuclear fusion devices. Strategies for H damage mitigation are proposed and discussed.

The thermal characteristics of the spent fission fuel from a hybrid fusion-fission system (LIFE) capable of extremely high burnups are described. The waste has higher thermal output per unit mass of heavy metal, but lower thermal output per unit energy generated. Plausible designs for interim storage containers and cooling configuration can remove the heat without exceeding fuel temperature limits. Calculations show that a spent LIFE fuel repository would perform within the limits established for the proposed repository at Yucca Mountain.

In our past work, two types of nanometer-sized oxide particle were found in the commercial ODS (Oxide Dispersion Strengthened) ferritic alloy MA957. They were nonstoichiometric Y-, Ti-, O-enriched clusters (2 ∼ 15 nm) and stoichiometric Y2Ti2O7 (15 ∼ 35 nm). The particles which have an important role for the strengthening were confirmed to be nonstoichiometric Y-, Ti-, O-enriched clusters (denoted by YxTiyOz in this work). Based on that result, the stability and evolution mechanism of oxide particle at high temperatures were studied in this work. After annealing (1473 K × 1 h), most of the non-stoichiometric YxTiyOz particles were stable and their size increased very little. However, some these particles apparently became Y2Ti2O7 particles with increasing size and yttrium content. They possibly contributed to the growth of Y2Ti2O7.

Density functional theory and Monte(Carlo methods were used to investigate the solid( solution behavior of actinide dioxides (AcO2). The end(members of interest include: ZrO2, ThO2, UO2, NpO2, and PuO2; all have the isometric fluorite structure. Ab initio and subsequent Monte( Carlo simulations are used to calculate the excess enthalpy of mixing (ΔHexcess), excess Gibbs free energy of mixing (ΔGexcess), and excess configurational entropy (ΔSexcess) for the above solid(solution series. From ΔGexcess, phase diagrams are derived and miscibility gaps identified. All of the binaries of the aforementioned end(members were studied; however, this paper focuses on the U1(xZrxO2 and Np1(xUxO2 binaries. About 25 at.% Zr can be in solid solution with the UO2 matrix above 1500 K, while Np is completely miscible in the UO2 matrix. Partial cation ordering was observed at all temperatures for the U1(xZrxO2 binary. The Np1(xUxO2 binary approaches perfect cation disorder at high temperatures (2000 K). The cation ordering scheme is not identified in this study because the number of cation(cation interaction parameters was limited by the single unit cell from the ab initio calculations.

The quantity, radioactivity, and isotopic characteristics of the spent fission fuel from a hybrid fusion-fission system capable of extremely high burnups are described. The waste generally has higher activity per unit mass of heavy metal, but much lower activity per unit energy generated. The very long-term radioactivity is dominated by fission products. Simple scaling calculations suggest that the dose from a repository containing such waste would be dominated by 129I, 135Cs, and 242Pu. Use of such a system for generating energy would greatly reduce the need for repository capacity

Corrosion of ferritic steels, including oxide dispersion strengthened (ODS) variants, in high temperature molten fluoride salts may limit the life of advanced reactors, including some hybrid systems that are now under consideration. In some cases, the steel may be protected through galvanic coupling with other less noble materials with special neutronic properties such as beryllium. This paper reports the development of a model for predicting corrosion rates for various ferritic steels, with and without oxide dispersion strengthening, in FLiBe (Li2BeF4) and FLiNaK (Li-Na-K-F) coolants at temperatures up to 800 °C. Mixed potential theory is used to account for the protection of steel by beryllium, Tafel kinetics are used to predict rates of dissolution as a function of temperature and potential, and the thinning of the mass-transfer boundary layer with increasing Reynolds number is accounted for with dimensionless correlations. The model also accounts for the deceleration of corrosion as the coolants become saturated with dissolved chromium and iron. Electrochemical impedance spectroscopy has been used for the initial in situ study of an ODS ferritic steel in high-temperature molten fluoride salt environments, with the complex impedance spectra obtained at its open circuit corrosion potential (OCP) interpreted in terms of the basic components of the equivalent circuit, which include the electrolyte conductivity, the interfacial charge transfer resistance, and the interfacial capacitance. Such in situ measurement techniques may provide valuable insight into the degradation of materials under realistic conditions.