Effect of temperature on graphite oxidation behavior

The temperature dependence of oxidation behavior for the graphite IG-11, used in the HTR-10, was investigated by thermogravimetric analysis in the temperature range of 400–1200 °C. The oxidant was dry air (water content <2 ppm) with a flow rate of 20 ml/min. The oxidation time was 4 h. The oxidation results exhibited three regimes: in the 400–600 °C range, the activation energy was 158.56 kJ/mol and oxidation was controlled by chemical reaction; in the 600–800 °C range, the activation energy was 72.01 kJ/mol and oxidation kinetics were controlled by in-pore diffusion; when the temperature was over 800 °C, the activation energy was very low and oxidation was controlled by the boundary layer. Due to CO production, the oxidation rate increased at high temperatures. The effect of burn-off on activation energy was also investigated. In the 600–800 °C range, the activation energy decreased with burn-off. Results of low temperature tests were very dispersible because the oxidation behavior at low temperatures is sensitive to inhomogeneous distribution of any impurity, and some impurities can catalyse graphite oxidation.     

UKAEA Reactor Group studies of irradiation-induced creep in graphite

A review is presented of the experimental and theoretical studies of fast neutron irradiation creep in reactor graphite carried out by the UKAEA Reactor Group. The studies have covered the effect of varying graphite type, oxidation, stress level and boron doping. The results are shown to accord better with a dislocation pinning-unpinning model of dislocation glide, rather than the Cottrell model often assumed.     

Nuclear reaction studies of 15N implanted spheroidal graphite cast iron using an ion microprobe

The relative concentrations of the matrix and graphite nodules of a spheroidal graphite cast iron implanted with ¹⁵N (200 keV, 2 × 10⁷ ions/cm², 440 K) have been studied by nuclear microprobe using the ¹⁵N(p, αγ)¹²C reaction at 898 keV.A beam diameter of less than 20 μm enabled individual nodules to be investigated independently to the matrix for two samples of differing microstructure. The matrix of the first sample was fine pearlite and that of the second sample was martensite.Significant diffusion of the implanted nitrogen out of the matrix into the nodules had taken place in the pearlitic sample, resulting in broadening of the nitrogen distribution. This is believed to be due to the large number of grain boundaries present in the matrix. Diffusion in the hardened iron was limited to a migration towards the surface, whilst there was no appreciable migration into the nodule.     

Wear performance and activity reduction effect of Co-free valves in PWR environment

Co-based hardfacing alloys exposed to PWR primary coolant may be replaced with Co-free alloys to lower occupational radiation exposure. To evaluate the viability of Co-free hardfacing alloys, we conducted hot-water tests for gate valves hardfaced with NOREM™ 02 (Fe-base), Deloro™ 50 (Ni-base), and Stellite™ 6 (Co-base). Using a high flow test loop, on–off cycling tests were conducted in 280 °C water. It was observed that NOREM 02 exhibited galling and excessive leak after 1000 cycle test whereas no leakage was developed with Deloro 50 after 2000 cycles. To estimate the activity reduction effect of Co-free hardfacing alloys, an existing activity transport model was modified. It is found that the main contributor of Co activity buildup is the corrosion of steam generator (SG) tubing. The Korean Next Generation Reactor (APR-1400) tubed with alloy 690 having a reduced cobalt impurity allowance is expected to have 73% lower Co activity on SG surface compared with the case of alloy 600 tubing. The complete replacement of Stellite 6 with Co-free hardfacing alloys is expected to cut additional 5% of activity which may be too small to justify the risk of galling and leakage development as revealed by the hot-water test.     

Channelling studies on the lattice location of B atoms in graphite

In order to obtain information on the lattice location of B atoms in graphite, channelling experiments have been performed at room temperature with a proton beam of an energy of 0.65-0.77 MeV for the 〈0 0 0 1〉 axial channel in highly oriented pyrolytic graphite (HOPG) crystals doped with 0.32 at.% B. The B atoms are detected by measuring α-particles which are emitted as a result of a nuclear reaction ¹¹B(p,α)αα. It is clearly demonstrated that most of B atoms are shadowed behind the 〈0 0 0 1〉 C atomic rows. Taking account of the already reported experimental results on a change of lattice parameters by B-doping, it is concluded that most of B atoms are located at substitutional sites. It is also observed that B-doping introduces lattice strain on the c-plane. In addition, the presence of a small portion of interstitial B atoms is suggested.     

Release behavior of tritium from graphite heavily irradiated by neutrons

The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of samples heavily irradiated by neutrons. The lithium impurities in graphite were estimated as the source of tritium formation. The main chemical form of released tritium was hydrogen accompanied by a small quantity of methane. No other hydrocarbons could be detected. Tritiated water was always measured, but the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule. The release spectrum of tritium in isochronal annealing was shifted to a higher heating temperature with the increase of the neutron fluence received by the graphite crystal. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.     

The various creep models for irradiation behavior of nuclear graphite

Graphite is a widely used material in nuclear reactors, especially in high temperature gascooled reactors (HTRs), in which it plays three main roles: moderator, reflector and structure material. Irradiation-induced creep has a significant impact on the behavior of nuclear graphite as graphite is used in high temperature and neutron irradiation environments. Thus the creep coefficient becomes a key factor in stress analysis and lifetime prediction of nuclear graphite. Numerous creep models have been established, including the visco-elastic model, UK model, and Kennedy model. A Fortran code based on user subroutines of MSC.MARC was developed in INET in order to perform three-dimensional finite element analysis of irradiation behavior of the graphite components for HTRs in 2008, and the creep model used is for the visco-elastic model only. Recently the code has been updated and can be applied to two other models—the UK model and the Kennedy model. In the present study, all three models were used for calculations in the temperature range of 280–450 °C and the results are contrasted. The associated constitutive law for the simulation of irradiated graphite covering properties, dimensional changes, and creep is also briefly reviewed in this paper. It is shown that the trends of stresses and life prediction of the three models are the same, but in most cases the Kennedy model gives the most conservative results while the UK model gives the least conservative results. Additionally, the influence of the creep strain ratio is limited, while the absence of primary creep strain leads to a great increase of failure probability.     

X-ray photoelectron and Raman spectroscopic studies of MeV proton irradiated graphite

Poly-crystalline graphite samples were irradiated using 2.25 MeV H⁺ ions with a fluence of 2 × 10¹⁷ ions/cm². Magnetic ordering in highly oriented pyrolytic graphite samples have been reported earlier under the similar irradiation conditions [Esquinazi et al., Phys. Rev. Lett. 91 (2003) 227201]. In that study, the authors attribute the observed irradiation induced magnetic ordering to the formation of a mixed sp²-sp³ hybridized carbon atoms. In the present study, we report the X-ray photoelectron and Raman spectroscopic studies on pristine and irradiated samples. Irradiated samples are found to show an increased number of sp³ hybridized carbon atoms. However, the Raman spectrum, specially the second order data, do indicate that the nature of the graphene lattice structure has been preserved in the irradiated samples. The mechanisms for the irradiation induced enhancement in sp³ hybridization are discussed.     

The modeling of graphite oxidation behavior for HTGR fuel coolant channels under normal operating conditions

The oxidation of graphite in normal operating conditions is a very important factor when evaluating the service time of the graphite structural material in a high temperature gas-cooled reactor (HTGR). This paper deals with the modeling of graphite oxidation by steam in the helium channel of a fuel block. The FEM software COMSOL is used: the turbulent flow of the coolant is simulated by using the k-? model and the chemical reaction is expressed by the Langmuir-Hinshelwood equation. Calculations were carried out for steam pressures around 1 Pa and for different temperature distributions. The influence of burn-off and the diffusion in graphite porosities were both considered in the oxidation. Results show that oxidation mainly occurred on the graphite surface at the bottom of the core because of the higher temperature. The thickness of graphite with a burn-off higher than 8% was about 1 mm at the core base. Less than 15% of steam was consumed in the coolant channel of the fuel assemblies. Calculations also showed that the mean gasification rate in one channel for the second service time was larger than the first service time.     

地质环境中Am的化学行为研究进展

评述了近年来与高放废物深部地质处置有关的地质环境中Am化学的研究进展,讨论了Am在地下水中的存在价态、溶解度、配合反应(包括水解反应)、胶体的形成等对迁移的影响.提出了近期和长期应开展的若干研究课题.     

Deuterium implantation in graphite

A study has been made of deuterium ion implantation in polycrystalline graphite over the temperature range 100 K to 450 K. Deuterium depth profiles are obtained using a method based on observing the energy spectra of the tritons and protons from D-D reactions initiated with a 200 keV probing deuteron beam. In marked contrast to the low adsorption for gaseous hydrogen previously reported, graphite is found here to retain large concentrations of implanted deuterium with atomic loadings exceeding unity in the deeper regions of a cold target. The amount of deuterium retained continues to decrease with increasing temperature until at 150°C the retention has fallen to approximately half that at −160°C. No evidence is found for a layer of surface deuterium of the type seen in metals such as copper. Scanning electron micrographs of irradiated areas show no observable surface deformation for doses of up to 2.4 × 10¹⁹ deuterons/cm².     

Effect of environment, mechanical conditions, and materials characteristics on AE behavior during corrosion fatigue processes of an austenitic stainless steel

Acoustic emission (AE) behavior during corrosion fatigue (CF) processes has been studied in a commercial grade 304 austenitic stainless steel with special attention to the effect of environment, mechanical conditions, and materials characteristics. Precracked compact tension specimens were tested under cyclic tension-tension load, polarized potentiostatically in 1N H₂SO₄ + 0.5M HaCl or 1N H₂SO₄ solutions at room temperature. Valuable AE signals discriminated from environmental and mechanical noises by source location were monitored at the loading phase near the maximum load. The experimental results showed that the AE activity observed in corrosive environments increased significantly with the acceleration of crack growth rates, compared with that observed in air (inert environment). Detailed SEM observations found cleavage-like cracking, intergranular-like cracking, separations, etc. on the crack surfaces. The AE sources which provided the high AE activity during CF crack growth were shown to be such microcracking processes as cleavage-like cracking, intergranular-like cracking, and separations caused by the cooperation of environments (intergranular corrosion or hydrogen), mechanical conditions (triaxial stress), and materials characteristics (non-metallic inclusions or carbides precipitated along grain boundaries by sensitization). The effect of these three factors on the AE behavior and cracking mechanisms is discussed in detail.     

Experimental study on the oxidation behavior and microstructural evolution of NG-CT-10 and NG-CT-20 nuclear graphite

NG-CT-10 and NG-CT-20 are newly developed grades of nuclear-grade graphite from China. In this study,their oxidation behaviors were experimentally investigated using thermal gravimetric analysis. Microstructural evolution before and after oxidation was investigated using scanning electron microscope, mercury intrusion, and Raman spectroscopy. The apparent activation energy of NG-CT-10 nuclear graphite is 161.4 k J/mol in a reaction temperature range of 550–700 °C and that of NG-CT-20 is153.5 kJ/mol in a temperature range of 550–650 °C. The activation energy in the inner diffusion control regime is approximately half that in the kinetics control regime. At high temperatures, the binder phase is preferentially oxidized over the filler particles and small pores are generated in the binder. No new large or deep pores are generated on the graphite surfaces. Oxygen can diffuse along the boundaries of filler particles and through the binder phase, but cannot diffuse into the spaces between the nanocrystallites in the filler particles. Filler particles are oxidized starting at their outer surfaces, and the sizes of nanocrystallites do not decrease following oxidation.