Characterization of crud deposited on fuel rods under HWC environment in Kuosheng Nuclear Power Plant

Under normal water chemistry conditions, the oxygen and hydrogen peroxide produced by water radiolysis in the coolant of boiling water reactors(BWRs) can lead to intergranular stress corrosion cracking in the constituent materials of plant components. This fact has led to the wide-scale adoption of hydrogen water chemistry(HWC) in the nuclear industry to counteract these effects.This study seeks to characterize the metallic composition and the surface properties of the constituent materials of plant components in order to determine their effects on the accumulation of chalk river unidentified deposits(crud) on fuel rods in the BWR Unit-1 of the Kuosheng Nuclear Power Plant in Taiwan. Inductively coupled plasma-atomic emission spectroscopy was used to calculate the concentrations of surface crud and gamma spectrometry was used to determine the radioactivity of the corrosion products, as well as their axial distribution across the surface of the fuel rods. X-ray diffraction analysis and scanning electron microscopy/energy-dispersive X-ray spectroscopy were used to identify the crystalline phase and morphology of the crud as irregular shapes and flakes. The amount of crud deposited during the fourth fuel cycle exceeded that of the third fuel cycle due to extended burn-up time. Our analytical results indicate that the implementation of HWC had no significant effect on the characteristics of subsequent crud.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

太钢1549mm热连轧宽度控制原理及应用

全面系统地介绍了太钢1549粗轧宽度控制的原理及应用、影响产品宽度的因素、头尾短行程的控制及改进.液压AWC直接影响到产品的宽度精度,在热连轧带钢中起着至关重要的作用.     

Thermal decomposition of hydrazine in sub- and supercritical water at 25 MPa

Supercritical water flow-through test facility (SCW-TF) for the study of hydrothermal fluids is described. The hydrodynamic behavior of the flow-through reactor is examined from ambient to supercritical water conditions by performing residence time distribution measurements. The results indicate that at 25 MPa, the employed reactor configuration exhibits plug flow behavior with a small extent of dispersion over the temperature range from 298 to 773 K. The experimentally determined effective volume of the reactor was used for the calculation of mean residence times of the fluid in the “hot zone” of the flow-through system. The thermal stability of hydrazine in aqueous solution was examined along the 25 MPa isobar from 473 to 725 K. The obtained first-order rate constant for the thermal decomposition of hydrazine increases from 3.73 × 10⁻⁴ s⁻¹ at 473 K to about 0.31 s⁻¹ at 725 K.     

Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.     

Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The corrosion fatigue crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was systematically characterized under simulated boiling water reactor normal water and hydrogen water chemistry conditions by low-frequency fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C.In this paper, the observed synergistic effects of environmental, material and loading parameters on the environmental acceleration of fatigue crack growth in low-alloy RPV steels are discussed in the context of the Ford-Andresen model. Additionally, the adequacy and conservatism of the current “ASME XI reference fatigue crack growth curves” of the ASME Boiler & Pressure Vessel Code are critically reviewed and assessed on the basis of the gathered experimental data base and this model. Based on the observed cracking behaviour and the Ford-Andresen model, a simple time-domain superposition model is suggested, which could reduce most of the undue conservatism and eliminate uncertainties of the existing codes and therefore serve as a basis for the development of improved reference fatigue crack growth curves.