Canada has participated in the Generation IV nuclear energy systems focusing on the Supercritical Water-cooled Reactor (SCWR) system. The 64-element fuel rods channel is used in the Canadian SCWR. There are few publicly available experimental studies for the flow in the fuel channels with multiple fuel rods. To date, the CFD simulation reported in open literature for rod bundle flow in the 64-element SCWR is relatively scarce. The cladding surface temperature is of key importance in assessing the safety of the reactor. Since the inhomogeneities in the bundle cross-section can present complex flow phenomena, a CFD study can provide substantial insight into the flow physics. In this work, the full-scale CFD simulation of the supercritical water flow in the 64-element rod bundle was performed. The results suggest the possibility of the presence of gap vortices in the flow subchannels. Higher streamwise velocities and normal Reynolds stresses always exist at the center subchannel regions. The circumferential cladding surface temperature distribution is extremely non-uniform and there is a large difference between the maximum cladding surface temperatures for different fuel rods. Graphical Abstract

To improve the understanding of the microstructural stability of F82H under fusion reactor environments, the synergistic effects of irradiation and thermal aging on the instability of M₂₃C₆ particle was investigated utilizing ion accelerator irradiation. Results show that the W-rich Laves phase growing along the grain boundaries was observed in the specimen after thermal aging at 773 K up to 10,000 hr. A bilayer contrast of the M₂₃C₆ particles consisting of an amorphous-rim and inner crystalline core was observed in the specimen irradiated by 10.5 MeV-Fe³⁺ at 623 K, but not in the specimen irradiated at 673 K, demonstrating that the critical temperature of the radiation-induced amorphization (RIA) in thermally aged M₂₃C₆ particles falls in the range of 623–673 K. These results show that thermal aging seems to play an insignificant role in the RIA behavior of M₂₃C₆ particles. Regarding the radial distribution function analysis, a relatively poor crystallinity was observed in the unirradiated thermally aged M₂₃C₆ particles, while the crystallinity was improved in the specimen irradiated at 673 K, which is ascribed to enhanced diffusion at elevated temperatures. Results indicate that the temperature-dependent point defect generation and recovery is the key factor related to the occurrence of RIA in M₂₃C₆ particles. GRAPHICAL ABSTRACT

In designing a steam generator in nuclear power plant, it is necessary to evaluate the flow-induced vibration of a U-shaped tube bundle from the viewpoint of safety engineering. CFD technology is used for predicting velocity and void fraction distributions along the U-tubes to evaluate the force acting on the tube from the surrounding two-phase flow. This study validated a newly developed thermal-hydraulic simulation method based on the drift-flux model to simulate various operating conditions under flexible power operations, including partial load conditions. Constitutive equations for two-phase flow were implemented into the platform of a general-use commercial code, ANSYS fluent, through user-defined functions. Experimental data measured in the slice model U-tube bundle apparatus with boiling Freon two-phase flow, which was a simulant of steam-water under the prototypic pressure and thermal conditions, was utilized to validate the developed simulation method. The predicted void fraction and velocity distribution were compared with the experimental results. In the cold-side region, a low void fraction occurred in a partial load condition. It is considered to be caused by the liquid downward flow outside the bundle. It was confirmed that the developed simulation method could reproduce such phenomena in the secondary side of a steam generator.

When assisting emergency responses to a nuclear accident through atmospheric dispersion simulations, it is necessary to provide the prediction results and their uncertainties. This study develops an estimation method using machine learning for uncertainty in forecasted plume directions. The difference in plume directions derived from the meteorological forecast and analysis inputs was considered as the uncertainty in forecasted plume direction. Bayesian machine learning was used to predict the uncertainty based on the accumulated uncertainty estimation result in past cases. A three-day forecast simulation was conducted every day from 2015 to 2020, considering a hypothetical release of ¹³⁷Cs from a nuclear facility to create training and test datasets for the machine learning. The findings reveal that the rate of good predictability was greater than 50% even in the forecast 36 h later when investigating the effectiveness of the Bayesian model on uncertainty prediction. The frequency of miss prediction of higher uncertainty was low (0.9%−7.9%) throughout the forecast period. However, the rate of over-prediction of uncertainty increased with forecast time up to 31.2%, which is acceptable as a conservative estimation. These results show that the Bayesian model in this study effectively estimates the uncertainty of plume directions predicted through atmospheric dispersion simulations. GRAPHICAL ABSTRACT

The irradiation hardening behavior and the void swelling behavior of the Japanese reduced-activation ferritic/martensitic steel, i.e. F82H IEA heat, were investigated by using single- and dual-ion irradiation experiments in this study. For irradiation hardening of F82H, a map of the hardness ratio before/after ion irradiation was obtained for 270–500 °C up to 80 dpa. The peak temperature of irradiation hardening was at about 350°C. On the other hand, it was shown that an extra hardening occurred in helium co-implanted F82H for 400–430 °C except for the temperatures less than 400 °C and more than 450 °C. For the temperature dependence of void swelling in helium co-implanted F82H, it was confirmed that the peak swelling temperature was identified at about 470 °C. That amount of swelling (average value) was 0.4% at 20 dpa. GRAPHICAL ABSTRACT

Dynamic probabilistic risk assessment (DPRA) of nuclear power plants (NPPs) has become one of the most critical research areas, especially in the aftermath of the 2011 Fukushima Daiichi nuclear accident. Uncertainty in NPP behavior is key when considering its safety under different operating conditions. Such uncertainty typically results from operation parameters, system conditions, and modeling assumptions. This study integrates the system dynamics (SD) modelling approach with an uncertainty analysis method to quantify the dynamic probabilistic risk in NPPs. To demonstrate the approach’s applicability, the average fuel temperature is used to estimate the probability of reactor core damage under different transients, representing perturbations in reactivity and steam valve coefficient. A Monte Carlo simulation is employed to investigate the effect of uncertainties associated with the different model parameters. A global sensitivity analysis demonstrates that the total delayed neutron fraction, the heat transfer coefficient from fuel to coolant, the coolant temperature coefficient of reactivity, and the fuel temperature coefficient of reactivity are the primary controllers of the plant response variability under the transients considered. In summary, the integration of SD modelling and uncertainty analysis presents an effective DPRA approach that overcomes the limitations of static counterparts while minimizing the computational resources required. Graphical Abstract

There are two failure criteria in seismic probabilistic risk assessment of a nuclear power plant to define component failure caused by an earthquake. The first is in terms of peak ground acceleration and peak ground acceleration capacity. The second is in terms of seismic response and component capacity. These criteria are closely related, but they are not equivalent. One can derive a different fragility curve from the second criterion. First, this study analytically shows the relation of these failure criteria, pointing out that a fragility curve may not represent the uncertainty of a failure probability derived from the second criterion. In addition, we propose a probability density function of a failure probability based on the second criterion using the transformation of a random variable method applied to probability density functions of seismic response and component capacity. Then, we derive a new fragility curve based on the derived probability density function. Also, we show that the derived probability density function links to a likelihood function based on a mean fragility curve used in the literature. Finally, we discuss the non-identifiability of uncertainties of response and capacity, and we propose a Bayesian model utilizing a local response to overcome the non-identifiability.

In the use of coal fly ash (FA) mixed cement, the assessment of the leachabilities of toxic elements in FA is crucial. This study evaluated the leaching behavior of arsenic and selenium in 40 types of raw FA (R-FA), aging-treated FA (A-FA), and cement-mixed FA (C-FA). The mean leaching concentrations from R-FAs were 0.03 mg/L for arsenic and 0.09 mg/L for selenium. The leaching of arsenic and selenium was suppressed by the aging treatment, and more significantly suppressed by the cement mixing under 0.01 mg/L for both arsenic and selenium. The pH for R-FAs ranged from 7.6 to 12.9, and that for C-FAs ranged from 11.1 to 12.3 with a smaller pH range due to cement mixing. Furthermore, the results of the leaching experiment with time for R-FAs and C-FAs indicated that the leaching of arsenic and selenium was suppressed by the formation of secondary compounds with calcium due to the supply of calcium from the cement. These results indicate that the common FA cement and low alkaline cement used in the actual site can sufficiently suppress the leaching of arsenic and selenium because the ratio of cement amount to FA is higher than that examined in this study. GRAPHICAL ABSTRACT

For a subcritical state of Kyoto University Reactor (KUR), Rossi-α and Feynman-α analyses were carried out using a water-surrounded fission counter assigned as reactor startup channel, where the Rossi-α analysis had a long time-interval range from 0.5 µs to 2 ms and the Feynman-α analysis had a wide gate-width range from 1 µs to 5 ms. The result in a shorter time range than several microseconds shows clearly a rise in a positive correlation amplitude resulting from spurious count, which is a well-known characteristic of fission counter. A time constant of the generating process of the spurious count and a probability of occurrence of a spurious count subsequently to a neutron count can be determined from these analyses. The time constant is around 1 µs, and the occurrence probability is around 0.1%, which is one digit smaller than that evaluated from a previous Feynman-α analysis using a graphite-surrounded fission counter. The considerable difference between the present and the previous occurrence probability indicates that the probability must strongly depend on surroundings of fission counter.

T-22 alloy is applied to the steam generator of the high-Temperature Reactor–Pebblebed Modules (HTR-PM). In this study, a corrosion experiment of T-22 alloy was carried out for 50 h under the environment of air, vacuum, and impure helium at ultra-high temperature (950°C). After the experiment, the corrosion behaviors of T-22 alloy were characterized and analyzed by electronic balance, scanning electron microscope (SEM), energy-dispersive spectroscopy (EDS), carbon sulfur analyzer, and gas chromatography. The results show that the T-22 alloy was seriously corroded and destroyed under the air ingress accident; in the vacuum environment, the alloy appeared mass loss and slight decarburization; in the impure helium environment, the microclimate reaction would occur on the surface of T-22 alloy, and the oxide on the surface of the alloy would be destroyed and the alloy decarburized obviously; CO₂ in the ppm range will accelerate the oxidation and decarburization of the alloy.