Band 87, Heft 2

In the present study, the effect of perforated hollow cylinders and twisted tapes on fluid flow and heat transfer behavior of a circular tube is presented for the Reynolds number ( Re ) range of 6000–40,000. The diametrical ratio ( DR ) and perforation index ( PI ) of compound devices are varied to determine the Nusselt number ( Nu ) and friction factor ( f ) for the combined perforated hollow cylinder (PHC) and twisted tape (TT) insert. The maximum value of Nusselt number ratio is found to be 3.19 and the minimum value of the friction factor ratio is obtained as 2.24. The PHC–TT combination provides the maximum thermal enhancement factor (TEF) of 2.43 for PI  = 8%, and DR  = 0.5.

In this study, the pool boiling performance of oxide nanofluid was investigated, the heating surface is a 5 × 30 mm stainless steel heating surface. Three kinds of nanofluids were selected to explore their critical heat flux (CHF) and heat transfer coefficient (HTC), which were TiO 2 , SiO 2 , Al 2 O 3 . We observed that these nanofluids enhanced CHF compared to R·O water, and Al 2 O 3 case has the most significant enhancement (up to 66.7%), furthermore, the HTC was also enhanced. The number of bubbles in nanofluid case was relatively less than that in R·O water case, but the bubbles were much larger. The heating surface was characterized and it was found that there were nano-particles deposited, and surface roughness decreased. The wettability also decreased with the increase in CHF.

For post-Fukushima nuclear power plants, there has been interested in accident-tolerant fuel (ATF) since it has better tolerant in the event of a severe accident. The fully ceramic microencapsulated (FCM) fuel is one kind of the ATF materials. In this study, the small modular pressurized water reactor (PWR) loading with FCM fuels was investigated, and the modified Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor (CANDLE) burnup strategy was successfully applied to such compact reactor core. To obtain ideal CANDLE shape, it’s necessary to set the infinity or enough length of the core height, but that is impossible for small compact core setting infinity or enough length of the core height. Due to the compact and finite core, the equilibrium state can only be maintained short periods and is not obvious, other than infinitely long active core to reach the long equilibrium state for ideal CANDLE. Consequently, the modified CANDLE shape would be presented. The approximate characteristics of CANDLE burnup are observed in the finite and compact core, and the power density and fuel burnup are selected as main characteristic of modified CANDLE burnup. In this study, firstly, lots of optimization schemes were discussed, and one of optimization schemes was chosen at last to demonstrate the modified CANDLE burnup strategy. Secondly, for chosen compact small rector core, the modified CANDLE burnup strategy is applied and presented. Consequently, the new characteristics of this reactor core can be discovered both in ignition region and in fertile region. The results show that application of CANDLE burnup strategy to small modular PWR loading with FCM fuels suppresses the excess reactivity effectively and reduces the risk of small PWR reactivity-induced accidents during the whole core life, which makes the reactor control more safety and simple.

Opinion mining and sentiment analysis based on social media has been developed these years, especially with the popularity of social media and the development of machine learning. But in the community of nuclear engineering and technology, sentiment analysis is seldom studied, let alone the automatic analysis by using machine learning algorithms. This work concentrates on the public sentiment mining of nuclear energy in German-speaking countries based on the public comments of nuclear news in social media by using the automatic methodology, since compared with the news itself, the comments are closer to the public real opinions. The results showed that majority comments kept in neutral sentiment. 23% of comments were in positive tones, which were approximate 4 times those in negative tones. The concerning issues of the public are the innovative technology development, safety, nuclear waste, accidents and the cost of nuclear power. Decision tree, random forest and long short-term memory networks (LSTM) are adopted for the automatic sentiment analysis. The results show that all of the proposed methods can be applied in practice to some extent. But as a deep learning algorithm, LSTM gets the highest accuracy approximately 85.6% with also the best robustness of all.

Centrifuge cascade hydraulics research is very important for the cascade system design and safety analysis. The one-dimensional centrifugal cascade dynamic hydraulics calculation program can be used to achieve the rapid verification of the cascade system dynamic operation or accident condition. While the computational fluid dynamics (CFD) program is mainly used to analyze the local three-dimensional fluid phenomena such as the centrifugal cascade pipe and equipment. To comprehensively utilize the advantages of the two simulation methods, based on the one-dimensional calculation software SimuWorks of centrifugal cascade hydraulics and three-dimensional flow field calculation software Fluent, a one-dimensional and three-dimensional coupling simulation program for centrifugal cascade hydraulics was developed by using the API functions provided by SimuWorks simulation software and the user-defined function (UDF) of Fluent. The coupling method is used to analyze the movement of solid particles in the main feed pipe of the separation stage. By comparing the calculation results of the coupling program with the calculation results of the traditional one-dimensional dynamic hydraulics program, it can be seen that the coupling program can correctly predict the influence of feeding orifice plate clogging on the system parameters and can more intuitively show the flow field characteristics in the important parts of the system. The motion characteristics of the solid particles obtained in the simulation verification are consistent with the experimental analysis results, which further verifies the accuracy of the coupling program. The coupling program can provide a new calculation method and analytical tool for the study of centrifugal cascade hydraulics.

The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

The thermodynamics and kinetics of bromide ion-isotopic exchange reactions performed by using fresh and UV radiation degraded industrial grade anion exchange resin Duolite A-638 resins were studied by application of radio analytical tracer technique. The reaction rate k in min −1 for the fresh resin decreases sharply with decrease in wavelength of UV radiations. Thus for 0.200 M labeled bromide ion solution maintained at a constant temperature of 30.0 °C, the k value for fresh resin (0.367 min −1 ), decreases to 0.335 min −1 for λ 384 UV radiation degraded resin, which further decreases to 0.273 min −1 for λ 284 UV radiation degraded resin. Under identical experimental conditions, the thermodynamic parameters like energy of activation (−1.65 kJ/mol), enthalpy of activation (−4.24 kJ/mol), free energy of activation (64.85 kJ/mol), and entropy of activation (−0.229 kJ/K/mol) calculated for the fresh resin increases to −1.61 kJ/mol, −4.19 kJ/mol, 64.92 kJ/mol, and −0.228 kJ/K/mol respectively for λ 384 UV radiation degraded resin; which further increases to −1.60 kJ/mol, −4.18 kJ/mol, 65.17 kJ/mol, and −0.228 kJ/K/mol respectively for λ 284 UV radiation degraded resin. Increase in thermodynamic parameters calculated here for the fresh and degraded resins suggests that decrease in wavelength of UV radiations has catastrophic effect on the resin making the bromide ion-isotopic exchange reactions thermodynamically less feasible. The impact of UV radiation on resin degradation was supported by their characterization study using Fourier-transform infrared spectroscopy (FTIR) and Scanning Electron Microscopy (SEM) techniques.

Radioactive waste is generated from activities that utilize nuclear materials such as nuclear medicine or power plants. Depending on their half-life, they emit radiation continuously, ranging from seconds to millions of years. Exposure to ionizing radiation can cause serious harm to humans and the environment. Therefore, special attention is paid to the management of radioactive waste in order to deal with its large quantity and dangerous levels. Current treatment technologies are still being developed to improve efficiency in reducing the hazard level and waste volume, to minimize the impact on living organisms. Thus, the aim of this study was to provide an overview of the global radioactive waste treatment technologies that have been released in 2019–2021.

Three samples of 99.97% pure rolled tungsten were annealed at 700 °C, 1000 °C, and 1400 °C in a vacuum of about 10 −6  mbar for 1 h. Then the effects of annealing were investigated. For this purpose, defects and their structure in the three annealed samples and the as-received sample, were studied, characterized, and compared by the positron annihilation lifetime spectroscopy (PALS) technique. A large number of mono-vacancy and cavity defects were determined in the as-received sample. The results show that by increasing the temperature, the density, and size of the defects drastically decreased, and their structure changed to dislocation/vacancy loops. The structure and status of defects in annealed samples are almost constant. τ ave with a 40% decrease in 1400 °C compared to as-received, indicates the beginning of removing defects from tungsten at the temperature of 1400 °C.

Cobalt-60 is an artificial radioisotope with a wide range of applications in many industries for its high intensity gamma-rays and longer half-life. It may be produced in research reactors from the stable isotope cobalt-59 by neutron activation when a fission neutron is absorbed in the Co-59 nucleus. The ETRR-2 research reactor is designed with an in-core neutron flux trap that accommodates space for placement of an irradiation box, which makes it suitable for the production of radioisotopes as an in-core fixed facility. This simulation study compares the production yield of Co-60 for different Co-59 pencil configurations within the irradiation box inside the in-core neutron flux trap. IAEA reference for Co-60 specifications was used to set the comparison criteria for generating different configurations. Reactor geometry and calculations were carried out using OpenMC Monte Carlo code to obtain the effective multiplication factor K eff , the flux distribution, and the production yield of Co-60 with respect to time. Results show that the ETTR-2 is capable of producing Co-60 with a specific activity of 83.89 Ci g −1 in 263 ± 14 days following the placement of eight Co-59 pencils with 0.77 cm diameter, 32 cm in height, and 74.506 g each. It was also shown that a higher production rate may be achieved by decreasing the diameter of the pencils and increasing their number while Co-59 mass is kept constant.

One speed, time-independent and homogeneous medium neutron transport equation is solved for second order scattering using the Anlı-Güngör scattering function which is a recently investigated scattering function. The scattering function depends on Legendre polynomials and the t parameter which is defined on the interval [−1,  1]. A half-space albedo problem is examined with the F N method and the recently developed SVD method. Albedo values are calculated with two methods and tabulated. Thus, the albedo values for the Anlı-Güngör scattering are compared with these methods. The behaviour of the scattering function is similar to İnönü’s scattering function according to calculated results.