Band 87, Heft 5

Pool boiling takes place when a stagnant liquid is placed on a surface above the saturation temperature of the liquid. Thermal studies have attracted a large number of researchers working in different segments of the industry to meet the heating and cooling demands of each individual. The pool boiling is a highly effective technique of convective heat transfer to maintain the system operation under safe temperature and safe temperature limits for better performance and longevity. This study investigates the effect of engraved concentric circles on aluminium disc in de-ionized water. The experiments are conducted by varying pitch ratio ( P ) and indentation ratio ( I ) of engraved concentric circles in the range of 0.0125–0.05, and 0.002–0.008, respectively. The maximum enhancement in the boiling heat transfer coefficient is found to be 21.11% more than that of a plain surface corresponding to the pitch ratio of 0.0125 and indentation ratio of 0.008.

In the current work, the pressure drop of the refrigerant flow in smooth and micro-fin pipes has been modeled with artificial neural networks as one of the powerful machine learning algorithms. Experimental analyses have been evaluated in two groups for the numerical model such as operation parameters/physical properties and dimensionless numbers used in two-phase flows. Feed forward back propagation multi-layer perceptron networks have been developed evaluating the practically obtained dataset having 673 data points covering the flow of R22, R134a, R410a, R502, R507a, R32 and R125 in four different pipes. The outputs acquired from the artificial neural network have been evaluated with the target ones, and the performance factors have been estimated and the prediction accuracy of the network models has been resourced comprehensively. The results revealed that the neural networks could predict the pressure drop of the refrigerant flow in smooth and micro-fin pipes between 10% deviation bands.

In a serious accident, after the core of a nuclear reactor melts and collapses, In-Vessel Retention in the External Reactor Vessel Cooling (IVR-ERVC) is an effective technology to maintain the integrity of lower head by reducing heat load on it. The various factors affecting the melting of the lower head have been widely studied. The mass of the molten metal layer may affect the consequences of the accident, since it is where the focusing effect occurs. However, the related research is still absent. In this paper, we systematically calculated the heat transfer behavior and melting process under different metal layer heights conditions. The temperature distribution, the velocity distribution, the heat flux of the outer wall of the Reactor Pressure Vessel (RPV), and the change of the thickness of the RPV were obtained through Large Eddy Simulation (LES). Interestingly, the heat flux increases with the metal layer height at first and achieve the maximum in the middle height. These results increase the understanding towards the serious accidents.

The main objective for reactor safety is to keep the fuel in a safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accidents (DBAs), one of them is the loss of flow accident (LOFA), is required for assessing reactor safety. In this research, the safety aspects of 22 MW MTR research reactor under steady state and during loss of flow accident is studied. The flow transients considered include fast loss of flow accident (FLOFA) and slow loss of flow accident (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The analysis is done using PARET, a neutronics-hydrodynamics-heat transfer code. The transients were initiated from a full power with a flow trip point at 85% nominal. The calculated parameters are the temperatures of different components (fuel, clad and coolant) as a function of time for the hot channel. The results indicate that in both accidents the calculated maximum cladding surface temperature for the hottest channel of the reactor core does not exceed the allowable safety limit and the fuel integrity is maintained.

The prediction of heat flux leading to the Onset of Flow Instability (OFI) phenomena is an important consideration in the design of Material Testing Reactors (MTR) due to the possibility of flow excursion during postulated accident. From the thermal-hydraulic point of view, OFI is the critical phenomenon limiting MTR reactor power. In a previous work, an empirical correlation is developed to predict the subcooling at OFI in narrow vertical rectangular channels simulating a coolant channel of MTR. In the present work, an innovative model to determine the heat flux leading to OFI in MTR reactors is introduced based on the previous correlation. The developed model gives a very low deviation of only 1.65% from the experimental data of Whittle & Forgan that covers a wide range of MTR operating conditions. The heat flux leading to OFI is also predicted by both Whittle & Forgan and Fabrega correlations for comparison. The present model is then applied on the IAEA 10 MW MTR generic reactor to predict the Best-Estimate (BE) and Best-Estimate-Plus-Uncertainty (BEPU) Onset of Flow Instability Ratio (OFIR) and the power leading to OFI as well as the bubble detachment parameter under different coolant velocities and inlet temperatures. The model is also used to predict both the OFIR and bubble detachment parameter in the reactor under unprotected Loss-of-Flow transient for exponential flow decay with a time constant of 1.0 s (fast LOFA), 10, 15 and 25 s (slow LOFA) from a power level of 10 MW. For BEPU calculation, a combined statistical method with direct propagation of errors is adapted to treat the uncertainty factors for fuel fabrication and measured parameters in the BEPU calculation. The model results is analyzed and discussed.

The phenomenon of saturated steam jet injection in subcooled quiescent water has many practical applications including in heat exchangers, steam jet pumps, steam dumping systems in nuclear plants, etc. The experimental setup is designed and fabricated indigenously to investigate this phenomenon at lower mass fluxes ∼120 and 150 kg/m 2  s. The steam jet of conical shape has been observed for all the test conditions. The recorded axial temperature distribution showed that near the nozzle region, the temperature is governed by the saturated condition of steam while the later region is dependent on the water pool temperature. The maximum temperature is observed to be at the center of the jet. It has been found that the dimensionless penetration length of the steam jet in water is directly dependent on both the temperature of the water pool and the mass flux of steam. The dimensionless jet length has been found in the range ∼1.54–2.02 and 2.07–2.19 for mass fluxes ∼120 and 150 kg/m 2  s, respectively. The average heat transfer coefficient has been found in the range ∼1.97–2.37 MW/m 2  K.

This research focuses on verifying the importance of the ADS and the CMT, by using the ASYST code. We evaluated the role of these two components by postulating the failure of the ADS as a single failure approach and the failure of the CMT with ADS failure as multiple failures approach during hypothetical SBLOCA conditions. These accidents acted as confounding factors distorting the AP1000 PSS. We investigated the reactor and safety system behavior during the SBLOCA. We evaluated the importance and effectiveness of two components in reducing and mitigating the consequences of the accident. We checked the effectiveness of these components by comparing the importunity-related issues with and without these components during the accidents. We found that the ADS decreased the pressure, allowing natural circulation to quench the reactor core during the LOCA. During the failure of ADS, the vapor bubbles formed in the reactor vessel covering the fuel rods increased their temperature. The CMT borated water feeding quenched the actinides decay heat. The non-existence of the CMT resulted in decreasing the RCS. ASYST was compared to NOTRUMP to validate it capability to analyze thermal phenomena during accidents. It was found that in the AP1000, the ADS and CMT were considered as the overall importunity of the others PSS.

Safe and efficient storage of spent fuel elements is an important aspect of the safety and economy of nuclear reactors. The present work investigates the thermal-hydraulic behaviour of the cooling process for the nuclear spent fuel stored in the material testing reactor auxiliary pool. The parameters affected by the spent fuel cooling accuracy, the decay power of spent fuel and the initial temperature of the coolant pool are studied. These parameters are simulated by developing a model using thermal-hydraulic computational fluid dynamics, ANSYS FLUENT 17.2 Code. The developed model is evaluated by the previous measurements; an experimental test rig is designed and constructed to investigate the thermal-hydraulic behaviour of the natural circulation cooling of the nuclear spent fuel. The present study uses the validated model to simulate numerically the forced convection heat transfer for spent fuel pools. Various coolant velocities and decay powers are examined. Also, the thermal-hydraulic behaviour of the nuclear spent fuel is studied in transient mode; the initial temperature is raised to 338K. The results show the spent fuel cooling improves as the coolant velocity increases. A good agreement was identified after comparing experimental results with the investigated model.

Nowadays, nuclear reactors became extremely fascinating not only for most of the nuclear communities but also for the prominent energy suppliers to fix the global warming effects worldwide. This paper presents a review of the last advances, applications, and challenges of nuclear reactors. Different types and classifications are introduced. Advantages and disadvantages are discussed for best decision-making. Next, nuclear safety is also discussed as the most important challenging subject to develop nuclear reactors worldwide. They are specially mentioned to find the key solution for the future of nuclear energy. A brief review of nuclear roadmaps is compared with other clean green technologies as well. Estimated prospects for projects timelines and progressions of new nuclear reactors are also presented and discussed briefly. Studies confirmed that nuclear reactors are not only required for developing non-electrical applications or even high-tech systems but also they are extremely profitable to restrict global warming effects. Finally, the solution is to enhance the markets of the nuclear reactors, especially the matured Gen III+ Pressurized Water Reactors (PWRs) to resolve short-term problems as well as advanced futuristic developing Small Modular Reactors (SMRs) for the mid-term and long-term strategies. Moreover, research reactors especially advanced Multi-Purpose Research Reactors (MPRR) are necessary tools to develop both nuclear power plants and other advanced technologies as well as the modern Micro Modular Reactors.

The water level control system implicated in the nuclear steam generator has played an essential role in unexpected shutdowns of the power plant. According to reports, about 25% of the emergency power blackouts are caused by improper level control systems. The effectiveness of optimization methods in designing a controller is currently proved in different disciplines. The novelty of this paper is the proportional integral derivative (PID) controller tuning of nuclear steam generator by particle swarm optimization (PSO) and genetic algorithm (GA) for the lowest steady-state error, overshoot, undershoot, and settling time. Different types of the cost function are employed to obtain the controller gains. The integral of the absolute error (IAE), square error (ISE), time-weighted average error (ITAE), time-weighted square error (ITSE), and a weighted function based on overshoot, undershoot, and settling time are used. The gain scheduling of optimized PIDs is used to have an entire operating range control system. The desired load-following and stability of the optimized PID controller are investigated under both time and frequency domains using trajectory tracking, disturbance rejection, and Nichols chart criterion.

Nuclear terrorism is studied by the complex algorithm with Analytic Hierarchy Process (AHP) and neural networking method. For the modeling, the modified crisis response is incorporated with the AHP level structure of five detections. It is applied to select the nine ranks with an input layer, hidden layer, and output layer in a neural networking algorithm containing the operator’s thinking. For 60 years, the range of the secure operation is between 0.01972 in the 18th year and 0.02099 in the 41st year, which means the highest range of security is about 1.064 times higher than that of the lowest value in this study. For the graphics, the highest and lowest values are seen as detection #5 and #1 respectively. The five sites differ in a dynamic manner where the #5 site is more than that of #1. This anti-terrorism study could be applied to energy or chemical plants in the future.

Democratic Republic of Congo (DRC) has a TRIGA mark II research reactor called TRICO II, its design power is 1.00 MW. The reactor was in extended shutdown state since November 2004. The DRC government has decided to resume its operation using the last uploaded core. One of the safety features to be determined before putting the spent fuel into the reactor core is the calculation of its excess reactivity, radionuclide inventories as well as its discharge burn-up. The spent fuel was modeled and simulated by using Monte Carlo software, MCNPX code. The input data and the horizontal and vertical modeling for the fuel pins, control rods and moderator were done. The model results were validated by calculating the effective delayed neutron fraction ( β eff ) and the worth of the control rods. The results of the criticality and fuel burn-up were compared with the reference design parameters and with the experimental measurements and it were found in good agreement. The calculations showed that the last uploaded core has 47.00 g of 235 U which represent only 2% of fissile materials. The depletion analysis results showed that the highest radio-activities come from 151 Sm, 137 Cs, 90 Y, 90 Sr and 85 Kr.