Volume 87, Issue 4

A heat pipe transfers heat effectively between two solid surfaces by incorporating the principles of the transition of phase and thermal conductivity. The study aims to investigate the thermal characteristics of a cylindrical heat pipe and the various factors affecting its performance. The effect of different working fluids, i.e., water, ethanol, and methanol, wick material, i.e., copper and stainless steel, and angle of inclination varied from varied between 0° and 90°. The fill volume is also varied from 20 to 40% to analyze the thermal resistance and effective thermal conductivity of the heat pipe. The optimum value of angle of inclination is found to be 60° at 30% fill volume of working fluid irrespective of the wick material.

The main objective of the presented work is to investigate the effectiveness of homogenization of the GFR-2400 core in simulating its neutronic characteristics. To explore and evaluate the neutronic behavior of the large scale Gas cooled Fast Reactor GFR-2400 core, two computer models (homogeneous and heterogeneous) were designed using the MCNPX code. The designed heterogeneous model has been validated by comparing its results with a previously published paper. The results of both models were compared with each other to study the effect of fuel homogeneity on the radial flux and power distribution through the core. As part of a safety analysis of the reactor core for both designs, the reactivity worth of control rods, Effective delayed neutron fraction ( ß eff ), Doppler constant and Depressurization effect have been analyzed and compared. The variations of the effective multiplication factor ( k eff ), the minor actinides concentration and the most important fission products as a function of operation time have been investigated for the presented designs.

Success criteria analysis plays a key role in the development of realistic probabilistic safety/risk assessment (PSA/PRA) model because it provides supporting information regarding the response of complex nuclear power plant systems to different accident conditions. The current paper performs plant specific success criteria analysis for steam generator tube rupture (SGTR) accident in a typical pressurized water reactor (PWR) and demonstrates implementation of the obtained best estimate results on a risk model which has been initially developed based on expert judgment and general plant design data. The modifications on the risk model include configuration of the safety systems as well as the event tree structure. The updated PSA model shows 50% reduction in the plant core damage frequency (CDF) in comparison to the base case risk model. This highlights the importance of success criteria analysis for the development of a realistic PSA model in risk informed applications.

With the release of reports such as NUREG/CR-6850, the calculation of ignition frequency concerning general large-scale Pressurized Water Reactors (PWRs) has matured. However, there is currently a lack of methods for calculating the ignition frequency in the case of Small Modular Light Water Reactors (SMRs). By studying the calculation of ignition frequency reported in NUREG/CR-6850, we propose a method for calculating the ignition frequency in SMRs, which is based on the generic ignition frequency at the component level. The problem of counting ignition sources is discussed in detail, and we determine the component-level ignition frequencies due to different types of ignition sources. In order to improve the calculation of the ignition frequency in SMRs, this paper provides two methods, which are based on the prior lognormal distribution and the prior Gamma distribution, respectively, for updating ignition frequency. We also consider the effect of the accumulation of fire events on the posterior mean determined by both methods. In this paper, we suggest that the method based on the prior lognormal distribution should be used in the initial stage of updates of ignition frequency. As data for plant-specific events accumulate, the approach based on the prior Gamma distribution should be considered.

The accurate calculation of pressure drop of evaporators/condensers are crucial and related with the pumping power, performance coefficient and energy consumption in a refrigeration equipment. This work aligns frictional pressure drop models/correlations with the experimental outcomes of boiling pressure drop of R134a in horizontal smooth and microfin copper tubes with equivalent outer diameter of 9.52 mm. The pressure drop through the test tube is obtained with a differential pressure transducer directly. Effective parameters are specified for smooth and microfin tubes and the most compatible models/correlations, 12 for smooth tubes and 9 for microfin ones, are determined accurately in relation to the consequences of investigation during intermittent and annular flow regime. Moreover, new two-phase multipliers have been developed by using regression analyses of 182 data points based on Lockhart-Martinelli parameter for each test tubes separately, and their predictability are found to be better than others in the literature as novel ones. Average errors of the developed empirical correlations are 11% for smooth and for 7% for microfin tubes. Finally, the measured data is given for the validation issues of researchers who can benefit from most of the investigated pressure drop models with tolerable accuracy regarding with their HEX design analyses.

This paper presents a newly developed steady state core thermal hydraulic model (named SSTH-RR10 model) for upgrading the Egyptian first Research Reactor (ETRR-1), from its original power of 2 MW th to a higher level of 10 MW th , by considering different types of nuclear fuels. The SSTH-RR10 model is capable to predict and calculate, by means of a developed computer program, all the steady state thermal hydraulic parameters for the defined core configuration for each fuel type at 10 MW th . Three different fuel types were investigated: the reference fuel EK-10 rod type, the MTR plate type, and the IRT-4M ducted type. For each fuel type, the distribution of central fuel, clad, and coolant temperatures for average and hot channels of the core were predicted in the axial direction. Power distributions and pressure gradients were predicted as well. Moreover, the program calculates the safety limits and margins against the critical phenomena encountered such as the Onset of Nucleate Boiling (ONB), Departure from Nucleate Boiling (DNB), and the Onset of Flow Instability (OFI). Results of the SSTH-RR10 program for benchmarks of powers of 2 and 10 MW th are verified by comparing it with the published results of the International Atomic Energy Agency (IAEA), and those published for other programs such as PARET code, and very good agreement is found. The safety margins against ONB and DNB were evaluated in which the minimum DNB ratio was found to be about 3.1, which gives a sufficient margin against the DNB. The present work gives confidence in the model results and applications.

The direct contact condensation (DCC) of steam in subcooled water encounters in wide range of the industrial applications. On one side, it is an efficient and rapid, mass and heat transfer phenomenon. But, on the other side, it may generate condensation induced water hammers (CIWH) events which may cause high pressure peaks resulting in severe damage to the mechanical systems. This computational study intends to explore the underlying physics of CIWH events while injecting subcooled water into steam filled horizontal pipe section. The Reynolds number is varied from, Re w  = 60,750 to 646,900, to study the flow regimes (stratified and slug), onset of CIWH and local flooding conditions. The results have been compared with the published data and found in good agreement. It has been observed that for Re w  = 182,300, flow remains stratified. However, the flow regime changes from stratified to slug flow at Re w  = 303,850–646,900, possibly due to the onset of CIWH. Extensive steam pockets have been observed at Re w  = 303,850, which may be considered as onset of CIWH. Local flooding condition is also started at Re w  = 303,850 and is observed to be shifted upstream with the increase in Reynolds number. This study is considered to be useful for the safe design and economical operation of the relevant systems in nuclear and other related industry.

In this study, graphene nanofluids were used to explore the effect of various concentrations on boiling heat transfer of downward-facing heating. Five concentrations of graphene nanofluids were prepared for pool boiling heat transfer experiments. The experimental results show that when the mass concentration is 10 mg/L, the maximum enhancement of the CHF is up to 76.1%. In order to explore the mechanism of graphene nanofluid enhancing boiling heat transfer, after the experiment, the wettability and roughness of the heating surface were measured and the heating surface was characterized by a scanning electron microscope (SEM) and electronic differential system (EDS). The results show that the wettability is enhanced and the surface roughness is reduced. In addition, boiling curves (the curves of heat flux with surface superheat) and the curves of heat transfer coefficient with heat flux at different concentrations have also been observed to further explore the mechanism of enhanced heat transfer.