Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

A Design Proposal and Assessment of In-Vessel Retention Strategy for 2nd Generation PWR

In order to satisfy the higher safety requirements after the Fukushima Accident, in-vessel retention (IVR) strategy is suggested to be a measure for the improvement for the operating 2nd generation Pressurized Water Reactor (PWR). In this paper, a design of passive cavity flooding system isproposed for the 2nd generation modified PWR with reference to the design of in-containment refueling water storage tank (IRWST)for AP1000, and the IVR assessmentis evaluated with the risk oriented accident analysis methodology (ROAAM). Seven representational accident sequences are selected for the IVR assessment and analyzed by lumped-parameter safety analysis computer code. Accident analysis focuses on four key parameters forthe assessment, that is, decay power, zirconium oxidation fraction, and both the mass of oxidic pool and metal layer. The values of them are obtained through analysis and used for the assessment. The probability density distributions of those parameters are determined by combining the analysis results and engineering judgment. The success probability of IVR from the viewpoint of thermal failure is predicted using a program written by the Matlab code. Furthermore, some sensitivity analysis and parametric studies are investigated to support the assessment. The assessment result shows that the success probability of the cavity injection system is higher than 99%. Detail analysis about system reliability and feasibility is needed for further work.

Towards Optimal In-Core Fuel Management of Thorium-Plutonium-Fuelled PWR Cores

Plutonium is a significant proliferation concern as well as a major contributor to the long-term toxicity of nuclear waste. Partial incineration in PWRs with uranium-MOX fuel is often considered to mitigate these concerns. Thorium-MOX is an alternative fuel with superior material properties and higher plutonium destruction rates, as shown in multiple feasibility studies. However, the core performance and operational characteristics (e.g. discharge burn-up, feasibility of controlling the core) are ultimately dependent on the core loading pattern (LP) and burnable poison (BP) design. In this paper, the LP for Th-Pu fuel of various compositions is optimized for (1) discharge burn-up, (2) radial form factor (RFF), (3) cycle length, (4) moderator temperature coefficient (MTC), and (5) reactivity swing over cycle. Maximizing the cycle length makes the discharge burn-up and reactivity swing worse due to placement of once- and twice-burnt fuel near the core periphery. It also makes the MTC less negative. The harder neutron spectrum of Th-Pu fuel compared to conventional U fuel favours the use of distributed integral burnable poisons to control the reactivity swing over the cycle. This leads to a significant amount of dissimilarity between LPs with relatively similar performance measures, and between optimal LPs for different Pu loadings in the fuel. The RFF can vary throughout the cycle but a careful placement of the assemblies can mitigate this. The cycle reactivity swing is controlled using enriched soluble boron, which makes the MTC worse, and this constrains feasibility for high Pu loading in the fuel.

Stress, Strain, and Energy at Fracture of Degraded Surfaces: Study of Replicates of Rough Surfaces

Due to the aging of structures, the issues of plant life management and license extension are receiving increasing emphasis in many countries. Understanding failure of structures due to random roughness on surfaces at early stages of degradation is therefore crucial. It has been shown that even slightly sinusoidal roughness can increase stress concentration by a factor of 2 or 3, which can be critical for a brittle component due to the significant reduction of its load-carrying capacity, even with slight roughness. A more in-depth fracture analysis of surfaces possessing random roughness is needed in order to more profoundly understand, and hence develop models that will predict more accurately, failure of structural materials exposed to degrading, in-service conditions. Using a technique previously developed and successfully applied, replicates of random rough surfaces, imprinted with various levels of degradation, and at three distinct auto correlation lengths, were realized and mechanical testing was performed on them. The stress, strain and energy at fracture are reported. Finite element analysis was carried out to elucidate experimental results. Besides the expected reduction of energy at fracture with degradation, a relaxation region was observed where the energy slightly increases. This phenomenon implies that even after degradation has progressed there is a local maximum of energy at fracture due to the competing effect of tendons and growth of pits. The results find applications on the early stage of maintenance of surfaces of structures in service.

Development of Numerical Simulation for Jet Breakup Behavior in Complicated Structure of BWR Lower Plenum: 1 — Preliminary Analysis of Jet Breakup Behavior in Complicated Structure by TPFIT

In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred as can be seen at Fukushima Daiichi nuclear power plant accident. Especially, it is important to estimate the behavior of molten core jet in the lower part of the containment vessel at severe accident. In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes, instrument guide tubes and core support plate. However, it is difficult to evaluate these effects on molten core jet experimentally. Therefore, we considered that multi-phase computational fluid dynamics approach is the best way to estimate the effects on molten core jet by complicated structure. The objective of this study is to develop the evaluation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum. So we are developing a simulation method to estimate the behavior of molten core jet falling down through the core support plate to the lower plenum of the BWR. The method has been developed based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Thus, in this study, we are carrying out experimental works by use of multi-phase flow visualization technique. In the experiments, time series of interface shapes are observed by high speed camera and velocity profiles in/out of the jet will be measured by the PIV method. In this paper, the outline of the developing method based on the TPFIT was explained. And, the developing method was applied to preliminary experiment with/without modeled complicated structures. As the results, predicted interface shapes were almost agreed with measured data. However, predicted falling down velocity of the jet was lower than measured data. We considered causes of this underestimation and improved the method and simulation conditions to resolve this problem.

Experimental and Numerical Investigation of H2-Air Deflagration in the Presence of Concentration Gradients

A set of experiments performed at Karlsruhe Institute of Technology (KIT) in the framework of the LACOMECO European project is devoted to flame propagation in an obstructed large scale facility A3 (of 8 m height and 33 m3 volume) with initially vertical hydrogen concentration gradients. Almost linear positive and negative (relative to gravity) concentration gradients are created prior to ignition in the range from 4% to 13%, and the process of flame acceleration is investigated depending on hydrogen concentration gradient and ignition positions. In this paper we describe the A3 facility and analyse the experimental data obtained during the project. The results of numerical simulation performed using Europlexus code are presented together with the critical discussions and conclusions.

Analysis of Early Stage Accident at TEPCO’s Fukushima Daiichi NPS

The operation of the Isolation Condenser (IC) played an important role in the progression of the accident at the unit 1 of Fukushima Daiichi nuclear power station. Analyses of Unit 1 accident in the early stage (prior to the occurrence of core melt) were performed using plant dynamics analysis code RELAP5/MOD31 and the results were compared with measured data. In the RELAP5 code, analysis scope of target is not a severe accident. But, the detailed simulation in the early stage of accident including the plant control system behavior is possible. Moreover, some sensitivity analyses were conducted and the reactor behavior under IC operations was examined.

Evaluation of Damaging Effects on a NPP Integrity Subjected to Beyond Design Earthquake

The aim of this study is to evaluate the influence of the damaging or failure effects induced by seismic loadings on the loading bearing capability of the overall nuclear structures, systems and components (SSCs). In particular it was analyzed the behaviour of the SSCs in terms of their required safety functions and capacity assuming the occurrence of an earthquake having magnitude beyond the design basis value (in agreement with the “stress tests” suggestions, presently foreseen by the European and International Associations). To the purpose, quite refined FEM models were set up and implemented considering suitable materials behaviour and constitutive laws for the reactor materials, particularly for the concrete material behaviour (which could suffer failure and damage mechanisms during the ground shaking). Several type of damaging mechanics were considered. The obtained results seemed to confirm the overall containment reliability even if with minor upgrading actions in operating procedures and design characteristics. Moreover they allowed to appropriately check the NPP containment strength reserve.

Experiment and Numerical Analysis of Control Method of Natural Circulation by Injection of Helium Gas

This study is to investigate a control method of the natural circulation of the air by the injection of helium gas. A depressurization is the one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, the air is predicted to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of the air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomenon during the depressurization accident of the VHTR, it is important to develop the method for prevention of air ingress during the accident. In this study, the air ingress process is discussed by comparing the experimental and analytical results of the reverse U-shaped channel which has parallel channels. The experiment of the natural circulation using a circular tube consisted of the reverse U-shaped type has been carried out. The vertical channel is consisted of the one side heated and the other side cooled pipe. The experimental apparatus is filled with the air and one side vertical tube is heated. A very small amount of helium gas is injected from the top of the channel. The velocity and the mole fraction of each gas are also calculated by using heat and mass transfer numerical analysis of multi-component gas. The result shows that the numerical analysis is considered to be well simulated the experiment. The natural circulation of the air has very weak velocity after the injection of helium gas. About 780 seconds later, the natural circulation suddenly produces. The natural circulation flow of the air can be controlled by the method of helium gas injection. The mechanism of the phenomenon is found that mole fraction is changed by the molecular diffusion and the very weak circulation.

Model Passivated Carbon Electrodes for Fluorine Generation in MSRs and the Nuclear Fuel Cycle

Elemental fluorine, F2, is used in the nuclear fuel cycle for the isotopic separation of uranium-235 and 238, as well as for the purification of LiF-BeF2 in molten salt reactors. F2 is generated on an industrial scale by an electrochemical process using carbon electrodes in a KF-2HF molten salt. Carbon electrodes are used for industrial F2 generation due to its chemical stability, high conductivity, and relatively low cost. One of the main issues faced when using carbon electrodes in this chemical system is passivation through the formation of C-F compounds on the surface of the electrode. This results in a loss of anode wettability to the electrolyte and diminished charge transfer rate. The voltage needed for the fluorine evolution reaction increases which negatively impacts the safety of the system, increases the operating costs, and leads to faster degradation of the electrode. The degradation of electrical properties during passivation is progressive, eventually leading to electrode deactivation. The process of deactivation begins with a passivating C-F layer at potentials above the equilibrium potential (2.92 V). The layer is both non-wetting to the KF-2HF media and insulating. Deactivation begins with inhibited F2 bubble detachment, formation of a persistent gas layer, and finally deactivation as the electrode surface is completely covered by a thick, insulating C-F layer causing charge transfer to cease. Only a small current is able to flow, even at high potentials (up to 9 V), indicating F2 generation is completely inhibited. The purpose of this study is to manufacture and test model carbon electrodes and, to examine the non-wetting properties of a partially fluorinated surface. The electrodes will be prepared by mixing PTFE-particles with Vulcan carbon powder and then pressing to form pellets. These electrodes should have a reproducible surface for electrochemical performance studies that will lead to a better understanding of the surface chemistry. The research will develop novel electrodes with a goal to minimize the voltage required for F2 production. This will enhance the efficiency in the overall process and lower the manufacturing costs for F2. Carbon electrodes with different PTFE-content (20 w.% and 35 w.%) were synthesized. Electrochemical fluorination was then carried out at different potentials in the F2 generation region (4 to 8 V) in molten KF·2HF electrolyte at ∼90 °C. The electrochemical behaviour of the carbon-PTFE electrodes was examined and compared for both fluorine passivated and non-passivated graphite, amorphous carbon, and vitreous carbon electrodes. The growth of the electrical double-layer capacitance between the carbon electrodes and the KF·2HF molten salt was studied. The effects of composition of fluorinated and non-fluorinated carbon on electrode performance are presented.