Juliana P. Duarte

Human Reliability Modeling of Radiotherapy Procedures by Bayesian Networks and Expert Opinion Elicitation

Abstract The purpose of this paper is to highlight and model the most important steps in cases of human failure in radiotherapy (teletherapy and brachytherapy) procedures by identifying possible modes of human failure. An approach via Bayesian networks (BNs) to model and highlight the most relevant steps of teletherapy and brachytherapy was used. Finally, as a technique for the quantification of BNs, an expert opinion elicitation procedure was used since no database is available. In the case of teletherapy, observing only the stages of prescription, planning, and execution, it appears that the step that most increases the success probability, after consideration of preventive measures, is execution. This is in agreement with cases of errors and accidents reported in the literature, considering that more than 50% of these cases are related to the implementation phase. Related to brachytherapy, the most relevant factor was the use of equipment, whose increase in success probability after consideration of preventive measures was 17.2%, demonstrating the importance of a continuous specific training. It is important to mention that the purpose of this study was not to calculate the risk associated with radiotherapy treatments but rather to check how accident prevention influences the success procedure and observe the relationship among all stages. An uncertainty analysis was performed of the expert data by considering that data scattering followed a normal or a lognormal distribution, due to data ranges considered. This analysis revealed that data scattering was better represented by normal distributions, and the results are consistent with pointwise estimates initially made.

Generation IV Nuclear Systems: State of the Art and Current Trends with Emphasis on Safety and Security Features

Fifty years ago, on June 26, 1954, in the town of Obninsk, near Moscow in the former USSR, the first nuclear power plant was connected to an electricity grid to provide power. This was the worlds first nuclear power plant to generate electricity for a power grid, and produced around 5 MWe [1]. This first nuclear reactor was built twelve years after the occurrence of the first controlled fission reaction on December 2, 1942, at the Manhattan Engineering Dis‐ trict, in Chicago, Illinois, US. In 1955 the USS Nautilus, the first nuclear propelled submar‐ ine, equipped with a pressurized water reactor (PWR), was launched. The race for nuclear technology spanned several countries and soon commercial reactors, called first generation nuclear reactors, were built in the US (Shippingport, a 60 MWe PWR, operated 1957-1982, Dresden, a boiling water reactor, BWR, operated 1960-1978, and Fermi I, a fast breeder reac‐ tor, operated 1957-1972) and the United Kingdom (Magnox, a pressurized, carbon dioxide cooled, graphite-moderated reactor using natural uranium).

Numerical Investigation of Single-Phase Heat Transfer in a 2 × 2 Rod Bundle with Spacer Grids for a High Pressure Heat Transfer Facility

Abstract Numerical simulation has been widely used in nuclear reactor safety analyses to gain insight into key phenomena. This paper compares simulations of a single-phase steady flow in a 2 × 2 rod bundle with spacer grids among different codes based on the high pressure heat transfer facility at University of Wisconsin. The detailed computational fluid dynamics modeling methodology was developed using FLUENT to help in the facility design and pretest analyses. After comparison between different turbulence models, the Standard k-ω was chosen to simulate the effect of unheated solid walls and grid spacers. It was found that solid walls had a small influence on the flow and heat transfer behavior. We note the effect of rod-to-wall gap needs be taken into account if it is larger than half of the gap between the rods. We compared the simulations of FLUENT, COBRA-TF, and TRACE to determine the position of thermocouples to be used in the planned experiments. An investigation was performed on the effect of bending angles of the grid spacer mixing vanes. Results showed that a larger bending angle results in higher turbulence mixing and locally higher Nusselt numbers downstream of the mixing vanes. Also, a small change of the bending angles results in a notable difference in the temperature distributions of the main flow.

Hydraulic and Heated Equivalent Diameters Used in Heat Transfer Correlations

Abstract Hydraulic and heated equivalent diameters are approximations to account for different flow geometries in thermal-hydraulic analyses. Most of the empirical models used in single- and two-phase flow heat transfer are based on experiments in heated tubes and extrapolated to complex geometries by means of the equivalent diameters. For heat transfer calculations, as a general rule, the heated equivalent diameter must be used for bundle geometries and the hydraulic equivalent diameter for annulus geometries. The use of both diameters in different correlations is discussed and clarified in this technical note.

An Initial Assessment of Minimum Film Boiling Temperature Correlations and Comparison to Experimental Data

The minimum film boiling temperature (TMFB) separates the unstable transition boiling region, where the liquid can contact the heated surface, and the post-CHF regime, where vapor can prevent the liquid from contacting the heated wall. This paper presents a review of minimum film boiling temperature correlations and describes a new technique to measure TMFB. The experiment proposed was simulated by TRACE and the same behavior was observed experimentally.Copyright

Coupling of a Lumped Parameter and a Finite Difference Model for Estimation of a Reactivity-Induced Transient in a PWR with Annular Fuel Rods

This paper develops a finite difference and a semianalytical model to evaluate the thermal behavior of fuel rods during a hypothetical reactivity-induced transient (absorber rod ejection) in a pressurized water reactor (PWR). The calculations are carried out for two different reactor core designs, namely, for a typical PWR core with typical fuel rods and for a modified PWR core containing annular fuel rods. The overall dimensions of the core internals and fuel assemblies are unchanged, and the total number of fuel assemblies is the same in both designs. The finite difference code was verified on the results provided by the semianalytical model. In the calculations, two point models were used (neutron kinetics to compute the fission power of the reactor; power balance of coolant in the active core region). The point models were coupled to the one-dimensional, finite difference heat conductivity model to calculate the radial temperature profile in the solid and the annular cylindrical pellets (UO2). The latter are the constituent part of annular fuel rods cooled both on their external and internal surfaces. The fuel, cladding, and fluid temperatures were evaluated for the annular and the solid fuel design in the hot spot, where the maximum allowable power rating of rods is 2.5 times higher than the average one. It was assumed that before the transient, (a) the reactor with annular fuel runs continuously at higher nominal power (150%) and at higher nominal coolant flow rate (150%) at the core inlet than the reference reactor with solid fuel (100%) and (b) the modified and the reference reactors have the same coolant temperature at the core inlet and the same temperature rise of coolant along their core. These conditions correspond to a 150% power uprate of the reference reactor. The coupled models require limited computational resources only. The calculated results showed that during the transient, in the annular fuel pellet, the temperatures peaked at considerably lower values, even at 150% power, than in the solid pellets at 100% power. These evaluations show that in the case of the reactivity-induced transient analyzed, the annular fuel cooled from both sides has a better safety performance than the solid fuel cooled only on its external surface.

Cold Water Injection and Rod Ejection Analysis of Annular Fueled PWRs by a Hybrid Lumped Parameter Model

Annular fuels are being studied to increase the power of advanced third-generation reactors by 50%. This paper aimed to analyze transient scenarios through a hybrid lumped parameter-finite difference model in a pressurized water reactor with annular fuel. The model used in this work is more detailed than the double lumped parameter one, but still simple enough to model some transients in PWR fuels, as rod ejection accident and cold water insertion accident. The heat transfer equations are solved by the numerical semi-implicit Crank-Nicolson method together with point kinetics equations with six groups of delayed neutrons and a lumped parameter model for the reactor coolant.The model takes into account in an approximate way the hot spot by using a composed peaking factor equal to 2.5. The reactivity feedback is taken into account by considering the Doppler effect of fuel temperature, and also moderator temperature variation. The results were compared with solid fuel performance and showed that the annular fuel reached considerable lower fuel temperature profiles even for 150% power, as compared to 100% power for solid fuel, thus showing that this kind of fuel has a better safety performance for the transients analyzed. The rod ejection accident showed that feedback effects can lead the reactor to a new safe steady state condition.Copyright