Experimental Study on Jet Breakup Behavior With Surface Solidification

For the safety design of the Fast Breeder Reactor (FBR), the Post Accident Heat Removal (PAHR) is required when a hypothetical Core Disruptive Accident (CDA) occurs. In the PAHR, it is strongly required that the molten core material can be cooled down and solidified by the sodium coolant in the reactor vessel. There is high possibility for molten material to be ejected as a liquid jet into sodium coolant in the reactor vessel. In order to estimate whether the molten material jet is completely solidified by sodium coolant or not, it is necessary to understand the interaction between molten core material and coolant such as jet breakup and fragmentation behavior in coolant. The jet breakup behavior is the phenomenon that the front of molten material breaks up in coolant. To clarify the mechanism of jet breakup and fragmentation during the CDA for the FBR, it is necessary to understand the correlation between jet breakup lengths and size distribution of fragments when molten material jet interacting with coolant. The objective of the present study is to clarify the dominant factor of the jet breakup length and the size distribution of fragments experimentally. Molten jet of U-alloy 138 is injected into water as simulated core material and coolant by free-fall. The density ratio of core material and coolant is almost same as that of the real FBR system. The jet breakup behavior as interaction of molten material with coolant is observed with high speed video camera. Front velocity of the molten material jet is estimated by using the image processing technique. It suddenly decreases when the jet fall into the coolant. The jet breakup length estimated from observed images is compared with the breakup theories to understand the effect of experimental parameters for the jet breakup length. The solidified fragments are gathered and classified in size, and the mass in each size is measured. Median diameter is obtained from the mass distribution of the fragments. In comparison with interfacial instabilities, the median diameter of fragments shows the independent of relative velocity. The jet breakup lengths and median diameters compared with existing theories is discussed.

Thermal Aspects of Using Uranium Mononitride Fuel in a SuperCritical Water-Cooled Reactor at Maximum Heat Flux Conditions

Generation IV nuclear reactor technology is increasing in popularity worldwide. One of the six Generation-IV-reactor types are SuperCritical Water-cooled Reactors (SCWRs). The main objective of SCWRs is to increase substantially thermal efficiency of Nuclear Power Plants (NPPs) and thus, to reduce electricity costs. This reactor type is developed from concepts of both Light Water Reactors (LWRs) and supercritical fossil-fired steam generators. The SCWR is similar to a LWR, but operates at a higher pressure and temperature. The coolant used in a SCWR is light water, which has supercritical pressures and temperatures during operation. Typical light water operating parameters for SCWRs are a pressure of 25 MPa, an inlet temperature of 280–350°C, and an outlet temperature up to 625°C. Currently, NPPs have thermal efficiency about of 30–35%, whereas SCW NPPs will operate with thermal efficiencies of 45–50%. Furthermore, since SCWRs have significantly higher water parameters than current water-cooled reactors, they are able to support co-generation of hydrogen. Studies conducted on fuel-channel options for SCWRs have shown that using uranium dioxide (UO2) as a fuel at supercritical-water conditions might be questionable. The industry accepted limit for the fuel centerline temperature is 1850°C and using UO2 would exceed this limit at certain conditions. Because of this problem, there have been other fuel options considered with a higher thermal conductivity. A generic 43-element bundle for an SCWR, using uranium mononitride (UN) as the fuel, is discussed in this paper. The material for the sheath is Inconel-600, because it has a high resistance to corrosion and can adhere to the maximum sheath-temperature design limit of 850°C. For the purpose of this paper, the bundle will be analyzed at its maximum heat flux. This will verify if the fuel centerline temperature does not exceed 1850°C and that the sheath temperature remains below the limit of 850°C.

3D Simulation of Molten Metal Freezing Behaviour Using Finite Volume Particle Method

Understanding the freezing behavior of molten metal in flow channels is of importance for severe accident analysis of liquid metal reactors. In order to simulate its fundamental behavior, a 3D fluid dynamics code was developed using Finite Volume Particle (FVP) method, which is one of the moving particle methods. This method, which is fully Lagrangian particle method, assumes that each moving particle occupies certain volume. The governing equations that determine the phase change process are solved by discretizing its gradient and Laplacian terms with the moving particles. The motions of each particle and heat transfer between particles are calculated through interaction with its neighboring particles. A series of experiments for fundamental freezing behavior of molten metal during penetration on to a metal structure was also performed to provide data for the validation of the developed code. The comparison between simulation and experimental results indicates that the present 3D code using the FVP method can successfully reproduce the observed freezing process such as molten metal temperature profile, frozen molten metal shape and its penetration length on the metal structure.

Depletion Calculation for a Nodal Reactor Physics Code

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

Effect of Pin and Subassembly Heterogenity in Sodium Void Worth Calculations

Metal fuelled sodium cooled fast reactors are known to have high breeding ratio and short doubling time. Due to these features they play a very important role in the energy scenario, where higher power growth is required. Large sodium cooled fast reactors have positive sodium void coefficient, which is considered to be undesirable feature even though reactor safety can be established for all design based accidents like loss of flow and transient over power accidents. These types of fast reactors, which have harder neutron spectra are having higher sodium void coefficient compared to ceramic fuelled fast reactors. In many of the safety analysis the total sodium void is calculated and it is used in the safely evaluation. However the sodium in the metal fuelled reactor has got three parts, namely bonding sodium, coolant sodium and the sodium in the inter space of subassembly hexagonal cans. In the reactor accident scenario the behavior of these three components of sodium will be different and will effect the sequence of the accident. The finer details, of the fuel sub assembly, are modeled in to Monte Carlo code and the sodium void coefficient is calculated for each of the component for the fuel zones. This study will be helpful in improving safety of the reactor and also reducing the conservatism in the safely features.

A Comparison of Different Communication Tools for Distance Learning in Nuclear Education

Nuclear Engineering Education has seen a recent surge in activity in the past 10 years in Canada due in part to a Nuclear Renaissance. The Nuclear Industry workforce is also aging significantly and requires a significant turnover of staff due to the expected retirements in the next few years. The end result is that more students need to be prepared for work in all aspects of the Nuclear Industry. The traditional training model used for nuclear engineering education has been an option in an existing undergraduate program such as Chemical Engineering, Engineering Physics, or Mechanical Engineering with advanced training in graduate school. The education model was mostly lecture style with a small number of experimental laboratories due to the small number of research reactors that could be used for experimentation. While the traditional education model has worked well in the past, there are significantly more advanced technologies available today that can be used to enhance learning in the classroom. Most of the advancement in nuclear education learning has been through the use of computers and simulation related tasks. These have included use of industry codes, or simpler tools for analysis of the complex models used in the Nuclear Industry. While effective, these tools address the analytical portion of the program and do not address many of the other skills needed for nuclear engineers. In this work, a set of tools are examined that can be used to augment or replace the traditional lecture method. These tools are Mediasite, Adobe Connect, Elluminate, and Camtasia. All four tools have recording capabilities that allow the students to experience the exchange of information in different ways. The students now have more options in how they obtain and share information. Students can receive information in class, review it later at home or while in transit, or view/participate it live at a remote location. These different options allow for more flexibility in delivery of material. The purpose of this paper is to compare recent experiences with each of these tools in providing Nuclear Engineering Education and to determine the various constraints and impacts on delivery.

EPR™: A Reactor to Support Economical Fuel Cycles

The EPR™ reactor has been designed by AREVA to support economical fuel cycles. The progress in the reactor and systems design improves the reactor safety, and allows the EPRTM reactor to support the large range of high performance fuel management strategies covering cycle length from 12 to 24 months. Different fuel management strategies with 12, 18 and 24 month cycles are described. Economic analyses are performed to illustrate the low uranium consumption and the high fuel cycle performance compared with the fuel managements implemented in most current traditional PWR reactors.

The Progress of ICRP New Recommendations

In order to probe into the usage of the Recommendations of the ICRP, through comparative analysis of low-dose-rate radiation-induced stochastic effects of a nominal risk coefficient, radiation weighting factor, tissue weighting factor as well as the the implementation of changes on the radiological protection system, analysis of the international on Radiological Protection fundamental recommendations of the Committee on the latest changes in radiological protection and development, and that these changes can not affect the existing radiation protection of China’s basic policy and standards.

Modeling the Spectral History in the Depletion of a PWR Core

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.

Particle Behaviors Under Acceleration Condition in Solid-Air Two Phase Flow

This paper presents a numerical study of the particle behaviors under acceleration conditions in the solid-air two-phase flow by means of a combined two-dimensional model of computational fluid dynamics and discrete element method (CFD-DEM). The simulation model provides the information regarding the particle distribution behaviors within the calculation region and the particle run-out rate from the calculation region under different parameter conditions, such as particle size, initial particle loading and particle acceleration condition. The results demonstrate that the particle run-out rate is directly influenced by the particle size and the initial loading condition. The particle acceleration in the horizontal direction adversely affects the particle run-out rate when the initial particle loading condition is dispersed and uniform. However, this adverse effect disappears when the initial particle loading condition becomes concentrate and partial.