CFD Analysis of Vortex Shedding in a Circular Cylinder: Flow Induced Vibration Design

Flow past a blunt body, such as a circular cylinder, usually experiences boundary layer separation and very strong flow oscillations in the wake region behind the body at a discrete frequency that is correlated to the Reynolds number of the flow. The periodic nature of the vortex shedding phenomenon can sometimes lead to unwanted structural vibrations. The effect of vibrating instability of a single cylinder is investigated in a uniform flow using the power of computational methods. Fluid structure coupling procedure predicts the fluid forces responsible for structural vibrations. An implicit approach to the solution of the unsteady two-dimensional Navier-Stokes equations is used for computation of flow parameters. Calculations are performed in parallel using a domain re-meshing/deforming technique with efficient communication requirements. Results for the unsteady shedding flow behind a circular cylinder are presented with experimental comparisons, showing the feasibility of accurate, efficient, time-dependent estimation of shedding frequency and resulting vibrations.

Financing Strategies for a Nuclear Fuel Cycle Facility

To help meet the nation’s energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs computed for the government case. The uncertainty in operations, leading to lower than optimal processing rates (or annual plant throughput), is the most detrimental issue to achieving low unit costs. Conversely, lowering debt interest rates and the required return on investments can reduce costs for private industry.

Root Cause of Thermal Sleeve Loosening in a Korean Standard Nuclear Power Plant

The safety injection (SI) nozzle of a 1000MWe-class Korean standard nuclear power plant (KSNP) is fitted with thermal sleeves (T/S) to alleviate thermal fatigue. Thermal sleeves in KSNP #3 & #4 in Yeonggwang (YG) & Ulchin (UC) are manufactured out of In-600 and fitted solidly without any problem, whereas KSNP #5 & #6 in the same nuclear power plants, also fitted with thermal sleeves made of In-690 for increased corrosion resistance, experienced a loosening of thermal sleeves in all reactors except KSNP YG #5-1A, resulting in significant loss of generation availability. An investigation into the cause of the loosening of the thermal sleeves only found out that the thermal sleeves were subject to severe vibration and rotation, failing to uncover the root cause and mechanism of the loosening. In an effort to identify the root cause of T/S loosening, three suspected causes were analyzed: (1) the impact force of flow on the T/S when the safety SI nozzle was in operation, (2) the differences between In-600 and In-690 in terms of physical and chemical properties (notably the thermal expansion coefficient), and (3) the positioning error after explosive expansion of the T/S as well as the asymmetric expansion of T/S. It was confirmed that none of the three suspected causes could be considered as the root cause. However, after reviewing design changes applied to the Palo Verde nuclear plant predating KSNP YG #3 & #4 to KSNP #5 & #6, it was realized that the second design modification (in terms of groove depth & material) had required an additional explosive energy by 150% in aggregate, but the amount of gunpowder and the explosive expansion method were the same as before, resulting in insufficient explosive force that led to poor thermal sleeve expansion. T/S measurement data and rubbing copies also support this conclusion. In addition, it is our judgment that the acceptance criteria applicable to T/S fitting was not strict enough, failing to single out thermal sleeves that were not expanded sufficiently. Furthermore, the T/S loosening was also attributable to lenient quality control before and after fitting the T/S that resulted in significant uncertainty. Lastly, in a flow-induced vibration test planned to account for the flow mechanism that had a direct impact upon the loosening of the thermal sleeves that were not fitted completely, it was discovered that the T/S loosening was attributable to RCS main flow. In addition, it was proven theoretically that the rotation of the T/S was induced by vibration.

Operation and Control of the PBMR Demonstration Power Plant

A large interest in High Temperature Gas-cooled Reactors (HTGR) has been shown in recent years. HTGR power plants show a number of advantages over existing technology including improved safety, modular design and high temperatures for process heat applications. HTGR plants with closed loop direct cycle power conversion units have unique transient responses which is different from existing nuclear plants as well as conventional non-nuclear power plants. The operation and control for a HTGR power plant therefore poses new and different challenges. This paper describes the modes of operation for the Pebble Bed Modular Reactor (PBMR) demonstration plant. The PBMR demonstration plant is an advanced helium-cooled, graphite-moderated HTGR consisting of a closed loop direct cycle power conversion unit. The use of transient analysis simulation makes it possible to develop effective control strategies and design controllers for use in the power conversion unit as well as the reactor. In addition to plant controllers the operator tasks and operational technical specifications can be developed and evaluated making use of transient analysis simulation of the plant together with the control system. The main challenges in the operation and control of the reactor and power conversion unit are highlighted with simulation results. Control strategies in different operating regions are shown and results for the power conversion unit start-up transition and the loss of the grid connection during power operation are presented.

Silicon Carbide and Uranium Oxide Based Composite Fuel Preparation Using Polymer Infiltration and Pyrolysis

Ceramic composite pellets consisting of uranium oxide, U3O8, particles in a silicon carbide matrix are fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, spherical particles of depleted uranium oxide, in the form of U3O8, are dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900°C under a continuous flow of ultra high purity (UHP) argon. Pyrolysis of AHPCS produces near-stoichiometric amorphous SiC at 900°C. Multiple polymer infiltration and pyrolysis (PIP) cycles are required to minimize open porosity and densify the silicon carbide matrix, in order to enhance the mechanical strength of the material. Structural characterization is carried out after first pyrolysis to investigate chemical interaction between U3O8 and SiC. The physical and mechanical properties are also quantified, and it is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix. Furthermore, the processing technique involves lower energy requirements than conventional sintering processes currently in practice.

Assessment of HTGR Helium Compressor Analysis Tool Based on Newton-Raphson Numerical Application to Throughflow Analysis

This study describes the development of a computer program for analyzing the off-design performance of axial flow helium compressors, which is one of the major concerns for the power conversion system of a high temperature gas-cooled reactor (HTGR). The compressor performance has been predicted by the aerodynamic analysis of meridional flow with allowances for losses. The governing equations have been derived from Euler turbomachine equation and the streamline curvature method, and then they have been merged into linearized equations based on the Newton-Raphson numerical method. The effect of viscosity is considered by empirical correlations to introduce entropy rises caused by primary loss sources. Use of the method has been illustrated by applying it to a 20-stage helium compressor of the GTHTR300 plant. As a result, the flow throughout the stages of the compressor has been predicted and the compressor characteristics have been also investigated according to the design specification. The program results show much better stability and good convergence with respect to other throughflow methods, and good agreement with the compressor performance map provided by JAEA.

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Rebalancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined.

Application of RFID to High-Reliability Nuclear Power Plant Construction

In nuclear power plant construction, countless variety of parts, products, and jigs more than one million are treated under construction. Furthermore, strict traceability to the history of material, manufacturing, and installation is required for all products from the start to finish of the construction, which enforce much workforce and many costs at every project. In an addition, the operational efficiency improvement is absolutely essential for the effective construction to reduce the initial investment for construction. As one solution, RFID (Radio Frequent Identification) application technology, one of the fundamental technologies to realize a ubiquitous society, currently expands its functionality and general versatility at an accelerating pace in mass-production industry. Hitachi believes RFID technology can be useful of one of the key solutions for the issues in non-mass production industry as well. Under this situation, Hitachi initiated the development of next generation plant concept (ubiquitous plant construction technology) which utilizes information and RFID technologies. In this paper, our application plans of RFID technology to nuclear power is described.

An Example of an INPRO Assessment of an INS in the Area of Economics

Following a resolution of the General Conference of the IAEA in the year 2000 an International Project on Innovative Nuclear Reactors and Fuel Cycles, referred to as INPRO, was initiated. INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contribution to meeting the energy needs of the 21st century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. This paper discusses the example presented in the manual for performing an INPRO assessment in the area of economics. The example considers a private utility, operating in a liberalized market that is planning for an additional supply of power of about 600 MWe within a time frame of 10 years. Two nuclear options, an LWR and a HWR, are considered as well as a gas-fired plant using liquefied gas as fuel.

Design and Testing of D.C. Conduction Pump for Sodium Cooled Fast Reactor

DC Conduction pump immersed in sodium forms a part of Failed Fuel Location Module (FFLM) of 500 MWe Fast Breeder Reactor (PFBR) currently under construction. FFLM housed in control plug of the reactor, is used to locate the failed fuel sub-assembly due to clad rupture in the fuel pin. The DC conduction pump sucks the sodium from the top of fuel sub-assemblies through the selector valve and pumps the sodium to hold up for detecting the presence of delayed neutrons. Presence of delayed neutron is the indication of failure in the sampled fuel sub-assembly. The DC Conduction Pump was chosen because of its low voltage operation (2 V) where argon/alumina ceramic can provide required electrical insulation even at operating temperature of 560°C without much complication on the manufacturing front. Sampling of sodium from top of different sub-assemblies is achieved by operation of selector valve in-conjunction with the drive motor. FFLM requires the pump to be immersed in sodium pool at ∼560°C located above the fuel sub-assemblies in the reactor. The Pump of 0.36 m3/h capacity and developing 1.45 Kg/ cm2 pressure was designed, manufactured and tested. The DC Conduction Pump has a stainless steel duct filled with liquid sodium, which is to be pumped. The stainless steel duct is kept in magnetic field obtained by means of electromagnet. The electromagnet is made of soft iron and the coil made of copper conductor surrounds the yoke portion of electromagnet. The external DC source of 2000 Amps, 2 Volt is used to send current through sodium placed in the stainless steel duct and the same current is sent through copper coil of electromagnet for producing required magneto motive force, which in turn produces required magnetic field. The interaction of current in sodium (placed in stainless steel duct) and magnetic field produced by the electromagnet in the duct region produces pumping force in the sodium. Electromagnet, copper coil, stainless steel duct, copper bus bar etc. are encapsulated in stainless steel shell. Hydraulic characteristics, efficiency, cavitation free operation at operating temperatures was ascertained by conducting tests in sodium loop called Large Component Test Rig (LCTR). The pump was also endurance tested for 750 hrs. The performance tests on DC Conduction Pump indicate that the pump meets the target specification at reactor operating condition. This paper deals with design, construction and performance testing of DC Conduction Pump.