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.

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.

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.

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.

A Development of Ceramics Cylinder Type Sulfuric Acid Decomposer for Thermo-Chemical Iodine-Sulfur Process Pilot Plant

The hydrogen production method applying thermo-chemical Iodine-Sulfur process (IS process) which uses a nuclear high temperature gas cooled reactor is world widely greatly concerned from the view point of a combination as a clean method, free carbon dioxide in essence. In this process, it is essential a using ceramic material, especially SiC because a operation condition of this process is very corrosive due to a sulfuric acid atmosphere with high temperature and high pressure. In the IS process, a sulfuric acid decomposer is the key component which performs evaporating of sulfuric acid from liquid to gas and disassembling to SO2 gas. SiC was selected as ceramic material to apply for the sulfuric acid decomposer and a new type of binding material was also developed for SiC junction. This technology is expected to wide application not only for a sulfuric acid decomposer but also for various type components in this process. Process parameters were provided as design condition for the decomposer. The configuration of the sulfuric acid decomposer was studied, and a cylindrical tubes assembling type was selected. The advantage of this type is applicable for various type of components in the IS process due to manufacturing with using only simple shape part. A sulfuric acid decomposer was divided into two regions of the liquid and the gaseous phase of sulfuric acid. The thermal structural integrity analysis was studied for the liquid phase part. From the result of this analysis, it was investigated that the stress was below the strength of the breakdown probability 1/100,000 at any position, base material or junction part. The prototype model was manufactured, which was a ceramic portion in the liquid phase part, comparatively complicated configuration, of a sulfuric acid decomposer. The size of model was about 1.9m in height, 1.0m in width. Thirty-six cylinders including inlet and outlet nozzles were combined and each part article was joined using the new binder (slurry binder) and calcinated. Final polishing of the flange faces established in the entrance nozzles was also satisfactory. Many parts were joinable using new technology (new binder). For this reason, new technology is applicable to manufacture of not only a sulfuric acid decomposer but the instruments in the IS process, or other chemical processes.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 13—Study on Heat Transfer of Direct Condensation of Steam on Subcooled Water Jet

Characteristics of thermal-hydraulic phenomena in the steam injector were examined. In experiments, a water jet from a nozzle of 5 mm diameter flowed into the condensing test section pipe concentrically. The inner diameter of the condensing section was 7, 10, or 20 mm and the length was 105 mm. Steam flowed into the peripheral space between the water jet and the inner wall of the test section and condensed on the ware jet surface. The radial and the axial distributions of velocity and temperature of the water jet were measured. Analyses by using the STAR-CD code were also performed. The temperature measured in the central portion of the water jet was higher than the predicted assuming the ordinary turbulent flow in a pipe. The temperature measured in the peripheral region was lower than the predicted. The radial temperature distribution measured was flatter than the predicted. When the steam condensation rate was large, the measured radial velocity distribution in the water jet was flatter than the predicted. In the case that the steam velocity was quite high, the velocity measured in the peripheral region was higher than that in the center portion. These results implied that the steam condensing on the water jet brought momentum in the water jet to result in more effective radial transport of heat and momentum. The STAR-CD code analyses to allow the interface between the wall that simulated the steam flow part and the water flow that stood for the water jet to move, i.e. creating momentum in-flux at the water jet interface, provided better results to support the experimental results. To increase the interfacial friction had a minor effect on the radial velocity distribution in the tested range.

SCC in Shroud Support Cylinder Vertical Weld Lines (Tokai-2)

SCC was found outer surface of shroud support cylinder vertical weld lines (V8) made of Nickel based alloy, alloy 182, for Tokai-2 (BWR-5) operated since 1978. Three SCC among 4 weld lines of V8 were observed. The material of Alloy182 was known to have SCC potential in BWR environment. Based on the result of finite element method analysis, it was estimated that tensile circumferential stress generated at the upper corner of cylinder corresponding crack location when H7 welded. The integrity assessment against seismic load in design was performed using shell model in finite element method analysis. It assumed conservatively that four vertical through wall cracks along whole length of V8 weld lines and four horizontal through wall partial cracks along H7 weld line at intersections of V8 and H7. The collapse load was estimated by twice slope method, a kind of limit load analysis. As the result of the integrity assessment, a critical horizontal through wall crack length along H7 weld line was about 1400 mm per one crack (about 70% of all circumferences). SCC growth was evaluated to reach the critical length after about 20 years, where maximum crack growth rate, 63mm/year was assumed. It is not judged that the immediate repair of vertical SCC is necessary.