Algorithm Development and Computer Graphic Simulation of an Articulated Transporter/ Manipulator System

The University of Florida is part of a multi-university research effort, sponsored by the U.S. Department of Energy, which is underway to develop and deploy an advanced semi-autonomous robotic system for use in nuclear power stations. The robotic system being designed by the Florida/Odetics team can be described . as an Articulated Transporter/Manipulator System (ATMS) which has several unique motion and transport capabilities. The ATMS will be capable of performing tasks in radioactive hazardous environments to reduce occupational radiation exposure of plant personnel and to increase the availability of the plant. This paper will describe the key design and control features of the ATMS with emphasis placed on the implementation of specific motion control algorithms

An Innovative High Thermal Conductivity Fuel Design

Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. UO2 has the advantages of a high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation. The main disadvantage of UO2 is its low thermal conductivity. During a reactor’s operation, because the thermal conductivity of UO2 is very low, for example, about 2.8 W/m-K at 1000 oC [1], there is a large temperature gradient in the UO2 fuel pellet, causing a very high centerline temperature, and introducing thermal stresses, which lead to extensive fuel pellet cracking. These cracks will add to the release of fission product gases after high burnup. The high fuel operating temperature also increases the rate of fission gas release and the fuel pellet swelling caused by fission gases bubbles. The amount of fission gas release and fuel swelling limits the life time of UO2 fuel in reactor. In addition, the high centerline temperature and large temperature gradient in the fuel pellet, leading to a large amount of stored heat, increase the Zircaloy cladding temperature in a lost of coolant accident (LOCA). The rate of Zircaloy-water reaction becomes significant at the temperature above 1200 oC [2]. The ZrO2 layer generated on the surface of the Zircaloy cladding will affect the heat conduction, and will cause a Zircaloy cladding rupture. The objective of this research is to increase the thermal conductivity of UO2, while not affecting the neutronic property of UO2 significantly. The concept to accomplish this goal is to incorporate another material with high thermal conductivity into the UO2 pellet. Silicon carbide (SiC) is a good candidate, because the thermal conductivity of single crystal SiC is 60 times higher than that of UO2 at room temperature and 30 times higher at 800 oC [3]. Silicon carbide also has the properties of low thermal neutron absorption cross section, high melting point, good chemical stability and good irradiation stability. Silicon carbide is expected to form a conductive lattice in UO2 for heat to flow out of the fuel pellet, and the thermal conductivity of SiC is anticipated to control the thermal conductivity of the fuel pellet. In this research, the effect of the SiC additive on the neutronic properties of a UO2 pellet was simulated by CASMO-3, a multi-group two-dimensional transport theory code. Three methods were studied to incorporate SiC into UO2. Firstly, silicon carbide whiskers were mixed with UO2 particles and then hot press sintered to achieve dense pellets. Secondly, a polymer precursor, allylhydridopolycarbosilane (AHPCS), was used to generate a SiC coating on UO2 particles prior to the hot press sintering process. Thirdly, chemical vapor deposition (CVD) process was used to coat UO2 particles with a SiC layer prior to the sintering process. To avoid a reaction that occurs between UO2 and SiC at 1377 oC [4], a two stages low temperature sintering method was used to sinter the mixture of SiC whiskers and UO2 particles or the SiC coated UO2 particles at 1200 oC. The sintered pellets were characterized by X-ray diffraction (XRD) and Scanning Electron Microscope (SEM), and the thermal conductivity of the sintered pellets was to be measured by laser flash method at Idaho National Laboratory. The centerline temperatures of the sintered pellets at the reactor operating condition were calculated based on the measured thermal conductivity

Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel

The Development of Models to Optimize Selection of Nuclear Fuels through Atomic-Level Simulation

Demonstrated that FRAPCON can be modified to accept data generated from first principles studies, and that the result obtained from the modified FRAPCON make sense in terms of the inputs. Determined the temperature dependence of the thermal conductivity of single crystal UO2 from atomistic simulation

AECL/US INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors -- Fuel Requirements and Down-Select Report

The U.S. Advanced Fuel Cycle Program and the Atomic Energy Canada Ltd (AECL) seek to develop and demonstrate the technologies needed to minimize the overall Pu and minor actinides present in the light water reactor (LWR) nuclear fuel cycles. It is proposed to reuse the Pu from LWR spent fuel both for the energy it contains and to decrease the hazard and proliferation impact resulting from storage of the Pu and minor actinides. The use of fuel compositions with a combination of U and Pu oxide (MOX) has been proposed as a way to recycle Pu and/or minor actinides in LWRs. It has also been proposed to replace the fertile U{sup 238} matrix of MOX with a fertile-free matrix (IMF) to reduce the production of Pu{sup 239} in the fuel system. It is important to demonstrate the performance of these fuels with the appropriate mixture of isotopes and determine what impact there might be from trace elements or contaminants. Previous work has already been done to look at weapons-grade (WG) Pu in the MOX configuration [1][2] and the reactor-grade (RG) Pu in a MOX configuration including small (4000 ppm additions of Neptunium). This program will add to the existing database by developing a wide variety of MOX fuel compositions along with new fuel compositions called inert-matrix fuel (IMF). The goal of this program is to determine the general fabrication and irradiation behavior of the proposed IMF fuel compositions. Successful performance of these compositions will lead to further selection and development of IMF for use in LWRs. This experiment will also test various inert matrix material compositions with and without quantities of the minor actinides Americium and Neptunium to determine feasibility of incorporation into the fuel matrices for destruction. There is interest in the U.S. and world-wide in the investigation of IMF (inert matrix fuels) for scenarios involving stabilization or burn down of plutonium in the fleet of existing commercial power reactors. IMF offer the potential advantage for more efficient destruction of plutonium and minor actinides (MA) relative to MOX fuel. Greater efficiency in plutonium reduction results in greater flexibility in managing plutonium inventories and in developing strategies for disposition of MA, as well as a potential for fuel cycle cost savings. Because fabrication of plutonium-bearing (and MA-bearing) fuel is expensive relative to UO{sub 2} in terms of both capital and production, cost benefit can be realized through a reduction in the number of plutonium-bearing elements required for a given burn rate. In addition, the choice of matrix material may be manipulated either to facilitate fuel recycling or to make plutonium recovery extremely difficult. In addition to plutonium/actinide management, an inert matrix fuel having high thermal conductivity may have operational and safety benefits; lower fuel temperatures could be used to increase operating and safety margins, uprate reactor power, or a combination of both. The CANDU reactor offers flexibility in plutonium management and MA burning by virtue of online refueling, a simple bundle design, and good neutron economy. A full core of inert matrix fuel containing either plutonium or a plutonium-actinide mix can be utilized, with plutonium destruction efficiencies greater than 90%, and high (>60%) actinide destruction efficiencies. The Advanced CANDU Reactor (ACR) could allow additional possibilities in the design of an IMF bundle, since the tighter lattice pitch and light-water coolant reduce or eliminate the need to suppress coolant void reactivity, allowing the center region of the bundle to include additional fissile material and to improve actinide burning. The ACR would provide flexibility for management of plutonium and MA from the existing LWR fleet, and would be complementary to the AFCI program in the U.S. Many of the fundamental principles concerning the use of IMF are nearly identical in LWRs and the ACR, including fuel/coolant compatibility, fuel fabrication, and fuel irradiation behavior. In addition, the U.S. and Canada both have interest in development of Generation IV SCWR (supercritical water reactor) technology, to which this fuel type would be applicable for plutonium and MA management. An inert matrix fuel with high thermal conductivity would be particularly beneficial to any SCWR concept. Given these similarities, it is proposed that a joint project be conducted within the framework of a U.S.-Canada INERI project on IMF. This report will present analysis of the inert matrix fuel compositions of interest for application to US and Canadian light water reactor fuel cycles, report on the development of fabrication procedures for these compositions, and provide an overview of the test and demonstration plan for these fuel systems

Final Closeout Report University Research Program in Robotics for Environmental Restoration and Waste Management

The report covers the 2003-04 contract period, with a retrospective of the 11 years for the contract, from 1993 to 2004. This includes personnel, technical publications and reports, plus research laboratories employed. Specific information is given in eight research areas, reporting on all technology developed and/or deployed by the University of Florida

University Research Program in Robotics - "Technologies for Micro-Electrical-Mechanical Systems in directed Stockpile Work (DSW) Radiation and Campaigns", Final Technical Annual Report, Project Period 9/1/06 - 8/31/07

The University Research Program in Robotics (URPR) is an integrated group of universities performing fundamental research that addresses broad-based robotics and automation needs of the NNSA Directed Stockpile Work (DSW) and Campaigns. The URPR mission is to provide improved capabilities in robotics science and engineering to meet the future needs of all weapon systems and other associated NNSA/DOE activities

UNIVERSITY RESEARCH PROGRAM IN ROBOTICS, Final Technical Annual Report, Project Period: 9/1/04 - 8/31/05

The University Research Program in Robotics (URPR) Implementation Plan is an integrated group of universities performing fundamental research that addresses broad-based robotics and automation needs of the NNSA Directed Stockpile Work (DSW) and Campaigns. The URPR mission is to provide improved capabilities of robotics science and engineering to meet the future needs of all weapon systems and other associated NNSA/DOE activities

HoxD Expression in the Fin-fold Compartment of Basal Gnathostomes and Implications for Paired Appendage Evolution

The role of Homeobox transcription factors during fin and limb development have been the focus of recent work investigating the evolutionary origin of limb-specific morphologies. Here we characterize the expression of HoxD genes, as well as the cluster-associated genes Evx2 and LNP, in the paddlefish Polyodon spathula, a basal ray-finned fish. Our results demonstrate a collinear pattern of nesting in early fin buds that includes HoxD14, a gene previously thought to be isolated from global Hoxregulation. We also show that in both Polyodon and the catsharkScyliorhinus canicula (a representative chondrichthyan) late phaseHoxD transcripts are present in cells of the fin-fold and co-localize with And1, a component of the dermal skeleton. These new data support an ancestral role for HoxD genes in patterning the fin-folds of jawed vertebrates, and fuel new hypotheses about the evolution of cluster regulation and the potential downstream differentiation outcomes of distinct HoxD-regulated compartments