Data Report on Post-Irradiation Dimensional Change of AGC-1 Samples

This report summarizes the initial dimensional changes for loaded and unloaded AGC-1 samples. The dimensional change for all samples is presented as a function of dose. The data is further presented by graphite type and applied load levels to illustrate the differences between graphite forming processes and stress levels within the graphite components. While the three different loads placed on the samples have been verified [ ref: Larry Hull’s report] verification of the AGC-1 sample temperatures and dose levels are expected in the summer of 2012. Only estimated dose and temperature values for the samples are presented in this report to allow a partial analysis of the results

Microstructural Characterization of Next Generation Nuclear Graphites

This article reports the microstructural characteristics of various petroleum and pitch based nuclear graphites (IG-110, NBG-18, and PCEA) that are of interest to the next generation nuclear plant program. Bright-field transmission electron microscopy imaging was used to identify and understand the different features constituting the microstructure of nuclear graphite such as the filler particles, microcracks, binder phase, rosette-shaped quinoline insoluble (QI) particles, chaotic structures, and turbostratic graphite phase. The dimensions of microcracks were found to vary from a few nanometers to tens of microns. Furthermore, the microcracks were found to be filled with amorphous carbon of unknown origin. The pitch coke based graphite (NBG-18) was found to contain higher concentration of binder phase constituting QI particles as well as chaotic structures. The turbostratic graphite, present in all of the grades, was identified through their elliptical diffraction patterns. The difference in the microstructure has been analyzed in view of their processing conditions

Nitride Fuel Development Using Cryo-process Technique

A new cryo-process technique has been developed for the fabrication of advanced fuel for nuclear systems. The process uses a new cryo-processing technique whereby small, porous microspheres (<2000 µm) are formed from sub-micron oxide powder. A simple aqueous particle slurry of oxide powder is pumped through a microsphere generator consisting of a vibrating needle with controlled amplitude and frequency. As the water-based droplets are formed and pass through the microsphere generator they are frozen in a bath of liquid nitrogen and promptly vacuum freeze-dried to remove the water. The resulting porous microspheres consist of half micron sized oxide particles held together by electrostatic forces and mechanical interlocking of the particles. Oxide powder microspheres ranging from 750 µm to 2000 µm are then converted into a nitride form using a high temperature fluidized particle bed. Carbon black can be added to the oxide powder before microsphere formation to augment the carbothermic reaction during conversion to a nitride. Also, the addition of ethyl alcohol to the aqueous slurry reduces the surface tension energy of the droplets resulting in even smaller droplets forming in the microsphere generator. Initial results from this new process indicate a lower impurity contamination in the final nitrides due to the single feed stream of particles, material handling and conversion are greatly simplified, a minimum of waste and personnel exposure are anticipated, and finally the conversion kinetics may be greatly increased because of the small oxide powder size (sub-micron) forming the porous microsphere. Thus far the fabrication process has been successful in demonstrating all of these improvements with surrogate ZrO2 powder. Further tests will be conducted in the future using the technique on UO2 powders

THE NEXT GENERATION NUCLEAR PLANT GRAPHITE PROGRAM

Developing new nuclear grades of graphite used in the core of a High Temperature Gas-cooled Reactor (HTGR) is one of the critical development activities being pursued within the Next Generation Nuclear Plant (NGNP) program. Graphite’s thermal stability (in an inert gas environment), high compressive strength, fabricability, and cost effective price make it an ideal core structural material for the HTGR reactor design. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermo-mechanical design of the structural graphite in NGNP is based. The NGNP graphite R&D program has selected a handful of commercially available types for research and development activities necessary to qualify this nuclear grade graphite for use within the NGNP reactor. These activities fall within five primary areas; 1) material property characterization, 2) irradiated material property characterization, 3) modeling, and 4) ASTM test development, and 5) ASME code development efforts. Individual research and development activities within each area are being pursued with the ultimate goal of obtaining a commercial operating license for the nuclear graphite from the US NRC

Structural Ceramic Composites for Nuclear Applications

A research program has been established to investigate fiber reinforced ceramic composites to be used as control rod components within a Very High Temperature Reactor. Two candidate systems have been identified, carbon fiber reinforced carbon (Cf/C) and silicon carbide fiber reinforced silicon carbide (SiCf/SiC) composites. Initial irradiation stability studies to determine the maximum dose for each composite type have been initiated within the High Flux Isotope Reactor at Oak Ridge National Laboratory. Test samples exposed to 10 dpa irradiation dose have been completed with future samples to dose levels of 20 and 30 dpa scheduled for completion in following years. Mechanical and environmental testing is being conducted concurrently at the Idaho National Laboratory and at Pacific Northwest National Laboratory. High temperature test equipment, testing methodologies, and test samples for high temperature (up to 1600º C) tensile strength and long duration creep studies have been established. Specific attention was paid to the architectural fiber preform design as well as the materials used in construction of the composites. Actual testing of both tubular and flat, "dog-bone" shaped tensile composite specimens will begin next year. Since there is no precedence for using ceramic composites within a nuclear reactor, ASTM standard test procedures will be established from these mechanical and environmental tests. Close collaborations between the U.S. national laboratories and international collaborators (i.e. France and Japan) are being forged to establish both national and international test standards to be used to qualify ceramic composites for nuclear reactor applications

AGC-1 Pre-Irradiation Data Report Status

The Next Generation Nuclear Plant (NGNP) Graphite R&amp;D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All samples in the experiment will be fully characterized before irradiation, irradiated in the Advanced Test Reactor (ATR), and then re-examined to determine the irradiation induced changes to key materials properties in the different graphite grades. The information generated during the AGC experiment will be utilized for NRC licensing of NGNP reactor designs, shared with international collaborators in the Generation IV Information Forum (GIF), and eventually utilized in ASME design code for graphite nuclear applications. This status report will describe the process the NGNP Graphite R&amp;D program has developed to record the AGC1 pre-irradiation examination data

Data Report on Post-Irradiation Dimensional Change of AGC-1 Samples

This report summarizes the initial dimensional changes for loaded and unloaded AGC-1 samples. The dimensional change for all samples is presented as a function of dose. The data is further presented by graphite type and applied load levels to illustrate the differences between graphite forming processes and stress levels within the graphite components. While the three different loads placed on the samples have been verified [ ref: Larry Hull’s report] verification of the AGC-1 sample temperatures and dose levels are expected in the summer of 2012. Only estimated dose and temperature values for the samples are presented in this report to allow a partial analysis of the results

AGC-1 Pre-Irradiation Data Report Status

The Next Generation Nuclear Plant (NGNP) Graphite R&amp;D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All samples in the experiment will be fully characterized before irradiation, irradiated in the Advanced Test Reactor (ATR), and then re-examined to determine the irradiation induced changes to key materials properties in the different graphite grades. The information generated during the AGC experiment will be utilized for NRC licensing of NGNP reactor designs, shared with international collaborators in the Generation IV Information Forum (GIF), and eventually utilized in ASME design code for graphite nuclear applications. This status report will describe the process the NGNP Graphite R&amp;D program has developed to record the AGC1 pre-irradiation examination data

Impurity Element Diffusion in Pressure Vessel Steels at Reactor Ambient Temperatures

U of I OnlyRestricted to UIUC communit

Dynamics of the Thirteenth-Century Depopulation of the Northern San Juan: The View from Cedar Mesa.

Included here is a presentation made by the authors at the Society for American Archaeology meeting in St. Louis, MO, April 15, 2010