Download or read book Comparison Between the Strength Levels of Baseline Nuclear Grade Graphite and Graphite Irradiated in AGC 2 written by and published by . This book was released on 2015 with total page 39 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report details the initial comparison of mechanical strength properties between the cylindrical nuclear-grade graphite specimens irradiated in the second Advanced Graphite Creep (AGC-2) experiment with the established baseline, or unirradiated, mechanical properties compiled in the Baseline Graphite Characterization program. The overall comparative analysis will describe the development of an appropriate test protocol for irradiated specimens, the execution of the mechanical tests on the AGC-2 sample population, and will further discuss the data in terms of developing an accurate irradiated property distribution in the limited amount of irradiated data by leveraging the considerably larger property datasets being captured in the Baseline Graphite Characterization program. Integrating information on the inherent variability in nuclear-grade graphite with more complete datasets is one of the goals of the VHTR Graphite Materials program. Between "sister" specimens, or specimens with the same geometry machined from the same sub-block of graphite from which the irradiated AGC specimens were extracted, and the Baseline datasets, a comprehensive body of data will exist that can provide both a direct and indirect indication of the full irradiated property distributions that can be expected of irradiated nuclear-grade graphite while in service in a VHTR system. While the most critical data will remain the actual irradiated property measurements, expansion of this data into accurate distributions based on the inherent variability in graphite properties will be a crucial step in qualifying graphite for nuclear use as a structural material in a VHTR environment.

Download or read book Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration that is capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is endeavoring to minimize the conservative estimates of as-manufactured mechanical and physical properties in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered "candidate" grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding process. An analysis of the comparison between each of these grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in variability in properties within each of the grades that will ultimately provide the basis for the prediction of in-service performance. The comparative performance of the different types of nuclear-grade graphites will continue to evolve as thousands more specimens are fully characterized from the numerous grades of graphite being evaluated.

Download or read book Radiation Damage in Graphite written by J. H. W. Simmons and published by Elsevier. This book was released on 2013-10-22 with total page 264 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear Energy, Volume 102: Radiation Damage in Graphite provides a general account of the effects of irradiation on graphite. This book presents valuable work on the structure of the defects produced in graphite crystals by irradiation. Organized into eight chapters, this volume begins with an overview of the description of the methods of manufacturing graphite and of its physical properties. This text then presents details of the method of setting up a scale of irradiation dose. Other chapters consider the effect of irradiation at a given temperature on a physical property of graphite. This book discusses as well the changes in dimensions produced by irradiation and the effects of irradiation on the mechanical properties of graphite. The final chapter deals with the accumulation of stored energy, which is one of the main problems caused by the irradiation of graphite in nuclear reactors. This book is a valuable resource for physicists and chemical physicists.

Download or read book Baseline Graphite Characterization written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Next Generation Nuclear Plant Project Graphite Research and Development program is currently establishing the safe operating envelope of graphite core components for a very high temperature reactor design. To meet this goal, the program is generating the extensive amount of quantitative data necessary for predicting the behavior and operating performance of the available nuclear graphite grades. In order determine the in-service behavior of the graphite for the latest proposed designs, two main programs are underway. The first, the Advanced Graphite Creep (AGC) program, is a set of experiments that are designed to evaluate the irradiated properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences, and compressive loads. Despite the aggressive experimental matrix that comprises the set of AGC test runs, a limited amount of data can be generated based upon the availability of space within the Advanced Test Reactor and the geometric constraints placed on the AGC specimens that will be inserted. In order to supplement the AGC data set, the Baseline Graphite Characterization program will endeavor to provide supplemental data that will characterize the inherent property variability in nuclear-grade graphite without the testing constraints of the AGC program. This variability in properties is a natural artifact of graphite due to the geologic raw materials that are utilized in its production. This variability will be quantified not only within a single billet of as-produced graphite, but also from billets within a single lot, billets from different lots of the same grade, and across different billets of the numerous grades of nuclear graphite that are presently available. The thorough understanding of this variability will provide added detail to the irradiated property data, and provide a more thorough understanding of the behavior of graphite that will be used in reactor design and licensing. This report covers the development of the Baseline Graphite Characterization program from a testing and data collection standpoint through the completion of characterization on the first billet of nuclear-grade graphite. This data set is the starting point for all future evaluations and comparisons of material properties.

Download or read book Statistical Comparison of the Baseline Mechanical Properties of NBG 18 and PCEA Graphite written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR), a graphite-moderated, helium-cooled design that is capable of producing process heat for power generation and for industrial process that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is endeavoring to minimize the conservative estimates of as-manufactured mechanical and physical properties by providing comprehensive data that captures the level of variation in measured values. In addition to providing a comprehensive comparison between these values in different nuclear grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons and variations between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between the two grades of graphite that were initially favored in the two main VHTR designs. NBG-18, a medium-grain pitch coke graphite from SGL formed via vibration molding, was the favored structural material in the pebble-bed configuration, while PCEA, a smaller grain, petroleum coke, extruded graphite from GrafTech was favored for the prismatic configuration. An analysis of the comparison between these two grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in variability in properties within each of the grades that will ultimately provide the basis for the prediction of in-service performance. The comparative performance of the different types of nuclear grade graphites will continue to evolve as thousands more specimens are fully characterized from the numerous grades of graphite being evaluated.

Download or read book Defect Evolution in High temperature Irradiated Nuclear Graphite written by Steve Johns and published by . This book was released on 2020 with total page 113 pages. Available in PDF, EPUB and Kindle. Book excerpt: "Graphite has historically been used as a moderator material in nuclear reactor designs dating back to the first man-made nuclear reactor to achieve criticality (Chicago Pile 1) in 1942. Additionally, graphite is a candidate material for use in the future envisioned next-generation nuclear reactors (Gen IV); specifically, the molten-salt-cooled (MSR) and very-high-temperature reactor (VHTR) concepts. Gen IV reactor concepts will introduce material challenges as temperature regimes and reactor lifetimes are anticipated to far exceed those of earlier reactors. Irradiation-induced defect evolution is a fundamental response in nuclear graphite subjected to irradiation. These defects directly influence the many property changes of nuclear graphite subjected to displacing radiation; however, a comprehensive explanation for irradiation-induced dimensional change remains elusive. The macroscopic response of graphite subjected to displacing irradiation is often modeled semi-empirically based on irradiation data of specific graphite grades (some of which are obsolete). The lack of an analytical description of the response of nuclear graphite subjected to irradiation is due in part to the complex microstructure of synthetic semi-isotropic graphites. Chapter One provides a general overview of the application, processing, and irradiation-induced property changes of nuclear graphite. The key properties affected by displacing irradiation include, but are not limited to, coefficient of thermal expansion (CTE), irradiation creep, and irradiation-induced dimensional change. Additionally, historical models of radiation damage in nuclear graphite, including their inadequacies in accurately describing property changes, are discussed. It should be noted that a comprehensive explanation for all irradiation-induced property change is beyond the scope of this work, which is focused on the evolution of novel atomic-level defects in high-temperature irradiated nuclear graphite and the implications of these defects for the current understanding of irradiation-induced dimensional change. Chapter Two is focused on the development of a novel oxidation-based transmission electron microscopy (TEM) sample-preparation technique for nuclear-grade graphite. Conventionally, TEM specimens are prepared via ion-milling or a focused ion beam (FIB); however, these techniques require the use of displacing radiation and may result in localized areas of irradiation damage. As a result, distinguishing defect structures created as artifacts during sample preparation from those created by electron- or neutron-irradiation can be challenging. Bulk nuclear graphite grades IG-110, NBG-18, and highly oriented pyrolytic graphite (HOPG) were oxidized using a new jet-polishing-like setup where oxygen is used as an etchant. This technique is shown to produce self-supporting electron-transparent TEM specimens free of irradiation-induced artifacts; thus, these specimens can be used as a baseline for in situ irradiation experiments as they have no irradiation-induced damage. Chapter Three examines the dynamic evolution of defect structures in nuclear graphite IG-110 subjected to electron-irradiation. As use of fast neutrons for irradiation experiments is dangerous, expensive, and time consuming, electron-irradiation is arguably a useful surrogate; however, comparisons between the two irradiating particles is also discussed. In situ video recordings of specimens undergoing simultaneous heating and electron-irradiation were used to analyze the dynamic atomic-level defect evolution in real time. Novel fullerene-like defect structures are shown to evolve as a direct result of high-temperature electron-irradiation and cause significant dimensional change to crystallites. Neutron-irradiated nuclear graphite IG-110 was supplied by Idaho National Laboratory as part of the Advanced Graphite Creep capsule experiments (AGC-3). Chapter Four reports the preliminary characterization of IG-110 neutron-irradiated at 817°C to a dose of 3.56 displacements per atom (dpa). Shown is experimental evidence of a 'ruck and tuck' defect occurring in high-temperature neutron-irradiated nuclear graphite. The 'ruck and tuck' defect arises due to irradiation-induced defects. The interaction of these defects results in the buckling of atomic planes and the formation of a structure composed of two partial carbon nanotubes. The "buckle, ruck and tuck" model was first theoretically predicted via computational modeling in 2011 as a plausible defect structure/mechanism occurring in high-temperature neutron-irradiated graphite by Prof. Malcolm Heggie et al. Chapter Four shows the first direct experimental results to support the "buckle, ruck and tuck" model. Chapter Five further characterizes nuclear graphite IG-110 neutron-irradiated at high temperature (>=800 °C) at doses of 1.73 and 3.56 dpa. Results show further evidence to support the 'buckle, ruck and tuck' model and additionally show the presence of larger concentric shelled fullerene-like defects. Fullerene-like defects were found to occur in disordered regions of the microstructure including within nanocracks (Mrozowski cracks). These results agree with high-temperature electron-irradiation studies which showed the formation of fullerene-like defects in-situ and give additional validity to the use of high-flux electron-irradiation as a useful approximation to neutron-irradiation. Furthermore, Chapter Five gives valuable insight to unresolved quantitative anomalies of historical models of graphite expansion and may improve the understanding of current empirical and theoretical models of irradiation-induced property changes in nuclear graphite."--Boise State University ScholarWorks.

Download or read book AGC 2 Specimen Post Irradiation Data Package Report written by and published by . This book was released on 2015 with total page 167 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Advanced Reactor Technology (ART) Graphite R & D program is conducting an extensive graphite irradiation program to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[,] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

Download or read book AGC 2 Graphite Preirradiation Data Package written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The NGNP Graphite R & D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

Download or read book AGC 2 Disassembly Report written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Next Generation Nuclear Plant (NGNP) Graphite Research and Development (R & D) Program is currently measuring irradiated material properties for predicting the behavior and operating performance of new nuclear graphite grades available for use within the cores of new very high temperature reactor designs. The Advanced Graphite Creep (AGC) experiment, consisting of six irradiation capsules, will generate irradiated graphite performance data for NGNP reactor operating conditions. The AGC experiment is designed to determine the changes to specific material properties such as thermal diffusivity, thermal expansion, elastic modulus, mechanical strength, irradiation induced dimensional change rate, and irradiation creep for a wide variety of nuclear grade graphite types over a range of high temperature, and moderate doses. A series of six capsules containing graphite test specimens will be used to expose graphite test samples to a dose range from 1 to 7 dpa at three different temperatures (600, 900, and 1200°C) as described in the Graphite Technology Development Plan. Since irradiation induced creep within graphite components is considered critical to determining the operational life of the graphite core, some of the samples will also be exposed to an applied load to determine the creep rate for each graphite type under both temperature and neutron flux. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR). AGC-1 and AGC-2 will be irradiated in the south flux trap and AGC-3-AGC-6 will be irradiated in the east flux trap. The change in flux traps is due to NGNP irradiation priorities requiring the AGC experiment to be moved to accommodate Fuel irradiation experiments. After irradiation, all six AGC capsules will be cooled in the ATR Canal, sized for shipment, and shipped to the Materials and Fuels Complex (MFC) where the capsule will be disassembled in the Hot Fuel Examination Facility (HFEF). During disassembly, the metallic capsule will be machined open and the individual samples removed from the interior graphite body containing the samples. Samples removed from the capsule will be loaded in a shipping drum and shipped to the Idaho National Laboratory (INL) Research Center (IRC) for initial post-irradiation examination (PIE) and storage for any future testing at the newly completed Carbon Characterization Laboratory (CCL). All work was performed under an ASME NQA-1-2008;1a-2009 compliant quality assurance program.

Download or read book AGC 2 Irradiation Data Qualification Final Report written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf, 491 lbf, or 589 lbf. Loads were consistently within 5% of the specified values throughout the experiment. Stack displacement increased consistently throughout the experiment with total displacement ranging from 1 to 1.5 inches. No anomalous values were identified. During reactor outages, a set of pneumatic rams are used to raise the stacks of graphite creep specimens to ensure the specimens have not become stuck within the test train. This stack raising was performed after all cycles when the capsule was in the reactor. All stacks were raised successfully after each cycle. The load and displacement data are Qualified.

Download or read book Status of the NGNP Graphite Creep Experiments AGC 1 and AGC 2 Irradiated in the Advanced Test Reactor written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

Download or read book Strength of Irradiated Graphite written by and published by . This book was released on 1979 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Experimental data on the mechanical strength of nuclear graphite subjected to fast neutron irradiation are reviewed. At fluences below the turnaround point, the mean tensile, flexural, or compressive strength, S, increases over the unirradiated value, S0, in a manner related to the irradiation-induced change in Young's modulus, E: S/S0 = (E/E0)/sup K/. The exponent, k, takes a value between 0.5 and 1, depending on the graphite grade, the irradiation temperature, and the method used for determining E.A value of 0.5 for k would be expected from the Griffith--Irwin theory of fracture if neither the critical flaw size nor the effective surface energy of a crack is altered by irradiation; a value of 1 would be expected if the strain at failure remains constant. At higher fluences, when expansion starts, strength values decrease. The effect of irradiation on the statistical spread of the strength measurements depends on the graphite grade and the neutron fluence. At fluences below the turnaround point, the coefficient of variation of strength determinations for standard reactor grades is little changed, but at higher fluences, and for fine-grained, high strength materials, the coefficient of variation may increase. Although specimens for different locations in a graphite log, or from different logs of the same grade, vary in strength, under the same conditions of irradiation the fractional increase in strength of these specimens will be similar. After irradiation, graphite appears to have a slightly greater resistance to cyclic fatigue, as measured by the homologous stress limits for survival to a specified number of cycles.

Download or read book Nuclear Graphite written by R. E. Nightingale and published by Academic Press. This book was released on 2013-10-02 with total page 566 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear Graphite focuses on the development and uses of nuclear graphite, including machining practices, manufacture, nuclear properties and structure, radiation, and electrical resistance. The selection first discusses the applications of graphite in the nuclear industry, machining practices, and manufacture. Discussions focus on early, current, and future applications of graphite, impregnation, graphitization, purification, general machining techniques, and equipment and methods in the nuclear industry. The book then examines the structure and nuclear and properties of graphite. The text evaluates radiation-induced structural and dimensional changes; radiation effects on electrical and thermal properties; and radiation effects on mechanical properties. Topics include radiation effects on crystal structure, electrical resistance, thermoelectric power, magnetoresistance, coefficient of friction, irradiation under stress, and elastic moduli of nuclear graphite. The book also ponders on stored energy, annealing radiation effects, and gas-graphite systems. The selection is a dependable source of data for readers interested in the applications of nuclear graphite.

Download or read book AGC 3 Graphite Preirradiation Data Analysis Report written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen's 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to provide specimens with similar neutron dose levels.

Download or read book An Empirical Prediction of Temperature Gradients in Moderator Graphite Within Operating Nuclear Reactors written by Robert L. Carter and published by . This book was released on 1962 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1975-04 with total page 1082 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Graphite Gamma Scan Results written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report documents the measurement and data analysis of the radio isotopic content for a series of graphite specimens irradiated in the first Advanced Graphite Creep (AGC) experiment, AGC-1. This is the first of a series of six capsules planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphites. The AGC-1 capsule was irradiated in the Advanced Test Reactor (ATR) at INL at approximately 700 degrees C and to a peak dose of 7 dpa (displacements per atom). Details of the irradiation conditions and other characterization measurements performed on specimens in the AGC-1 capsule can be found in "AGC-1 Specimen Post Irradiation Data Report" ORNL/TM 2013/242. Two specimens from six different graphite types are analyzed here. Each specimen is 12.7 mm in diameter by 25.4 mm long. The isotope with the highest activity was 60Co. Graphite type NBG-18 had the highest content of 60Co with an activity of 142.89 μCi at a measurement distance of 47 cm.