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Title: Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program

Abstract

This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billetmore » to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.« less

Authors:
 [1];  [1];  [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1364481
Report Number(s):
INL/EXT-16-39604
TRN: US1703334
DOE Contract Number:
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; COMPARATIVE EVALUATIONS; GRAPHITE; TEST REACTORS; TENSILE PROPERTIES; CREEP; PHYSICAL PROPERTIES; graphite; mechanical properties; PCEA; physical properties

Citation Formats

Carroll, Mark Christopher, Cottle, David Lynn, and Rohrbaugh, David Thomas. Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program. United States: N. p., 2016. Web. doi:10.2172/1364481.
Carroll, Mark Christopher, Cottle, David Lynn, & Rohrbaugh, David Thomas. Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program. United States. doi:10.2172/1364481.
Carroll, Mark Christopher, Cottle, David Lynn, and Rohrbaugh, David Thomas. 2016. "Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program". United States. doi:10.2172/1364481. https://www.osti.gov/servlets/purl/1364481.
@article{osti_1364481,
title = {Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program},
author = {Carroll, Mark Christopher and Cottle, David Lynn and Rohrbaugh, David Thomas},
abstractNote = {This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billet to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.},
doi = {10.2172/1364481},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 8
}

Technical Report:

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  • 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,more » 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.« less
  • Here we report the physical properties of a series specimens machined from pilot scale (~ 152 mm diameter x ~305 mm length) grade PCEA recycle billets manufactured by GrafTech. The pilot scale billets were processed with increasing amounts of (unirradiated) graphite (from 20% to 100%) introduced to the formulation with the goal of determining if large fractions of recycle graphite have a deleterious effect on properties. The properties determined include Bulk Density, Electrical Resistivity, Elastic (Young s) Modulus, and Coefficient of Thermal Expansion. Although property variations were observed to be correlated with the recycle fraction, the magnitude of the variationsmore » was noted to be small.« less
  • High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration 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 establishing accurate as-manufactured mechanical and physical property distributions 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 individualmore » 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 the overall variability in properties within each of the grades that will ultimately provide the basis for predicting in-service performance. The comparative performance of the different types of nuclear-grade graphites will naturally continue to evolve as thousands more specimens are fully characterized with regard to strength, physical properties, and thermal performance from the numerous grades of graphite being evaluated.« less
  • 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 ofmore » 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.« less
  • We studied the sorption of neptunium onto tuffs characteristic of the proposed nuclear waste repository at Yucca Mountain, Nevada. The neptunium was in the Np(V) oxidation state under oxidizing conditions in groundwaters from two wells located close to the repository site (J-13 and UE-25 p No.1). We used devitrified, vitric, zeolitic (with emphasis on clinoptilolite-rich samples), and calcite-rich tuffs characteristic of the geology of the site. Neptunium sorbed well onto calcite and calcite-rich tuffs, indicating that a significant amount of neptunium retardation can be expected under fractured-flow scenarios because of calcite coating of the fractures. Neptunium sorption onto clinoptilolite-rich zeoliticmore » tuffs in J-13 well water (pH from 7 to 8.5) was moderate, increased with decreasing pH, and correlated to surface area and amount of clinoptilolite. Neptunium sorbed poorly onto zeolitic tuffs from UE-25 p No.1 groundwater (pH from 7 to 9) and onto devitrified and vitric tuffs from J-13 and UE-25 p No.1 waters (pH from 7 to 9). Iron oxides appeared to be passivated in tuffs, not seeming to contribute to the observed neptunium sorption, even though neptunium sorption onto synthetic iron oxide is significant.« less