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Burchell, Timothy D.

Publications and source records attributed to Burchell, Timothy D..

USDOE ART Graphite – Selection and Acquisition Strategy

The previous Selection and Acquisition Strategy document dates from 2007. Since then, significant changes have occurred with reactor vendors and with the graphite manufacturers. There have also been changes in the membership of the Gen IV International Forum. These factors were considered when formulating a revised selection and selection strategy. In summary, the new graphite selection and acquisition plan is that no graphite billets will be purchased until there is new graphite production (NBG-18, PCEA). Existing supplies of NBG-17, NBG-18, PCEA, IG-110, 2114 are considered enough for future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Mersen Grade 2114: Tensile Strength from Brazilian Disc Testing

The ASTM Brazilian disc graphite strength (splitting tensile strength, σ sts ) test method (ASTM D8289) is of interest because the small-specimen geometry is compatible with that of environmental effects specimens, thus allowing for environmental effects such as irradiation, irradiation creep, or oxidation on tensile strength to be investigated. The Brazilian disc strength of Mersen grade 2114 graphite (billet 116310 slab 1) is reported and compared with strength data previously obtained using larger cylindrical ASTM dog-bone specimens and with Brazilian disc strength data obtained from billet 116310 slab 5.

36 MATERIALS SCIENCE↗

Mersen Grade 2114: A Comparison of Tensile Strength Data

The ASTM Brazilian disc graphite strength (splitting tensile strength, σ sts ) test method (ASTM D8289) is of interest because the small-specimen geometry is compatible with that of environmental effects specimens, thus allowing for environmental effects such as irradiation, irradiation creep, or oxidation on tensile strength to be investigated. The Brazilian disc strength of Mersen grade 2114 graphite is reported and compared with strength data previously obtained using larger cylindrical ASTM dog-bone specimens.

36 MATERIALS SCIENCE↗

Grade 2114: Flexure Strength and Elastic Properties

This document reports flexural strength, bulk density, and elastic constants data (the last determined by two separate methods) for Mersen grade 2114, billet 116310. These data are needed to support the design of graphite core components. This technical memorandum is responsive to work package AT-19OR03050405, “Graphite Materials Properties—ORNL.”

36 MATERIALS SCIENCE↗

Effects of High-Temperature Irradiation on the Microstructure of Nuclear Graphite

This report summarizes the nondestructive microstructure characterization undertaken at Oak Ridge National Laboratory (ORNL) to understand the effects of high-temperature neutron irradiation on the structure of nuclear graphite. Nuclear graphite specimens were irradiated in the ORNL High Flux Isotope Reactor (HFIR) in the High Temperature Vehicle (HTV) with target irradiation temperatures of 900ºC, 1200ºC, and 1500ºC [1, 2]. The experiment was irradiated in the HFIR flux trap for two irradiation cycles, providing the specimens in the highest-fluence positions a total exposure of 4.5×10 21 n/cm 2 [E>0.1 MeV], equivalent to radiation damage of ~3.3 dpa (displacements per atom) according to the conversion factor in Burchell [3].

36 MATERIALS SCIENCE↗

Experimental Plan for NBG-18 Graphite Bi-axial Strength Testing in the Second and Third Stress Quadrants

The plan for test system's shake down and eventual (funding permitting) bi-axial testing of grade NBG-18 graphite in the second and third stress quadrant are reported. Testing in the first and fourth stress quadrants was previously reported. Details of the specimen and test system design are given along with the anticipated stress ratios to be tested, including specimen numbers are reported. The estimated test pressures needed to attain the planned stress ratios are reported.

36 MATERIALS SCIENCE↗

THE USA’S ADVANCED REACTOR TECHNOLOGIES (ART) GRAPHITE R&D PROGRAM

The USA DOE Advanced Reactor Technologies (ART) graphite R&D program supporting HTR design is one of the longest running nuclear graphite R&D programs. Significant progress has been made in several areas of research within this extensive program including unirraidated and irradiated testing of nuclear graphite, oxidation results and modeling, fracture behavior, irradiation damage studies, and several international collaboration studies. The latest results in these research areas as well as a discussion of a new direction in nuclear graphite irradiation for the USA program will be presented. Recently, the USA made a decision to support moderate to high dose irradiation of nuclear graphite (0.5 to 15 dpa). This will provide irradiation data for graphite HTR core components up to 15 dpa for numerous nuclear graphite grades. The irradiation plan and a tentative schedule for this new irradiation direction will be presented in detail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Review of Finite Element Method Models for Nuclear Graphite Applications

Graphite components serve as structural support and as a neutron moderator in multiple nuclear power stations and is also a prospective material in Generation IV reactor concepts. The environment (e.g. high temperature and intense radiation) within a graphite core changes the geometry, microstructure and material properties of the graphite components, that could limit the lifetime (i.e. safe operation) of a reactor. Some of the assessments made for graphite reactor core components include Finite Element Method (FEM) simulations. These simulations are used to estimate the stress state and strain state and probability of failure of irradiated graphite components. FEM also forms part of the validation of measurements or as a complementary technique to experimental procedures in graphite research. A comprehensive overview of FEM methodologies is provided in this review, including: FEM software, basics steps to set up a FEM model, and analytical solutions used to validate simulations for this field of study. In conclusion, the objective of this paper is to inform researchers in the different available FEM techniques and tools used in the research of nuclear graphite.

42 ENGINEERING↗

Report on the ASME design code development effort for Composite Core Components for High Temperature Reactors

This report summarizes the activities and status of the ASME design code development effort for Composite Core Components for High Temperature Reactors during Fiscal Year 2018. The subsection drafts develop for ASME BPVC Division 5 Section III, namely Subsection HA Subpart B and Subsection HH Subpart B, are near publication. It is likely that Subsection HH Subpart B will be accepted during the next ASME review meeting. Subsection HA Subpart B may have to continue through more review iterations before final acceptance. Two new standards for fiber-reinforced advanced ceramic composites have been published and a report on the technology gap for Molten Salt Reactor applications of SiC/SiC composites was completed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

XRD and SANS evaluation of HOPG and Polycrystalline Graphite

As the primary construction material used in gas-cooled reactor cores, either of high-temperature (HTR) or very high-temperature (VHTR) design, nuclear graphite must be well characterized and variations in its properties upon high-energy neutron irradiation must be known, understood, and predict ed as accurately as possible before a reactor is built and operational. However, even after more than 70 years of operation of graphite-moderated reactors and intense research on graphite’s behavior in an extreme environment created by high temperatures, irradiation, mechanical stress, and chemically reactive (oxidizing) gaseous environment, questions still remain.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation with temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

HTV Post Irradiation Examination (PIE) Results and Analysis

This report is responsive to 2017 Idaho National Laboratory (INL) Memorandum Purchase Order No. 00181805, statement of work 13687 “HTR Graphite R&D – FY ORNL 2017 MPO”, Deliverable 12A “issue final report on the HTV PIE to INL”, due June 15 th 2017.The HTV (High Temperature Vessel) was a HFIR (High Flux Isotope Reactor) high temperature target rod capsule in support of the DOE’s (Department of Energy) High Temperature Reactor (HTR) program. The HTV capsule contained eight sub-capsules, each with nine graphite specimens. The sub-capsules were designed to operate at ~900°C, ~1200°C or ~1600°C. The actual irradiation temperature was established by interrogating SiC and graphite TMs during PIE (post-irradiation examination. The capsule dose ranged from 1.49 dpa to 3.34 dpa, and varied with specimen position within the capsule the maximum dose occurring at the reactor mid-plane and the minimum dose occurring at the core periphery. The irradiated graphite specimens were examined during PIE to establish their high temperature dimensional and elastic (Young’s) modulus changes. The PIE results are reported and analyzed here. The dimensional and modulus changes were shown to have a strong thermal dependence. The dimensional changes were small, being of the order of a few percent or less, ranging from a maximum shrinkage of -3.1% (PCEA WG, T irr =1500°C at 3.3 dpa) to +0.131% (2114 AG, T irr = 869 at 3.2 dpa). The specimen volume shrinkages varied from –0.81 (2114, T irr = 840 at 3.1 dpa) to -9.15% (PCEA, T irr =1500°C at 3.3 dpa). The volume changes were larger at higher Irradiation temperatures. Of the six grades examined, only 2114 moved into the net positive dimensional swelling situation. Significantly, the volume change for 2114 remained negative at the doses and temperature tested here. Dimensional and volume changes were explained in terms of the temperature dependence of the crystal growth rate, the anisotropy of the crystal growth (crystallographic c-axis growth and crystallographic a-direction shrinkage) and the thermal closure of aligned micro-porosity. The modulus changes were much larger, being of the order of a few tens of percent. A trend of lesser fractional increases of Young’s Modulus with an increasing T irr was noted. The minimum increase in Young’s Modulus was 20% (H-451 WG, T irr =1500°C at 3.3 dpa) and the maximum 71% (IG-110, T irr = 840(actual) at 3.3 dpa). The rise in Young’s modulus is attributed to dislocation pinning by lattice defects produced by neutron irradiation and the temperature dependency is related to the irradiation temperature influence on the number of defect available to pin dislocations.

36 MATERIALS SCIENCE↗

Developments in carbon materials

The following carbon-based materials are reviewed and their applications discussed: fullerenes; graphite (synthetic and manufactured); activated carbon fibers; and carbon-carbon composites. Carbon R&D activities at ORNL are emphasized.

Burchell, Timothy D.↗