Development of Constitutive Equations for Nuclear Grade Graphite for Space Applications I Final Report
Room temperature mechanical behavior of various chemical and experimental graphites
GRAPHITE↗
SEARCH · Search NASA
Results for “nuclear graphite”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Graphite for Advanced Nuclear Reactors: Deployment Readiness Review
Historically, graphite has been used in numerous reactor technologies, including research/test reactors and, commercially, in advanced gas reactors. These non-metallic materials can play a key role as internal core structures, reflectors, or neutron moderators, making them important for deployment of certain advanced reactor technologies. This report explores the industry readiness for graphite material deployment, documenting Codes and Standards applicable to their design, qualification, and manufacturing. Discussions also examine the manufacturing processes, aging/degradation, inspection techniques, and disposal options. The report covers the current status, documents gaps, and proposes some conclusions about approaches to managing some of the current gaps.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Viewing is understanding: Graphite microstructure effects on infiltrated molten salt distribution revealed by 3D neutron tomography
Molten salt infiltration in the pore network of nuclear graphite may cause unwanted changes to graphite's local structure and mechanical and thermal properties. A detailed and comprehensive understanding of molten salt intrusion (distribution across sample cross section and penetration depth) is needed to assess its effects. Here, in this work, we report on an improved methodology for the use of neutron imaging (computed tomography) to evaluate salt penetration and distribution of a wide range of graphite grades with diverse microstructures. Neutron tomography data were acquired on the same graphite sample before and after salt intrusion; the 3D reconstructed volumes were digitally co-registered and subtracted. The difference in neutron attenuation coefficient represents direct visualization of FLiNaK (LiF–NaF–KF) salt distribution in the salt-impregnated graphite samples. This improved methodology was applied to investigate the effect of exposure times (12 h and 336 h) and of graphite microstructure when exposed to FLiNaK at 750 °C and 3 bar (gauge) pressure, starting from flowing argon at near atmospheric pressure. The results show that medium-grained and fine-grained graphites evolve to equilibrium at significantly different rates: fast salt uptake in medium-grained graphites produces salt deposits throughout the volume of graphite specimens, whereas salt infiltration in fine-grained graphites is much slower and limited to exposed surfaces.
FLiNaK infiltration↗
AGC-4 Specimen Post-Irradiation Examination Data Interim Report
This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Low-Temperature Annealing of Nanoscale Defects in Polycrystalline Graphite
Polycrystalline graphite contains multi-scale defects, which are difficult to anneal thermally because of the extremely high temperatures involved in the manufacturing process. In this study, we demonstrate annealing of nuclear graphite NBG-18 at temperatures below 28 °C, exploiting the electron wind force, a non-thermal stimulus. High current density pulses were passed through the specimens with a very low-duty cycle so that the electron momentum could mobilize the defects without heating the specimen. The effectiveness of this technique is presented with a significant decrease in electrical resistivity, defect counts from X-ray computed tomography, Raman spectroscopy, and nanoindentation-based mechanical characterization. Such multi-modal evidence highlights the feasibility of nanoscale defect control at temperatures about two orders of magnitude below the graphitization temperature.
Liu, Gongyuan↗
HDG-1 Graphite Preirradiation Data Package Report
This report documents all pre-irradiation examination material-property measurement data for graphite specimens that are going to be used within the first high dose graphite (HDG) -1 irradiation capsule. The two new HDG capsules signify a major change to the AGC Experiment. HDG-1 and HDG-2 will replace the last two Advanced Graphite Creep (AGC) capsules (AGC-5 and AGC-6) which were designed to irradiate graphite at the extreme upper operational temperatures for a very-high-temperature reactor (VHTR) design, 1100°C. These very high temperature AGC-5 and AGC-6 capsules have been repurposed to re-irradiated specimens (from AGC-2, AGC-3, and AGC-4) at the lower temperatures of 600°C and 800°C. HDG-1 will be irradiated at 600°C and HDG-2 will be irradiated at 800°C. By re-irradiating the previous AGC specimens a total maximum neutron dose of around 15 dpa (displacements per atom) can be achieved for all major graphite grades at irradiation temperatures of 600°C and 800°C. Specimens in the HDG-1 capsule are made up of previously irradiated specimens from the AGC-2 capsule and unirradiated specimens prepared for the now discontinued AGC-5 capsule. Utilizing the irradiated specimens, a maximum neutron dose of around 15 dpa is anticipated. These new maximum dose levels will provide irradiated material property data over a total neutron dose range of 1-15 dpa at a temperature of 600°C when combined with the previous AGC-1 and AGC-2 irradiation data. This will provide quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades for use within high temperature reactor designs. Similar to previous AGC test trains, HDG-1 includes the major graphite grades (IG-110, NBG-17, NBG-18, PCEA, and 2114) as well as adding the very fine-grain grade IG-430 which is of interest to the Molten Salt Reactor (MSR) designs. Also new to the HDG-1 capsule are 90 smaller geometry specimens designated as pencil specimens. These specimens take up only one third the space of a standard creep size specimen. This increased number of specimens will enhance property measurement statistics because they will provide 3 times the control specimen data at a position that would otherwise only have a single measurement.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of mesopores in superfine grain graphite neutron-irradiated at high fluence
Microstructural changes induced by neutron irradiation of superfine grain graphite G347A (Tokai Carbon, Japan) were examined by nitrogen adsorption at 77K and by three microscopy techniques (SEM, TEM and FIB-SEM tomography). The specimens were irradiated at doses of up to 30 dpa, covering stages before and after the turnaround fluence at three temperatures (300, 450, 750 °C) of their irradiation envelope. The initial graphite densification at low fluences did not produce any detectable effect in the pore size range (<350 nm) measured by gas adsorption. However, graphite irradiated at high fluences, after turnaround, showed severe structural changes. At all three temperatures and high irradiation fluences, gas adsorption revealed significant increase of the volume of narrow mesopores (<5–20nm) and up to five times increase of BET surface area, both in linear relationship with the relative volume expansion. Analysis of microscopy images showed multiplication of fine macropores (>50 nm) at high irradiation fluences and more structural changes on multiple scales, from nanometers to microns. This work demonstrates the unique ability of gas adsorption techniques to analyze open pores with sizes between sub-nanometer and sub-micron in bulk nuclear graphite, with supporting microscopy results.
I. Contescu, Cristian↗
Nuclear rockets based on graphite reactor technology - Their applications and development status.
Nuclear rockets based on graphite reactor technology noting applications and development status
Finger, H. B.↗
Nuclear rockets based on graphite reactor technology - Their applications and development status
Development and mission applications of nuclear rockets based on graphite reactor technology
MISSION PLANNING↗
AGC-2 Graphite Preirradiation Data Analysis Report
This report describes the specimen loading order and documents all preirradiation examination material property measurement data for graphite specimens contained within the Second Advanced Graphite Capsule (AGC 2) irradiation capsule. The AGC 2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens in order to garner quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear grade graphites. This testing will ascertain the in service behavior of the graphite for pebble bed and prismatic very high temperature reactor designs. Similar to the First Advanced Graphite Capsule (AGC 1) preirradiation examination report, material property tests were conducted on specimens from 18 nuclear grade graphite types. However, AGC 2 tested an increased number of specimens (i.e., 512) prior to loading them into the AGC 2 irradiation assembly. All AGC 2 specimen testing was conducted at Idaho National Laboratory from July 2009 to August 2010. This report also details the specimen loading methodology for graphite specimens inside the AGC 2 irradiation capsule. The AGC 2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid plane. This provides similar specimens with and without an applied load. Analysis in this document utilizes the neutron flux profile calculated for the AGC 2 capsule design, the capsule dimensions, and the size (i.e., length) of the selected graphite specimens to create a stacking order that produces “matched pairs” of graphite specimens above and below the AGC 2 capsule elevation mid point, thus providing specimens with similar neutron dose levels.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The 13C nuclear magnetic resonance in graphite intercalation compounds
The (13)C NMR chemical shifts of graphite intercalation compounds were calculated. For acceptor types, the shifts come mainly from the paramagnetic (Ramsey) intra-atomic terms. They are related to the gross features of the two-dimensional band structures. The calculated anisotropy is about -140 ppm and is independent of the finer details such as charge transfer. For donor types, the carbon 2p pi orbitals are spin-polarized because of mixing with metal conduction electrons, thus there is an additional dipolar contribution which may be correlated with the electronic specific heat. The general agreement with experimental data is satisfactory.
Tsang, T.↗
C-13 nuclear magnetic resonance in graphite intercalation compounds
The C-13 NMR chemical shifts of graphite intercalation compounds have been calculated. For acceptor types, the shifts come mainly from the paramagnetic (Ramsey) intra-atomic terms. They are related to the gross features of the two-dimensional band structures. The calculated anisotropy is about - 140 ppm and is independent of the finer details such as charge transfer. For donor types, the carbon 2p pi orbitals are spin-polarized because of mixing with metal-conduction electrons, thus there is an additional dipolar contribution which may be correlated with the electronic specific heat. The general agreement with experimental data is satisfactory.
Tsang, T.↗
HDG-1 Fiber Bragg grating data analysis
The main goal of the High dose graphite 1 Advanced test reactor experiment was to study nuclear grade graphite at high fluences. Additional supplementary optical fiber instrumentation was added to this long duration experiment for instrumentation development purposes. The supplementary instrumentation consisted of two pure silica core, fluorine doped cladding optical fibers each etched with 9 fiber Bragg gratings, one fiber being heat treated for 9 hours at 750 C and 16 hours at 750 C, the other being heat treated for 24 hours at 550 C and 48 hours at 650 C. Fiber Bragg gratings are known to have issues of measurement drift when in high temperature and high radiation environments like what is encountered in the Advanced test reactor. At the culmination of this experiment, the optical fibers saw ~1.3E21 n/cm2 total fluence, which is at the highest fluences that fiber Bragg gratings have been studied to date. Reported here is the analysis of this data including radiation induced shift and changes in sensitivity.
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
Reliability Assessment of Graphite Specimens under Multiaxial Stresses
An investigation was conducted to predict the failure strength response of IG-100 nuclear grade graphite exposed to multiaxial stresses. As part of this effort, a review of failure criteria accounting for the stochastic strength response is provided. The experimental work was performed in the early 1990s at the Oak Ridge National Laboratory (ORNL) on hollow graphite tubes under the action of axial tensile loading and internal pressurization. As part of the investigation, finite-element analysis (FEA) was performed and compared with results of FEA from the original ORNL report. The new analysis generally compared well with the original analysis, although some discrepancies in the location of peak stresses was noted. The Ceramics Analysis and Reliability Evaluation of Structures Life prediction code (CARES/Life) was used with the FEA results to predict the quadrants I (tensile-tensile) and quadrant IV (compression-tension) strength response of the graphite tubes for the principle of independent action (PIA), the Weibull normal stress averaging (NSA), and the Batdorf multiaxial failure theories. The CARES/Life reliability analysis showed that all three failure theories gave similar results in quadrant I but that in quadrant IV, the PIA and Weibull normal stress-averaging theories were not conservative, whereas the Batdorf theory was able to correlate well with experimental results. The conclusion of the study was that the Batdorf theory should generally be used to predict the reliability response of graphite and brittle materials in multiaxial loading situations.
Sookdeo, Steven↗
Program For Optimization Of Nuclear Rocket Engines
NOP is versatile digital-computer program devoloped for parametric analysis of beryllium-reflected, graphite-moderated nuclear rocket engines. Facilitates analysis of performance of engine with respect to such considerations as specific impulse, engine power, type of engine cycle, and engine-design constraints arising from complications of fuel loading and internal gradients of temperature. Predicts minimum weight for specified performance.
Plebuch, R. K.↗
HDG-1 Experiment Irradiation Monitoring Data Qualification Final Report
SUMMARY The U.S. Department of Energy (DOE) Advanced Reactor Technologies (ART) Graphite Research and Development (GRD) Program is conducting a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. This report documents the qualification of irradiation monitoring data for the fifth experiment, High Dose Graphite-1 (HDG-1). Qualified monitoring data are required by the ART program to support the design and licensing of the first high-temperature reactor (HTR) nuclear plant. Data are classified as Qualified if they meet the usage requirements described in the experiment planning and quality assurance (QA) documents, Failed if they do not meet those requirements and provide no usable information, or Trend if they do not fully meet all requirements but still provide useful information subject to an assessment of how any deficiencies may affect a particular use of the data. HDG-1 irradiation began with Advanced Test Reactor (ATR) Cycle 168B on August 24, 2020, and concluded after Cycle 173C on January 27, 2025. The HDG-1 capsule was removed from the reactor core twice—during core internal change (CIC) Cycle 170A and powered axial locator mechanism (PALM) Cycle 172A—to prevent overheating of the graphite specimens during high-power PALM cycles. The capsule was therefore irradiated during a total of seven normal ATR cycles: 168B, 169A, 171A, 171B, 173A, 173B, and 173C. Irradiation monitoring data evaluated in this report include thermocouple (TC) temperature, gas flow rate, gas moisture, gas pressure, specimen load, and graphite stack displacement. Temperature. A total of 14,508,065 TC temperature records were captured. Of these, 13,901,785 (95.8%) are Qualified and 606,280 (4.2%) are Failed. The principal source of failed temperature data was the instrument failure of TC-9 (Zone 2) on June 24, 2024, and TC-10 (Zone 1) on July 5, 2024, near the end of Cycle 173A, which resulted in 595,554 Failed readings. An additional 379 missing values and 10,347 slightly negative values from TC-13 during ATR outages are also Failed. Neither TC-9 nor TC-10 was used as a temperature-control TC, and their failures did not compromise capsule condition monitoring. Correlation analysis of all 13 TCs found no evidence of virtual junction formation. Control chart analysis revealed clear downward drift of approximately 80°C for TC-6 (Zone 3) relative to other stable TCs, and possible downward drift of approximately 60°C for TC-13 relative to the Zone 5 control TC (TC-1), though TC-13 remained consistent with the Zone 2 control TC (TC-12). Gas flow. A total of 20,088,090 gas flow rate records were captured. Of these, 19,941,463 (99.3%) are Qualified and 146,627 (0.7%) are Failed due to missing values. All argon, helium, and total gas flow data were within expected ranges throughout the irradiation. Gas moisture. A total of 1,116,005 outlet gas moisture values were captured. Of these, 1,101,421 (98.7%) are Qualified and 14,584 (1.3%) are Failed, comprising 14,556 out-of-range values and 28 missing values. The out-of-range moisture values exceeded 22,000 ppmv for approximately 1 week at the beginning of Cycle 173A, when accumulated moisture evaporated after the capsule was retrieved from water storage during PALM Cycle 172A and reinserted into the east flux trap. Moisture levels returned to below 25 ppmv for the remaining three cycles, and the transient high-moisture event did not affect the integrity of specimen irradiation. Gas pressure. A total of 7,812,035 gas pressure values were captured. Of these, 6,642,048 (85.0%) are Qualified and 1,169,987 (15.0%) outlet pressure values are Failed, comprising 718,537 zero outlet pressure values due to sensor failure from Cycle 168B through Cycle 171B, 54,550 missing values, and 396,900 too-low outlet pressure values, ranging from 1.1 to 1.6 psia after sensor replacement during Cycle 173A. Load. A total of 6,696,030 load values were captured. Of these, 6,694,580 (99.98%) are Qualified and 1,450 (0.02%) are Failed due to missing values. Applied loads to the six specimen stacks were stable throughout the irradiation. Stack displacement. A total of 6,696,030 displacement values were captured. Of these, 5,713,297 (85.32%) are Qualified and 3,781 (0.06%) are Failed due to missing values. Stack displacement increased consistently throughout the irradiation, reaching approximately 3.08 in. for Channels 5 and 6 by the end of irradiation. 978,952 (14.62%) substantially elevated displacements observed for Channel 6 beginning in Cycle 171A and for Channel 5 beginning in Cycle 173A are assigned Trend status. Raising pressure. A total of 1,115,999 raising pressure values were captured. Of these, 1,115,430 (99.95%) are Qualified and 569 (0.05%) are Failed due to missing values. Ram pressure. A total of 6,696,030 ram pressure values were captured. Of these, 6,692,249 (99.95%) are Qualified and 3,484 (0.05%) are Failed due to missing values. Stack raising was perf
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Summary of Graphite Data Stored within NDMAS
The Graphite Technology Development Project provides data to support the design of graphite core components within specific reactor service conditions of the next generation of high-temperature, gas-cooled nuclear reactors. Physical, mechanical, and thermal properties of nuclear grade graphite were characterized for specimens that were unirradiated, irradiated, and irradiated under various stress conditions. The material properties include diameter, length, mass, density, compressive strength, tensile strength, flexural strength, modulus, resistivity, thermal diffusivity, and thermal expansion coefficient. Baseline graphite specimens are unirradiated from different grades (2114, IG-110, NBG-17, NBG-18, and PCEA) and different types used in different characterization tests (compressive, flexural, tensile, one-inch cylinder, and quarter-inch cylinder). The Advanced Graphite Creep (AGC) irradiation specimens are from a much larger number of grades and cylinder types (creep, piggyback, and pencil). For baseline graphite, characterization data for 7,756 specimens extracted from thirteen graphite billets were captured to the NDMAS database. For AGC experiments, four irradiation campaigns have been completed: AGC-1, AGC-2, AGC-3, and AGC-4. The ongoing HDG-1 (High-Dose Graphite) experiment, which began irradiation with Cycle 168B on August 26, 2020, includes specimens previously irradiated in AGC-2 in addition to the specimens originally destined for AGC-5.. Besides the characterization data, the AGC data includes irradiation monitoring and physics data representing the irradiation conditions of AGC specimens. Currently, all data for AGC-1, AGC-2, and AGC-3 have been captured to the NDMAS database. Only AGC-4 pre-irradiation and irradiation monitoring data have been captured, and HDG-1 pre-irradiation data are in process of being captured for unirradiated specimens. To date, a total of 38,149 material property records have been captured into NDMAS database for baseline and AGC specimens. All characterization data are qualified for use according to their perspective data verification reports. For the AGC irradiation campaigns, a total of 41,502 qualified physics calculation records were added for AGC-1, AGC-2, and AGC-3. Finally, a total of 173,698,505 AGC irradiation monitoring data records (thermocouple temperature, gas flow rate, gas pressure, gas moisture, applied load, and specimen displacement) have been captured to NDMAS; the majority of those records (~94%) are qualified data records.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Microstructural Observations on Advanced Multilayer Protective Coatings for Nuclear Thermal Propulsion
The present research discusses the microstructures of a new multilayered coating concept proposed to solve the “midrange corrosion” problem observed in NbC and ZrC-coated graphite (Gr) nuclear fuels during the NERVA/Rover programs. The concept envisions designing a compliant multilayered coatings architecture to accommodate the thermal expansion mismatch strains between the ZrC outer coating and the Gr/(U,Zr)C fuel matrix along with using a diffusion barrier to minimize carbon and hydrogen diffusion, respectively. Multilayered coatings were deposited on Gr substrates by chemical vapor deposition (CVD). Graphite disks, as well as the inner channels of a 19-hole hexagonal rod 152.4 mm long, were coated by CVD in proof-of concept process demonstration studies. Coated disk specimens were thermally-cycled between ambient and 1900 K for as long as 20 cycles with a 1h hold time at the temperature using both stepped and non-interrupted thermal cycling methods. Detailed cross-sectional microstructural analyses were conducted on the thermally-cycled coated disks. These results showed that the coatings were intact after 20 thermal cycles at 1900 K although the bonding of the overlay Mo coating appeared to be influenced by the oxygen content in the layer. Seven transverse sections were cut along the length from the coated hexagonal rod for microstructural studies. All the nineteen channels were observed to have been coated in all the sections. This paper reports the results of detailed microstructural analyses, compositional line scans and x-ray dot maps conducted on three coated channels from a mid-section of the rod. The measured values of carbon (C) were observed to be exceptionally high in all the specimens, which were attributed to hydrocarbon contamination of the specimen surfaces. It was observed that the oxygen (O) levels were high in all the coating layers even after approximately correcting for the observed high values of C. The source of O contamination of the CVD reactor system was attributed to either the presence of oxychloride impurities in the precursor salts or a pin hole leak in the coating system. It is concluded that high values of O result in poor bonding of the Mo overlay to the ZrC layer. These trials have established that it is possible to deposit multilayered coatings by CVD in 19-hole hexagonal rods 152.4 mm long similar to the NERVA fuel rod design although the quality of the coatings and the uniformity of their thicknesses require further improvement.
Nuclear thermal propulsion, multilayer coatings, m↗