Athanasia Tzelepi

Polycrystalline graphite has a unique combination of high-temperature properties that has made it the material of choice for many industrial applications. Several nuclear reactor designs that operate between 500°C and 1,000°C include graphite components. These components must maintain their integrity even at the 1,800°C they could be exposed to during an accident. The operational behavior of these graphites during both proof testing of as-manufactured material and postirradiation examination must be determined by measuring physical, mechanical, and thermal properties. For reasons of expense and practicality the properties are measured in (or near to) ambient conditions. It is essential that the measured properties may be extrapolated reliably to high temperatures. Laboratory testing at elevated temperatures therefore provides data for (1) defining temperature-dependent extrapolation curves, (2) informing conceptual models that help to establish confidence in ambient-temperature test methods, and (3) inputs into numerical simulations of operating conditions. The properties of interest for this paper are selected on the basis of current ASTM standards to include those most relevant to current and future fission reactor operation. The effects of fast neutron irradiation on the high-temperature behavior are presented in general terms, and the conventional understanding of the mechanisms behind both the inert and irradiated behavior are outlined. Areas for further research are then highlighted, the findings of which would support design, qualification, operation, and safety monitoring of graphite-moderated nuclear reactors.

Postirradiation examination (PIE) of graphite samples trepanned from the UK reactor cores has been carried out for more than 50 years. Due to the nature of the material, there are sample size and geometry restrictions and no standard test methods to cover measurements on this material. Nevertheless, these measurements are used to support the continued operation of the UK reactor cores, and hence a large program of trials is carried out to provide confidence that each method is accurate and reproducible. These trials typically involve a study of size effects using virgin graphite and simulant materials for the irradiated and oxidized graphite, but a corresponding study with irradiated samples is usually not possible. This paper combines the work of two UK studies to investigate the size effect of the PIE test methods on irradiated graphite on the basis of the characterization of graphite used in the advanced gas-cooled reactors (AGRs) and Magnox reactors. The AGR study focused on the static Young's modulus, three-point bend strength of unnotched and notched beams, and the work of fracture. The Magnox study focused on the coefficient of thermal expansion, diametral compression, and flexural strength. The two studies used large irradiated graphite samples from “installed sets” (i.e., precharacterized graphite samples installed in the reactor cores before the start of operation for monitoring purposes). Large Magnox samples that were trepanned from the reactor core after shutdown were also tested. The purpose of these investigations was to relate the graphite measurements normally undertaken on small trepanned samples to property values obtained using standard test methods on irradiated material. The sample selection was such that it covered as wide a range of dose and radiolytic weight loss as possible. This paper outlines the methodology, results, and conclusions for each of the studies and provides some guidelines for similar studies on new graphites.

Coefficient of thermal expansion (CTE) gives a key insight into the thermomechanical properties of a material, representing the dimensional expansion of a material for a given temperature change. It is of particular relevance to graphite components within reactors, providing insight into the thermal strains experienced at operating temperatures. In combination with other internal strains and external loads, these thermal strains can have a direct effect on component performance, as well as operational lifetimes. Hence, a good understanding of how CTE changes with irradiation is essential for any graphite reactor component in both existing and newly built reactors. This understanding relies on accurate measurements of small specimens with noncompliant geometries. The most commonly used method of measuring CTE is push-rod dilatometry, favored for its simplicity, and this technique is covered by ASTM E228. This and other CTE measurement standards place specific size requirements on samples, which can be difficult to achieve for irradiated graphite samples recovered from reactors or irradiated in material test reactors. The National Nuclear Laboratory (NNL) has used this technique to routinely measure the CTE of small samples trepanned from Magnox and advanced gas-cooled reactor (AGR) graphite over the past three decades as part of the graphite core monitoring campaigns. The samples measured at the NNL are not compliant with the geometry requirements mandated by ASTM E228. This paper examines two studies on nonstandard compliant geometry AGR samples. The first is a comparison between dilatometry and electronic speckle pattern interferometry, and the second is a dilatometry study conducted on samples of similar material but differing geometries. The results are used to discuss typical challenges and aspects of the measurements that require addressing to demonstrate confidence in dilatometry and optical interferometry CTE measurements on noncompliant geometries.

Gas permeability describes the rate at which gases can pass through a material under pressure-driven flow. It is a property of interest for graphite components within gas-cooled reactors, indicating the level of permeation of cooling gases within the components and hence the rate of reaction between coolant gases and the graphite. These gas-graphite reactions are responsible for radiolytic oxidation in reactors in the United Kingdom and have the potential for thermal oxidation in high-temperature reactors (HTRs), which can limit reactor lifetimes. It is also important for understanding vulnerabilities in reactors (e.g., the extent of thermal oxidation in HTRs during air-ingress fault scenarios). At the time of writing no specific ASTM standard exists for the measurement of permeability or diffusivity on graphite, but ASTM C781, Standard Practices for Testing Graphite Materials for Gas-Cooled Nuclear Reactor Components, recommends using ASTM C577, Standard Test Method for Permeability of Refractories, with a transport gas of helium for the measurement of permeability. The National Nuclear Laboratory (NNL) has conducted permeability measurements on graphite samples trepanned from both advanced gas-cooled reactor (AGR) and Magnox cores over the past three decades as part of reactor monitoring campaigns. This paper reviews permeability measurements on irradiated and virgin AGR graphite conducted using the “vacuum leak” methodology and a novel method currently in development at the NNL that does not impose the same stringent sealing requirements on samples. The results are used to examine the size dependence of samples, virgin material variation, and the evolution of permeability with radiolytic oxidation.

One of the key issues that must be addressed when a sampling and measurement program underwrites the performance of full-size graphite components is the understanding of the influence of material type and specimen size on the measured property. Because of sampling constraints, specimen sizes (geometry and/or volume) may not be compliant with measurement standards and this issue is investigated in the context of graphite strength. Theoretical considerations on stress distribution, together with a review of experimental work based upon Gilsocarbon nuclear graphite IM1-24, are presented for the measurement of strength by three-point bend. A broader historical review of the effects of specimen size on strength, covering studies published in the open literature and hitherto restricted studies published within the U.K. nuclear industry, is also presented, together with some considerations of three-point bend measurement consistency across multiple facilities and some basic trending of three-point bend data. Whereas the findings of this experimental work apply specifically to the grade of graphite selected for investigation, the methods employed in the study provide a model for general application. Furthermore, the review of the wider literature gives some indication of the sensitivity of flexural strength to specimen geometry, volume, and graphite grade.

The ASTM Standard C769-09 covers a procedure for measuring the sonic velocity in manufactured carbon and graphite which can then be used to obtain Young’s modulus. There are four main assumptions in this standard that require further investigation (i) homogeneity and isotropy of the tested specimens, (ii) transducer frequency that is sufficiently high to provide the required level of timing accuracy yet not affected by frequency dependent attenuation in graphite, (iii) specimen lateral dimensions much larger than the wavelength of the transmitted pulse so that dispersion and hence, distortion of the propagated pulse is minimised and, (iv). specimen lateral dimensions much larger than the wavelength of the transmitted pulse so that Young’s modulus can be calculated from the sonic velocity. This paper presents the experimental and theoretical work undertaken to provide the technical basis or compensate for the above assumptions. Specifically, the experimental work investigated sample size (length and diameter), coupling, equipment, and signal analysis effects on measured velocity; the theoretical work investigated the relation between sonic velocity and Young’s modulus at intermediate sample diameters. The experimental results on as-manufactured gilsocarbon graphite samples show that the effect of the lateral dimension is within the stated accuracy and reproducibility limit of the technique. Regarding the sample length, the experimental results provide conflicting evidence and this suggests that the parameters of the experimental setup, e.g., transducer type and frequency and type of excitation pulse could affect the results. This finding is also confirmed by the theoretical investigation, which shows that, in certain cases, the distortion of the propagated pulse is severe enough that the apparent wave speed depends considerably on sample length and on the area of the sample surface in contact with the transducer.