Graphite Creep Research Articles

Graphite creep negation during flash spark plasma sintering under temperatures close to 2000 °C

Graphite creep has high importance for applications using high pressures (100 MPa) and temperatures close to 2000 °C. In particular, the new flash spark plasma sintering process (FSPS) is highly sensitive to graphite creep when applied to ultra-high temperature materials such as silicon carbide. In this flash process taking only a few seconds, the graphite tooling reaches temperatures higher than 2000 °C resulting in its irreversible deformation. The graphite tooling creep prevents the flash spark plasma sintering process from progressing further. In this study, a finite element model is used to determine FSPS tooling temperatures. In this context, we explore the graphite creep onset for temperatures above 2000 °C and for high pressures. Knowing the graphite high temperature limit, we modify the FSPS process so that the sintering occurs outside the graphite creep range of temperatures/pressures. 95% dense silicon carbide compacts are obtained in about 30 s using the optimized FSPS.

Historical experiment to measure irradiation-induced creep of graphite

This paper presents historical results of graphite irradiation-induced creep experiments that were performed at Oak Ridge National Laboratory from the 1950's to the 1970's. These experiments were performed at temperatures from 150°C to 1000 °C, and bend stresses ranging from 500 to 5000 psi (∼3.3–34.5 MPa). The experimental setup utilized in-situ measurement of specimen displacement, on-line applied stress control, and the ability to change stress during the experiment. The different stress conditions showed that the primary creep strain and the steady-state creep rates both have a linear stress dependence. The temperature range used in this work resulted in trends that have not be previously presented in the literature: 1) a linear dependence of primary creep strain on temperature, and 2) the shape of steady state creep rate versus temperature (see graphical abstract). The maximum dose in the specimens was 0.9 dpa, which is sufficient to achieve steady-state creep without the structural changes that alter the observed creep behavior. The results from this experiment provide evidence that dispels that the pinning-unpinning model describes the mechanism of irradiation creep in graphite. Instead these results suggest a dislocation climb mechanism is the probable mechanism for creep within the crystalline regions.

Creep of graphite blocks under irradiation in RBMK reactors

Computer simulation of the stress–strain state of graphite blocks in prolonged thermal and neutron irradiation is considered, in a three-dimensional formulation. Allowance is made for the anisotropy of the graphite’s thermophysical properties, as well as crack initiation and growth. Creep of the material is taken into account, in the absence of plastic deformation, and its influence on the properties of the graphite blocks is assessed.

Surface morphology and microstructure evolution of IG-110 graphite after xenon ion irradiation and subsequent annealing

IG-110 graphite samples were polished and irradiated with Xe ions at various fluences, then annealed at high temperatures up to 1100 °C. After irradiation, small hills were found on the polished surfaces, indicating an anisotropic swelling induced by irradiation. Around 30% swelling at a fluence of 2 × 1015 ions/cm2 was characterized using atomic force microscopy. Severe swelling of the graphite crystallites caused stresses between adjacent crystallites, but leaved no intergranular cracks on the polished surface, which was ascribed to irradiation-induced creep of graphite. The pore morphology was affected by the anisotropic swelling. We found many contracted pores but only one expanded pore, indicating a decreased porosity induced by irradiation. After annealing at 1100 °C, TEM characterization showed clearly increased lattice order and decreased width of the (002) diffraction arc, indicating the annihilation of dislocations and recovery of basal plane rotations. Annealing-induced recrystallization of damaged graphite led to recovery of the crystallites' swelling and many small cracks appearing on the samples' surfaces.

Modeling irradiation creep of graphite using rate theory

We have examined irradiation induced creep of graphite in the framework of transition state rate theory. Experimental data for two grades of nuclear graphite (H-337 and AGOT) have been analyzed to determine the stress exponent (n) and activation energy (Q) for plastic flow under irradiation. We show that the mean activation energy lies between 0.14 and 0.32 eV with a mean stress-exponent of 1.0 ± 0.2. A stress exponent of unity and the unusually low activation energies strongly indicate a diffusive defect transport mechanism for neutron doses in the range of 3–4 × 1022 n/cm2.

Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

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 three different compressive loads applied to the top half of three diametrically opposite pairs 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.

Significance of primary irradiation creep in graphite

Traditionally primary irradiation creep is introduced into graphite analysis by applying the appropriate amount of creep strain to the model at the initial time-step. This is valid for graphite components that are subjected to high fast neutron flux fields and constant stress fields, but it does not allow for the effect of movement of stress locations around a graphite component during life, nor does it allow primary creep to be applied rate-dependently to graphite components subject to lower fast neutron flux.This paper shows that a differential form of primary irradiation creep in graphite combined with the secondary creep formulation proposed by Kennedy et al. performs well when predicting creep behaviour in experimental samples.The significance of primary irradiation creep in particular in regions with lower flux is investigated. It is shown that in low flux regions with a realistic operating lifetime primary irradiation creep is significant and is larger than secondary irradiation creep.

Irradiation induced creep of graphite

The status of graphite irradiation induced creep strain prediction is reviewed and major creep models are described. The ability of the models to quantitatively predict the irradiation induced creep strain of graphite is reported. Potential mechanisms of in-crystal creep are reviewed as are mechanisms of pore generation under stress. The case for further experimental work is made and the need for improved creep models across multi-scales is highlighted.

Investigation of radiation creep of reactor graphite

The results of investigations of the radiation creep of GR-280 graphite under a high compression load (about 15 MPa) after irradiation in a BOR-60 reactor at 520°C to fast-neutron fluence 1.2·10 22 cm −2 are presented. It is shown that the fluence dependence of the creep deformation, calculated using the standard relation as the difference of the change in the dimensions of loaded and control samples, is anomalous. The linear thermal expansion coefficients of loaded and control samples are found as functions of the neutron fluence under the same conditions. It is noted that the linear thermal expansion coefficient of the samples irradiated under a load is much higher than that of the control samples. Simmons’ theorem is used to take account of the effect of a load on the linear thermal expansion coefficient, and the dimensional changes of graphite exposed to radiation and the dependence of the true creep deformation on the neutron fluence are calculated. It is shown that these dependences are close to linear in the experimental fluence range (0.4‐1.2)·10 22 cm −2 . Creep is an important characteristic of graphite, determining its behavior in a reactor. It gives rise to relaxation of the stresses in the graphite components of the structure, and in most cases it has a favorable effect on the negative processes occurring in the graphite elements under neutron irradiation. The creep rate is a complicated function of the irradiation conditions and the changes of the physicomechanical properties under irradiation. The study of creep under reactor irradiation is complicated by the fact that the deformation of the loaded samples is determined by the creep as well as the dimensional changes as a result of the radiation evolution of the structure, which depends strongly on the physical properties of the graphite, first and foremost, the linear thermal expansion coefficient and the elastic modulus. Irradiation Conditions and Materials. The samples were fabricated from GR-280 graphite. They consisted of 8 mm in diameter and 30 mm long cylinders cut along and across the direction of extrusion of the blocks [1]. A setup consisting of 32 control samples on eight stages (four per stage) and 12 samples under an axial compression load was developed to irradiate the loaded samples. The stress in the loaded samples was 15 MPa on average. The samples were irradiated in the core of the BOR-60 reactor at 520°C in the neutron-fluence range (0.4‐1.2)·10 22 cm ‐2 (here and below the fluence is given for neutrons with energy >0.18 MeV). Silicon carbide monitors were used to determine the temperature of the samples to within ±25°C in the course of irradiation [2], and the fast-neutron fluence was determined according to the induced activity of 54 Fe and 93 Nb monitors [3].

Computational investigation of the stress-strain state of graphite blocks in light of substantiation of RBMK service-life extension

To increase the service life of RBMK-1000 power-generating units to 45 years, it is necessary to substantiate lifetime of the reactor structures operating under irradiation. Specifically, computational work is needed to substantiate the strength and lifetime characteristics. The characteristic features of each type of reactor, and not only the general behavior of graphite under irradiation, are important for making more accurate predictions. It is desirable to rely on the finite-element model with three-dimensional finite elements, since graphite possesses anisotropic properties in different directions. This article describes a method and program, developed on the basis of the finite-element method, that takes account of the nonuniformity of the neutron field and temperature distributions over the cross-section of a graphite block and the nonuniform anisotropic shrinkage (swelling) and creep of graphite.

The analysis of irradiation creep experiments on nuclear reactor graphite

The phenomenon of irradiation creep in graphite is essential to the design and construction of graphite-moderated reactors. Recent experimental results have shown large apparent reductions in creep rate for creep strains greater than ±1%. It is shown that this is not a true reduction in creep rate, but is an artifact resulting from changes in the properties of the stressed samples, which modify their dimensional changes under irradiation compared to the unstressed control samples and this is not accounted for in the standard definition of creep strain.

Irradiation creep in graphite—some new considerations and observations

Irradiation creep in reactor graphite is normally taken to be linear visco-elastic in character, but modified by any structural changes in the graphite, such as those due to radiolytic oxidation or dimensional changes. Corrections for these changes are made to the creep rate using the well established correlations of creep rate with the appropriate elastic compliance. Measurements of the changes in thermal-expansion coefficient with creep strain show that small creep strains perturb the graphite structure significantly, and the standard definition of creep strain is not strictly correct. In this situation, corrections to the creep rate by measurements on unstressed control samples become inaccurate with increasing creep strain. New measurements are reported of thermal-expansion coefficient changes due to creep strain, both parallel and perpendicular to the applied stress, using samples from the earlier creep studies in the BR-2 reactor. The effect of these studies on calculations of reactor-brick internal stresses is considered.

295. Out-of-pile bench test of a 1200°c graphite creep irradiation test

295. Out-of-pile bench test of a 1200°c graphite creep irradiation test

UKAEA Reactor Group studies of irradiation-induced creep in graphite

A review is presented of the experimental and theoretical studies of fast neutron irradiation creep in reactor graphite carried out by the UKAEA Reactor Group. The studies have covered the effect of varying graphite type, oxidation, stress level and boron doping. The results are shown to accord better with a dislocation pinning-unpinning model of dislocation glide, rather than the Cottrell model often assumed.

Radiation creep of graphite. An introduction

Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted.

Uni axial tensile graphite creep capsules with continuous strain registration

Two irradiation devices are described for the in pile measurement of tensile irradiation creep of graphite. In each machine a single sample is maintained under a controlled load by a pneumatic bellows system. Irradiation creep is measured relative to unstressed reference shells which surround the stressed sample. This differential strain is detected by linear displacement transducers, and recorded automatically by the out of pile installation. Irradiation temperatures are in the 800 to 1100°C range, and the stresses up to 60% of the U.T.S. One machine has been specifically designed for a flux change experiment, other irradiation parameters remaining fixed. Temperature control is achieved through varying gas mixtures in control gas gaps. The paper details the design principles of the machines and gives an account of the cold and hot commissioning tests, with particular reference to the accuracy of the in pile measuring system. Finally, the early irradiation experience is evaluated.

Graphite high temperature creep rigs

A description is given of two high temperature tensile creep rigs, for irradiating pyrocarbon and graphite specimens. PIRITHOOS is a creep rig operating at 1100°C and utilizing three in line pyrocarbon specimens. These have different cross sections giving three stress values. Unstressed specimens are placed close to the tensile ones. Dimensional measurements: length, thickness, width are made in hot cells, after each reactor shut down. FLACH is a graphite creep rig allowing continuous length measurement to be made in comparison with the length of two reference specimens. These rigs consist of two main parts: the creep capsule including: specimens, loading bellows, microwave measuring apparatus, which is introduced into a standard gas gap furnace regulating the temperature by gas mixture and electrical heating.

Reloadable graphite creep facility for tensile or compressive experiments

An irradiation test consisting of four experimental periods was prepared to evaluate the creep characteristics at elevated temperatures and to high fast neutron fluences of a special core graphite for the High Temperature Reactor. The material is a cylindrical stressed sample which is surrounded by a number of unstressed half-shell reference pieces. All the samples are irradiated at about 900°C in four consecutive steps leading to a total fast fiuence of 1.0, 2.0, 3.5, and 5.5 × 10 21 n · cm −2 (DNE). Irradiations are carried out in core positions of the High Flux Reactor of the Euratom Joint Research Centre at Petten/The Netherlands. To meet the experimental requirements a Graphite Creep Facility was designed and manufactured which enables the creep samples to be subjected to either tensile or compressive loads during the irradiation merely by changing the specimen can assembly. The load control modes can be either (a) steady tensile or compressive, or (b) continuous cycling at predetermined rates. The facility is re-loadable out of the thimble and the specimen can may be removed during reactor shut-down to permit intermediate dimensional measurements or replacement of the specimen. The paper describes special design features of the Creep Facility e.g. (a) specimen stressing mechanism, (2) load control system, (3) reloadability, and concludes with a description of the measuring apparatus which detects the dimensional changes of the stressed samples and the unstressed reference pieces. Experimental results of the first irradiation step are presented.

Design, operation, and initial results from a series of graphite creep irradiation experiments

A series of irradiation tests was designed to evaluate the creep characteristics at elevated temperatures and to high fast fluences of various graphites of interest to HTGR designers. The series encompasses the irradiation of 28 specimens, each 15.24 mm (0.6 in.) diam × 25.4 mm (1 in.) long, at 900°C to incremental exposures of 1,2,4, and 8 × 10 21 neutrons/cm 2 ( E > 0.18 MeV ); 28 similar specimens at 600°C to the same exposures; and 28 other similar specimens at 1250°C under the same conditions. A compressive stress of 13.79 MPa (2000 psi) is applied to 20 of the specimens in each test by means of a metal bellows expanded by gas pressure against the specimen columns. Eight of the stacked specimens in each test are stressed to 20.68 MPa (3000 psi) by a reduction in diameter. Special features of the capsules are described which include (1) movable center-line thermocouples which measure the temperature profile along the axis of the capsule, (2) linear variable differential transformer type load cells to monitor the applied load, and (3) computerized temperature control designed to achieve accurate longitudinal temperatures over the 0.508 m (20 in.) length of the test specimen columns. The operational characteristics of the first capsule, which is currently being irradiated and is one of a total of twelve in the series to be irradiated over a four-year period, are presented. The performance of the Oak Ridge Research Reactor Data Acquisition and Control System (ORRDACS) is briefly described.

Fast neutron induced creep in pyrocarbon under constant stress

Abstract Pyrocarbon is used as a coating material in the fuel of high-temperature nuclear reactors, and a thorough understanding of its irradiation behaviour includes a knowledge of its ability to creep under fast neutron irradiation. An experiment is described which demonstrates fast neutron-induced creep of a pyrolytic carbon under constant applied stress. This differs from previous work which has obtained creep ductility data from restrained shrinkage tests. The specimens were centre-loaded discs freely supported at the rim, thus subjected to a constant biaxial bend stress. On each specimen, elastic and plastic strains were produced and measured using the same geometry and loading arrangement, to allow the creep strain to be expressed simply in terms of initial elastic strain units. Results were obtained on specimens of initial density 1.95 g/cm and 1.64 g/cm 3 up to a fast neutron dose of 4 × 10 20 n/cm 2 (DNE) at a temperature of 1000°C. The low-density specimens showed both the greater shrinkage and the greater creep strain, and average creep rates were 0.5 and 1.0 elastic units per 10 20 n/cm 2 (DNE) for the high and low-density specimens respectively. These constant-stress creep results are shown to be consistent with other data on pyrocarbon. They differ from graphite creep data in that the two pyrocarbons give creep strains per unit initial elastic strain which depend on their initial densities.