Carbon Materials for Advanced Technologies Episode 13 ppsx

460 dependent on temperature and time. This effect has been accounted for through the concept of equivalent temperature. If two irrahations are performed at hfferent levels of fast neutron flux, 4, and a2, identical damage will be caused if the two irradiation temperatures are related by where k is Boltzmann's constant and A is an activation energy determined experimentally. Usually, one of the flux levels would pertain to a standard position in a materials test reactor. As discussed by Burchell [58] experimental evidence suggests that flux level or rate effects are significant only at low to moderate irradiation temperatures (<400"C). 3.2 Dimensional changes A principal result of carbon atom displacements is crystallite dimensional change. Interstitial defects will cause crystallite growth perpendicular to the layer planes (c- axis direction), whereas coalescence of vacancies will cause a shmkage parallel to the layer plane (a-axis direction). The damage mechanism and associated dimensional changes are illustrated in Fig. 6. Radiation-induced dimensional changes can be very large, exceedmg 60% in well-ordered graphite materials. Pryrolytic graphite has frequently been used to study the neutron irradiation- induced dimensional changes of the graphite crystallite [57,59]. Price [60] conducted such a study. Figure 7 shows Price's data for crystallite shmkage in the a-direction for neutron doses up to - 12 dpa. Price's samples were graphitized at a temperature of 2200-3300°C prior to being irradiated at 1300-1500°C. The a- axis shrinkage increased linearly with dose for all of the samples, but the magnitude of the shrinkage at any given dose decreased with increasing graphitization temperature. Similar trends were noted for the c-axis expansion. The effect of graphitization temperature on irrabtion-induced dimensional change accumulation can be attributed to thermally induced improvements in crystal perfection. Higher graphitization temperatures reduce the initial number of lattice defect sites which are available to trap irradiation-induced vacancies, and thus reduce the rate of damage accumulation. Polygranular graphites possess a polycrystalline structure, usually with significant texture resulting from the method of forming during manufacture. Consequently, structural and dimensional changes in polygranular graphites are a function of the crystallite dmensional changes and the graphite's texture. In polygranular graphite, thermal shrinkage cracks formed during manufacture and that are preferentially aligned in the crystallographic a-direction, initially accommodate the c-direction expansion, so mainly a-direction contraction will be observed. The 46 1 graphite thus undergoes a net volume shrinkage. With increasing neutron dose (displacements), the incompatibility of crystallite dimensional changes leads to the generation of new porosity oriented parallel to the basal planes, and the volume shrinkage rate falls, eventually reaching zero. The graphite then begins to swell at an increasing rate with increasing neutron dose because of the combined effect of c-axis growth and new porosity generation. The graphite thus undergoes a volume change "turnaround" into net growth which continues until the generation of cracks and pores in the graphite, due to differential crystal strain, eventually causes total disintegration of the graphite. COLLAPSW VACANCY VACANCY EXPANSION Fig. 6. Radiation damage in graphite showing the induced crystal dimensional strains. Impinging fast neutrons displace carbon atoms from their equilibrium lattice positions, producing an interstitial and vacancy. The coalescence of vacancies causes contraction in the a-direction, whereas interstitials may coalesce to form dislocation loops (essentially new graphite planes) causing c-direction expansion. Fig. 7. High-temperature neutron irradiation a-axis shrinkage behavior of pyrolytic graphite showing the effects of graphitization temperature on the magnitude of the dimensional changes [60]. 462 Irradiation-induced dimensional damage data for GraphNOL N3M are shown in Fig. 8. N3M is a molded graphite and thus the filler coke particles are preferentially aligned in the radial direction. Consequently, the crystallographic a- direction is predominantly aligned in the radial direction (perpendicular to forming) direction. Therefore, the a-direction irradiation-induced shmkage is more apparent in the radial direction, as indicated by the radial data (both 600 and 875°C) in Fig. 8. 1-0- 6009: RADIAL I -3.5 I I I I I I I 0 5 10 (5 20 25 30 35 40 FLUENCE (dpa) Fig. 8. Neutron irradiation induced dimensional changes for GraphNOL N3M graphite irradiated a 600 or 875 "C [6 11. Note that the radial dimensional changes exceed the axial changes due to textural effects. A general theory of dimensional change in graphite due to Simmons [62] has been extended by Brocklehurst and Kelly [ 171. A detailed account of the treatment of dimensional changes in graphite can be found in Kelly and Burchell's analysis of H-451 graphite irradiation behavior [63]. 3.3 Stored energy The irradiation induced displacement processes previously described can cause an excess of energy (associated with the vacancylinterstitial pairs) in the graphite crystallites. The release of this stored energy (or Wigner energy, after the physicist who fist postulated its existence [21]) was historically the first problem of radiation damage in graphite to manifest itself. When an interstitial carbon atom and lattice vacancy recombine, their excess energy is given up. If sufficient damage has accumulated in the graphite, the release of this stored energy can result in a rapid rise in temperature. Stored energy accumulation was found to be 463 0.6 particularly problematic in the early (air-cooled) graphite moderated reactors, which operated at relatively low temperatures. Figure 9 shows the rate of release of stored energy with temperature, as a function of temperature, for graphite samples irradiated at 30°C to three different doses (0.01, 0.1, and 0.6 dpa) in the Hanford K reactor. The release curves are characterized by a peak occurring at -200°C which is associated with the recombination of single interstitials and vacancies. With increasing neutron dose, the 200°C peak becomes broader and the maximum release rate is reduced. The release rate exceeds the specific heat, thus under adiabatic conditions the graphite would rise sharply in temperature. For ambient temperature irradiations it was found that the stored energy could attain values up to 2720 J/g, which if released adiabatically would cause a temperature rise of some 1300°C [7]. The uncontrolled release of stored energy from graphite, causing a sharp rise in core temperature, was of great concern to the operators of the early air-cooled (low-temperature) graphite reactors. In order to limit the total amount of stored energy it became necessary to periodically anneal the graphite. The core temperature was raised sufficiently, by nuclear heating or inserted electrical heaters, to "trigger" the release of stored energy from the graphite. The release then self-propagated slowly through the core, raising the graphite moderator temperature and thus partially annealing the graphite core. It was during such a reactor anneal that the Windscale (U.K.) Reactor accident occurred in 1957 [24]. I I I I I I I I EXWSURES IN Mwd/At & dpabpprox) - i- a3 9 'O 02 u) 0. i 0 300 400 500 100 200 ANNEALING TEMPERATURE 0 Fig. 9. Stored energy release curves for CSF graphite irradiated at -30°C in the Hanford K reactor cooled test hole [64]. Note, the rate (with temperature) of stored energy release (JKgK) exceeds the specific heat and thus under adiabatic conditions self sustained heating will occur. 464 The accumulation of stored energy in a graphite is both dose and irradiation temperature dependent. With increasingly higher irradiation temperatures the total amount of stored energy and its peak rate of release diminish, such that above an irradiation temperature of -300°C stored energy ceases to be a problem. Excellent accounts of stored energy in graphite can be found elsewhere [7,62,64,65]. 3.4 Eflects on mechanical and physical properties The physical properties of carbon and graphite materials are drastically altered by irradiation damage. For example, low dose irradiation (<<1 dpa) can increase the strength of a graphite by up to 80% while simultaneously reducing the thermal conductivity by more than an order of magnitude. Graphite is a phonon conductor of heat. The temperature dependence of thermal conductivity is shown in Fig. 10 for various pyrolytic graphites in the unirradiated condition. The substantial improvements in thermal conductivity caused by thermal annealing, andor compression annealing, are attributable to increased crystal perfection and size of the regions of coherent ordering (crystallites). This minimizes the extent of phonon-defect scattering and results in a larger phonon mean free path. With increasing temperature the dominant phonon interaction becomes phonon-phonon scattering (Umklapp processes). Therefore, the observed reduction of thermal conductivity with increasing temperature, and the convergence of the curves in Fig. 10, are attributed to the dominant effect of Umklapp scattering in reducing phonon mean free path. 1200 -a- COMPRESSION ANNEALED E \ -*- ANNEALED 3000% u 3 900 E 700 800 > 3 0 J (r w I- 2 600 500 1400 I 300 200 II Ill I II 0 200 400 600 800 10oO12oO4~ 4W 4800 TEMPERATURE ("c) Fig. 10. The temperature dependence of thermal conductivity for pyrolytic graphite in three diffment conditions [66]. The reduction of thermal conductivity with increasing temperature is attributed to increasing Umklapp scattering of phonons. 465 The mechanism of thermal conductivity and the degradation of thermal conductivity have been extensively reviewed [57-591. The increase of thermal resistance due to irradiation damage has been ascribed to the formation of [67]: (i) submicroscopic interstitial clusters, containing 4 f 2 carbon atoms; (ii) vacant lattice sites, existing as singles, pairs, or small groups; and (iii) vacancy loops, which exist in the graphite crystal basal plane and are too small to have collapsed parallel to the hexagonal axis. The contributions of collapsed lines of vacant lattice sites and interstitial loops to the increased thermal resistance is negligible. The reduction in thermal conductivity due to irradiation damage is temperature and dose sensitive. At any irradiation temperature, the decreasing thermal conductivity will reach a "saturation limit." This limit is not exceeded until the graphite undergoes gross structural changes at very high doses. The "saturated" value of conductivity will be attained more rapidly, and will be lower, at lower irradiation temperatures. The effect of radiation damage on the thermal conductivity of carbon materials is discussed extensively here by Snead in his chapter on "Fusion Reactor Applications." In graphites, the neutron irradiation-induced degradation of thermal conductivity can be very large, particularly at low temperatures. Bell et al. [65] report that the room temperature thermal conductivity of PGA graphite (the Magnox core graphite) is reduced by more than a factor of 70 when irradiated at 155°C to a dose of -0.6 dpa. At an irradiation temperature of 355°C the thermal conductivity of PGA was reduced by less than a factor of 10 at doses twice that obtained at 155°C. Above 600°C the reduction of thermal conductivity is less significant. For example, Kelly [7] reports the degradation of PGA at a high temperature. At an irradiation temperature of 600°C and a dose of - 13 dpa, the thermal conductivity was only degraded by a factor of -6. Moreover, at a irradiation temperatures of 920°C and 1150°C the degradation was minimal (less than a factor of 4 at -7 dpa). The thermal expansion of polygranular graphites is controlled by the thermal closure of aligned internal porosity. Irradiation-induced changes in the pore structure (see earlier discussion of structural changes) can therefore be expected to modify the thermal expansion behavior of carbon materials. The behavior of GraphNOL N3M (Fig. 11) is typical of many fiie-textured graphites [61], which undergo an initial increase in the coefficient of thermal expansion followed by a steady reduction to a value less than half the unirradiated value of - 5 x 1 0-6 O C'. Similar behavior is reported by Kelly [7] for the AGR moderator graphite (grade IM1-24). The electrical resistivity of graphite will also be affected by radiation damage. The mean free path of the conduction electron in an unirradiated graphte is relatively large, being limited only by crystallite boundary scattering. Neutron irradiation introduces: (i) scattering centers, which reduce charge carrier mobility; (ii) electron traps, which decreases the charge carrier density; and (iii) additional spin 466 resonance. The net effect of these changes is to increase the electrical resistivity on irrahation, initially very rapidly, with little or no subsequent change to relatively high fluence [58,61]. A subsequent decrease at very high neutron doses may be attributed to structural degradation. Fig. 11. Neutron irradiation-induced changes in the coefficient of thermal expansion of GraphNOL N3M at irradiation temperatures of 600 and 875°C [61]. The mechanical properties of graphites are substantially altered by radiation damage. In the unirradiated condition, polygranular graphites behave in a brittle fashion and fail at relatively low strains. The stress-strain curve is non-hear, and the fracture process occurs via the formation of sub-critical cracks, which coalesce to produce a critical flaw [9,10]. When graphites are irradiated the stress-strain curves become more linear, the strain to failure is reduced, and the strength and elastic modulus increased. As shown in Fig. 12, there is a rapid rise in strength attributed to dislocation pinning at irradiation-induced lattice defect sites. This effect has largely saturated at doses >1 dpa. Above - 1 dpa a more gradual increase in strength occm due to structural changes within the graphite. For polygranular graphites the dose at which the maximum strength is attained loosely corresponds with the volume change turnaround dose, indicating the importance of pore generation in controlling the high-dose strength behavior. The strain behavior of polygranular graphite subjected to an externally applied load is largely controlled by shear of the component crystallites. As with strength, irradiation-induced changes in Young’s modulus are the combined result of in- crystallite effects, due to low fluence dislocation pinning, and superimposed 467 structural changes external to the crystallite. The effects of these two mechanisms are generally considered separable, and related by: (EEo)irradiated = (EEo)pinnmg (E’Eo)structwe (2) Where E& is the ratio of the irradiated to unirradiated elastic modulus. The dislocation pinning contribution to the modulus change is due to relatively mobile small defects and is thermally annealable at -2000°C. Figure 13 shows the irradiation-induced elastic modulus changes for GraphNOL N3M. The low dose change due to dislocation pinning (dashed line) saturates at a dose 4 dpa. 1 60 rn 6W’C 60 I 10 20 FLUENCE (dpa) Pig. 12. Neutron irradiation-induced strength changes for GraphNOL N3M temperatures of 600 and 875°C [61]. iotr I I I I I 5 40 45 20 25 30 0. FLUENCE (dpd at irradiation Fig. 13. Neutron irradiation-induced Young‘s modulus changes for GraphNOL N3M at irradiation temperatures of 600 and 875°C [61]. 468 The elastic modulus and strength are related by a Griffith theory type relationship. GE strength, u = (-)1'2 ITC (3) where G is the fracture toughness or strain energy release rate (J/m2)>, E is the elastic modulus (Pa), and c is the flaw size (m). Thus, irradiation-induced changes in u and E (in the absence of changes in [G/c]) should follow u Eln. High dose data reported recently by Ishiyama et al. [68] show significant deviation from this relationship for grade IG-110 graphite, indicating that changes in G and or c must occur. 3.5 Radiation creep Graphite will creep under neutron irradiation and stress at temperatures where thermal creep is normally negligible. The phenomenon of irrahation creep has been widely studied because of its significance to the operation of graphite moderated fission reactors. Indeed, if irradiation induced stresses in graphite moderators could not relax via radiation creep, rapid core disintegration would result. The observed creep strain has traditionally been separated into a primary reversible component (e,) and a secondary irreversible component (e2), both proportional to stress and to the appropriate unirradiated elastic compliance (inverse modulus) [69]. The total irradiation-induced creep strain (€3 is thus: Ec = 61 f E2 (4) or, Eo = (O/Eo)[l - exp(-by)] + (K/Eo)ay (5) where E,, the unirradiated Young's Modulus, b is a constant, y is the neutron dose, and K is the irradiation creep coefficient. Kelly [7] has reported that values of -4 x for b and 0.23 x lo-*' for K apply to U.K. data taken over a range of irradiation temperatures (300-650°C). At high fluences Eq. (5) must be modified to account for structural changes occurring in the graphite: where - is the initial secondary creep rate and S(y) is the "structure factor" normally deduced from Young's Modulus changes ascribed to structural effects l"d:.1, 469 (i.e., S(y) = (E/E,) where E is the Young's Modulus at fluence y and E ,is the Young's modulus after the initial increase due to dislocation pinning). Oh et al. [70] have reported the creep coeEcient of IG- 1 10 graphite and shown it to be reasonably linear with temperatures over the range 300-1400°C at low to moderate fluences (< 2 dpa). Kennedy et aZ. [71] have reported the irradiation creep rate of a German graphite in tension and compression for creep strains in excess of 3.5%. Their data show the creep rate decreasing at higher fluences (>6 dpa) where the creep strain exceeds - 1%. Kelly and Burchell [72] attempted to rationalize the disparity between Kennedy et aZ.'s data indicating a reducing creep rate and the more commonly reported constant creep rate. They concluded that the reported reduction in creep rate was not a true reduction, but rather an artifact of changes in the properties in the stressed sample which modified their dimensional change under irradiation compared to the unstressed control samples. Based upon the success of their analysis at linearizing creep rate data, Kelly and Burchell proposed a redefinition of irradiation creep strain as 'Ithe difference in dimensions between a stressed sample and a sample with the same properties as the stressed sample irradiated unstressed'' [72]. 4 Radiolytic Oxidation In reactor designs which utilize inert gas coolants (typically helium), the only process which alters the properties of the graphite is irradiation damage. However, in carbon dioxide-cooled reactors graphite properties are also changed by the process of raholytic oxidation. Complete reviews of radiolytic oxidation and its effects on graphite properties may be found in the literature [73-751. Here, radiolytic oxidation of graphte is briefly reviewed and its consequences for reactor design and operation discussed. 4. I ne mechanism of radioljtic oxidation The simplest description of the reaction responsible for the radiolytic oxidation of graphite is: CO, + radiation energy -+ C02* (activated state) CO,* + C (graphite) -+ 2 CO. In reality the situation is considerably more complicated. The exact nature of the activated stated (oxidizing species) has been the subject of intense study [73-751, but is now generally accepted to be the negatively charged ionC03- [73,75,76]. The oxidation reaction occurs at temperatures far below those at which thermal oxidation becomes significant and, although the reaction is slow, it can lead to significant mass loss from the moderator during its lifetime. The oxidation reaction [...]... therefore desirable, and assures control rod availability under all conceivable reactor conditions With this goal in mind, efforts have been directed in the U.S.A 1921 and Japan [93,94] toward the development of carbon- carbon (C/C) composite control rods A C/C composite material comprises a carbon or graphite matrix that has been reinforced with carbon or graphite fibers Multidirectionally reinforced... reactive carbon deposits which arise from the gas phase inhibition reaction discussed in Section 4.1 Therefore, it behooves the reactor operator to have a reliable assessment of the amount and distribution of the reactive carbon deposit in the reactor core 5 Other Applications of Carbon in Fission Reactors The overwhelming majority of carbon utilized in nuclear reactors is in the form of graphite for the... thus preferred over graphites for many critical applications, such as control rods 5.5 Carbon insulation materials Because of their low thermal conductivity, high temperature capability, low cost, and neutron tolerance, carbon materials make ideal thermal insulators in nuclear reactor environments For example, the HTTR currently under construction in Japan, uses a baked carbon material (Sigri, Germany... on carbon and graphite, Chem Phys Carbon, 1968,4, p 1 Kelly, B.T., Nuclear reactor moderator materials In Materials Science and Technology: Nuclear Materials, Part 1 (VCH Weinheim, 1994) pp 365-417 Heintz, E.A., Influence of coke structure on the properties of the carbon- graphite artefact, FUEL, 1985,64, 1192 1196 Tucker, M.O., Rose, A.P.G., and Burchell, T.D., The fracture of polygranulagraphites, Carbon, ... fuel matrix materials Once fabricated, the fuel particles are combined with a matrix material containing a pitch or resin binder, and graphite or carbon filler Fuel element designs usually fall into two categories, referred to as prismatic fuel elements or spherical fuel elements The former arrangement was used in the U.S.A for the Peach Bottom and Fort St.Vrain HTGRs [Fig 14(a)], and in Japan for the... 92 Strizak, J.P., Effects of oxidation on the strength of C/C composites for GT-MHR control rods In Proceedings of 22nd Biennial Con$ on Carbon Pub American Carbon Society, 1995, pp 760-761 93 Ishiyama, S and Eto, M., Recent R&D of C/C Composite Control Rod for HTGRs In Proceedings of 22nd Biennial Con$ on Carbon Pub American Carbon Society, 1995, pp 762-763 94 Eto, M., Ishiyama, S., and Ugachi, H.,... into eq 2 for the tensile stress, u* can be approximated by substituting u = uin horizontal plane, 0 = 0 where u, is the stress level required for inelastic deformation and microcrack formation In reality, the stress above the inelastic zone exceeding uin is redistributed giving a somewhat larger monotonic inelastic zone Hence, for monotonic loading, under conditions of small scale inelastic deformation,... ofthe Fifth Conference on Carbon, Vol 2, Pergamon Press, Oxford, 1963, pp 347 386 Cadwell, J.J., McEachem, D.W., Read, J.W., Simon, W.A and Walker, R.F., Operational testing highlights of Fort St Vrain In Proceeding of the Symposium on Gas-Cooled Reactors with Emphasis on Advanced Systems, Vol 1, IAEA-SM200158, IAEA, Vienna, 1976, pp 151 163 Walker, R.E and Johnston, T.A., Fort Saint Vrain nuclear power... applications of carbon are noteworthy, and are briefly discussed here 5.I Activated carbon Gaseous fission products are produced during reactor operation, notably iodlne (in elemental form and as methyl iodide), krypton, and xenon Accidental leakage of these gasses could occur from the reactor core or primary coolant circuit during operation Therefore, these gasses are trapped in activated carbon beds... distribution of pore sizes for grade H-451 graphite Moreover, a calibration exercise was performed to determine a single value of particle critical stress intensity factor for the Burchell model [3] Most recently, the model was successfully validated against experimental tensile strength data for several graphites of widely different texture [4,19] Tucker and McLachlan [20] reported the performance of a model . reactive carbon deposit in the reactor core. 5 Other Applications of Carbon in Fission Reactors The overwhelming majority of carbon utilized in nuclear reactors is in the form of graphite for. toward the development of carbon- carbon (C/C) composite control rods. A C/C composite material comprises a carbon or graphite matrix that has been reinforced with carbon or graphite fibers diffraction studies on carbon and graphite, Chem. Phys. Carbon, 1968,4, p. 1. Kelly, B.T., Nuclear reactor moderator materials. In Materials Science and Technology: Nuclear Materials, Part