Fission-enhanced self-diffusion of uranium in UO2 and UC

In-pile self-diffusion measurements in stoichiometric UO₂ sinters and single crystals and in arc-cast stoichiometric UC have been performed using the thin layer condition and ²³³U as tracer. The nominal irradiation temperature was 900°C. The resulting diffusion coefficients $\text{D1}$ of $\text{1.5 × 10}{}^{\mathrm{?16}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UO₂ and $\text{2.2 × 10}{}^{\mathrm{?17}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UC for a fission rate $\text{S}$ of 1 × 10¹³f/cm³ · sec represent radiation enhanced diffusion and are higher by factors of 10³ to 10⁴ than (extrapolated) coefficients of thermal diffusion. The data are of immediate relevance for understanding and predicting such important quantities as in-pile sintering and densification, diffusion controlled creep and fission gas behavior in the outer zones of the fuel. They are at the upper limit of expected values.     

Neutron damage and fission product gas release in the initial burst region of uranium monocarbides

Thermal neutron damage and fission product gas (¹³³ Xe) release in a burst region of uranium monocarbides were studied. After neutron irradiation, the electrical resistivity was measured from room temperature to 800° C. Three recovery stages were revealed in the resistivity of UC irradiated to $\text{4.0 × 10}{}^{16}$ nvt. The lattice parameter of UC with the same irradiation also showed three stages of recovery up to 1050°C. The initial burst of Xe from UC was studied in a dose range between $\text{1.6 × 10}{}^{15}$ and $\text{2.9 × 10}{}^{18}$ nvt. The burst occurred in three steps for lightly irradiated specimens, while there were two steps of the burst in heavily irradiated specimens. The activation energies for each burst step were calculated. From the results obtained here, we concluded that the burst was correlated with the recovery of damage in the neutron-irradiated UC.     

Heterodiffusion de l'or,de l'argent et du cuivre dans les phases cubiques δ et ε du plutonium

The solute diffusion at infinite dilution of ¹⁹⁸Au and ^110mAg in cubic phases of Pu has been studied using the serial sectroning method. The solute diffusion coefficients in the b.c.c. ? phase can be expressed by: $\text{D}{}^{\mathrm{Au}}{}_{\mathrm{?Pu}}\text{= 5,7 × 10}{}^{\mathrm{?5}}$ exp(?10300/RT) cm²/s and $\text{D}{}^{\mathrm{Ag}}{}_{\mathrm{?Pu}}\text{= 4,9 × 10}{}^{\mathrm{?5}}\text{exp(?9600/RT) cm}{}^2\text{/s}$. The solute diffusion mechanism is interstitial of the dissociative type in both cases. These experiments confirm the activated interstitial model which has been proposed for self diffusion of ?Pu. Indeed the solute diffusion coefficients of Au and Ag are near of the self diffusion coefficients of Pu. The mechanisms are therefore interstitial in both cases. In the f.c.c. δ phase of Pu where self diffusion takes place by a vacancy mechanism, the solute diffusion coefficients of Au and Ag are near of the self diffusion coefficients of δ Pu. Solute diffusion takes place also by a vacancy mechanism. On the other hand, the extrapolation at infinite dilution of experiments of solute diffusion of Cu in ?Pu (Matano-Wagner coupling) gives the following results: $\text{D}{}^{\mathrm{Cu}}{}_{\mathrm{?Pu}}\text{= 1 × 10}{}^{\mathrm{?3}}\text{exp(?12300/RT)}$ cm²/s. The solute diffusion mechanism is interstitial of the dissociative type. In the ? phase the smaller the atomic radius the faster the migration: r_Co < r_Cu < r_?Pu < r_Ag = r_Au, and D^Co_?Pu > D^Cu_?Pu >D^Pu_?PU > D^Ag_?Pu ≈ D^Au_?Pu.     

Grain-boundary diffusion in uranium dioxide: The correlation between sintering and creep and a reinterpretation of creep mechanism

Previously published data on the final stage sintering kinetics of stoichiometric uranium dioxide are correlated with a reinterpretation of low-stress creep behaviour of identical material (data on both processes by the present authors). For both processes the rate-controlling diffusional flux is considered to be that of uranium ions along grain boundaries. The effective diffusion coefficient for uranium ion diffusion along grain boundaries, $\text{D}{}_{\mathrm{GB}}$, is estimated to be: $\text{D}{}_{\mathrm{GB}}\text{= 1.38(}{}_{\mathrm{÷}}{}^{\mathrm{x}}\text{5) × 10}{}^{\mathrm{?6}}\text{exp ? [(2.39 ± 0.8) × 10}{}^5\text{/8.31T] m}{}^2\text{/s}$. Comparisons are made between this value and those previously measured by radio-tracer methods.     

Density,thermal expansion of stainless steel and interfacial properties of UO2-stainless steel above 1690 K

The sessile drop method was used for the determination of the density in liquid state. The results for stainless steel 1.4970 using uranium dioxide as substrate material in the temperature range 1690 K (liquidus temperature $\text{T}{}_1$) $\text{< T < 2120 K}$ are $\text{ρ = 6.82 × 10}{}^3\text{? 10.25 × 10}{}^{\mathrm{?1}}\text{(T ? T}{}_1\text{) kg/m}{}^3$, and $\text{α = 1.50 × 10}{}^{\mathrm{?4}}\text{K}{}^{\mathrm{?1}}$. Below 1690 K the linear thermal expansion is given by $\text{Δl/l}{}_0\text{= 0.002}{}_{04}\text{+ 7.11}{}_0\text{× 10}{}^{\mathrm{?6}}\text{T + 7.734 × 10}{}^{\mathrm{?9}}\text{T}{}^2$. Using the same method but not correlated with the density measurements the following interfacial properties of the system UO₂-stainless steel have been determined: surface energy of liquid steel $\text{γLv = 1.19 ? 0.57 × 10}{}^{\mathrm{?3}}\text{(T ? T}{}_1\text{) J/m}{}^2$ and interfacial energy of liquid steel against UO₂$\text{γSL = 1.57 ? 2.01 × 10}{}^{\mathrm{?3}}\text{(T ? T}{}_1\text{) J/m}{}^2$, the results yield a contact angle $\text{θ = 0°}$ at $\text{T= 2515 K}$. Using literature data for the compressibility of liquid UO₂, an estimate of the surface energy of UO₂ in liquid state was performed. The estimated value at the melting point is: $\text{γLV = 0.522 J/m}{}^2$. The mean value of the experimental data given by several authors is $\text{0.513 ± 0.085 J/m}{}^2$. The estimated temperature dependence of the surface energy of liquid UO₂ is given by $\text{dγLV/dT = ?0.19 × 10}{}^{\mathrm{?3}}\text{J/m}{}^2$.     

Thermally activated deformation of Zircaloy-4

Tensile tests were carried out on Zircaloy-4 over the temperature range 298–798 K. Yield stress values at the strain rates $\text{1.33 × 10}{}^{\mathrm{?4}}\text{s}{}^{\mathrm{?1}}$ and $\text{6.67 × 10}{}^{\mathrm{?4}}\text{s}{}^{\mathrm{?1}}$ were used to determine the activation parameters. A peak in activation volume ($\text{V}{}_{\mathrm{app}}\text{= 3100 b}{}^3$) was observed at about 690 K; outside this temperature range the activation volumes became almost independent of temperature ($\text{V}{}_{\mathrm{app}}\text{= 200?300 b}{}^3$). The peak in activation volume was explained in terms of a basic rate controlling mechanism and dynamic strain aging. This analysis indicated that the peak could be ascribed to the negative value of the strain rate sensitive solute strengthening term $\text{M}$ and that the mechanism based on the non-conservative motion of jogs appeared to be more favored as the basic rate controlling mechanism of Zircaloy-4 than an impurity mechanism     

Neutron sputtering of gold

The sputtering yield of gold bombarded with fission neutrons was studied in the CP-5 reactor. The experiment was designed to investigate sputtering patterns in addition to the sputtering yield. The investigation was carried out on bulk single crystals and near liquid helium temperature. Overall sputtering yields ranged from $\text{1 × 10}{}^{\mathrm{?3}}\text{to}\text{6 × 10}{}^{\mathrm{?3}}$ sputtered gold atoms per incident neutron for doses ranging from $\text{2.1 × 10}{}^{17}\text{to}\text{5.5 × 10}{}^{17}\text{(}\text{nvt > 0.1 MeV}\text{)}$. These results are roughly an order of magnitude larger than many previously reported sputtering yields. Sputtering patterns have not been pursued yet, due to prohibitively high background radiation.     

The oxygen snoek damping and an abnormal peak of neutron-irradiated vanadium

Measurements of low-frequency internal friction and electron microscope observations were made on neutron-irradiated vanadium with various oxygen contents. Irradiation was carried out at about 60°C to a fast fluence of $\text{2 × 10}{}^{17}$ or $\text{5 × 10}{}^{19}$ n/cm² ($\text{E ? 1 MeV}$). The oxygen Snoek damping was decreased by irradiation and post-irradiation annealing below 200 or 250° C, while it began to recover by annealing above this temperature. Complete recovery was attained by 30 min anneal at 450°C in the case of the lower fluence, whereas in the other case it was not observed after the same treatment. The results of electron microscope observations were consistent with those of internal friction measurements. The specimens irradiated to $\text{5 × 10}{}^{19}$ n/cm² showed an abnormal peak after annealing above 250°C near the nitrogen Snoek temperature. The height of this peak, $\text{P}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$, was expressed as $\text{P}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$ ∝ exp ($\text{2.72 × 10}{}^3\text{/RT}$) $\text{Q}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$, where $\text{Q}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$ the heiβht of the oxygen Snoek damping after each annealing. The mechanism for radiation-anneal hardening and the abnormal peak were considered in the light of these experiments.     

Diffusion of carbon in zirconium and some of its alloys

Diffusion of carbon in zirconium, zircaloy-2 and Zr- 2.5% Nb has been studied in the temperature range 873–1523K for zirconium and zircaloy-2 and 753–1523K for Zr-2.5% Nb alloy, using the residual activity technique. The diffusivities (in m²/s) in the α and β phases could be represented by $\text{D}{}_{\mathrm{C/\alpha -Zr}}\text{(873–1123K) = (2.00 ± 0.37) × 10}{}^{\mathrm{?7}}\text{exp [}\text{?(151.59 ± 2.51)}\text{RT}\text{]}$$\text{D}{}_{\mathrm{C/\alpha -Zircaloy-2}}\text{(873–1043K) = (1.41 ± 0.32) × 10}{}^{\mathrm{?7}}\text{exp [}\text{?(158.99 ± 3.14)}\text{RT}\text{]}$$\text{D}{}_{\mathrm{C/\alpha -Zr-Nb-alloy}}\text{(753–873K) = (4.68 ± 0.88) × 10}{}^{\mathrm{?7}}\text{exp [}\text{?(159.98 ± 2.91)}\text{RT}\text{]}$$\text{D}{}_{\mathrm{C/\beta \; Zr}}\text{((1143–1523K) = (8.90 ± 1.60) × 10}{}^{\mathrm{?6}}\text{exp [}\text{?(133.05 ± 1.46)}\text{RT}\text{]}$$\text{D}{}_{\mathrm{C/\beta \; Zircaloy-2}}\text{(1263–1523K) = (2.45 ± 0.61) × 10}{}^{\mathrm{?5}}\text{exp [}\text{?(150.29 ± 1.72)}\text{RT}\text{]}$$\text{D}{}_{\mathrm{C/\beta \; Zr-Nb\; alloy}}\text{(1143–1523) = (1.70 ± 0.42) × 10}{}^{\mathrm{?5}}\text{exp [}\text{?(158.20 ± 2.09)}\text{RT}\text{]}$The activation energies are given in kJ/mole. In the phase transition region, the diffusivities could be represented by the empirical relation: D = D_α^Cα · D_β^Cβ, where C_α, C_β are the concentrations of the two phases in the alloy and D_α, D_β are the extrapolated values of diffusion co-efficients in the α and β phases respectively.The results have been explained in terms of the interstitial mechanism of diffusion.     

Silver diffusion in zirconium

The diffusion coefficients of Ag in both α and β phases of Zr are reported. The temperature dependence of the diffusion coefficients may be expressed by and and is strongly dependent on the phase transition. The general diffusion behaviour is analyzed in terms of the Hägg and the Anthony and Turnbull rules, indicating that Ag diffuses as substitutional in Zr.     

Steady-state creep of zircaloy-4 fuel cladding from 940 to 1873 K

Steady-state creep rates of as-received zircaloy-4 fuel cladding have been determined from 940 to 1073 K in the α-Zr range, from 1140 to 1190 K in the mixed (α + β) phase region and from 1273 to 1873 K in the β-Zr phase region. Strain rates of between 10^?6 and 10^?2/s were determined under constant uniaxial load conditions. Assuming that creep rates can be described by a power law-Arrhenius equation, the creep rate for α-phase zircaloy-4 is given by: $\text{ge}{}_{\mathrm{ss}}\text{?}\text{= 2000 σ}{}^{5.32}\text{exp}\text{(?284 600/kT) s}{}^{\mathrm{?1}}$; for the β-phase zircaloy-4 by: $\text{ge}{}_{\mathrm{ss}}\text{?}\text{= 8.1 σ}{}^{3.79}\text{exp}\text{(?142 300/kT) s}{}^{\mathrm{?1}}$; and for the mixed (α + β) phase of zircaloy-4 (for creep rates ?3 × 10^?3 s^?1) by: $\text{ge}{}_{\mathrm{ss}}\text{?}\text{= 6.8 × 10}{}^{\mathrm{?3}}\text{σ}{}^{1.8}\text{exp}\text{(?56 600/kT) s}{}^{\mathrm{?1}}$. For the both the α and β phases, the activation energies for creep are in agreement with those of self-diffusion. For the mixed (α + β) phase region, the low creep rate range is controlled by grain boundary sliding at the α/(α + β) phase boundary.     

Contribution a l'etude des cinetiques de l'oxydation a l'air du hafnium

The oxidation kinetics in air of Van Arkel hafnium were studied between 750 and 950°C. Comparison of the results obtained by thermogravimetry and microhardness measurements has allowed us to characterise the overall oxidation kinetics which are of parabolic type and also the two separate kinetics, likewise parabolic, of growth of the oxide film and of dissolution of oxygen in the metal underlying the oxide. For each of these three kinetics the rate constants are respectively given by the following relations: $\text{K}{}_{\mathrm{p}}\text{= 3.1 × 10}{}^8\text{exp (?46 000/RT)}$, $\text{K}{}_1\text{= 1.4 × 10}{}^6\text{exp (?36 000/RT)}$, $\text{K}{}_2\text{= 3.4 × 10}{}^{10}\text{exp (?59 000/RT)}$.These results are in agreement with a classical diffusion model of oxygen, both in the oxide and in the metal, set up on the basis of Wagner's general theory. The activation energy of diffusion of oxygen in hafnium is $\text{53.0 ± 2}$ kcal/mole.     

Transport properties of U0.7Ce0.3O2 − x

We have determined a number of transport properties of $\text{U}{}_{0.7}\text{Ce}{}_{0.3}\text{O}{}_{\mathrm{2-x}}$ at 1273 K for various deviations from stoichiometry and compared them with available results on $\text{(UPu)O}{}_{\mathrm{2\; ?\; x}}$. They are: the electrical conductivity, Seebeck coefficient, effective charge number and chemical diffusion coefficient.A very characteristic behaviour is observed for the electronic properties of $\text{(UCe)O}{}_{\mathrm{2\; ?\; x}}$. A p-type conduction for all the studied deviations from stoichiometry (up to $\text{x = 3 × 10}{}^{\mathrm{?2}}$) is interpreted in terms of a high electronic disorder in the stoichiometric compound. Electronic disorder at stoichiometry is probably less important in $\text{(UPu)O}{}_{\mathrm{2\; ?\; x}}$, which presents a sharp p-n transition at $\text{x = 5 × 10}{}^{\mathrm{?3}}$.Ionic transport properties obtained on $\text{(UCe)O}{}_{\mathrm{2\; ?\; x}}$ indicate an approximate proportionality between the ionic conductivity resulting from oxygen ions transport and the deviation from stoichiometry. Results available on $\text{(UPu)O}{}_{\mathrm{2\; ?\; x}}$ do not appear to be compatible with ours and some arguments are presented which cast doubt on their validity.     

Irradiation induced creep of ceramic nuclear fuels

This contribution gives a review of the experimental results and accompanying theoretical considerations. The mechanisms considered for irradiation creep are: relaxation of elastic stresses by fission spikes, promotion of dislocation slide by thermal spikes, preferential, stress-orientated nucleation of dislocation loops and preferential growth of dislocation loops. A survey over the irradiation creep rates attributed to steady-state creep shows $\text{ε}{}_{\mathrm{irr}}\text{~ σ · F}$ for oxide fuel in the stress and fission rate ranges of $\text{σ = 10–50}\text{MN/m}{}^2$ and $\text{F = 3 × 10}{}^{12}\text{–1 × 10}{}^{14}\text{f/cm}{}^3\text{· s}$ at burnups < 3%. There seems to be a continuous increase of the irradiation creep rate of oxide fuels with increasing temperature. However, that increase cannot be directly interpreted through a thermally activated process. It seems that the irradiation creep rate will also depend on fuel porosity, on plutonium distribution in mechanically blended UO₂-PuO₂, but not substantially on the plutonium content per se. Some results were already given for carbide and nitride fuels, which show the irradiation creep rate to be lower by about a factor of 10 than for oxide fuel under comparable conditions. Primary irradiation creep has been observed up to $\text{(3–5) × 10}{}^{19}\text{f/cm}{}^3$ and could prevail up to $\text{1 × 10}{}^{20}\text{f/cm}{}^3$.     

Uranium self-diffusion in stoichiometric uranium monocarbide

Uranium-233 self-diffusion was studied in stoichiometric uranium monocarbide; the diffusion coefficients can be represented by the expression $\text{D}{}^1{}_{\mathrm{U}}\text{= (2.58}{}^{\mathrm{+4.54}}{}_{\mathrm{?1.64}}\text{) × 10}{}^3\text{exp(?}\text{157000 ± 5000}\text{RT}\text{)cm}{}^2\text{/s}$ in the temperature range from 1900 to 2260°C. Self-diffusion of uranium and carbon, in stoichiometric and hypostoichiometric uranium carbides occurs by a vacancy mechanism. It is shown that the kinetics and mechanism are similar to those for self-diffusion in the group IVb and Vb transition element monocarbides; for these B1 (fcc) type carbides in general, the activation energy for diffusion is empirically correlated to the congruent melting temperature: for carbon $\text{ΔH}{}^{\mathrm{C}}{}_{\mathrm{m}}\text{～- 31 T}{}_{\mathrm{m}}$ and for the metal, $\text{(ΔH}{}^{\mathrm{M}}{}_{\mathrm{f}}\text{+ ΔH}{}^{\mathrm{M}}{}_{\mathrm{m}}\text{) ～-54 T}{}_{\mathrm{m}}$. In addition, the activation entropies correlate well to the temperature derivative of the shear modulus; this behaviour is consistent with the familiar Zener model. Microstructural effects have probably been the principal underlying cause of the preponderant discordance in the results of other investigators.     

Mass transport of chromium and nickel in monel-400

Diffusion of⁵¹Cr and ⁶³Ni in Monel-400 has been studied in the temperature ranges 1023–1573 K for volume diffusion and 713–1123 K for grain boundary diffusion, respectively. The volume diffusion coefficients can be represented by: ${}^{\mathrm{D}}\text{Cr/Monel-400 =}{}^{\mathrm{(0.31\; \pm \; 0.08)}}\text{× 10}{}^{\mathrm{?4}}$ exp[$\text{-(249.2 ± 2.5 kJ/mole)/RT]m}{}^2$. ${}^{\mathrm{D}}\text{Ni/Monel-400 =}{}^{\mathrm{(0.68\; \pm \; 0.16)}}\text{× 10}{}^{\mathrm{?4}}$ exp[$\text{-(261.0 ± 4.1 kJ/mole)/RT]m}{}^2$. Grain boundary diffusivities evaluated by Whipple and Suzuoka's method were found to be in good agreement but Fisher's analysis yielded values of diffusion coefficients that were 2 to 4 times lower. The grain boundary diffusion coefficients evaluated by Whipple and Suzuoka's models can be represented by: ${}^{\mathrm{Dg}}\text{Cr/Monel-400 =}{}^{\mathrm{1.8\; \times \; 10}}{}^{\mathrm{?5}}$ expl[$\text{-(1.47.8 kJ/mole)/RT]m}{}^2$. ${}^{\mathrm{Dg}}\text{Ni/Monel-400 =}{}^{\mathrm{1.8\; \times \; 10}}{}^{\mathrm{?5}}$ exp[$\text{-(156.7kJ/mole)/RT]m}{}^2$. Segregation of ⁵¹Cr and⁶³Ni has been studied by autoradiography. It was observed that at lower temperatures (≤900 K) material transport in specimens with 3&#x0303;00 μm grain diameter is mainly through the grain boundaries.     

Beryllium diffusion in zirconium

The diffusion coefficients of ⁷Be in both α and β phases of Zr, are reported. The temperature dependence of the diffusion coefficient in the α phase may be expressed by $\text{D}{}^{\mathrm{Be}}{}_{\mathrm{\alpha -Zr}}\text{= 0.33 exp(?31900/RT) cm}{}^2\text{/s}$. The measured values in the β phase are in agreement with previously literature reported data, which give a temperature dependence expressed by $\text{D}{}^{\mathrm{Be}}{}_{\mathrm{\beta -Zr}}\text{= 8.33 × 10}{}^{\mathrm{?2}}\text{exp(?31800/RT) cm}{}^2\text{/s}$. Be diffusion in Zr, which is consistent with an interstitial-like behavior, is analyzed in terms of the Anthony and Turnbull conditions, and atomic size criteria. It is concluded that the latter is a very important parameter when assessing the possibility of significant interstitial-like dissolution.     

Fluage du pyrocarbone sous irradiation

Three massive samples of pyrocarbon were irradiated at 1100°C for a maximum fast-neutron dose of $\text{1.6 × 10}{}^{21}$ DNE. They were subjected to stresses in the range $\text{1.33 × 10}{}^2\text{–2 × 10}{}^2$ Kg/cm². The pyrocarbon was deposited from methane in a rotating furnace. Its density, its isotropy, its structure according to X-rays and TEM relate closely to its homologue deposited from methane in fluidised conditions. A study of creep under irradiation showed that a brief stage of primary creep is followed by a stage which is linear with respect to both stress and fast-neutron dose. Creep is thus well represented by an expression of the form $\text{? = Kσφ}$, where $\text{K}$ is $\text{2 × 10}{}^{\mathrm{?25}}$ (Kg · cm^?2. DNE)^?1, which is a value ten times greater than previously estimated. Irradiation is accompanied by densification, a slight increase in anisotropy and a reduction in $\text{L}{}_{\mathrm{c}}$ (apparent crystallite size measured along the $\text{c}$ axis). The variation of these parameters with dose does not, however, differ appreciably between the three creep samples and the unstressed sample.     

Emissivity of zirconium alloys in air in the temperature range 100–400°c

A technique is described for measuring the emissivity of materials in air. This technique was used to determine both the normal total emissivity ($\text{?}{}_{\mathrm{nt}}$) and the total hemispherical emissivity ($\text{?}{}_{\mathrm{ht}}$) of both unoxidized and oxidized samples of Zircaloy-2, Zr/2.5wt.% Nb and other zirconium alloys in the temperature range 100–400°C. Temperature had a negligible effect on the values obtained. However, $\text{?}{}_{\mathrm{nt}}$ values increased from $\text{0.158 ± 0.025}$ for the range of unoxidized zirconium alloys examined to $\text{0.6 ± 0.08}$ for the alloys when they were oxidized to have a $\text{2.0 × 10}{}^{\mathrm{?6}}\text{m}$ thick oxide.     

Consideration of the simulation of in-pile radial temperature profiles in mixed-oxide fuel pins

Temperature profiles similar to those existing in fuel rods under irradiation have been simulated by passing electrical current through cylindrical pellets. The comparison between calculated and measured temperatures in pellets heated in a thermal gradient shows: (1) The values of thermal conductivity obtained by different authors in isothermal experiments and extrapolated to temperatures up to 2700°C are not in agreement. Therefore, the calculation of the temperature of the fuel leads to errors which vary between 1 and 16% depending on the data for λ used. For central temperatures above 1900°C the values of Schmidt better suit the calculations, especially if the oxygen content of the fuel is smaller than 2.00. (2) The published data of electrical conductivity are in open disagreement. The values of activation energy are generally higher than those deduced from the present investigation. It has been assumed that the activation energy $\text{E}$ in the equation $\text{σ = A exp(?E(T)/kT)}$ varies with temperature as $\text{E(T)= E}{}_0\text{(1?1.94 × 10}{}^{\mathrm{?4}}\text{T)}$ when $\text{E}{}_0\text{= 0.58}$ eV if O/M = 2.00, and $\text{E(T) = E}{}_0\text{(1?2.21× 10}{}^{\mathrm{?4}}\text{T)}$ when $\text{E}{}_0\text{=0.91}$ eV if O/M = 1.94.