Damage in neutron-irradiated molybdenum: (I). Characterization of as-irradiated microstructure

The as-irradiated microstructure of molybdenum, irradiated in the EBR-II reactor at six different temperatures in the range 430–1000°C (0.24–0.44 $\text{T}{}_{\mathrm{m}}$) to a fast neutron fluence of ≈ 1 × 10²² n · cm^?2 ($\text{E > 1 MeV}$), has been characterized as black spot clusters, loops, rafts, voids (random and ordered) and dislocations. Present results show that both the void density, $\text{N}$_v and the void size, $\text{d}{}_{\mathrm{v}}$, are independent of irradiation temperature in the range 430–700° C. Above the 700° C irradiation temperature the void density decreases and the void size increases exponentially with increasing irradiation temperature and they have been expressed empirically as $\text{N}{}_{\mathrm{<mtext>v</mtext>}}\text{= 3.6 × 10}{}^{20}\text{exp}\text{(?26.9 T/T}{}_{\mathrm{<mtext>m</mtext>}}\text{)}$, $\text{d}{}_{\mathrm{<mtext>v</mtext>}}\text{= 1.5}\text{exp}\text{(9.44 T/T}{}_{\mathrm{<mtext>m</mtext>}}\text{)}$, where $\text{T/T}{}_{\mathrm{<mtext>m</mtext>}}$ is the irradiation temperature presented as a fraction of the melting point. The void density of all available published data has been used to show that the void density is (a) a strong function of irradiation temperature for a constant number of displacements per atom (dpa) and (b) a function of reactor power and spectrum when normalized to dpa.     

High-temperature specific heat of UO2 ThO2, and A12O3

The high-temperature specific heat of solid UO₂, ThO₂, and Al₂O₃ can be represented by an equation of the form $\text{C}{}_{\mathrm{p(s)}}\text{= 3}{}_{\mathrm{n}}\text{RF(?}{}_{\mathrm{D}}\text{/T) + dT}{}^3$, (1) where ?_D is the Debye temperature, F(?_D/T) is the Debye function, $\text{d}$ represents contributions of the anharmonic vibrations within the lattice, and $\text{n}$ denotes the number of atoms per molecule. In the liquid the corresponding equation is $\text{C}{}_{\mathrm{p(1)}}\text{= 3}{}_{\mathrm{n}}\text{RF(?}{}_{\mathrm{D}}\text{/T) +}\text{h}\text{T}{}^2$, (2) where h is the anharmonic term. It is shown that for Al₂O₃ and UO₂, where experimental data for the liquid phase are also available, $\text{d}\text{h}$ has the same value, Indicating that both materials behave identically. If we compare the thermodynamic relationship $\text{C}{}_{\mathrm{p}}\text{? C}{}_{\mathrm{v}}\text{= Vα}{}^2\text{KT}$, (3) where V is the volume, $\text{α}$ the volume expansion coefficient, and $\text{K}$ the bulk modulus, with equation (1), It follows that d must be equal to $\text{Vα}{}^2\text{K}\text{T}{}^2$; the value of $\text{Vα}{}^2\text{K}\text{T}{}^2$ is calculated in the temperature region where d was obtained; within experimental error they are equal.     

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 $\text{dγLV/dT = ?0.19 × 10}{}^{\mathrm{?3}}\text{J/m}{}^2$.     

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.     

The diffusion of fission products in thorium metal

A study was performed of the diffusion in α-thorium of fission products representing impurity atoms with a diversity of size and valance differences with respect to the solvent lattice. The atoms were recoil injected into thorium disks. Diffusion coefficients were determined for ¹³³Xe by monitoring its release during annealing, and for the other isotopes by post-annealing concentration profile analysis. The Arrhenius constants $\text{d}{}_{\mathrm{o}}\text{(cm}{}^2\text{/sec)}$, resp. $\text{Q(kcal/mole)}$ were obtained for the diffusion coefficients where. $\text{D = D}{}_{\mathrm{o}}$exp$\text{(?Q/RT)}$;${}^{99}\text{Mo: 7.6 × 10}{}^{\mathrm{?4}}$ and $\text{37.6;}{}^{132}\text{Te: 1.32 × 10}{}^6$ and $\text{95.6;}{}^{133}\text{I: 2.7 × 10}{}^{\mathrm{?1}}$ and $\text{66.2;}{}^{133}\text{Xe: 3.6 × 10}{}^2$ and $\text{82.3;}{}^{140}\text{Ba: 2.3 × 10}{}^{\mathrm{?2}}$ and $\text{59.4;}{}^{141}\text{Ce ?}{}^{143}\text{Ce: 1.8 × 10}{}^{\mathrm{?2}}$ and 60.0. The fission product diffusion behavior, in general, fit either the vacancy or the substitutional-interstitial diffusion mechanisms for impurity atoms in a fcc metal. Both valence and ionic radius correlations were found. The data indicate low rates of diffusion for the operating temperatures at which α-thorium-based fuel might be used.     

Isotopic enrichment of tritium by using host-guest chemistry

The isotopic enrichment of tritium in the liquid-liquid extraction system, butylammonium/crownether was investigated using water solution of butylammonium iodide and chloroform solution of dicyclohexyl-18-crown-6. Tritium was enriched in the ammonium-crown complex which was extracted into the organic phase. Both the enthalpy and entropy changes were positive for the direction of negative free energy change in the following equilibrium equation: $\text{R-NH}{}_2\text{T}{}^{\mathrm{+}}{}_{\mathrm{(aq)}}\text{+ R-NH}{}^{\mathrm{+}}{}_3\text{L}{}_{\mathrm{(org)}}\text{= R-NH}{}_2\text{T}{}^{\mathrm{+}}\text{L}{}_{\mathrm{(org)}}\text{+ R-NH}{}^{\mathrm{+}}{}_{\mathrm{3(aq)}}$ where L represents the ligand of dicyclohexyl-18-crown-6 and R-NH⁺₃, butylammonium ion, respectively. The value of entropy effect (TΔS) exceeded that of enthalpy change around room temperature. Single stage isotope separation factors obtained were 12 for n-butylammonium and 50 for tertbutylammonium as guest compounds at 55°C.     

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.     

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.     

Plasma cooling by metal snow: I. Origin of the effect and elimination procedures

In many plasma devices, steel or inconel walls are exposed to large flux densities ?₁ of atomic hydrogen particles which, soon after the reduction of the surface oxides and carbides has started, penetrate into the lattice. The stationary hydrogen concentration $\text{c}{}_1$ in the lattice is expressed as function of ?₁, of the wall temperature $\text{T}{}_{\mathrm{<mtext>w</mtext>}}$ and of the surface roughness factor σ. It is found to be much larger than in an H₂ surrounding. Dissolved atoms recombining on internal surfaces (e.g. at grain boundaries) within the solid can then build up a considerable pressure within resulting gas pockets; should depend strongly on $\text{T}{}_{\mathrm{<mtext>w</mtext>}}$. Near room temperature, the computed values are such that surface-near pockets should crack open, releasing locally high pressure H₂ gas and some metal dust (impurity source) and increasing σ. The expected distribution of $\text{c}{}_{\mathrm{<mtext>H</mtext>}}$ within the sponge-like structure which is expected to result from a prolonged exposure to ?₁ at low $\text{T}{}_{\mathrm{<mtext>w</mtext>}}$ is derived. Means of avoiding the plasma contamination by dust release are pointed out.     

Thermochemistry of the alkali metal and alkaline earth-actinide complex oxides

After a brief discussion of the various techniques used for the preparation of actinide complex oxides, the present status of the thermochemistry of these compounds is reviewed. Perovskiterelated compounds with general formulae $\text{M}{}^{\mathrm{II}}{}_3\text{An}{}^{\mathrm{VI}}\text{O}{}_6$, $\text{M}{}^{\mathrm{II}}{}_2\text{M'}{}^{\mathrm{II}}\text{An}{}^{\mathrm{VI}}\text{O}{}_6$, $\text{M}{}^{\mathrm{I}}\text{An}{}^{\mathrm{V}}\text{O}{}_3$ and $\text{M}{}^{\mathrm{II}}\text{An}{}^{\mathrm{IV}}\text{O}{}_3$, as well as compounds of the type $\text{M}{}^{\mathrm{II}}\text{An}{}^{\mathrm{VI}}\text{O}{}_4$, $\text{M}{}^{\mathrm{I}}{}_2\text{An}{}^{\mathrm{VI}}\text{O}{}_4$, $\text{M}{}^{\mathrm{I}}{}_2\text{An}{}^{\mathrm{VI}}{}_2\text{O}{}_7$ and $\text{M}{}^{\mathrm{I}}{}_4\text{An}{}^{\mathrm{VI}}\text{O}{}_5$, are especially considered as, in the case of those species, thermodynamic data are available for compounds of several actinides and/or several alkali and alkaline earth metals.The stabilities of the complex oxides are discussed with respect to the parent binary oxides and to the aqueous ions; trends as a function of the size of the alkali or the alkaline earth cation are presented. Suggestions for synthesis of some analogous compounds with heavier actinides are also discussed.     

Etude par compression a hautes temperatures de la deformation plastique du bioxyde d'uranium polycristallin

We have studied the compression of stoichiometric, sintered polycrystalline UO₂ as a function of strain rate $\text{?e (7 × 10}{}^{\mathrm{?4}}\text{/min < ?e < 7 × 10}{}_{\mathrm{?2}}\text{/}\text{min}\text{)}$, and temperature $\text{T (600 < T < 1700°}\text{C}\text{)}$. The brittle-to-ductile transition temperature is about 1000°C and we have studied by fractography the characteristics of the fracture at 600 and 800°C. In the plastic deformation range, two types of behaviour have been observed, (i) At low stress ($\text{5 < < 10}\text{kg/mm}{}^2$), the dépendance of strain rate with flow stress σ is $\text{?e = (}\text{cte}\text{/T) σ}{}^{\mathrm{4,5}}\text{exp}\text{-(ΔH/RT)}$ with $\text{ΔH = 96}\text{kcal/mol}$. (The deformation is probably by dislocation climb.) (ii) At high stress ($\text{σ > 10}\text{kg/mm}{}_2$), the deformation is heterogeneous and one observes a compression yield point, the magnitude of which decreases with increasing temperature. In this range, the analysis of the activation parameters has not allowed us to establish a very satisfactory correlation with any of the available deformation schemes.     

Etude thermodynamique et structurale a haute temperature du systeme Zr-C-O. Diagramme de phases a 1555°C

Direct measurements were made of equilibrium pressures of CO over solid phases of the Zr-C-O system, at 1555 °C, along isotherms characterising the reduction of zirconia by carbon. The results, together with the results of X-ray and chemical analysis of quenched specimens, have led us to the following equation describing the reduction of zirconia by carbon: . For this reaction, $\text{ΔG}{}_{\mathrm{T}}{}^0\text{= 151000 ? 78.76T (cal)}$. The homogeneity limits of $\text{ZrC}{}_{\mathrm{y}}\text{O}{}_{\mathrm{z}}$ have been determined as ZrC_0.98 for the carbon side and ZrC_0.73O_0.14 for the zirconia side. An isothermal section of the Zr-C-O phase diagram at 1555°C is presented, and the principal thermodynamic characteristics of some zirconium oxycarbides evaluated.     

Study of 3He+-bombarded palladium using nuclear magnetic resonance techniques

The nuclear magnetic resonance (NMR) from ion-implanted ³He atoms has been observed in ³He⁺-bombarded palladium. Two 1 μm palladium thin films, one on each side of a copper foil substrate, were bombarded at 75°C with 75 and 140 keV ³He⁺-ions to a fluence of $\text{～10}{}^{18}{}^3\text{He}{}^{\mathrm{+}}\text{/}\text{cm}{}^2$ for each energy. Spin-lattice ($\text{T}{}_1$) and spin-spin ($\text{T}{}_2$) relaxation times were measured at 10, 20, and 35 MHz in the temperature range 1–4K. The ³He nuclear relaxation data indicate that the implanted atoms after the 75°C bombardment are situated in small clusters within the Pd thin films.     

On the simulation of CTR neutron radiation damage by heavy ion irradiation

The primary recoil spectrum in nobium under 33 MeV self-ion bombardment is calculated and compared with a calculation by Logan et al. of the same spectrum under neutron irradiation in a controlled thermonuclear reactor. The two spectra are similar, which suggests that a close simulation of CTR neutron radiation damage by ion bombardment will be possible. The displacement rate per projectile is found to be $\text{3.7 × 10}{}^5$ times higher for ions than for neutrons, allowing very high neutron fluxes, up to about $\text{10}{}^{18}\text{cm}{}^{\mathrm{?2}}\text{· sec}{}^{\mathrm{?1}}$, to be simulated with existing accelerators.     

Thermodynamic and phase behaviour in the U-Rh-C system

A galvanic cell using a single crystal of CaF₂ as electrolyte has been used to measure the free energy of formation of the compounds occurring in the U-Rh-C system. Arc-melted samples were subjected to varying annealing conditions, then metallographic and electron probe micro-analysis were used for phase identification. In binary U-Rh system five compounds were found, URh₃, U₃Rh₅, U₃Rh₄, URh and U₂Rh. The free energy of formation values obtained for these compounds were:$\text{ΔG}{}_{\mathrm{<mtext>f</mtext>}}{}^0\text{(URh}{}_3\text{) = ?282.44 + 0.108TkJ/mol}$, $\text{ΔG}{}_{\mathrm{<mtext>f</mtext>}}{}^0\text{(}\text{U}{}_3\text{Rh}{}_5\text{) = ?623.02 + 0.252T}\text{kJ/mol}$, $\text{ΔG}{}_{\mathrm{<mtext>f</mtext>}}{}^0\text{(}\text{U}{}_3\text{Rh}{}_4\text{) = ?547.69 + 0.226T}\text{kJ/mol}$, $\text{ΔG}{}_{\mathrm{<mtext>f</mtext>}}{}^0\text{(}\text{URh}\text{) = ?152.81 + 0.0672T}\text{kJ/mol}$, $\text{ΔG}{}_{\mathrm{<mtext>f</mtext>}}{}^0\text{(}\text{U}{}_2\text{RhC}{}_2\text{) = ?303.38 + 0.0329T}\text{kJ/mol}$, all for the range 1000–1200 K and $\text{ΔG}{}_{\mathrm{<mtext>f</mtext>}}{}^0\text{(}\text{U}{}_2\text{Rh}\text{) = 182.08 + 0.085T}\text{kJ/mol}$ for 893–978 K. However, for U₂Rh the upper temperature range is not necessarily the transformation temperature of this compound.     

Cross sections for the photoionization of H2(X1Σg+, vi = 0–14) with the formation of H2+(X2Σg+, vf = 0–18), and vibrational overlaps and Rn-centroids for the associated vibrational transitions

Cross sections for the photoionization of $\text{H}{}_2\text{(X}{}^1\text{Σ}{}_{\mathrm{g}}{}^{\mathrm{+}}\text{)}$, initially in vibrational levels v_i = 0–14, with the production of $\text{H}{}_2{}^{\mathrm{+}}\text{(X}{}^2\text{Σ}{}_{\mathrm{g}}{}^{\mathrm{+}}\text{)}$ in vibrational levels v_f = 0–18 are tabulated for the full vibrational array at 24 photon wavelengths ranging from 912 Å to 450 Å. The associated vibrational overlap integrals 〈v_ifv_f〉 and R-centroids, $\text{〈v}{}_{\mathrm{i}}\text{|R}{}^{\mathrm{n}}\text{|v}{}_{\mathrm{f}}\text{〉}\text{〈v}{}_{\mathrm{i}}\text{|v}{}_{\mathrm{f}}\text{〉}\text{, n = 1}\text{and}\text{2}$ are also presented together with accurate curve fits of the bound-free ($\text{H}{}_2\text{-H}{}_2{}^{\mathrm{+}}\text{+ e}$) electronic matrix elements.     

Existence of hypostoichiometric chromium sesquioxide at low oxygen partial pressures

The nonstoichiometric composition of $\text{Cr}{}_2\text{O}{}_{\mathrm{3\pm x}}$ was measured by means of thermogravimetry in the range of $\text{1173 ≦ T/K ≦ 1318}$ and . The compositional deviation from stoichiometry, $\text{x}$, in the hyperstoichiometric Cr₂O_3+x phase was observed to be smaller than 2 × 10^?4, irrespective of temperatures, provided that the hyperstoichiometric Cr₂O_3+x exists. The existence of the hypostoichiometric Cr₂O_3?x phase was first established in this study in the region of low oxygen partial pressure below 10^?5 Pa. From the oxygen partial pressure dependence of $\text{x}$ in Cr₂O_3?x, the defect structure was discussed with the neutral chromium interstitials in the composition near stoichiometry and with the triply charged ones far from stoichiometry. The partial molar enthalpy and entropy of oxygen of Cr₂O_3?x showed the complex compositional dependences, suggesting the change of the type of the predominant defect.     

Zum verhalten von tritium in reaktorgraphiten

An experimental investigation of the behaviour of tritium and hydrogen in HTR graphites is described. The diffusion and adsorption kinetic isotherms were measured using a specially developed vacuum apparatus with a variable temperature/time function. Hydrogen was estimated using a mass spectrometer. Tritium was analysed quantitatively after catalytic oxidation using a liquid scintillator. The diffusion coefficients for tritium in reactor graphite were found to be as follows: $\text{D = 0.024 cm}{}^2\text{/sec}\text{exp}\text{(?2.78 eV/kT)}$ and the desorption energies $\text{E}{}^{\mathrm{I}}{}_{\mathrm{des}}\text{= 1.38 ± 0.07 eV}$, $\text{E}{}^{\mathrm{II}}{}_{\mathrm{des}}\text{= 2.07 ± 0.13 eV}$, $\text{E}{}^{\mathrm{III}}{}_{\mathrm{des}}\text{= 3.27 ± 0.07 eV}$. The results indicated that the actual tritium inventory is largely determined by the buffering action of the graphite core. In the start-up phase of an HTR, the tritium concentration in the coolant gas remains relatively low, despite an increased rate of production because the tritium is being adsorbed by the graphite. In later stages of operation, after equilibrium has been reached, the tritium level will be supplemented by material from the hot core regions, so that the coolant activity is not determined by the production rate alone.     

Plastic flow of ordered Zr3Al

Room-temperature tensile experiments established that ordered Zr₃Al of the Ll₂ type obeys the relationship $\text{σ}{}_{\mathrm{?}}\text{= σ}{}_{\mathrm{0,?}}\text{+ kd}{}^{\mathrm{<mtext>?</mtext><mtext>1</mtext><mtext>2</mtext>}}$ where σ_? is the flow stress at a given strain ?, σ_0,? is a strain-dependent frictional stress, $\text{d}$ is the average grain diameter, and $\text{k}$ is a strain-independent constant of magnitude $\text{(2.8 ± 0.3) kg/mm}{}^{\mathrm{<mtext>3</mtext><mtext>2</mtext>}}$. Zr₃Al flows by fine, planar slip, and is susceptible to intergranular cracking.     

Tritium solubility in SUS-316 stainless steel

Tritium solubility in SUS-316 stainless steel was determined with a gas absorption method, in which tritium gas diluted by protium was used. The tritium absorption experiments were carried out at temperatures of 703, 804 and 903 K under pressures of 10, 30, 50 and 100 torr of tritiated hydrogen gas. The radioactivity of tritium dissolved in the specimen was measured by the method of liquid scintillation counting.The tritium solubility was derived from the experimental data by taking into consideration of isotopic equilibrium among H₂, T₂ and HT molecules. The determined tritium solubility can be expressed by the equation: