Mechanisms of Helium Thermodesorption from Irradiated Boron Carbide

Abstract

Two limiting regimes of gas release from particles are studied: diffusion and activation. The conditions under which either the diffusion or the activation desorption regime occurs are determined. A model is proposed for the activation release of helium atoms from centers of accumulation and diffusion of these atoms toward the surface of a spherical particle of boron carbide powder. A quantitative estimate is obtained for the release of helium from a material during annealing, and the influence of the size of the powder grains and the rate of change of the temperature of the material on the thermodesorption of helium from irradiated boron carbide is estimated.Information reported on the destruction of the cladding of absorbing elements caused by excess gas pressure [1, 2] dictates the need for developing a computational program, based on this information, for simulating the behavior of (n, α) absorbers and rods in the control and protection system under irradiation, specifically, the development of mathematical models of helium release from boron carbide powder. The amount of helium released from irradiated boron carbide by heating has been determined experimentally many times [3][4][5], but there is still no satisfactory model for describing this process adequately [6]. A method for estimating the amount of helium released from irradiated boron carbide during annealing under nonstationary conditions has been proposed [7]. The model was constructed on the basis of a theory of annealing of radiation defects [8] under the assumption that the limiting stage of helium release is overcoming an activation barrier when gas atoms detach from accumulation centers. However, this mechanism meglects the influence of helium diffusion toward the surface of a powder particle after the atoms detach from the accumulation center. It would be useful to develop a mathematical model that would make it possible to generalize the activation mechanism of helium release [7] so as to include diffusion transfer.Boron carbide powder is now used in absorbing kernels in the VVÉR-1000 control and protection system. The particle size is about 10 µm, which is close to the travel distance of a thermal neutron in boron carbide. Consequently, it can be assumed that accumulation centers capture helium uniformly over the entire volume of a particle.A mathematical model of helium release from a powder particle in the process of annealing was constructed under the following assumptions:• all of the helium in the irradiated boron carbide is located in stationary accumulation centers, and the number of helium atoms occupying interstices is negligibly small compared with the total number in a particle; • when the powder is being heated helium atoms are released from all capture centers by an activation mechanism with a low probability of helium being captured by an accumulation center;

The influence of irradiation induced vacancies on the mobility of helium in boron carbide

“…The goal was, however, mainly phase stability and not the properties of isolated defects, nor their role in mass transport kinetics. In spite of a number of experimental works devoted to helium bubbles [11,12,13,14,15,16,17,18,19,20,21] and the related question of helium kinetics [20,22,23,24], not much is known on self-diffusion.…”

Influence of Irradiation on Mechanical Properties of Materials

“…As Eqs (2.7) has not an analytic solution, we illustrate a dependence pattern of the barrier density by a numerical analysis example of the Cauchy problem solution in which the parameter simulating values and the initial defect densities Here it is shown the dependence typical behavior of the barrier relative densities of the vacancy 10 / CC and interstitial 20 / CC types on the relative irradiate dose 0   where 0 is a measure scale of the defect densities (here it is equal 10 15 cm -3 ). The measure units are accepted relative to the selected scale τ 0 (fluence) the numerical value of which is determined by a specific problem (it is convenient to select a minimal fluence of the specific problem as the dose measure scale; for instance, in the ion irradiation the value of τ 0 can be equal 10 14 ion/cm 2 or 10 22 n/m 2 as the contemporary neutron fluence).…”

“…C  =0.5. The results of experimental studying radiation embrittlement effects and the temperature dependences of such durable material characteristics as specific elongation and yield strength have been given in a series of the works [10][11][12][13][14][15]. In Refs [12,15], it is shown that a deformation process connected with dislocation collective behavior in irradiated deformed materials is characterized by availability of the different structure deformation levels.…”