S.W. Tam

The trio experiment

The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.

Modeling of tritium behavior in ceramic breeder materials

Computer models are being developed to predict tritium release from candidate ceramic breeder materials for fusion reactors. Early models regarded the complex process of tritium release as being rate limited by a single slow step, usually taken to be tritium diffusion. These models were unable to explain much of the experimental data. We have developed a more comprehensive model which considers diffusion and desorption from the grain surface. In developing this model we found that it was necessary to include the details of the surface phenomena in order to explain the results from recent tritium release experiments. A diffusion-desorption model with a desorption activation energy which is dependent on the surface coverage was developed. This model provided excellent agreement with the results from the CRITIC tritium release experiment. Since evidence suggests that other ceramic breeder materials have desorption activation energies which are dependent on surface coverage, it is important that these variations in activation energy be included in a model for tritium release. 17 refs., 12 figs.

Advanced understanding of the tritium recovery process from the ceramic breeder blanket

Abstract The key to successful operation of a tritium breeder blanket is to understand the tritium transport and release characteristics and the role that hydrogen plays in this process. Indications are that grain size (surface-to-volume ratio) largely determine whether tritium release is limited by diffusion or desorption. That is, the larger the grain size the higher the probability that bulk diffusion will determine the release rate. For smaller grain size, the actions taking place on the grain surface become extremely important especially as regards the role that hydrogen plays in the overall process. Experimental studies have indicated that the presence of 0.1% H 2 in the helium purge gas enhances the release of tritium from the lithium ceramic. The tritium released has been found in the form of both HT and HTO. The ab initio calculations on the dissociative hydrogen chemisorption on lithium oxide surfaces provide one component of the quantitative basis for an understanding of the role of hydrogen in affecting the release of tritium from lithium ceramic breeders. These calculations suggest heterolytic adsorption of hydrogen onto the ceramic surface.

Ab initio calculations for dissociative hydrogen adsorption on lithium oxide surfaces

Abstract Dissociative hydrogen chemisorption on the Li 2 O surfaces of the (100), (110) and (111) planes has been investigated with ab initio Hartree-Fock calculations. Calculations for unrelaxed crystal Li 2 O structures indicated that except for the (100) surface, the (110) and (111) surfaces are stable. Results on the heterolytic sites of n -layer (110) (where n ⩾2) slabs and three-layer (111) slabs suggest that dissociative hydrogen chemisorption is endothermic. For a one-layer (110) slab at 100% surface coverage, the dissociative hydrogen chemisorption is exothermic, forming OH − and Li + H − Li + . This results also indicate that the low coordination environment in surface step structure, such as kinks and ledges, may play an important role in the hydrogen chemisorption process. On the homolytic sites of the (110) and (111) surfaces, there is no hydrogen chemisorption.

Enhanced tritium transport and release by solids modification

In order to improve the tritium release characteristics of lithium ceramics, we are investigating the effects of dopants on tritium transport and release. Prior work has suggested a correlation between tritium and lithium diffusion in lithium-containing ceramics. This correlation has led us to propose a mechanism for tritium diffusion in which the tritium diffuses in the form of a lithium-vacancy/tritium complex. If this is the case, one should be able to increase the tritium diffusivity by increasing the number of lithium vacancies and thereby increasing the number of lithium-vacancy/tritium complexes. The size of the increase in the diffusivity, however, will be dependent upon several parameters, including the binding energy of the lithium-vacancy/tritium complex. Our calculations indicate that, under conditions comparable to those in some in-pile irradiation experiments, a binding energy of around 84 kJ/mol should increase the diffusivity and lead to a decrease in the steady-state tritium inventory by about a factor of six.

Theory of high-temperature phase transitions in actinide oxides

Abstract A theory of phase transitions in actinide oxides which includes the effects of both Frenkel and Schottky defects has been developed. Coulomb interaction between charged defects and its effects on both defect formation energy and configurational entropy have been shown to be key factors in describing the phase transition. This simple model accounts for most of the experimental information available but predicts lower defect concentrations than have been observed in recent neutron-scattering experiments. An extended treatment which should account for that effect as well is suggested.

Saturation phenomena and percolative transitions in the high temperature thermal conductivity of γ-LiAlO2

The thermal conductivity of γ-LiAlO2 with porosities varying from 7% to 53% has been measured over a temperature range of 1088–1400 K. The conductivity was found to be insensitive to temperature in that regime. This saturation phenomenon was interpreted in terms of minimum conductivity. Thermal conductivity exhibited a non-linear dependence on porosity. It is suggested that the distribution of inter- and intra-granular porosity in the solid was instrumental in bringing about this non-linear behavior.

Theory of the temperature dependence of positron bulk lifetimes — implications for vacancy formation enthalpy measurements via positron experiments☆

Temperature dependent effects, which may have a bearing on determinations of vacancy formation enthalpies in metals by positron annihilation, have been observed in certain metals. These effects have been observed to occur both at temperatures below those at which positron annihilation is most sensitive to equilibrium vacancies and at temperatures well within the vacancy-sensitive region. The effect of thermal lattice displacements on positron lifetimes in metals was investigated to help understand these phenomena. (GHT)

Burn Behavior of a Polyurethane Foam Impact Limiter

We recently studied the burn behavior of a polyurethane (PU) foam-filled impact limiter — a design for both the Mixed-Oxide Fresh Fuel Package and the Hanford Unirradiated Fuel Package (HUFP) that have been certified for shipment of their authorized contents by the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy, respectively. In this paper, we examined the mechanisms of thermal degradation of PU foam by reaction type and flame temperature. Evidence suggests that, for an air-tight impact limiter, the pyrolysis reaction dominates initially in the limiter enclosure. The pyrolysis generates a large amount of highly flammable gases, creating the conditions necessary for a subsequent combustion reaction near the vent holes where air is abundant. The coupled heat release from the combustion of these flammable gases and oxygen drives the jet flames to burn at a very high temperature. The three-dimensional finite-element analysis code, ANSYS Mechanical, was used to model the HUFP package (with the PU foam-filled impact limiters) and compute the temperatures near the seal region of the containment boundary of the package. We examined the effects of various parameters — including fire duration, chimney flow, foam thickness loss, and jet flame temperature — on the thermal performance of the package. The results indicate that, under the simulated conditions, the O-ring seal temperature near the packaging containment boundary rose to varying degrees, but it did not exceed the seal temperature limit of 400°F (204°C). The remaining foam thickness is critical to maintain package safety.Copyright

Training in the Application of the ASME Code to Transportation Packaging of Radioactive Materials

The Department of Energy has established guidelines for the qualifications and training of technical experts preparing and reviewing the safety analysis report for packaging (SARP) and transportation of radioactive materials. One of the qualifications is a working knowledge of, and familiarity with the ASME Boiler and Pressure Vessel Code, referred to hereafter as the ASME Code. DOE is sponsoring a course on the application of the ASME Code to the transportation packaging of radioactive materials. The course addresses both ASME design requirements and the safety requirements in the federal regulations. The main objective of this paper is to describe the salient features of the course, with the focus on the application of Section III, Divisions 1 and 3, and Section VIII of the ASME Code to the design and construction of the containment vessel and other packaging components used for transportation (and storage) of radioactive materials, including spent nuclear fuel and high-level radioactive waste. The training course includes the ASME Code-related topics that are needed to satisfy all Nuclear Regulatory Commission (NRC) requirements in Title 10 of the Code of Federal Regulation Part 71 (10 CFR 71). Specifically, the topics include requirements for materials, design, fabrication, examination, testing, and quality assurance for containment vessels, bolted closures, components to maintain subcriticality, and other packaging components. The design addresses thermal and pressure loading, fatigue, nonductile fracture and buckling of these components during both normal conditions of transport and hypothetical accident conditions described in 10 CFR 71. Various examples are drawn from the review of certificate applications for Type B and fissile material transportation packagings.Copyright