C.E. Johnson

Thermodynamic review and calculations—alkali-metal oxide systems with nuclear fuels, fission products, and structural materials☆

Abstract This paper considers the phase equilibria of alkali metal oxides and their combinations with other oxides relevant to nuclear fuels, fission products, and structural materials. The other oxides include those of the lanthanides, the actinides, iron, nickel, aluminum, silicon, as well as those of periodic table groups IIA, IVB, VB, VIB, and VIA. The alkali metal halides, chalcogenides, and hydroxides are also included. Techniques are developed to permit calculation of phase equilibria and Ellingham diagrams in ternary and higher-order systems. These techniques include estimation of previously unknown 298.15 K values of the enthalpies of formation and the entropies of many compounds.

Ceramic breeder material development

Lithium-based ceramics have long been recognized as promising tritium-breeding materials for fusion reactor blankets. In particular, their high thermal stability and chemical inertness are favorable safety characteristics. The most important qualification for a candidate ceramic breeder material is most likely, the ability to withstand the rigors of long-term irradiation at high temperature and under large temperature gradients. As a group, the lithium-based ceramics have shown good irradiation behaviour and excellent tritium release characteristics. Individual materials performance will depend upon the actual application, namely, pebble bed versus pellet concept, higher versus lower cooling temperature, etc. Recently Li2ZrO3 and Li2TiO3 were selected as the breeder material for the ITER breeding blanket due to their excellent tritium release behaviour at low temperature. Issues being addressed in support of current blanket design studies will be highlighted.

Ceramic breeder materials : status and needs.

The tritium breeding blanket is one of the most important components of a fusion reactor because it directly involves both energy extraction and tritium production, both of which are critical to fusion power. Because of their overall desirable properties, lithium-containing ceramic solids are recognized as attractive tritium breeding materials for fusion reactor blankets. Indeed, their inherent thermal stability and chemical inertness are significant safety advantages. In numerous in-pile experiments, these materials have performed well, showing good thermal stability and good tritium release characteristics. Tritium release is particularly facile when an argon or helium purge gas containing hydrogen, typically at levels of about 0.1%, is used. However, the addition of hydrogen to the purge gas imposes a penalty when it comes to recovery of the tritium produced in the blanket. In particular, a large amount of hydrogen in the purge gas will necessitate a large multiple-stage tritium purification unit, which could translate into higher costs. Optimizing tritium release while minimizing the amount of hydrogen necessary in the purge gas requires a deeper understanding of the tritium release process, especially the interactions of hydrogen with the surface of the lithium ceramic. This paper reviews the status of ceramic breeder research and highlights several issues and data needs.

Current experimental activities for solid breeder development

Lithium-containing ceramics are among the principal materials being considered for tritium production in future fusion reactors. To ensure development of a data base adequate for evaluation of solid breeder materials, ongoing experimental studies are focused on resolving critical issues related to thermodynamic, thermophysical, and mechanical behavior; to tritium transport and release; and to material response to a neutron environment.

Tritium behavior in lithium ceramics

Tritium is the principal fuel for future fusion power reactors. Unfortunately, tritium is not available naturally and so must be produced through transmutation of lithium. The current approach to fusion reactor breeder blanket design is to place lithium-containing ceramics in a blanket module that surrounds the fusion plasma. These materials have performed well in numerous in-pile experiments, showing good thermal stability and good tritium release characteristics. Tritium release is particularly facile when an argon or helium purge gas containing hydrogen, typically at levels of about 0.1%, is used. However, the addition of hydrogen to the purge gas imposes a penalty when it comes to recovery of the tritium produced in the blanket. Optimizing tritium release while minimizing the amount of hydrogen necessary in the purge gas requires a detailed understanding of the tritium release process, especially the interactions of hydrogen with the surface of the lithium ceramic.

Summary of experimental results for ceramic breeder materials

Abstract Lithium-containing ceramics were quickly recognized as promising tritium breeding materials for fusion reactor blankets, particularly because of their safety advantages. Relevant material properties were investigated to evaluate further their suitability. An extensive RD overall properties (baseline, thermal, mechanical); compatibility with structures and beryllium; tritium release characteristics; irradiation behavior; activation; reprocessing; waste disposal issues. As a result of this investigation, lithium-containing ceramics are considered to be excellent tritium breeding materials.

Fuel-cladding chemical interaction in uranium-plutonium oxide fast reactor fuel pins☆

Abstract A model of cladding attack is developed based on the formation of a cesium chromate on the cladding inner surface and the formation of a cesium-fuel compound at the cooler, outer edge of the fuel. These compounds are assumed to be in equilibrium via the gas-phase transport of cesium and/or cesium hydroxide across the fuel-cladding gap. Using thermochemical and kinetic data for the species involved, the model successfully predicts: 1. (a) the observed temperature dependency of cladding attack in EBR-II, DFR, Rapsodie, and Phenix pins; (b) the observed in-pile threshold temperature for cladding attack; 2. (c) the decreased incidence and decreased severity of cladding attack for sphere-pac fuel and for pellet fuel irradiated at low power levels; and 3. (d) the large difference in the temperature dependency of cladding attack between out-of-pile versus in-pile experiments. Furthermore, the model identifies the local temperature difference between the fuel outer surface and the cladding inner surface to be the single most important parameter governing cladding attack.

Tritium release from lithium titanate, a low-activation tritium breeding material

Abstract The goals for fusion power are to produce energy in as safe, economical, and environmentally benign a manner as possible. To ensure environmentally sound operation low-activation materials should be used where feasible. The ARIES Tokamak Reactor Study has based reactor designs on the concept of using low-activation materials throughout the fusion reactor. For the tritium breeding blanket, the choices for low activation tritium breeding materials are limited. Lithium titanate is an alternative low-activation ceramic material for use in the tritium breeding blanket. To date, very little work has been done on characterizing the tritium release for lithium titanate. We have thus performed laboratory studies of tritium release from irradiated lithium titanate. The results indicate that tritium is easily removed from lithium titanate at temperatures as low as 600 K. The method of titanate preparation was found to affect the tritium release, and the addition of 0.1% H 2 to the helium purge gas did not improve tritium recovery.

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.

Oxygen potential of irradiated urania--plutonia fuel pins

Abstract The oxygen-potential gradient in irradiated urania-plutonia fuel pins has been estimated from the distribution of molybdenum between the noble-metal-alloy inclusions and the oxide matrix as measured using an electron microprobe. The values of the oxygen potential were found to be more negative than would be computed for the fuel on the basis of the initial O/M ratio, thus indicating that the overall O/M of the fuel was decreased by irradiation. The O/M gradient has been computed and an oxygen material balance was determined between fuel oxide, fission products, and cladding. The nearly identical oxygen pressure profiles found for fuel pins irradiated for a range of values of the burnup indicates that the oxygen-potential gradient is establsihed early in the irradiation period, probably at the time the structure has been established and then remains nearly constant. The oxygen lost during the restructuring period is determined by the initial O/M of the fuel, the smear density and the temperature gradient. The oxygen lost during the burnup period is proportional to the burnup.