Tritium Recovery Experiment Research Articles

Influence of Lithium Mass Transfer on Tritium Behavior in Pebbles of Li2TiO3 with Excess Lithium

• Li 2 TiO 3 with excess Li was heated at 900 °C for 30 days in 1000 Pa H 2 /Ar flow. • Li mass loss was 0.7 wt%, grain size increased and BET surface area decreased. • The majority of evaporative Li was lost by first 3 days. • The influence of the 30 days heating on tritium behavior was not large. • Tritium recovery by isotope exchange with H 2 was effective at 600 °C and 900 °C. Tritium breeding ceramic materials are placed at high temperatures for a long period in a fusion DEMO reactor. Therefore, the understanding of Li mass loss phenomena and its influence on tritium behavior are important. In this study, the pebbles of Li 2 TiO 3 with excess Li were heated at 900 °C for 30 days in a 1000 Pa H 2 /Ar flow and tritium sorption and recovery experiments were carried out. Li mass loss by the heating was evaluated to be 0.7 wt%. The value of Li mass loss was almost same as that for 3 days heating at the same condition. Tritium sorption capacity for the heated pebbles at 600 °C and 900 °C were almost same as that for the pebbles as received. Tritium sorbed in the pebbles could not be recovered effectively by the 1000 Pa H 2 /Ar purge at room temperature and 300 °C but it could be recovered at 600 °C and 900 °C. The influence of the long-time heating on the behavior of tritium sorbed in the pebbles was not large.

A new blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

DT中性子照射による固体増殖ブランケット模擬体系からのトリチウ ム回収実験 -(1)

DT中性子照射による固体増殖ブランケット模擬体系からのトリチウ ム回収実験 -(1)

Recovery of Tritium Dissolved in Sodium at the Steam Generator of Fast Breeder Reactor

The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D2 gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment.

In situ tritium recovery from Li 2O irradiated in a fast neutron flux: BEATRIX-II, Phase II temperature-change canister

BEATRIX-II, Phase II was an in situ tritium recovery experiment designed to characterize the behavior of lithium ceramics irradiated to high burnup in a fast neutron flux. A thin-walled Li 2O ring specimen was irradiated at temperatures in the range from 510 to 640°C in helium sweep gases containing 0, 0.01 and 0.1% H 2 to a 5% lithium burnup. The tritium recovery behavior of Li 2O was characterized by using temperature transients and sweep gas composition changes to effect changes in the tritium inventory.

Changes in the microstructure and density of Li2O during irradiation in BEATRIX-II, phase I

Abstract BEATRIX-II, phase I was an in situ tritium recovery experiment carried out on Li2O to lithium burnups in excess of 4%. Solid Li2O specimens were irradiated under a temperature gradient in the range from 700 K at the outer diameter to 1270 K at the center. A ring specimen of Li2O operated at a nearly uniform temperature with the capability of changing the temperature in the range 780–920 K. Li2O single crystals with two different 6Li enrichments (0.07 and 1.8 at.%) were irradiated at 650 K in non-vented canisters. The microstructures of these specimens before and after irradiation were investigated using optical and scanning electron microscopy. Substantial microstructural change was found to have occurred in the solid specimen during irradiation and this change was characterized by the development of large columnar grains, lenticular pores and a center void. The development of columnar grains and the formation of the center void was attributed to the migration of lenticular pores up the thermal gradient by an evaporation-condensation process. Morphological changes on fracture surfaces were observed for both polycrystalline (solid and ring) specimens and single-crystal specimens. Measurements of density were used to characterize further the effects of irradiation on the open and closed porosity distributions in all the specimens. Although extensive microstructural and density changes occurred during irradiation, these changes had no significant impact on the integrity of Li2O during irradiation, and therefore these results support the use of Li2O as a fusion solid breeder.

Tritium release kinetics of solid lithium ceramics with irradiation defects

For the assessment of tritium release behaviors of lithium oxide under irradiation, a model reaction scheme was developed, which was based on the interactions of irradiation-induced defects with tritium. F 0 centers (oxygen vacancy trapping two electrons) and their partners, O 2 molecules, were considered to be produced under the blanket conditions. The model predictions were compared with the results of an in situ tritium recovery experiments with Li 2O, called BEATRIX-II, and the observed tritium release behaviors were successfully interpreted by the present model. The reactions were discussed, of which the equilibrium constant and rate constant were to be determined for the further development of the model.

BEATRIX-II: A multinational solid breeder materials experiment

BEATRIX-II is an in situ tritium recovery experiment in the fast flux test facility (FFTF) reactor designed to characterize the feasibility of utilizing solid breeder materials at extended burnups in a fast neutron flux. Although not yet complete, the BEATRIX-II experiments have already substantiated that the solid breeder selected for ITER, Li2O, has good irradiation stability and tritium recovery. Temperature stability, lithium transport, dimensional stability and tritium recovery issues of Li2O up to 5% Li burnup were addressed in this experiment, Temperature gradients far more severe than in the ITER design, 400 to 1000°C, were found to be essentially unchanged by burnup and produced no observable instability, either from swelling or lithium vapor transport. Temperature change experiments illustrated that lithium inventories do not appear to increase as a result of irradiation to burnups of 5%.

ITER R & D on breeder blanket materials

The International Thermonuclear Experimental Reactor (ITER) activity includes specific R&D in critical areas to support the ITER design. Several of the critical tasks defined as part of the ITER validating R&D are materials related. This paper presents a summary of the breeder/blanket materials related R&D conducted by the four participating parties; viz., European Community, Japan, Soviet Union and the United States. The current effort includes several subtasks in each of the following four task areas: 1. (1) ceramic breeder materials, 2. (2) lead-lithium breeder, 3. (3) aqueous salt solutions, 4. (4) neutron multiplier materials. The objective and scope of each subtask is stated. Although this work was initiated in late 1988, important results have been obtained in all four areas. The work on ceramic breeders includes properties measurements, compatibility studies, and tritium recovery experiments with Li 2O and selected ternary ceramics. Investigations on the lead-lithium breeder have focused primarily on properties, compatibility, and tritium extraction from the 17Li-83Pb eutectic alloy. The work on the aqueous salt breeders includes corrosion, solution chemistry, and radiolysis effects in a radiation environment. Investigations defined under the neutron multiplier task are focused on beryllium; however, lead in the lithium lead alloy also serves as a neutron multiplier. The schedule and planned future work in each area are summarized.

The interactions of tritium with irradiation defects in solid breeder materials

A post-irradiation annealing experiment was performed to determine the mechanism of reactions of tritium with irradiation-induced defects in a candidate solid breeder material, Li 2O. The chemical state distribution of tritium was observed to depend on neutron fluence and to vary with time and temperature during the annealing. The results were analyzed by taking into account the reactions of tritium with irradiation defects and the rate constants were determined. By using the rate constants, a simple model was developed to predict the effect of irradiation on the tritium recovery, that is the production of noncondensable tritium species, at in situ conditions of actual blanket systems. The prediction agreed very well with the observation in an in situ tritium recovery experiment.

Mistral: A comprehensive model for tritium transport in lithium-base ceramics: Part II: Comparison of model predictions with experimental results

A new tritium transport model called MISTRAL ( M odel for I nvestigative S tudies of T ritium R elease in L ithium Ceramics) has been developed to describe and predict the kinetics of tritium release in lithium ceramic materials for tritium breeding applications in fusion blankets. The model has transient capabilities and has been developed to analyze the full range of transient conditions produced in in-pile tritium recovery experiments and expected in fusion blankets. Calibration of the model against experiments has been done in parallel with its development in order to assess its predictive capabilities and to identify the ranges of potential applicability. The comparisons of the results available for lithium metasilicate and aluminate samples irradiated respectively in the two in-pile tritium recovery experiments LISA1 and MOZART are presented and discussed in this paper. They have been selected for the calibration of the codes as being good examples of various features relevant for tritium release analysis in ceramic breeders under different transient conditions such as change in temperature, purge gas composition and reactor power.

Mistral: A comprehensive model for tritium transport in lithium-base ceramics: Part I: Theory and description of model capabilities

MISTRAL is a theoretical model developed to describe tritium transport and release in fine-grained ceramic materials for tritium breeding applications in fusion blankets. The model includes as relevant physical processes tritium diffusion in the effective bulk (grains and grain boundaries), adsorption, recombination and desorption at the breeder surface and diffusion through the network of pores. A key improvement of the model, compared with those already in the literature, consists of a better characterization of the processes at the breeder surface and their linking to the bulk and pore regions. The sets of governing transport equations and corresponding boundary conditions are formulated together with the choice of the computational algorithm. A computer code with transient capabilities has been developed based on the model. It aims at describing tritium release for several transient conditions relevant for in-pile tritium recovery experiments and for fusion blankets. To assess the range of applicability of the model, several calculations have been performed and the results of the analysis are herein presented and discussed.

A Fast Neutron, In Situ Tritium Recovery Experiment on Solid Breeder Materials

An in situ tritium recovery experiment has been designed and is being fabricated for the irradiation of Li2O in the Fast Flux Test Facility, FFTF. Two in situ tritium recovery canisters will be irradiated with lithium atom burnups to 4%. One canister will provide fundamental data on tritium release as a function of temperature, gas composition, and flow rate. The other canister will contain solid pellet specimens with large (430°C) radial temperature gradients in order to provide integrated performance data.

The effect of fusion neutron irradiation on tritium recovery from lithium salts

The effect of 14 MeV neutron irradiation on tritium recovery was studied from the lithium salts LiF, LiCl, LiBr, Lil, LiF-BeF 2 and LiF-PbF 2. The chemical state distribution of tritium formed in the irradiated lithium salts was found to depend on the neutron fluence. The dependence was explained by taking into account the interactions of tritium with radiation defects. In the tritium recovery experiments, the effects of irradiation were apparently observed with the highly irradiated samples; the TX recovery with a helium sweep gas and the HT recovery with a hydrogen-containing sweep gas were well described by the reaction kinetics in which F centers play an important role. In the case of the less-irradiated samples, on the other hand, the tritium recoveries were found to follow a diffusion-controlled kinetics and were explained on the basis of atomic theory of diffusion.

Diffusion behaviors of tritium formed in lithium-lead alloys by 14 MeV neutron irradiation

Tritium recovery experiments were performed with Li-Pb alloys irradiated with 14 MeV neutrons. The effects of temperature, alloy composition, and neutron fluence were observed on tritium recovery. Tritium was released mainly as HT from the alloy and its release rate was controlled by bulk diffusion in the alloy. From the chemical analysis of tritium species in the alloy, it was found that free tritium atoms in the T 0 state diffuse more rapidly than the others such as tritide ions in the T − state. The diffusion mechanism for each species was discussed based on the obtained diffusion coefficients. It was also found that free tritium atoms might be trapped by radiation defects produced in the alloy.

In situ tritium recovery experiment from lithium oxide under high neutron fluence

In situ measurement of tritium release during irradiation provides a more realistic understanding of the tritium release mechanism than that obtained by the post-irradiation annealing method. The tritium release curves reached saturated values after irradiation of a few days at the reactor operation of 10 MW. The saturated values were in good agreement with the predicted values of the tritium generation rate obtained from neutronics calculation. Temperature dependence of tritium release under the temperature of 550°C at the pellet center was interpreted by the decomposition of LiOH(T) at the pellet surface. The ratio of tritium gas and tritiated water (HT/HTO) was strongly affected by the moisture generated from the piping system and the pellet. The retained tritium in Li 2O was 0.5–0.9 wppm after the final stage of the irradiation of 760°C and the burnup of 0.24% of total Li.