Tritium release experiments on Li4SiO4 pebbles from TRINPC-I experiments: Effects of water adsorption and hydrogen gas

Out-of-pile tritium release experiments under different water uptake contents and purge gas chemistry were performed on Li₄SiO₄. Water measurement was performed on samples under different experimental procedures. It was found that water was adsorbed on the sample during its transferring and storage process. A strong dependence of tritium release behavior on water uptake was determined. By doping H₂ in the sweep gas, the formation of water in orthosilicate was observed in addition to the isotope exchange reaction with H₂ gas. Thermal desorption peaks of the water formation reaction and H₂ isotope exchange reaction appeared at 668 °C and 463 °C, respectively, at ramping rate of 5 °C/min.     

Out-of-pile tritium release study on Li4SiO4 pebbles from TRINPC-I experiments

Out-of-pile tritium release examinations of irradiated Li₄SiO₄ pebbles were performed in TRINPC-I experiments for evaluating material performance and verifying the system design. To generate tritium the specimens were irradiated with neutrons. Li₄SiO₄ pebbles were made by a freeze-drying method. In the experiments, concentrations of tritium in the form of tritium gas (HT + T₂) and tritiated water (HTO + T₂O) in the outlet streams of a reactor tube were measured separately with an ionization chamber and a liquid scintillation radiometer. The results show that the percentage of tritium gas (HT + T₂) and tritiated water trapped by the breeder pebbles were about 72% and 19% of totally released tritium, respectively. Thus, more tritium was released in the form of tritium gas in this work. In addition to tritium trapped by the breeder pebbles, the amount of free tritium was also measured by breaking on-line a quartz capsule containing Li₄SiO₄ pebbles, the percentage of which was 9% of totally released tritium. The temperature peaks of tritium gas mainly appeared at about 477 °C and 654 °C, while the temperature peak of tritiated water appeared at about 402 °C, under which most of tritiated water released.     

氚增殖剂Li4SiO4 陶瓷小球的制备工艺

欧洲和中国聚变堆固态产氚包层（TBM）的氚增殖剂倾向于采用直径0.5～2mm的Li₄SiO₄陶瓷小球填充床。本工作探讨锂陶瓷小球的性能指标设计,研究挤压-滚圆、烧结法制备Li₄SiO₄小球的工艺可行性,测试分析小球的密度、直径、球形度、晶粒尺寸、压碎载荷等性能。研究表明：挤压-滚圆成型、1050℃无压烧结的Li₄SiO₄陶瓷小球密度为90.4%TD,堆积密度为52.9%TD;平均直径为0.95mm,标准偏差为0.15mm;球形度为1.10;平均压碎载荷为18.50N,标准偏差为2.76N;平均晶粒尺寸为14μm;相结构由Li₄SiO₄主晶相、少量Li₂SiO₃和Li₂Si₂O₅等组成。采用优化的挤压 滚圆、烧结工艺可制备出合格的Li₄SiO₄陶瓷小球产品。     

Fabrication and characterization of Li4SiO4 pebbles by melt spraying method

The solid breeder blanket concept proposed by the China features the tritium breeding ceramic as pebble beds in several submodules. The lithium orthosilicate (Li₄SiO₄) is considered as first candidate ceramic breeder materials fabricated by the melt spraying method, which is favorable to other processes in terms of density and recycling. The production process involves rapid quenching of the liquid droplets from the melt to room temperature which cause internal stresses and leads in some cases to formation of microcracks and the dispersion of mechanical properties. Molar ratio (Li/Si) of the pebbles was evaluated by ICP–OES. It is shown that the Li/Si ratio of the pebbles is slightly varying from batch to batch because of evaporation of lithium at high temperatures. The crush tests on single pebbles show that a mean value of 7.0 N was obtained in crush load experiments of 40 pebbles with a diameter of 1.0 mm. It results that heat treatment of pebbles improves the density and mechanical stability. The activation characteristics for the current composition of Li₄SiO₄ pebbles were assessed. The calculations were used to identify critical amounts of impurities and were compared to the results of pure material without impurities.     

Failure initiation and propagation of Li4SiO4 pebbles in fusion blankets

Lithium orthosilicate (Li₄SiO₄) pebbles are considered to be a candidate as solid tritium breeder in the helium cooled pebble bed (HCPB) blanket. These ceramic pebbles might be crushed during thermomechanical loading in the blanket. In this work, the failure initiation and propagation of pebbles in pebble beds is investigated using the discrete element method (DEM). Pebbles are simplified as mono-sized elastic spheres. Every pebble has a contact strength in terms of critical strain energy, which is derived from a validated strength model and crush test data for pebbles from a specific batch of Li₄SiO₄ pebbles. Pebble beds are compressed uniaxially and triaxially in DEM simulations. When the strain energy absorbed by a pebble exceeds its critical energy it fails. The failure initiation is defined as a given small fraction of pebbles crushed. It is found that the load level for failure initiation can be very low. For example, if failure initiation is defined as soon as 0.02% of the pebbles have been crushed, the pressure required for uniaxial loading is about 2.5 MPa. Therefore, it is essential to study the influence of failure propagation on the macroscopic response of pebble beds. Thus a reduction ratio defined as the size ratio of a pebble before and after its failure is introduced. The macroscopic stress–strain relation is investigated with different reduction ratios. A typical stress plateau is found for a small reduction ratio.     

Effect of water vapor on tritium permeation behavior in the blanket system

It is found that most hydrogen supplied to the purge gas changed to water vapor due to the water formation reaction in the early stage of the blanket operation and that physical or chemical adsorbed water is released in the high concentration into the blanket purge gas when the blanket temperature becomes higher than several hundreds of degrees K if the pre-treatment is not applied to the solid breeder materials. Effect of coexistence of water vapor in the purge gas on permeation behavior of hydrogen through F82H ferritic steel in the breeding part and palladium–silver (Pd–Ag) in the recovery part is discussed because use of them is generally considered for recovery of bred tritium from the solid blanket. Almost no decrease in permeation rate of F82H is observed in this study when water vapor exists in the blanket purge gas. The permeability of hydrogen isotopes through the Pd–Ag pipe gradually decreases when water vapor exists in the blanket purge gas. Properties required in estimation of the hydrogen permeated to the purge gas are experimentally obtained in this study.     

Diffusion analysis on the thermal release of tritium from a neutron absorbing material

Tritium released from neutron irradiated borosilicate glass was determined by a specially designed sampling system and a liquid scintillation counter at temperatures in the range of 200–700°C. It was found that the chemical form of tritium released was tritiated water (HTO, T₂O) for the most part. Tritium produced in the glass would react with oxygen to form OT and diffuse out by a similar mechanism as the molecular diffusion of water in glasses. The diffusion coefficient of tritiated water in borosilicate glass obtained is expressed by D (cm²/s) = 5.3 × 10⁻⁴ exp( −128 kJ/mol)/RT). It is concluded from the diffusion analysis that the greater part of tritium produced in a neutron absorber, which is made of borosilicate glass, would remain in the glass for a few years of irradiation.     

Thermal Conductivity Measurement with Silica-Coated Hot Wire for Li4SiO4 Pebble Bed

The effective thermal conductivity of a Li₄SiO₄ pebble bed was measured by the hot wire method. The bare and silica-coated Nichrome heaters were used as the hot wires. At 975 K, effective thermal conductivity was not measured correctly by the bare hot wire. This is due to the fact that the electrical signal of a bare thermocouple is distorted due to the electrical conductivity of Li₄SiO₄. Using a silica-coated hot wire, effective thermal conductivity can be measured at temperatures ranging from room temperature to 975 K. The effect of the coating layer on the measured effective thermal conductivity was estimated to be small and corresponded to the experimental data. The hot wire method with silica coating can be applied to other ceramic breeder materials.     

Isotope Exchange Capacity on Li4SiO4 and Comparison of Tritium Inventory in Various Solid Breeder Blankets

It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of a breeder material, absorption of water vapor into bulk of the grain, adsorption of water on surface of the grain, and exchange capacity of tritium to be trapped to surface of the grain together with two types of isotope exchange reactions for evaluation of the tritium inventory in a solid breeder blanket under various conditions. The isotope exchange capacity on the Li₄SiO₄ surface is experimentally obtained in this study. Most of the properties required for evaluation of the tritium inventory for various blanket materials have been already quantified by the present authors. Then it has become possible to compare the tritium inventory in solid breeder blankets packed with either Li₂O, LiAlO₂, Li₂ZrO₃, Li₂TiO₃ or Li₄SiO₄ using the calculation model previously presented by the present authors.     

Effect of sweep gas species on tritium release behavior from lithium titanate packed bed during 14 MeV neutron irradiation

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li₂TiO₃) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.     

Effect of outside tritium source on tritium balance of a D-T fusion reactor

Attainable tritium breeding ration in the blanket system must be larger than the required breeding ratio when no effective tritium resources from outside are expected. It is revealed recently that a considerable amount of tritium can be trapped to the re-deposition layer of the first wall materials and that the time constant of this phenomenon is rather long. Then, the tritium breeding ratio around 1.1 is required in the blanket system when 3 years is claimed for the tritium doubling time to prepare tritium for the initial inventory of a next reactor. Construction of an outside tritium supply is one of the possible ways to compensate the lack of tritium because it is generally considered that the attainable tritium breeding ratio in the solid breeder system is around 1.05. It is reported recently that a high-temperature gas-cooled reactor can produce 10 kg of tritium per year. The preferable amount of tritium production rate of the outer tritium supply is discussed in this study from the viewpoint of tritium balance in a D-T power reactor.     

Study of space-dependent fast-neutron spectra and tritium breeding ratio in different assemblies of Li, Li and natural Li

Using diffusion theory and the eigenfunction expansion method, a detailed space-dependent study of fast neutron spectra has been carried out in ⁷Li, ⁶Li and natural Li assemblies of dimensions 100 × 60 × 60 cm³ and 30 × 60 × 60 cm³. Values of total (space-averaged) tritium breeding ratio (TBR) have been obtained for these assemblies. BARC-892, a 27-group data-set was used for these calculations.

The results of TBR for the smaller assembly have been compared with the corresponding experimental and calculated values of Takahashi et al. (1984). We find that the present values of TBR are in reasonably good agreement with the above experimental results.     

D+注入Li4SiO4的表面化学状态及其热解吸行为

锂陶瓷氚增殖剂的氢同位素行为是聚变堆固态产氚包层关心的重要课题。本文将3 keV D⁺注入Li₄SiO₄,采用X射线光电子能谱在线分析注氘前后材料表面的化学状态,同时采用热解吸谱（TDS）实验技术,研究注氘后Li₄SiO₄中氢同位素的热解吸行为。实验结果表明：D⁺注入会改变Li₄SiO₄表面的化学环境,产生多种辐照缺陷和化学键合状态;氘滞留量和热解行为受注氘时样品的温度影响较大,可在一定程度上预测产氚包层中氚的滞留行为。     

大气氚释放剂量评价模式验证北大核心CSCD

基于秦山核电厂2014—2016年气载流出物氚的排放数据,采用联合国原子辐射影响科学委员会(UNSCEAR)推荐的比活度模型评价了秦山核电基地氚所致公众辐射剂量。并与同期秦山核电基地周围环境氚监测数据评价公众辐射剂量结果进行比较。基于流出物排放的评价结果与基于环境监测数据的评价结果相差不大,在同一水平。推荐在进行气载氚所致公众辐射剂量评价时采用该比活度模型。     

Influence of surface condition on deuterium release from Li2TiO3 pebble

Lithium titanate (Li₂TiO₃) pebbles were irradiated with D₃⁺ ions with energy of 5.0 keV, and the amounts of retained deuterium in the pebbles were measured by thermal desorption spectroscopy. In this research the irradiation/heating cycles were carried out repeatedly in order to investigate the influence of surface condition on deuterium release from Li₂TiO₃. The composition ratio of Li decreased with the increase of the number of the irradiation/heating cycle. Then, the desorption peaks of the gases contained deuterium atoms were shifted to higher temperature region, and the amount of desorbed gases in forms of water tended to increase. In addition, we carried out other experiments for the comparison. Comparing these results, we considered that the increase of the defects created by the irradiation was more responsible for the change in the desorption behavior by the irradiation/heating cycles than the lithium depletion. These results suggest that the tritium recovery efficiency would decrease with the increase of the defects and the damages especially at the low temperature region during the operation.     

Isothermal release of tritium from neutron-irradiated Li17Pb83

Isothermal release experiments were carried out to study the tritium recovery from lithium-lead alloy Li₁₇Pb₈₃ in which tritium was produced by irradiation with thermal neutrons. The experimental results indicate that the tritium recovery was incomplete within two hours at 200 °C. At temperatures above the melting point, the tritium release rates have been significantly increased and found to be controlled by the diffusion in the alloy. The determined diffusion coefficients of tritium in the alloy are 6.6 × 10⁻⁶, 7.8 × 10⁻⁶ and 9.5 × 10⁻⁶ cm²/s at 300, 400 and 500°C, respectively.