The effect of a micro bubble dispersed gas phase on hydrogen isotope transport in liquid metals under nuclear irradiation

The present work intend to be a first step towards the understanding and quantification of the hydrogen isotope complex phenomena in liquid metals for nuclear technology. Liquid metals under nuclear irradiation in, e.g., breeding blankets of a nuclear fusion reactor would generate tritium which is to be extracted and recirculated as fuel. At the same time that tritium is bred, helium is also generated and may precipitate in the form of nano bubbles. Other liquid metal systems of a nuclear reactor involve hydrogen isotope absorption processes, e.g., tritium extraction system. Hence, hydrogen isotope absorption into gas bubbles modelling and control may have a capital importance regarding design, operation and safety.Here general models for hydrogen isotopes transport in liquid metal and absorption into gas phase, that do not depend on the mass transfer limiting regime, are exposed and implemented in OpenFOAM® CFD tool for 0D–3D simulations. Results for a 0D case show the impact of a He dispersed phase of nano bubbles on hydrogen isotopes inventory at different temperatures as well as the inventory evolution during a He nucleation event. In addition, 1D and 2D axisymmetric cases are exposed showing the effect of a He dispersed gas phase on hydrogen isotope permeation through a lithium lead eutectic alloy and the effect of vortical structures on hydrogen isotope transport at a backward facing step.Exposed results give a valuable insight on current nuclear technology regarding the importance of controlling hydrogen isotope transport and its interactions with nucleation event through gas absorption processes.     

Phase-field modeling of gas bubbles and thermal conductivity evolution in nuclear fuels

A phase-field model was developed to simulate the accumulation and transport of fission products and the evolution of gas bubble microstructures in nuclear fuels. The model takes into account the generation of gas atoms and vacancies, and the elastic interaction between diffusive species and defects as well as the inhomogeneity of elasticity and diffusivity. The simulations show that gas bubble nucleation is much easier at grain boundaries than inside grains due to the trapping of gas atoms and the high mobility of vacancies and gas atoms in grain boundaries. Helium bubble formation at unstable vacancy clusters generated by irradiation depends on the mobilities of the vacancies and He, and the continuing supply of vacancies and He. The formation volume of the vacancy and He has a strong effect on the gas bubble nucleation at dislocations. The effective thermal conductivity strongly depends on the bubble volume fraction, but weakly on the morphology of the bubbles.     

Bubble nucleation process in a metal tritide modeled using a cellular automaton

The modeling of the ³He bubble nucleation phase that occurs during the aging of metal tritides such as palladium tritide is undertaken using a cellular automaton describing the material at the atomic scale. In that model, using simple rules of cell state change, the physical phenomena involved in the bubble nucleation, namely tritium diffusion, formation of ³He by radioactive tritium decay, ³He diffusion, ³He self-trapping, bubble growth, influence of pre-existing trapping sites such as vacancies, are taken into consideration. Calculations steps are related to physical time through the decay of tritium atoms into ³He atoms which is characterized by a half-life of 12.32 years.That work has shown that the bubble density and distribution are almost stable within a few days of aging – typically 5–10 –, whatever the input parameters, the system making created bubbles to grow thereafter instead of creating new ones. The reached bubble density is very dependent on the ³He mobility – related to temperature for instance – during the first days of aging. The smaller it is, the higher the density. With a larger density of trapping sites, bubbles appear earlier and their density is higher. These results are discussed and compared to available experimental and theoretical works on palladium tritide.     

充氚不锈钢中氦泡长大行为的加热效应

以室温贮存经历的充氚不锈钢试样为研究对象,计算了充氚不锈钢中氚、氦浓度的深度分布,利用透射电镜观察了充氚不锈钢在加热过程中氦泡的演化行为。结果表明:在氚压0.131MPa、780℃充氚8h后,不锈钢中氚在深度方向分布均匀,平均浓度为110μL/L;在空气室温环境下放置6a后,不锈钢中氚衰变的氦浓度在深度方向分布均匀,平均浓度为60μL/L;对充氚不锈钢加热处理后,在550℃/1h时效即可观察到氦泡;在950℃/1h和1050℃/1h时效时,氦泡明显长大,大的可达100nm,小的可达30nm,在晶界、晶内和位错处均可见氦泡。     

钯的氚老化效应

钯因其显著的氢同位素效应、抗毒化及良好的固氦特性,已广泛应用于氚工艺中。随着工作时间的延长,钯中衰变产生的³He将影响其应用性能。文章就氚老化对钯的p-C-T曲线、力学性能、微观结构的影响,及³He在钯中的微观行为进行了综述。氚老化导致坪压降低、坪斜增加、氚尾增加、力学性能下降。氚衰变产生的³He聚集形成氦泡,导致晶格膨胀,且在钯中形成自间隙原子簇、位错、位错环等结构缺陷。     

中国HCCB-TBM氚增殖单元三维多物理场耦合氚输运模拟分析

中国氦冷固态实验包层模块（HCCB-TBM）将在国际热核聚变实验堆（ITER）上安装测试,以验证其氚增殖能力与核热移出能力。HCCB-TBM中的氚输运与流体的传热和传质、氢同位素交换、结构材料的SORET效应密切相关。考虑以上物理因素,基于商业软件COMSOL完成了HCCB-TBM氚增殖单元多物理场耦合的氢同位素输运模拟分析。分析结果表明：球床吹洗气体中含氢有助于抑制氚渗透损失;当吹洗气体含氢浓度为4.66×10^-2 mol/m³时,产生约13.2倍的氚渗透阻止效应。     

Fuel hydrogen retention of tungsten and the reduction by inert gas glow discharges

Polycrystalline tungsten was exposed to deuterium glow discharge followed by He, Ne or Ar glow discharge. The amount of retained deuterium in the tungsten was measured using residual gas analysis. The amount of desorbed deuterium during the inert gas glow discharge was also measured. The amount of retained deuterium was 2–3 times larger compared with a case of stainless steel. The ratios of desorbed amount of deuterium by He, Ne and Ar glow discharges were 4.6, 3.1 and 2.9%, respectively. These values were one order of magnitude smaller compared with the case of stainless steel. The inert gas glow discharge is not suitable to reduce the fuel hydrogen retention for tungsten walls. However, the wall baking with a temperature higher than 700 K is suitable to reduce the fuel hydrogen retention. It is also shown that the use of deuterium glow discharge is effective to reduce the in-vessel tritium inventory in fusion reactors through the hydrogen isotope exchange.     

高温气冷堆闭式布雷顿间接循环中氚的来源及其影响

氚是氢的放射性同位素,影响环境和人体健康.目前,全球自然界中的氚主要来自人类的核活动.因此,需研究核反应堆中氚的来源及其影响.在高温气冷堆中,氚是一回路放射性的主要来源之一.由于高温气冷堆堆芯温度较高,不能忽视一回路中氚向外界和二回路渗透造成的污染问题.文章阐述了氚的物理和化学特性,高温气冷堆闭式布雷顿间接循环中氚的生成来源和释放途径,分析了氚对设备材料力学性能的影响,介绍了氚向环境释放的限值、控制措施及防止氚渗透的方法.     

Experimental study of efficiency of natural oxide layers for reduction of tritium permeation through Eurofer 97

Tritium permeation from the breeder through the helium coolant is a fundamental safety issue in the design of the HCLL (Helium coolant lithium lead) blanket system. The permeation of hydrogen isotopes through Eurofer in different conditions was deeply studied in the past, demonstrating that it is necessary to reduce this amount using tritium permeation barriers (TPB). A strong effort has been made to select the best technological solution for the realisation of tritium permeation barriers on complex structures not directly accessible after the completion of the manufacturing process, but after many years of activity for the qualification of different materials acting as TPBs it was demonstrated that these technologies are not yet mature for nuclear applications. An easier solution was identified in the nucleation and growth of natural oxides on the helium-exposed surface of cooling system components. The major objective of this work is the evaluation of the Permeation Reduction Factor (PRF) of natural oxides on Eurofer steel adding a known content of water and hydrogen to argon, used in substitution of helium. The PRF was measured on disk shaped specimens, in gas phase, using the PERI II apparatus, at a temperature of 550 °C. The oxide layer was produced in situ, in a well-defined range of hydrogen on water ratios. The obtained results are presented and discussed.     

充氚不锈钢中氦行为的PAL和TEM研究

对充氚和未充氚的抗氢-2(HR-2)不锈钢样品进行退火处理,利用正电子湮没寿命谱(PAL)以及透射电镜(TEM)等技术探讨不锈钢中氦和微缺陷的相互作用。未充氚样品中,退火温度对缺陷态的影响主要表现为偏聚物在晶界的析出。充氚样品实验中,退火温度小于300℃时,充氚不锈钢中的He原子主要通过自捕获机制在晶内缺陷处聚集成泡;热处理温度为300～600℃时,充氚不锈钢中的He原子主要通过热迁移的方式迁移至晶界导致晶界宽化,但晶界处无明显的He泡形成;热处理温度大于600℃时,热平衡空位开始发挥作用,与聚集在晶内缺陷处的He原子结合形成He泡,且随退火温度的升高,He泡有明显聚合长大的现象。     

Design and development of hydrogen isotope sensor in liquid PbLi

Indian LLCB – TBM uses liquid Lead-Lithium (Pb-Li) as tritium breeder, neutron multiplier and coolant. Tritium bred in liquid PbLi stream has to be recovered by tritium extraction system. Therefore, a reliable sensor with quick response time for measurement of hydrogen isotope is essential.A hydrogen isotope sensor in liquid Pb-Li, based on permeation of hydrogen isotopes through metal (sensor material) is designed. The capsule shaped sensor, made of iron membrane coated with Pd from inside (downstream side), allow hydrogen isotope to permeate through it. The design work mainly includes the selection of proper material, its thickness and surface conditions, which is to be supported by numerical calculations for optimization of maximum permeation flux, fast response and fabricability. The numerical calculation utilizes a physical model having recombination of two hydrogen isotope atoms at the surface and atomic diffusion through the bulk. In this work, design calculations based on numerical simulation and fabrication procedure of the hydrogen isotope sensor are presented.     

Fission gas swelling and long-range migration at low temperatures

Fission gas behavior at temperatures below ~ 1100°C is assumed to consist of gas bubble nucleation and coalescence through random motion of the bubbles and then complete bubble destruction by subsequent fission events. A model is proposed describing the gas behavior based on a modified form of Van der Waals gas law for very small bubbles. A bubble is assumed to move inversely to the cube of its radius. Long range migration of gas bubbles to grain boundaries is also predicted and the swelling due to gas motion in the grain boundaries is calculated.     

Mathematical models for tritium permeation analysis in liquid metal flows with helium bubbles

In LIBRETTO-2 test, evidence was obtained that helium bubbles nucleated and grew in the neutron irradiated PbLi probes. If such phenomenon occurs inside liquid metal (LM) breeding blanket channels, the study of its effect on tritium permeation and heat transfer in the near wall region will acquire utmost importance. The T4F research group has developed in the past a nucleation, growth and transport model for helium bubbles in LM flows, as well as a tritium transport model in such a multi-fluid system. In the present study, we are focused on the near-wall region analysis in order to obtain a wall function that allow reproducing the tritium permeation with coarse meshes and, hence, reduce the computational time. First, we perform some detailed CFD simulations of the near-wall region where bubbles might be attached. In these simulations, tritium diffusion processes as well as tritium recombination and dissociation are modelled. The analysis of such simulations allows us to further understand the complex phenomena and justify the use of simplified models. As a result, a new model for tritium transport across a LM–solid interface partially covered by helium bubbles is developed, implemented and validated. This simplified model can be seen as a wall function for the CFD simulation which substantially reduces computational time.     

Study on the tritium behaviors in the VHTR system. Part 1. Development of tritium analysis code for VHTR and verification

A tritium permeation analyses code (TPAC) has been developed at Idaho National Laboratory (INL) by using MATLAB SIMULINK package for analysis of tritium behaviors in the VHTR integrated with hydrogen production and process heat application systems. The modeling is based on the mass balance of tritium-containing species and hydrogen (i.e., HT, H₂, HTO, HTSO₄, and TI) coupled with a variety of tritium source, sink, and permeation models. The code includes: (1) tritium sources from ternary fission and neutron reactions with ⁶Li, ⁷Li ¹⁰B, and ³He; (2) tritium purification system; (3) leakage of tritium with coolant; (4) permeation through pipes, vessels, and heat exchangers; (5) electrolyzer for high temperature steam electrolysis (HTSE); and (6) isotope exchange for SI process. Verification of the code has been performed by comparisons with the analytical solutions, the experimental data, and the benchmark code results based on the Peach Bottom reactor design. The results showed that all the governing equations are well implemented into the code and correctly solved. This paper summarizes all the background, the theory, the code structures, and some verification results related to the TPAC code development at INL.     

Permeation Analysis of Tritium Through the Vessel Wall of a Titanium Hydride Storage Container

Abstract

A preliminary design for a stainless steel vessel for the long-term storage of hydrogen isotopes has been proposed. The immobilised hydrogen, as a titanium hydride, could be stored in a stainless steel vessel for this application. The vessel, as a primary package, is designed to form titanium hydride and to contain the hydrogen isotopes and helium-3 produced from the decay of tritium. In order to predict the possibility of contamination and the deterioration of the mechanical properties, a numerical diffusion analysis calculation of the hydrogen isotopes and helium inside the stainless steel vessel was carried out. Numerical results showed that a negligible amount of tritium would be released by permeation through a 0.7 cm thick vessel wall at normal conditions over the entire period of the storage. When the vessel is heated up to a temperature of 600^°C for the routine conditions of activation or exothermic hydriding, tritium loss or contamination would be of little concern. However, if the vessel were exposed to fire conditions with a temperature of 800^°C, permeation of hydrogen through the vessel wall would result in a serious increase in the amount of tritium escaping, in a very short time.     

Experimental design of tritium extraction loop from lead lithium eutectic

Effective tritium breeding achievable in Test Blanket Module (TBM) is a major issue for sustainable fusion energy program. Equally important is tritium extraction to recover and recycle tritium back to fusion reactor. Tritium extraction from lead lithium is much more complicated than from purge gas due to low tritium extraction efficiency in transfer step to gas phase and the limitations imposed on space and lead lithium inventory in port cell. Earlier investigations do suggest the preference of packed columns over bubble columns. Theoretical models based on axial dispersion plug flow in liquid and gas proposed for bubble columns and packed columns are reinvestigated for different boundary conditions.This paper highlights the critical issues of experimental design based on tritium extraction efficiency and its impact on recovery loop. Steady state closed loop for absorption and stripping of hydrogen isotopes using inert gas is designed along with the associated auxiliaries.     

Nucleation,growth and transport modelling of helium bubbles under nuclear irradiation in lead–lithium with the self-consistent nucleation theory and surface tension corrections

Helium (He) nucleation in liquid metal breeding blankets of a DT fusion reactor may have a significant impact regarding system design, safety and operation. Large He production rates are expected due to tritium (T) fuel self-sufficiency requirement, as both, He and T, are produced at the same rate. Low He solubility, local high concentrations, radiation damage and fluid discontinuities, among other phenomena, may yield the necessary conditions for He nucleation. Hence, He nucleation may have a significant impact on T inventory and may lower the T breeding ratio.A model based on the self-consistent nucleation theory (SCT) with a surface tension curvature correction model has been implemented in OpenFOAM^® CFD code. A modification through a single parameter of the necessary nucleation condition is proposed in order to take into account all the nucleation triggering phenomena, specially radiation induced nucleation. Moreover, the kinetic growth model has been adapted so as to allow for the transition from a critical cluster to a macroscopic bubble with a diffusion growth process.Limitations and capabilities of the models are shown by means of zero-dimensional simulations and sensitivity analyses to key parameters under HCLL breeding unit conditions. Results provide a good qualitative insight into the helium nucleation phenomenon in LM systems for fusion technology and reinforces the idea that nucleation may not be a remote phenomenon, may have a large impact on the system's design and reveals the necessity to conduct experiments on He cavitation.     

氦离子辐照烧结碳化硅损伤效应研究

利用中国科学院近代物理研究所320 kV高压平台提供的氦离子辐照烧结碳化硅,辐照温度从室温到1 000 ℃,辐照注量为1015~1017 cm-2。辐照完成后,进行退火处理,然后开展透射电子显微镜、拉曼光谱、纳米硬度和热导率测试。研究发现,烧结碳化硅中氦泡形核阈值注量低于单晶碳化硅。同时,氦泡形貌和尺寸与辐照温度、退火温度有关。另外,对辐照产生的晶格缺陷、元素偏析进行了研究。结果表明,辐照产生了大量的缺陷团簇,同时氦泡生长也会发射间隙子,在氦泡周围形成间隙型位错环。在晶界处,容易发生碳原子聚集。辐照导致材料先发生硬化而后发生软化,且热导率降低。     

Fusion tritium research facilities in KAERI

Korea Domestic Agency (KODA) is developing a nuclear fusion fuel storage and delivery system (SDS) as one of the Korean procurement packages. Korea Atomic Energy Research Institute (KAERI) is operating the following basic scientific research laboratories for an SDS and tritium supply study: a metal hydride bed preparation laboratory, hydrogen isotope recovery and delivery performance test rig, in-bed calorimetry performance test rig, and tritium shipping container integrity test facility. Furthermore, the development of a test blanket module (TBM) is required to test and validate the design concept of tritium breeding blankets relevant to fusion power plants. KAERI is also operating the following laboratories for the TBM research, such as a tritium extraction performance test rig, High-flux Advanced Neutron Application Reactor (HANARO), and Experimental Loop for Liquid Breeder (ELLI).