Safety assessment of a lithium target

This paper addresses the safety assessment of the lithium target of the International Fusion Materials Irradiation Facility (IFMIF) through evaluating the most important risk factors related to system operation and verifying the fulfillment of the safety criteria. The hazard assessment is based on using a well-structured Failure Mode and Effect Analysis (FMEA) procedure by detailing on a component-by-component basis all the possible failure modes and identifying their effects on the plant. Additionally, a systems analysis, applying the fault tree technique, is performed in order to evaluate, from a probabilistic standpoint, all the relevant and possible failures of each component required for safe system operation and assessing the unavailability of the lithium target system. The last task includes the thermal–hydraulic transient analysis of the target lithium loop, including operational and accident transients. A lithium target loop model is developed, using the RELAP5/Mod3.2 thermal–hydraulic code, which has been modified to include specific features of IFMIF itself. The main conclusions are that target safety is fulfilled, the hazards associated with lithium operation are confined within the IFMIF security boundaries, the environmental impact is negligible, and the plant responds to the simulated transients by being able to reach steady conditions in a safety situation.     

Tests of the cryogenic target for lithium and hydrogen isotopes extraction from the chamber of T-11M tokamak without its venting

T-11M lithium program is focused on a solution of technological issues of a steady-state tokamak with liquid lithium plasma facing components (PFC). Lithium, collected by the chamber wall of such tokamak is able to capture a considerable amount of tritium, which is unacceptable. In order to restrict the level of lithium deposited on the chamber wall and captured tritium it was suggested early to use a cryogenic target technique. Such target placed in the plasma of glow discharge (GDH, He or Ar) during the tokamak conditioning can play the role of collector of lithium and tritium atoms which were sputtered by GD bombardment of the wall. The collected lithium and tritium can be evacuated mechanically together with target from tokamak chamber through vacuum lock without venting. Cryogenic target, cooled by liquid nitrogen (LN), was installed in the T-11M and tested in different modes of wall conditioning and tokamak operations. The maximum speed of the lithium collection during GDH was 3.5 mg/h, that corresponds “to contamination” of wall by lithium during approximately 200 regular shots of T-11M which are equivalent to two-week regular operations. It was established that considerable part of lithium was collected in ionized state. On this basis it can be suggested the creation in tokamak chamber an equivalent ionic pump for extraction both lithium and tritium from chamber without venting during regular tokamak operation.     

The sensitivity of the PIGE analytical technique

The aim of this work is to provide a basis for the estimation of fluorine and lithium contributions to the background of gamma-ray spectra and of their influence on PIGE (Proton-Induced Gamma-Ray Emission) detection limits of lithium, boron, fluorine and sodium. Results for yields and background contributions of lithium and fluorine were obtained from measurements using samples of known composition. Detection limits were determined by extrapolation of those measurements and by calculations based on the inferred background values. Yields were converted in concentrations using the ERYA (Emitted Radiation Yield Analysis) code. This code divides the sample in sublayers parallel to its surface for the integration of the partial gamma-ray yields along the depth of the target.     

The European contribution to the development and validation activities for the design of IFMIF lithium facility

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility.The paper gives an overview of the status of the activities and of the main outcomes achieved so far.     

Conceptual design of the liquid metal laboratory of the TECHNOFUSION facility

The application of liquid metal technology in fusion devices requires R&D related to many phenomena: interaction between liquid metals and structural material as corrosion, erosion and passivation techniques; magneto-hydrodynamics; free surface fluid-dynamics and any other physical aspect that will be needed for their safe reliable operation. In particular, there is a significant shortage of experimental facilities dedicated to the development of the lithium technology. In the framework of the TECHNOFUSION project, an experimental laboratory devoted to the lithium technology development is proposed, in order to shed some light in the path to IFMIF and the design of chamber's first wall and divertors. The conceptual design foresee a development in two stages, the first one consisting on a material testing loop. The second stage proposes the construction of a mock-up of the IFMIF target that will allow to assess the behaviour of a free-surface lithium target under vacuum conditions. In this paper, such conceptual design is addressed.     

Assessment for explosion strength of liquid lithium droplet coolant interaction

Liquid lithium served as plasma-facing material for limiters and divertor target plates are useful for the improvement of plasma performances. However, the liquid lithium water interaction is a main concern for the safety of fusion reactors. The assessment of explosion strength is significant to the risk analysis of application of liquid lithium. An experiment of liquid lithium droplet coolant interaction has been conducted. The experimental result indicates that the mass and initial temperature of liquid lithium droplet has a significant impact on the explosion strength. The peak pressure rises with the increase of the mass and the temperature of liquid lithium. Based on the theory of shock wave overpressure and experimental data, an overpressure formula for liquid lithium droplet coolant interaction is fitted by the least-square method. The optimum values of undetermined coefficients are obtained and the model to assess explosion strength of liquid lithium droplet coolant interaction is established. The model is verified by experimental data and proved to be applicable. It reflects the influence of the mass and initial temperature of lithium on the explosion strength and also provides a novel method for the evaluation of explosion strength of liquid lithium droplet coolant interaction.     

Wetting properties of liquid lithium on select fusion relevant surfaces

Research into lithium as a plasma facing component material has illustrated its ability to engender low recycling operation at the plasma edge leading to higher energy confinement times. Introducing lithium into a practical fusion device would almost certainly require the lithium to be flowing to maintain a clean lithium surface for gettering. Several conceptual designs have been proposed, like the LiMIT concept of UIUC (Ruzic, 2011). Critical to the implementation of these devices is understanding the interactions of liquid lithium with various surfaces. For a device that relies on thermoelectric magnetohydrodynamic drive, such as the LiMIT concept, two of the critical interactions are the wetting of materials by lithium, which may be characterized by the contact angle between the lithium and the surface, and the relative thermopower between lithium and potential substrate materials.Experiments have been performed into the contact angle of liquid lithium droplets with various surfaces, as well as methods to decrease the contact angle of lithium with a given surface. The contact angle, as well as its dependence on temperature was measured. For example, at 200 °C, tungsten registers a contact angle of 130°, whereas above its wetting temperature of 350 °C, the contact angle is less than 80°. Glow discharge cleaning of the target surface as well as evaporation of a thin layer of liquid lithium onto the surface prior to performing wetting measurements were both found to decrease the wetting temperature.     

Main Directions and Recent Test Modeling Results of Lithium Capillary-Pore Systems as Plasma Facing Components

At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing materials. A solid CPS filled with liquid lithium will have a high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stressinduced cracking in steady state and during plasma transitions to provide the normal operation of divertor target plates and first-wall protecting elements. These materials will not be the sources of impurities inducing an increase of Zeef and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS under simulating conditions of plasma disruption on a hydrogen plasma accelerator MK-200 [-(10 - 15) MJ/m^2, - 50 μs] have been performed. The formation of a shielding layer of lithium plasma and the high stability of these systems have been shown. The new lithium limiter tests on an up-graded T-11M tokamak (plasma current up to 100 kA, pulse length -0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, deposited power of the limiter are investigated in these experiments. The first results of experiments are presented.     

Computational modelling of the IFMIF lithium target

The present work is devoted to the computational modelling of the process of beam action on a lithium target. The aim of the investigation is to determine the maximum values of temperature and pressure as well as general pattern of the process. The analysis is based on the compressible Euler equations with the stiffened gas equation of state with parameters corresponding to lithium. The energy influx allocation caused by the beam interaction with the target is described by the source term in the energy balance law. The formulated problem is solved numerically by a high-resolution Godunov-type method. The obtained results show a moderate rise in the lithium temperature and relatively large pressure variations.     

On the thermomechanical behavior of the European target assembly design of IFMIF-EVEDA lithium test loop under start-up transient scenarios

Uncoupled thermomechanical transient analyses have been carried out to investigate the behavior of IFMIF-EVEDA lithium test loop bayonet backplate target assembly under two selected start-up transient operational scenarios. The first transient scenario considered foresees that the target assembly, starting from the initial uniform temperature of 50 °C, is heated up uniquely by convective heat transfer with lithium, flowing from inlet to outlet nozzle at its reference nominal temperature and pressure, until its nominal steady state thermal field distribution is reached. The second transient scenario foresees, more realistically, that the target assembly, starting from the uniform temperature of 50 °C, is initially warmed-up by electric heaters mounted onto its main accessible surfaces and, subsequently, by convective heat transfer with lithium reference flow, until nominal steady state conditions are reached. Heaters have been supposed to operate in an on/off stepwise mode, resulting to be alternatively switched on and off in order to allow the target assembly thermal field to grow up minimizing thermal gradients. To this purpose, a parametric analysis has been performed to realistically assess, for each electric heater, its heat flux and duty-cycle. Numerical results obtained are presented and critically discussed.     

Progress in development and application of lithium based components for Tokamak

Experiments with lithium plasma facing components (PFCs) show promising results for the operation of hot plasma facilities and the general improvement of plasma parameters. The design and development of new tokamak plasma facing material (PFM) based on lithium capillary porous systems (CPS) are described in this paper.The recent progress in the development of limiters with different kinds of CPS is relevant for protection of tokamak PFCs from damage under normal operation, ELMs and disruptions. New PFM eliminates the lithium flux into plasma, its pollution and lithium accumulation.Here we present an overview of the design and the experimental tests of the liquid lithium limiters. These limiters are based on CPS with hard matrix from stainless steel mesh, molybdenum and tungsten. Different types of limiter have been taken into account: the horizontal and vertical rail type limiters with passive and active cooling for investigation the possibility to provide the closed lithium circulation in tokamak chamber; the ring CPS-based limiter for investigation of lithium behavior in limiter scrape-off layer (SOL).Here we also present the preliminary results of the application of the cryogenic techniques for lithium removal from the chamber wall after operation in hot plasma.     

流动锂液帘对堆芯等离子体影响的评估(英文)

运用零维模型评估了流动液态锂幕帘作为聚变实验增殖堆工程概要设计 (FEB-E) 第一壁对聚变等离子体的影响。得到了锂液帘工作温度对堆芯有效平均等离子体电荷?Zeff?,燃料稀释以及聚变功率之间的关系。表明在正常工作情况下,液态锂的蒸发对?Zeff?的影响不是很严重,但对燃料稀释和聚变功率的影响却较为敏感。作为一个例子,对较高功率密度的反剪切位形聚变实验增殖堆FEB-E设计方案 II,计算了液帘的流速与它表面最大温升的关系,结果表明:即便0.5m/s的低速流动液帘第一壁, 蒸发对聚变等离子体的影响也甚微。     

Functions of intrinsic magnetic field line structures on lithium ion transport in the LHD plasma periphery

Plasma discharge operation with lithium coating suggests that the lithium effectively control neutral particles in the plasma periphery, which can lead to improvement of plasma parameters. The effect of lithium coating on the large helical device (LHD) for a closed helical divertor configuration is discussed from viewpoints of neutral particle and impurity ion transport in the plasma periphery. It shows that the closed helical divertor configuration can enhance the neutral particle density in the divertor region, which is enough to achieve efficient particle control, and that it can effectively confine neutral lithium atoms near divertor plates. A one-dimensional impurity (lithium) ion transport analysis along magnetic field lines on divertor legs indicates that the friction force due to the plasma flow from the main plasma is dominant over the thermal force caused by the temperature gradient on the divertor legs, which prevents lithium ion contamination in the main plasma and excessive cooling of the plasma temperature in an ergodic layer. The analysis shows that the lithium coating is compatible with LHD plasma discharge operation for the closed helical divertor configuration.     

Experimental and Analytical Investigation of the Fast Reactor Passive Shutdown System: LIM

To enhance the inherent safety of the fast reactors, the lithium injection module (LIM) is proposed for inherent ultimate shutdown instead of conventional scram rod. LIM is composed of a refractory metal envelope in which 95% enriched ⁶Li is enclosed. In case the core outlet temperature exceeds the melting point of the freeze seal, ⁶Li is injected by a pneumatic mechanism from the top to bottom chamber to achieve negative reactivity insertion. This concept is attractive because the actuator has no moving parts and depends on the reliable physical property. In this paper, experimental and analytical verification of the LIM performance are presented. Real-time monitoring of LIM during reactor operation has been discussed as well.     

核电厂运行事件分析方法

核电厂运行事件数据是安全监督和运行经验反馈的重要依据。文中针对核安全监督和运行经验反馈的要求,介绍了运行事件的分析处理方法。     

自由表面液态锂偏滤器靶板物理过程研究

本文建立了一种高温液态锂蒸发、锂蒸气云等离子体运动、它对入射等离子体粒子屏蔽和锂蒸气云等离子体内的光子辐射和输运的综合物理模型.导出了与温度相关的蒸发功率.研究了静态液态锂表面在10 MJ/m2,1 ms高脉冲表面热负荷作用下考虑蒸发和不考虑蒸发两种情况下靶板温升并作了比较.结果表明定常自由表面液态锂靶板也可以取出大量表面热负荷.最后计算了入射到偏滤器靶室中的高能α粒子和弱相对论性电子在锂蒸气云等离子体中的能量沉积.     

New separator prepared by electron beam irradiation for high voltage lithium secondary batteries

Grafted separators, for which poly(ethylene glycol) borate acrylate (PEGBA) was grafted onto polyethylene (PE) separator, were newly prepared by electron beam irradiation. The grafted separators were characterized by FT-IR, energy dispersive X-ray spectrometer (EDS). The morphological changes of the grafted separators were investigated by scanning electron microscopy (SEM). The degree of grafting was increased with irradiation doses. The ionic conductivity of the grafted separator showed the highest value of 6.24 × 10⁻⁴ S cm⁻¹ at 10 kGy. In addition, its lithium ion transference number and electrochemical stability were enhanced to 0.53 and 4.8 V, respectively owing to anion trapping effect of the grafted unit. The Li ion cells using the grafted separator showed better cycle performances than that using conventional PE separator at various C-rates and high voltage operation conditions. It is suggested that this grafted separator can be a promising candidate for high voltage operation of lithium secondary batteries.     

The production of tritium in fission and fusion reactors

Conclusions It follows from the published data on the selection of lithium materials for the breeding zone of a thermonuclear reactor [43–45] that the tritium production involves large-scale radiochemical installations for the conversion of the irradiated material and the extraction of the pure isotope which is present in extreme dilution. When such radiochemical installations are built, a set of basic technological problems must be solved (continuously maintaining the purity of the lithium materials, removing the radiolysis products, stabilization of the physicochemical properties of the material, removing the corrosion products of the materials used for the construction); one must also solve problems related to personnel safety and the protection of the environment from the emission of the radioactive isotope. Problems related to the construction of apparatus for the processes occurring at a high temperature must be solved.When from this viewpoint lithium materials and their irradiation conditions are selected, one must recall that solid lithium materials (ceramics, alloys) and melts of salts have advantages over metallic lithium for the very reason that they eliminate operations involving a flammable material.The use of solid lithium materials or of molten systems as a neutron-absorbing material in the breeding zone of the blanket depends upon the characteristics of the tritium-breeding zone in terms of neutron physics and upon the thermophysical parameters.At the present time nuclear physics research, physicochemical research, and radiochemical research is done on systems with irradiated lithium materials. Methods are developed for separating the tritium which is present in various states at relatively low concentrations. The goal of this work is not only to obtain answers to the technological problems with the greatest reliability but also to effectively select a lithium material suitable for building the tritium cycle of a thermonuclear station as a whole.In the ensuing stage of the research work, when the blanket zone will be drafted, one must evaluate the various conditions of operation of the breeding zone with metallic lithium, diluted fluorine systems, and solid ceramic materials. At the present time, the selection of the blanket design and the construction of the breeding zone, as well as the selection of the lithium material providing the highest tritium production coefficient, are the basic problems among the engineering problems of thermonuclear fusion and the technological and economical aspects of the forthcoming developments.Translated from Atomnaya Énergiya, Vol. 44, No. 5, pp. 440–446, May, 1978.     

Overview of the Preliminary Safety Analysis of the National Ignition Facility

The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF.     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.