Measurement of implanted helium particle transport by a flowing liquid lithium film

Due to its low atomic number, low sputtering yield, high sputtered ion fraction and excellent thermal properties, liquid lithium has been proposed as a potential candidate for advanced plasma-facing components (PFC). Using a liquid material opens the possibility of a continuously flowing, self-regenerating plasma-facing surface with a small residence time. This would allow such component to handle very high heat loads that are expected. There are, however, multiple unanswered questions regarding how such a liquid PFC would interact with the plasma in the reactor. The issue of particle control is critical, and it can be a factor to determine the feasibility of these advanced concepts. Hydrogen and helium are important in this regard: hydrogen transport by a flowing PFC impacts the reactor fuel recycling regime and tritium inventory; helium transport can help quantify ash removal by the flowing PFC. The flowing liquid-metal retention experiment (FLIRE) was built at the University of Illinois to answer some of the questions regarding particle transport by flowing liquid films exposed to plasmas. Experimental results regarding helium transport by a flowing lithium film after irradiation with an energetic He ion beam are presented in this work. Retained fraction values up to 2% were measured for the experimental conditions, and the retention was found to increase linearly with implanted ion energy. A pure diffusion model was used to describe the helium transport by the Li film, and it was found that such model predicts a diffusion coefficient of (2.8 ± 0.6) × 10⁻¹¹ m²/s, based on the experimental retention measurements. Preliminary evidence of long-term trapping of helium will also be presented.     

Hydrogen and helium recycling from stirred liquid lithium under steady state plasma bombardment

For improved core performance via edge plasma-wall boundary control, solid and liquid lithium has been used as a plasma-facing material in a number of confinement experiments over the past several decades. Unfortunately, it is unavoidable that lithium is saturated in the surface region with implanted hydrogenic species as well as oxygen-containing impurities. For steady state operation, a flowing liquid lithium divertor with forced convection would probably be required. In the present work, the effects of liquid stirring to simulate forced convection have been investigated on the behavior of hydrogen and helium recycling from molten lithium at temperatures up to ∼350 °C. Data indicate that liquid stirring reactivates hydrogen pumping via surface de-saturation and/or uncovering impurity films, but can also induce helium release via surface temperature change.     

Management of a water leak on actively cooled fusion devices

ITER will be the most important machine equipped with actively cooled plasma facing components (PFCs). In case of abnormal events during a discharge, the PFC will be submitted to localized transient phenomena (high power densities, run away electrons, etc.), leading, in the worst case, to the degradation of the PFC wall and possibly to a water leak. In any case, a leak will have important consequences for the PFCs and equipment located in the vacuum vessel or connected to the ports such as seals, pumping systems or diagnostics.Considerable experience of these events has been gained at Tore Supra over a period of more than 10 years [J.J. Cordier, Ten years of maintenance on Tore Supra actively cooled components, in: Proceedings of the 21th Symp. of Fusion Technology (SOFT), Madrid, Spain, September, 2000.], which will be useful for the next step machines.This paper describes for each leak size type the procedures for maintaining save conditions in the vacuum vessel. It also presents the methods used at Tore Supra to drain-off the primary loop circuits and to identify the leaky PFC.     

Vacuum pumping requirement considerations for future fusion devices

The vacuum pumping requirements for a fusion device are dictated by a number of factors, including the materials used in construction of the device, the cleaning and conditioning techniques implemented, the operating conditions and device temperature, the plasma characteristics, and the fuel gases and impurities retained and released by the plasma-facing components (PFCs). In an attempt to derive guidelines for determining the vacuum pumping requirements of a generic fusion device, a study was undertaken to examine the vacuum pumping capabilities of existing large fusion devices, to review the cleaning and conditioning techniques now in use, and to catalog pertinent vacuum equipment that is now available or anticipated soon. In a survey of six large fusion devices [ASDEX, DIII-D, the Joint European Torus (JET), JT-60, the Tokamak Fusion Test Reactor (TFTR), Tore Supra], information was collected on PFC materials, cleaning and conditioning techniques, device operating temperatures, and vacuum pumping system characteristics and on the surface conditions and partial pressures necessary to achieve high-purity plasma discharges in each device. The results of the survey are reviewed to determine how the various factors affect vacuum pumping requirements.

The use of tritium in future fusion devices is expected to create a new set of problems for the vacuum pumping systems of these devices. These problems are addressed, and some possible solutions for the pumping and gas handling systems are identified. New vacuum equipment is under development, and the fusion community is in a position to influence its direction and emphasis. Information gained in studies of this type should be brought to bear on the development process to ensure that vacuum equipment will be available to meet the needs of future fusion devices.     

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.     

Design for a high power-density Astron reactor

A liquid lithium blanket surrounding the plasma volume is described. The liquid lithium flows along magnetic flux tubes at a high speed. There is no vacuum wall between the blanket and the plasma. The E-layer of relativistic particles within which the plasma is confined serves as a vacuum wall protecting the plasma from the lithium vapor, which is continuously produced at the surface of the blanket, by ionizing the lithium atoms and ejecting the same along open magnetic lines. The heat load at the surface of the blanket generated by 14 MeV neutrons can be several hundred MW per square meter.Work performed under the auspices of the U.S. Atomic Energy Commission.Deceased September 24, 1972.     

Investigation on Retention and Release Behaviors of Hydrogen and Helium in Vanadium Alloy

Vanadium alloy is proposed as an attractive candidate for first wall and blanket structural material of fusion reactors. The retention and release behaviors of hydrogen and helium in vanadium alloy may be an important issue. In the present work, 1.7 keV deuterium and 5 keV helium ions are respectively implanted into V-4Cr-4Ti and V-4Ti at room temperature. The retention and release of deuterium and helium are measured with thermal desorption spectroscopy (TDS). When the helium ion fluence is larger than 3 ×1017 He/cm2, the retained helium saturates with a value of approximately 2.5 ×1017 He/cm2. However, when the ion fluence is 1×1019 D/cm2, the hydrogen saturation in vanadium alloy does not take place. Experimental results indicates that hydrogen and helium retention in vanadium alloy may lead to serious problems and special attention should be paid when it is applied to fusion reactors.     

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.     

Analysis for hydrogen particle balance of plasma-wall system in the large helical device

The hydrogen particle balance of the plasma-wall system in the large helical device (LHD) is analyzed, using a zero dimensional model for plasma particles, neutrals in vessel and hydrogen inventory in wall. Based on the measurement of neutral gas pressure, plasma density and the pumping speed of the cryo-pumping system, it is found that the hydrogen retained in the wall desorbes with short and long time constant. The short term desorption is of order of 10²¹ atoms with a time constant of a few minutes, which is much smaller than the wall pumping for one shot, 10²² atoms. In a long time scale of about one experimental day, the wall absorbs significantly large amounts of hydrogen, up to 10²⁴ atoms. One of the possible reasons for the large wall pumping is a carbon deposition layer on the first wall surface. The effect of hydrogen retention on density control is also discussed.     

Deuterium ion-surface interactions of liquid-lithium thin films on micro-porous molybdenum substrates

Lithium has been utilized to enhance the plasma performance for a variety of fusion devices such as TFTR, CDX-U and NSTX. Lithium in both the solid and liquid states has been studied extensively for its role in hydrogen retention and reduction in sputtering yield. A liquid lithium diverter (LLD) was recently installed in the National Spherical Torus Experiment (NSTX) fusion reactor to investigate lithium applications for plasma-facing surfaces (PFS). Representative samples of LLD material were exposed to lithium depositing and simulated plasma conditions offline at Purdue University to study changes in surface chemical functionalities of Mo, O, Li and D. X-ray photoelectron spectroscopy (XPS) conducted on samples revealed two distinct peak functionalities of lithiated porous molybdenum exposed to deuterium irradiation. The two-peak chemical functionality noticed in porous molybdenum deviates from similar studies conducted on lithiated graphite; such deviation in data is correlated to the complex surface morphology of the porous surface and the correct “wetting” of lithium on the sample surface. The proper lithium “wetting” on the sample surface is essential for maximum deuterium retention and corresponding LLD pumping of deuterium.     

Line identification of extreme ultraviolet(EUV) spectra from low-Z impurity ions in EAST tokamak plasmas

Extreme ultraviolet(EUV) spectra emitted from low-Z impurity ions in the wavelength range of10–500 ? were observed in Experimental Advanced Superconducting Tokamak(EAST)discharges. Several spectral lines from K-and L-shell partially ionized ions were successfully observed with sufficient spectral intensities and resolutions for helium, lithium, boron, carbon,oxygen, neon, silicon and argon using two fast-time-response EUV spectrometers of which the spectral intensities are absolutely calibrated based on the intensity comparison method between visible and EUV bremsstrahlung continua. The wavelength is carefully calibrated using wellknown spectra. The lithium, boron and silicon are individually introduced for the wall coating of the EAST vacuum vessel to suppress mainly the hydrogen and oxygen influxes from the vacuum wall, while the carbon and oxygen intrinsically exist in the plasma. The helium is frequently used as the working gas as well as the deuterium. The neon and argon are also often used for the radiation cooling of edge plasma to reduce the heat flux onto the divertor plate. The measured spectra were analyzed mainly based on the database of National Institute of Standards and Technology. As a result, spectral lines of He Ⅱ, Li Ⅱ–Ⅲ, B Ⅳ–Ⅴ, C Ⅲ–Ⅵ, O Ⅲ–Ⅷ, Ne Ⅱ–Ⅹ,Si Ⅴ–Ⅻ, and Ar Ⅹ–XVI are identified in EAST plasmas of which the central electron temperature and chord-averaged electron density range in T_(e0)=0.6–2.8 keV and n_e=(0.5–6.0)×10~(19) m~(-3), respectively. The wavelengths and transitions of EUV lines identified here are summarized and listed in a table for each impurity species as the database for EUV spectroscopy using fusion plasmas.     

Engineering studies for the installation of an axi-symmetric metallic divertor in Tore Supra

Tore Supra (TS) has been designed to operate using technologies that allow long plasma operation (a few minutes), by means of superconducting magnets and actively-cooled high heat flux plasma facing components (PFCs). Actively cooled tungsten PFC will be used in the baffle area of the first ITER divertor. In order to validate such a technology fully (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axi-symmetric divertor in the tokamak Tore Supra has been studied [1]. With this second major upgrade, Tore Supra should be able to address the problematic of long plasma discharges with a metallic divertor.The proposed divertor is made up of two stainless steel casings containing a copper coil winding located at the top and bottom area of the vacuum vessel. These casings are firmly maintained by connection beams and protected by PFC. This paper describes the mechanical design of this major component and its integration in TS, the associated electromagnetic and thermomechanical analysis, the manufacturing issues and finally the integration of ITER representative PFCs.     

Status and prospect of the development of liquid lithium limiters for stellarotor TJ-II

Stellarator concept is considered as a promising approach for power fusion reactor development because it basically free from disruptions and other extreme thermal load events. However, the potential problem of impurity accumulation in stellarator plasma should be taken into account. Very promising results in density control, plasma reproducibility and confinement characteristics have been obtained with application of “lithiation” technology. The next step in the improvement of TJ-II Heliac plasma performance is the development and creation of two poloidal liquid lithium limiters (LL). Experimental possibilities, design, structural materials and main parameters of LL based on capillary-pore structure (CPS) filled with liquid lithium are considered. Understanding in hydrogen isotope interaction with liquid lithium surface is an important aspect of lithium technology development for fusion reactor application. Therefore study of deuterium sorption/desorption process on a lithium surface of LL is stipulated. The development of lithium CPS based devices decreasing intensity of plasma–wall interaction on a central “groove” of TJ-II vacuum camera is proposed as the further step in plasma performance improvement owing to decrease in impurity flux from the wall.     

Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.     

Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10^?7 mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10^?8 mbar was achieved inside the vacuum vessel.     

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.     

The KSTAR plasma facing components for 2010 operation

The KSTAR plasma facing components (PFCs) consist of inboard limiter, poloidal limiter, divertor, passive stabilizer and neutral beam armor. The main function of the PFCs is to define boundary of operating plasma and to protect the vacuum vessel and in-vessel components such as diagnostic components, in vessel control coil and several kinds of launchers for heating and current drive systems. The divertor is designed to enhance effective particle control to keep high quality plasma with various flexibilities in the shaping control for wide range of operational regime. The passive stabilizer that is made of CuCrZr alloy is designed to passively control the vertical position and MHD instabilities during operation as well as outer boundary of the plasma. Since fabrication has been started for all of the plasma facing components from middle of 2009, the inboard limiter, the divertor, and the passive stabilizer were successfully installed in the vacuum vessel, in turn. Moreover, one set of neutral beam armor and three strings of poloidal limiters were also installed according to the heating system that newly comes in 2010. All the PFCs tiles were baked to 200 °C and the PFC system showed no vacuum leakage and other mechanical troubles. In this paper, key features, fabrication, results of assembly, and baking of the KSTAR PFCs are summarized in detail.     

Rutherford backscattering spectroscopy and elastic recoil detection analysis with lithium ions – The better alternative to helium?

Rutherford backscattering spectroscopy (RBS) and elastic recoil detection analysis (ERDA) with lithium ions are compared to using helium ions. The availability and accuracy of backscattering cross-section and stopping power data for incident Li ions are reviewed, and energy broadening contributions due to detector resolution and energy loss straggling are discussed. Theoretical calculations of the depth resolution are compared with experimental data for RBS from Nb/Co multilayers and foil-ERDA from amorphous hydrogenated carbon multilayers. In RBS about the same or better depth resolution with Li than with He is achieved, while in ERDA for the detection of hydrogen isotopes the depth resolution is increased by a factor of about 1.5 compared to incident He.     

Hydrogen and helium trapping in tungsten under single and sequential irradiations

The trapping of hydrogen and helium in polycrystalline tungsten irradiated with 500 eV He⁺, H⁺ and D⁺ ions, individually or sequentially, has been measured by thermal desorption spectroscopy. Specimens irradiated with 500 eV He⁺ at 300 K show three He release peaks in the vicinity of ∼500, ∼1000, and ∼1200 K. The helium is thought to form He vacancy complexes or bubbles. Increasing the specimen temperature to 700 K does not significantly affect the trapping behavior of He. Sequential He⁺-D⁺ irradiation at 300 K results in the elimination of He release above 800 K. Instead, both D and He were released in the range 400-800 K. This is interpreted as interstitial D and He released from the near surface. Sequential He⁺-D⁺ irradiation at 700 K resulted in a reduced single He peak at ∼1000 K with very little release observed below 800 K; no D was trapped for irradiations at 700 K. Sequential D⁺-He⁺ irradiations at 300 K show that He trapping occurs in much the same manner as for the He⁺-only case while D retention is reduced at the near surface. Sequential D⁺-He⁺ irradiations at 700 K indicate that pre-irradiation with D⁺ has little or no effect on the subsequent trapping behavior of He.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.