Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basic phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. We plan a scaled-model test and a full-scale model test for the LOVE.     

Dust tracking techniques applied to the STARDUST facility: First results

An important issue related to future nuclear fusion reactors fueled with deuterium and tritium is the creation of large amounts of dust due to several mechanisms (disruptions, ELMs and VDEs). The dust size expected in nuclear fusion experiments (such as ITER) is in the order of microns (between 0.1 and 1000 μm). Almost the total amount of this dust remains in the vacuum vessel (VV). This radiological dust can re-suspend in case of LOVA (loss of vacuum accident) and these phenomena can cause explosions and serious damages to the health of the operators and to the integrity of the device. The authors have developed a facility, STARDUST, in order to reproduce the thermo fluid-dynamic conditions comparable to those expected inside the VV of the next generation of experiments such as ITER in case of LOVA. The dust used inside the STARDUST facility presents particle sizes and physical characteristics comparable with those that created inside the VV of nuclear fusion experiments. In this facility an experimental campaign has been conducted with the purpose of tracking the dust re-suspended at low pressurization rates (comparable to those expected in case of LOVA in ITER and suggested by the General Safety and Security Report ITER-GSSR) using a fast camera with a frame rate from 1000 to 10,000 images per second. The velocity fields of the mobilized dust are derived from the imaging of a two-dimensional slice of the flow illuminated by optically adapted laser beam. The aim of this work is to demonstrate the possibility of dust tracking by means of image processing with the objective of determining the velocity field values of dust re-suspended during a LOVA.     

Numerical study of air jet flow field during a loss of vacuum

Air leakage into tokamaks vacuum vessel during plasma burning or maintenance operations may lead to the fast pressurization of the vacuum vessel. A fraction of the dust inventory present in the vacuum vessel can be mobilized threatening the safety of staff and workers on site, the local population and the environment. A numerical analysis of the physical phenomena involved in such accidents is necessary in order to predict the thermal-fluid dynamics into the vacuum vessel after air ingress and consequent dust mobilization. Accuracy of the numerical results is also required in order to provide a sufficient margin in the design of the safety systems. The numerical simulation of Loss of Vacuum Accident (LOVA) scenarios is a challenging task for today numerical methods and models because it involves large volumes, multiphase flows ranging from highly supersonic to nearly incompressible and contemporary heat transfer. The drag force exerted on the dust by a moving fluid due to the viscous surface shear stress and pressure distribution around the dust particles depends mainly on the Reynolds number, i.e. property of the fluid (kinematic viscosity), its mean velocity and characteristic length of the geometry. For a fixed geometry, the key parameter for the dust mobilization is the velocity field of the continuous phase, and its thermodynamics properties, inside the vacuum vessel. In this contribution, the authors present and discuss the results of numerical simulations of air jet flow field during a LOVA with particular attention to the comparison with the experimental data and differences arising from the use of different types of grid resolution and turbulence models (Zero-Equation, k–ω and SST).     

Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.     

Simulation of a New Plasma Mitigation Method by Transferring Part Poloidal Magnetic Energy in J-TEXT

The magnitude of the damaging effect caused by plasma disruptions increases with the energy dissipated inside the vacuum vessel (VV) during plasma disruptions. To mitigate the disruption damage, a new method aiming at reducing the total energy dissipated inside the VV is proposed. By adding a group of coils on tokamak that are inductively coupled with plasma and conducting these coils during disruption, part of the poloidal magnetic energy can be coupled and be dissipated outside the VV, which leads to a reduction of the energy dissipated inside the VV. This new method is named poloidal magnetic energy transfer (MET) and the coils are named energy transfer coils. This method is firstly proposed and analyzed in J-TEXT. The principle of MET is described and the effect of MET is calculated with a simplified lump-parameter circuit. A 23 % reduction of the magnetic energy dissipated inside the VV can be obtained.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part I: Flow distribution and pressure drop

The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector #5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT^®. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10 m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ～50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping.     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

IWR-solution for the ITER vacuum vessel assembly

The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.     

Thermo-hydraulic design verification of the neutral beam liner for the ITER vacuum vessel

The ITER vacuum vessel has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping and access inside the vessel for maintenance. A neutral beam (NB) port of equatorial ports provides a path of neutral beam for plasma heating and current drive. An internal duct liner is built in the NB ports, and copper alloy panels are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. Global NB liner models for the upper panel and the lower panel have been developed, and flow field and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux are applied in the analyses for beam steering and misalignment conditions. For the upper panel, it is found that unbalanced flow distribution occurs in each flow path, and this causes poor heat transfer performance as well. In order to improve flow distribution and to reduce pressure losses, hydraulic analyses for modified cooling path schemes have been carried out, and design recommendations are made based on the analysis results. For the lower panel, local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying design coolant flow rate. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired from the analyses. Based on the analysis results, it is concluded that the lower panel for the NB liner is relatively well designed even though the given heat loads are very severe.     

Analysis of large-scale tests for AP-600 passive containment cooling system

All next-generation light water reactors utilize passive systems to remove heat via natural circulation and are significantly different from past and current nuclear plant designs. One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 h without action by the reactor operator. During a design-basis accident (DBA), i.e., either a loss-of-coolant or a main-steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annular space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-1D code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single-phase flow, transport equations for the k − two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-1D results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized.     

Vacuum system of SST-1 Tokamak

Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN₂ cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10^?4 mbar and 1.0 × 10^?5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10^?6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10^?5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ～3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10^?5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10 °C deviation was achieved with a net mass flow rate of 0.8 kg/s at 1.5 bar gauge inlet pressure and supply temperature of 230 °C at the heater end. Also during gas feed system installation, the pressure inside the VV was raised from 3.01 × 10^?5 mbar to 1.72 × 10^?4 mbar by triggering a pulse of lower amplitude of 25 voltage direct current (VDC) for 100 s to piezoelectric valve. This paper describes in detail the design and implementation of the various vacuum subsystems including relevant experimental results.     

Status of ITER blanket attachment design and related R&D

Main function of the ITER blanket system [1], [2], [3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R&D is also briefly described.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part II: Thermal-hydraulic effects of the nuclear heat load

The 3D Computational Fluid Dynamic (CFD) steady state analysis of the regular sector #5 of the ITER vacuum vessel (VV) is presented in these two companion papers using the commercial software ANSYS-FLUENT®. The pure hydraulic analysis, concentrating on flow field and pressure drop, is presented in Part I. This Part II focuses on the thermal-hydraulic analysis of the effects of the nuclear heat load. Being the VV classified as safety important component, an accurate thermal-hydraulic analysis is mandatory to assess the capability of the water coolant to adequately remove the nuclear heat load on the VV. Based on the recent re-evaluation of the nuclear heat load, the steady state conjugate heat transfer problem is solved in both the solid and fluid domains. Hot spots turn out to be located on the surface of the inter-modular keys and blanket support housings, with the computed peak temperature in the sector reaching ～290 °C. The computed temperature of the wetted surfaces is well below the coolant saturation temperature and the temperature increase of the water coolant at the outlet of the sector is of only a few °C. In the high nuclear heat load regions the computed heat transfer coefficient typically stays above the 500 W/m² K target.     

Design and construction of Alborz tokamak vacuum vessel system

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

Electromagnetic and structural analyses of the vacuum vessel and plasma facing components for EAST

During plasma disruptions, time-varying eddy currents are induced in the vacuum vessel (VV) and Plasma Facing Components (PFCs) of EAST. Additionally, halo currents flow partly through these structures during the vertical displacement events (VDEs). Under the high magnetic field circumstances, the resulting electromagnetic forces (EMFs) and torques are large. In this paper, eddy currents and EMFs on EAST VV, PFCs and their supports are calculated by analytical and numerical methods. ANSYS software is employed to evaluate eddy currents on VV, PFCs and their structural responses. To learn the electromagnetic and structural response of the whole structure more accurately, a detailed finite element model is established. The two most dangerous scenarios, major disruptions and downward VDEs, are examined. It is found that distribution patterns of eddy currents for various PFCs differ greatly, therefore resulting in different EMFs and torques. It can be seen that for certain PFCs the transient reaction force are severe. Results obtained here may set up a preliminary foundation for the future dynamic response research of EAST VV and PFCs which will provide a theoretical basis for the future engineering design of tokamak devices.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

Design and application of GDC on EAST Tokamak

On EAST Tokamak, DC glow discharge (GDC) is developed to clean the first wall of plasma. It can effectively control the recycling of H, C, O impurities and improve the wall conditions. There are four GDCs which distribute equally on the EAST Tokamak vacuum vessel wall. Each GDC is equipped with an anode, a stainless steel cover and four support legs. The anode is insulated from cover with Al₂O₃ ceramics. After a round of experiment, the value of insulation resistance of GDC decreases remarkably due to metallization. To protect the insulation parts and heighten the reliability, ceramic protection covers are used on the GDCs. The other measures which can heighten insulation grades are also taken. After upgrade, the insulation resistance of each GDC between anode and ground is raised highly. When the pressure reaches 4 Pa, H₂-GDC and He-GDC is strarted. Boronation and siliconization are also applied to the device wall conditioning. After GDC cleaning, the impurities and partial pressure of remainder gases in vacuum vessel (VV) is decreased greatly and vacuum degree of VV can reach high easily.     

Containment for the low temperature district nuclear-heating reactor

An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happened during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety.