Analysis of an ex-vessel break in the ITER divertor cooling loop

In the present work the integrated ECART code, developed for severe accident analysis in LWRs, is applied on the analysis of a large ex-vessel break in the divertor cooling loop of the international thermonuclear experimental reactor (ITER). A comparison of the ECART results with those obtained by Studsvik Nuclear AB (S), utilizing the MELCOR code, was also performed in the general framework of the quality assurance program for the ITER accident analyses. This comparison gives a good agreement in the results, both for thermal-hydraulics and the environmental radioactive releases. Mainly these analyses, from the point of view of the ITER safety, confirm that the accidental overpressure inside the vacuum vessel and the Tokamak cooling water system (TWCS) Vault is always well below the design limits and that the radioactive releases are adequately confined below the ITER guidelines.     

Progress of IRSN R&D on ITER Safety Assessment

The French “Institut de Radioprotection et de S?reté Nucléaire” (IRSN), in support to the French “Autorité de S?reté Nucléaire”, is analysing the safety of ITER fusion installation on the basis of the ITER operator’s safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.     

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.     

Experimental and numerical analysis of the air inflow technique for dust removal from the vacuum vessel of a tokamak machine

In fusion facilities, the dust production inside the plasma chamber is a concern from the viewpoint of both machine performance and safety. To the purpose of a correct handling of the experimental devices the problem of its removal must be properly solved. This work deals with the experiments carried out in the STARDUST facility by using as dust removal technique an air inflow into the volume representing the vacuum vessel. The goal was to evaluate the effectiveness of such an approach, less invasive as compared to all the others so far. These experiments, performed by using characterized carbon, tungsten and stainless steel dusts, show that the mobilization capability of the air inflow is between few percent and 100%, mainly depending on dust type of and deposit shape. The capture efficiency in a filter reached a maximum of about 7.5% in the STARDUST geometrical configuration. In conclusion, this simple and clean (from the radioactive point of view) removing technique needs particular care to be more effective and is not the perfect solution due to its low efficiency in the collection of removed powder in proper surfaces (i.e., filters). Nevertheless improvements are possible and worthwhile.     

Real-time measurement of hydrogen generation level during beryllium dust oxidation by steam depending on the dust arrangement geometry

The paper concentrates on the safety issues in the International Thermonuclear Experimental Reactor (ITER) and describes the experiment on the measurement of hydrogen generation rate in case of Ingress of Coolant Event (ICE)—leak inside the vacuum vessel during interaction between water and beryllium (Be) dust. The ICE situation in ITER was simulated in a facility; the active spectroscopy was used to define the hydrogen content by the dynamics of oxidant concentration at a sampling frequency up to 10 Hz. Hydrogen release in time at temperatures of 500-900 °C is investigated, and different versions of dust arrangement are considered, i.e. on the surface and in a slot between armoring tiles at different initial density. The obtained results are compared with the known experiments.     

A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium–steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.     

Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.     

Model of particle resuspension in turbulent flows

The graphite dust generated in an HTR/PBMR during normal reactor operation is deposited inside the primary system and becomes radioactive due to sorption of fission products. A significant amount of radioactive dust may be resuspended and released to the environment in case of LOCA. Therefore accurate particle resuspension models are required for HTR/PBMR safety analyses.Thermal-hydraulic safety analyses of HTR/PBMR type reactors are typically performed using computer codes such as FLOWNEX, MELCOR, or SPECTRA. None of these codes currently includes a well-tested mechanistic resuspension model.The resuspension model based on the Vainshtein model has been developed and implemented into the SPECTRA thermal-hydraulic system code. The resuspension model formulation has been extended in such way that other formulations, for example the Rock’n Roll model, may easily be defined and used within the general model framework.Several test calculations were performed, including comparisons of the numerical SPECTRA results with the analytical solutions obtained by means of MathCAD. Furthermore, comparisons with the experimental results of the Reeks and Hall, and STORM experiments were made. It was concluded that the model gives satisfactory results for a number of tests.     

Power dissipation and energy transfer during testing of the ITER toroidal field model coil

The testing of the ITER toroidal field model coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the LCT coil and 80 kA in the TFMC, respectively. The heat load of both coils, including the eddy current losses in the passive structures and the joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. A computer code, based on the full inductance and resistance matrices, has been developed with SIMULINK™. After validation with experimental data the code has been used to perform circuit analysis and to evaluate the power dissipation and energy transferred to the cryogenic plant and to the external power circuits.     

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.     

Fusion Safety Research and Development in JAERI

This paper describes the current status and future plans of the fusion safety research and development regarding to the developments of the dust removal system and safety analysis code and the thermofluid experiments in the Japan Atomic Energy Research Institute (JAERI) for a fusion experimental reactor. The containment of the radioactive material is the key to achieve fusion safety. In the event of accidents, the source terms need to be evaluated with sufficient accuracy. Therefore, in JAERI, the dust characterization have been investigated and the dust removal system using electric force has been developed and tested. A safety analysis code including both thermal and plasma transient analyses under the various event sequences has been developed. Moreover, the preliminary experiments of thermofluid transients in the vacuum vessel such as Ingress of Coolant Event (ICE) and Loss of Vacuum Event (LOVA) have been started and the experimental results using preliminary LOVA/ICE apparatus during 1995–1996 are summarized in this paper.     

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).     

Detailed nuclear analysis of ITER ELM coils

The ELM coils in ITER are intended to provide control of Edge Localized Modes (ELMs). These coils are located on the outboard side of ITER between the shield modules and vacuum vessel (VV) and are subject to high radiation levels. Detailed three-dimensional (3-D) models of the toroidal and poloidal legs of the ELM coil and surrounding region for the MCNP code were updated to reflect the latest design changes. Neutronics calculations were performed to determine a variety of radiation damage parameters for the ELM coils as well as the VV located behind them. Additionally, detailed CAD based models for the upper ELM coil region were used to perform a CAD based analysis using the DAG-MCNP5 code. The results show that the ELM coil will meet the specified material radiation limits. However, the nuclear heating on the vacuum vessel behind the poloidal multi-pipe manifolds will exceed the specified limit.     

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.     

Analysis of Aerosol Resuspension under Severe Accident by ECART Model

Utilizing ECART model in resuspension module of CABSA code developed by China Institute of Atomic Energy, the SR11 test in STORM project was calculated and the features of aerosol resuspension under severe accident of nuclear power plant were analyzed. The result shows that all the forces on aerosol increase with aerosol diameter, drag force and burst force increase faster than adhesive force and gravity force. Aerosol with larger diameter has greater resuspension rate. Fluid velocity above structure can affect drag force and burst force, the increment of velocity can enlarge both the forces, and make the resuspension rate increase. By taking this advantage, the drag force and burst force can be reduced by reducing the fluid velocity above structure, and finally decrease the release of fission product to the environment.     

严重事故下气溶胶再悬浮的ECART模型分析

利用中国原子能科学研究院开发的CABSA程序气溶胶再悬浮模块中的ECART模型，对STORM项目的SR11试验进行计算，分析了核电厂严重事故下的气溶胶再悬浮特性。结果表明：气溶胶所受各种力均随直径的增大而增大，其中使气溶胶悬浮的拖曳力和爆发力比使气溶胶附着在结构表面的黏着力和重力增长更快；直径大的气溶胶悬浮率更大；结构表面流体速度能够影响拖曳力和爆发力，速度增大会提高拖曳力和爆发力，最终导致悬浮率增加。利用该特点，可通过降低结构表面流速降低拖曳力和爆发力，从而减小悬浮率，最终减小裂变产物向空间的重新释放。     

Experimental and theoretical investigation of droplet emission from tungsten melt layer

Tungsten in form of macrobrush structure is foreseen as one of candidate materials for the ITER divertor and the dome. Melting of tungsten and the following melt motion and melt splashing are expected to be the main mechanisms of damage which determine the lifetime of plasma facing components. New experimental investigations of droplet emission from the W melt layer for the Edge Localised Mode (ELM)-like heat loads have been carried out at the plasma gun facility quasistationary plasma accelerators (QSPA-T). In these experiments the threshold for droplet emission and the distributions of velocity on emission angles and amplitude of the ejected droplets were determined. In the paper the main physical mechanism (the Kelvin–Helmholtz instability) of the melt splashing under the heat loads being applied at QSPA-T and those anticipated after the ITER transients is analyzed. These numerical simulations demonstrated a reasonable agreement with the experimental data on the droplet sizes and droplet velocities and allowed the projections upon the W melt splashing at ITER conditions.     

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。     

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 residual thermostress in tungsten after repetitive ELM-like heat loads

Brittle destruction of tungsten armour under action of edge localised modes of plasma instabilities (ELMs) in ITER is an important issue determining the lifetime of the divertor. Besides, cracking of the armour produces tungsten dust with characteristic size of 1–10 μm flying from the armour surface with velocities up to 10 m/s. Influx of the tungsten dust into the ITER confinement decreases the temperature of the plasma, reduces the thermonuclear gain and even may run the confinement into disruption.This paper describes experiments in QSPA-Kh50 plasma gun and modeling, which has been performed for providing more insight into the physics of tungsten cracking under action of ELMs and for confirmation of the important result on stabilization of the crack development at the tungsten armour surface, predicted in our previous paper – the same authors, 2010.The threshold value of energy density deposition for start of tungsten cracking has been measured as 0.3 MJ/m² after 5–10 shots. From analytical considerations three times smaller threshold value has been predicted with increasing number of shots.