Transient thermal hydraulic modeling and analysis of ITER divertor plate system

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m² plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.     

Safety analysis of ITER failures and consequences during maintenance

This study has been a first attempt at identifying potential worker overexposure situations during machine maintenance operations. The results indicate potential areas, or situations, where worker overexposure may be possible [A. Natalizio, T. Pinna, Safety analysis of failures and consequences during maintenance, ENEA Report, FUS-TN-SA-SE-R-170, June 2007, Frascati, Italy].The key findings obtained are as follows. Firstly, we have found no machine maintenance operations where the risk of worker overexposure is considered significantly large that immediate design attention is needed.Secondly, the most significant risk of worker overexposure is due to airborne releases of radioactivity from cooling water pipes and tubes that may not have been fully drained and dried, when they are cut, or inadvertently opened, by workers (frequency of pipe-cutting activities could be significantly high).Thirdly, the risk of overexposure from human error could also be significant. This varies from mistaking the machine sector, to mistaking the component to be maintained. This is analogous to working on a live electrical circuit, when it is believed to be dead (disconnected from the power source) because the worker has mistakenly selected the wrong circuit—a look-alike one. Similarly, consider the situation of a worker mistakenly preparing to work on a cooling water circuit that is still at pressure and temperature, instead of the one that has been drained and dried. The more look-alike situations there are in the facility, the greater the probability of committing this type of error.Fourthly, when consideration is given to human error, we believe that the aggregation of different diagnostics in the same port enhances the probability of human error. At the moment, these risks cannot be quantified. The task of quantifying those risks in the future should be considered.Finally, the transport of activated in-vessel components, including components of plasma-heating and current-drive systems, in non-shielded casks, could carry with it a significant risk of worker overexposure. In the context of ALARA, this approach requires a specific study to justify its use.Concluding, it is important to note that by having identified the possibility of an overexposure situation does not mean that it is probable. The calculation of probability awaits further studies of this nature, when the design reaches a more detailed level.     

Safety Analysis of ITER EDA Design by GEMSAFE

General Methodology of Safety Analysis and Evaluation for Fusion Systems (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) design in the stage of Engineering Design Activities (EDA) to identify Design Basis Events (DBEs) and the related safety features, which were compared with those of the ITER design in the stage of Conceptual Design Activities (CDA). As a result, 18 DBEs for the EDA design were selected in comparison with 25 DBEs for the CDA design. DBEs related to the fuel area were categorized in higher event category than those of the CDA design due to the increase of the mobile tritium contained in some components. It was necessary to reduce the inventory of the tritium absorbed in the tokamak dust in the EDA design as well as in the CDA design. Some measures were recommended to reduce mobile tritium dissolved in the coolant in the single cooling loop due to the increase of this estimated inventory.     

Physics of the conceptual design of the ITER plasma control system

The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints.In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma performance must apply multiple control functions simultaneously with a limited number of actuators. A sophisticated shared actuator management system is being designed to prioritize the goals that need to be controlled or weigh the algorithms and actuators in real-time according to dynamic control needs. The underlying architecture will be event-based so that many possible plasma or plant system events or faults could trigger automatic changes in the control algorithms or operational scenario, depending on real-time operating limits and conditions.     

Recommendations on the Nature and Level of U.S. Participation in the International Thermonuclear Experimental Reactor (ITER) Extension of the Engineering Design Activities

The Department of Energy (DOE) Office of Energy Research chartered through the Fusion Energy Sciences Advisory Committee (FESAC) a panel to address the topic of U.S. participation in an ITER construction phase, assuming the ITER Parties decide to proceed with construction. Given that there is expected to be a transition period of 3 to 5 years between the conclusion of the Engineering Design Activities (EDA) and the possible construction start, the DOE Office of Energy Research expanded the charge to include the U.S. role in an interim period between the EDA and construction.This panel has heard presentations and received input from a wide cross-section of parties with an interest in the fusion program. The panel concluded it could best fulfill its responsibility under this charge by considering the fusion energy science and technology portion of the U.S. program in its entirely. Accordingly, the panel is making some recommendations for optimum use of the transition period considering the goals of the fusion program and budget pressures.     

Development of the PACTITER code and its application to safety analyses of ITER Primary Cooling Water System

The PACTITER code derives from the PACTOLE code, developed by the CEA for predicting activated corrosion products (ACPs) in PWR primary circuits. The operating conditions, material compositions and water chemistry of the various Primary Heat Transfer Systems (PHTS) of the International Thermonuclear Experimental Reactor (ITER) made mandatory the adaptation of the PACTOLE code.PACTITER was developed on the basis of dedicated experiments, namely devoted to determine copper solubility and stainless steel release in the ITER primary cooling systems conditions, which are rather different from those in PWR (i.e. water chemistry and temperatures). The PACTITER code has been extensively used in support of the ITER Generic Site Safety Report (GSSR) in the field of accident analysis and worker collective dose assessment.     

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S.Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.     

ITER safety studies: The effect of two simultaneous perturbations during a loss of plasma control transient

The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. Conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer are used to determine the effects of such transients on wall integrity from a thermal point of view.In this contribution, progress in a “two simultaneous perturbations over plasma” approach to the analysis of the loss of plasma control transients in ITER is presented. The effect of variation in confinement time is now considered, and the consequences of this variation are shown over a n–T diagram. The study has been done with the aid of AINA 3.0 code. This code implements the same 0D plasma-1D wall scheme used in previous LOPC studies.The rationale of this study is that, once the occurrence of a loss of plasma transient has been assumed, and due to the uncertainties in plasma physics, it does not seem so unlikely to assume the possibility of finding a new confinement mode during the transient.The cases selected are intended to answer to the question “what would happen if an unexpected change in plasma confinement conditions takes place during a loss of plasma control transient due to a simultaneous malfunction of heating, or fuelling systems?”Even taking into account the simple models used and the uncertainties in plasma physics and design data, the results obtained show that the methodology used in previous analyses could probably be improved from the point of view of safety.     

ITER Safety Analyses with ISAS

Detailed analyses of accident sequences for the International Thermonuclear Experimental Reactor (ITER), from an initiating event to the environmental release of activity, have involved in the past the use of different types of computer codes in a sequential manner. Since these codes were developed at different time scales in different countries, there is no common computing structure to enable automatic data transfer from one code to the other, and no possibility exists to model or to quantify the effect of coupled physical phenomena. To solve this problem, the Integrated Safety Analysis System of codes (ISAS) is being developed, which allows users to integrate existing computer codes in a coherent manner. This approach is based on the utilization of a command language (GIBIANE) acting as a glue to integrate the various codes as modules of a common environment. The present version of ISAS allows comprehensive (coupled) calculations of a chain of codes such as ATHENA (thermal-hydraulic analysis of transients and accidents), INTRA (analysis of in-vessel chemical reactions, pressure built-up, and distribution of reaction products inside the vacuum vessel and adjacent rooms), and NAUA (transport of radiological species within buildings and to the environment). In the near future, the integration of S AFALY (simultaneous analysis of plasma dynamics and thermal behavior of in-vessel components) is also foreseen. The paper briefly describes the essential features of ISAS development and the associated software architecture. It gives first results of a typical ITER accident sequence, a loss of coolant accident (LOCA) in the divertor cooling loop inside the vacuum vessel, amply demonstrating ISAS capabilities.     

Thermal and mechanical analysis of the ITER plasma-position reflectometry antennas

The antennas of the ITER plasma-position reflectometer are the components most exposed to the plasma. High thermal loads can cause high temperatures and excessive stress, so the first constrains on the antenna design arise from thermal simulations results. Therefore, the first step of the analysis is to develop a finite element thermal model with ANSYS. Once the temperatures are kept at acceptable levels, structural analysis is performed to know the thermal stress. Simulations performed using different materials and support structure geometries are discussed. Further, it has been checked that the components can withstand the electromagnetic loads expected during disruptions and vertical displacement events. The stress due to these electromagnetic loads is calculated analytically as well as with ANSYS simulations. The proposed antenna arrangement is properly designed against thermal and mechanical loads.     

Modeling of a confinement bypass accident with CONSEN,a fast-running code for safety analyses in fusion reactors

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.     

Structural Analysis of ITER Lower In-Vessel-Viewing Port

A finite element model of the International Thermonuclear Experimental Reactor （ITER） in-vessel viewing port was developed by the ANSYS code in order to evaluate the stress level of this structure. The thermal, elastic and modal analyses were made in succession based on the loads designated by the ITER International team. The designed loads include electromagnetic loads, seismic loads, pressure, temperature and gravity. The preliminary results of the finite element analysis （FEA） show that the stress intensity exceeded the allowable stress and the maximum stress was concentrated in the geometric discontinuous region of the shroud stub extension （SSE）. Therefore, the SSE has been modified recently. For the modified structure, we found that the stresses do not exceed the allowable value for all load combinations. In addition the modal analysis results show that the natural frequencies of the IVV port structure are located in the typical diapason of seismic excitation.     

Structural Analysis of the ITER VV Lower Port Region

A structural analysis of the International Thermonuclear Experimental Reactor （ITER） vacuum vessel＇s lower port region was presented by means of a finite element analysis method. The purpose is to evaluate the stress and displacement level on this structure under various combinations of five designed loads, including the gravity of the vacuum vessel, seismic loads, electromagnetic loads, and possible pressure loads to ensure structural safety. The cyclic symmetry finite element model of this structure was developed by using ANSYS code. The re- sults showed that the maximum stress does not exceed the allowable value for any of the load combinations according to ASME code and the nine vacuum vessel （VV） supports have the ability to sustain the entire VV and in vessel-components and withstand load combinations under both normal as well as off-normal operation conditions. Stress mainly concentrates on the connecting region of the VV support and lower port stub extension.     

Prototype Engineering Test Platform of ITER Magnet Gravity Support

ITER magnet gravity support (GS) has been redesigned as a structure of pre- assembled multi-flexible plates instead of the original welded structure. In the past several years, engineering tests of the new structure have been proposed. A prototype engineering test plat- form is being developed. In order to apply the loads/load combinations onto the test mock-up, seven hydraulic bolt tensioners in three directions have been applied to simulate various loads (forces and moments), through which the deformation of bolts, flexible plates and clamp blocks, the stress distribution in the flexible plates, the friction between the contact surface, etc. can be monitored/tested. The measurement and control system includes seven sets of synchronization controller, a 16-channel strain gauge, 25 sets of displacement sensors, etc. Principles of EDC220 digital controller and development of multi-channel control software are also demonstrated.     

Redesign of gravity support system for ITER construction

In this article, we describe an alternative design for ITER gravity support, which use various connection bolts and shear keys to assemble all the parts, rather than welding them together. The finite element model (FEM) analysis of this structure shows that the maximum static stress intensity of all the components is within the stress limitation under ITER operation condition. No terrible stress concentration and large deformation would occur during normal operation and abnormal operation. The buckling analysis shows that the new designed structure is stable, and no destructive damage would occur. The fatigue simulation calculation shows that the fatigue life is up to 1,361,445 repetitions for normal operation, which is far larger than that of the ITER 30,000 times discharge requirement. Therefore, it can be concluded that the new designed structure is safe and can be utilized in the ITER construction.     

A first characterization of the quench detection system for ITER

The purpose of this paper is to assess the expected response of conventional and non-conventional quench detection sensors proposed for the ITER coils, and to be tested in the QUELL experiment in SULTAN. The assessment is based on simulation of thermohydraulic transients in the ITER coils for various operating conditions, and a tentative definition of the transfer functions of each sensor concept. It is shown that, for the investigated conditions, the co-wound voltage taps are more accurate than hydraulic systems and conventional voltage balance methods. The additional complication associated with the insertion of taps in the conductor is well offset by the low sensitivity to external disturbances.     

The Effect of Hydrostatic Pressure and Seismic Load on ITER Lower Cryopump Ports

The lower cryopump ports in International Thermonuclear Experimental Reactor （ITER） as a part of the vacuum vessel play many important roles. As the boundary of vacuum it must be ensured against structural damage. In this study a finite element model of the lower cryopump ports was developed by ANSYS code with a purpose to evaluate the stress and displacement level on it. Two kinds of loads were taken into account. One was the hydrostatic pressure including the normal operation pressure and test pressure. The other was the seismic load. The analysis results show that the peak stress does not exceed the allowable stress for either the hydrostatic pressure or the seismic load according to the ITER structural design criterion, which indicates that the structure has a good safety margin.     

Structural analysis and optimization for ITER upper ELM coil

ITER ELM coils are used to mitigate or suppress Edge Localized Modes (ELM), which are located between the vacuum vessel (VV) and shielding blanket modules and subject to high radiation levels, high temperature and high magnetic field. These coils shall have high heat transfer performance to avoid high thermal stress, sufficient strength and excellent fatigue to transport and bear the alternating electromagnetic force due to the combination of the high magnetic field and the AC current in the coil. Therefore these coils should be designed and analyzed to confirm the temperature distribution, strength and fatigue performance in the case of conservative assumption. To verify the design structural feasibility of the upper ELM coil under EM and thermal loads, thermal, static and fatigue structural analysis have been performed in detail using ANSYS. In addition, design optimization has been done to enhance the structural performance of the upper ELM coil.     

Preliminary Thermal Mechanical Analysis of the Equatorial Thermal Shield for ITER

Thermal analysis of the equatorial thermal shield for ITER is conducted in order to confirm that the cooling tube design was reasonable under both the plasma operational and the baking operational conditions. The structural performance was analyzed by means of the finite element software ANSYS. A comparison of the results with design requirements shows that the results of the simulation are within allowable design requirements, which indicates the feasibility and reliability of the equatorial thermal shield structure.     

Thermo-mechanical analysis of the thermal shield for ITER magnet feeder

The thermal shield for ITER magnet feeder plays the role of preventing thermal radiation from the warm components to the cool superconductor and supercritical helium system. Heat loads were calculated for thermal analysis, then finite element model was established by ANSYS code. Thermal analysis was performed in order to check the temperature distribution and pressure drop of the thermal shield under normal operation state. Different materials (steel or aluminum) for the thermal shield were also checked. Thermal stress analysis was performed based on the results of thermal analyses. Compared analysis results with design criteria, it is demonstrated that the results of the simulation are within allowable design requirements and the design scheme can be applied to the detailed design.