Shield Module Research Articles

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of the coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.

Development of load specifications for the design of ITER diagnostic system and port integration

ITER is the world’s largest fusion device currently under construction in South of France with 50+ diagnostic systems installed inside the port plugs, the interspace or the port cell region of various diagnostic ports. The plasma facing Diagnostic First Wall (DFW) and its supporting Diagnostic Shielding Modules (DSM) are designed to protect the front-end diagnostics from plasma neutron and radiation while providing apertures for diagnostic access to the plasma. The design of ITER port plug (PP) structure assembly including the DFWs, DSMs with in-port diagnostics is largely driven by the electromagnetic (EM) loads induced on these passive structural components during plasma disruptions, and the steady-state thermal loads from plasma volumetric and surface heating during normal operations. The design is significantly influenced by the dynamic response of in-port components attached to the port-specific DSM or PP closure plate as a result of transient EM loads induced on the PP assembly, vacuum vessel (VV) and the port extension during asymmetric plasma Vertical Displacement Events (VDEs). We investigate in this study the worst plasma disruption load cases for diagnostic systems of varying size and location, and summarize the steady-state thermal, transient EM and VV inertial loads for ITER diagnostic systems.

Neutronics analysis of the in-vessel components of the ITER plasma-position reflectometry system on the low-field side

The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations. For gap 4, the antennas are to be installed in-vessel between two blanket shield modules. The antennas and adjacent waveguides are directly exposed to the plasma through cut-outs in the blanket modules and are therefore subjected to high radiation doses from neutrons and photons, which may cause irradiation-induced changes in the material properties and compromise the integrity of these components. Here, we report on the neutronics analysis of the PPR system of gap 4 performed with the Monte Carlo simulation code MCNP6, with the objective of estimating the thermal loads in the system. The nuclear loads obtained in the neutronics analysis will be used as input in the thermal analysis of the in-vessel components of the ITER PPR system.

Update on the status of the ITER ECE diagnostic design

Considerable progress has been made on the design of the ITER electron cyclotron emission (ECE) diagnostic over the past two years. Radial and oblique views are still included in the design in order to measure distortions in the electron momentum distribution, but the oblique view has been redirected to reduce stray millimeter radiation from the electron cyclotron heating system. A major challenge has been designing the 1000 K calibration sources and remotely activated mirrors located in the ECE diagnostic shield module (DSM) in the equatorial port plug #09. These critical systems are being modeled and prototypes are being developed. Providing adequate neutron shielding in the DSM while allowing sufficient space for optical components is also a significant challenge. Four 45-meter long low-loss transmission lines transport the 70–1000 GHz ECE from the DSM to the ECE instrumentation room. Prototype transmission lines are being tested, as are the polarization splitter modules that separate O-mode and X-mode polarized ECE. A highly integrated prototype 200–300 GHz radiometer is being tested on the DIII-D tokamak in the USA. Design activities also include integration of ECE signals into the ITER plasma control system and determining the hardware and software architecture needed to control and calibrate the ECE instruments.

Structure and mechanical properties of improved cast stainless steels for nuclear applications

Casting of stainless steels is a promising and cost saving way of directly producing large and complex structures, such a shield modules or divertors for the ITER. In the present work, a series of modified high-nitrogen cast stainless steels has been developed and characterized. The steels, based on the cast equivalent of the composition of 316 stainless steel, have increased N (0.14–0.36%) and Mn (2–5.1%) content; copper was added to one of the heats. Mechanical tests were conducted with non-irradiated and 0.7 dpa neutron irradiated specimens. It was established that alloying by nitrogen significantly improves the yield stress of non-irradiated steels and the deformation hardening rate. Manganese tended to decrease yield stress but increased radiation hardening. The role of copper on mechanical properties was negligibly small. Analysis of structure was conducted using SEM-EDS and the nature and compositions of the second phases and inclusions were analyzed in detail. No ferrite formation or significant precipitation were observed in the modified steels. It was shown that the modified steels, compared to reference material (commercial cast 316L steel), had better strength level, exhibit significantly reduced elemental inhomogeneity and only minor second phase formation.

Outline Design of ITER Radial X-Ray Camera Diagnostic

The radial X-ray camera (RXC) is designed to measure the poloidal profile of plasma X-ray emission with high spatial and temporal resolution. Its primary diagnostic role includes measuring low (m, n) magnetohydrodynamic modes, sawteeth and disruption precursors, H-mode, edge-localized modes, and L-H transition. The RXC comprises two subsystems, i.e., in-port and ex-port cameras that view the outer and core regions, respectively, through vertical slots in the diagnostics shield module of an equatorial port plug. Detailed camera design is in progress including design of the camera structure, electronics, data acquisition and control, calibration, and pretest on the EAST tokamak. The sight path and neutron shielding have been optimized. The secondary vacuum, heat insulation, cooling, positioning, and calibration have been designed. The structure analysis results for the external camera indicate that even under five times gravity acceleration, the maximum stress was still below the allowable stress. The heat analysis results indicate that the maximum temperature on the detector box was ~56°C, which is within the detector operation temperature limit. The neutronics analysis results indicate that the detectors can be operated during the whole deuterium-deuterium phase without detector replacement. The electronics group and instrumentation and control group have also made good progress.

Liquid volume monitoring based on ultrasonic sensor and Arduino microcontroller

Incident of oil leakage and theft in oil tank often happens. To prevent it, the liquid volume insides the tank needs to be monitored continuously. Aim of the study is to calculate the liquid volume inside oil tank on any road condition and send the volume data and location data to the user. This research use some ultrasonic sensors (to monitor the fluid height), Bluetooth modules (to sent data from the sensors to the Arduino microcontroller), Arduino Microcontroller (to calculate the liquid volume), and also GPS/GPRS/GSM Shield module (to get location of vehicle and sent the data to the Server). The experimental results show that the accuracy rate of monitoring liquid volume inside tanker while the vehicle is in the flat road is 99.33% and the one while the vehicle is in the road with elevation angle is 84%. Thus, this system can be used to monitor the tanker position and the liquid volume in any road position continuously via web application to prevent illegal theft.

Results of availability imposed configuration details developed for K-DEMO

The Korean fusion demonstration reactor (K-DEMO) has completed a two year study looking at key Tokamak components and configuration options in preparation of a conceptual design phase. A key part of a device configuration centers on defining an arrangement that enhances the ability to reach high availability values by defining design solutions that foster simplified maintenance operations. To maximize the size and minimize the number of in-vessel components enlarged TF coils were defined that incorporate a pair of windings within each coil to mitigate pressure drop issues and to reduce the cost of the coils. A semi-permanent shield structure was defined to develop labyrinth interfaces between double-null plasma contoured shield modules, provide an entity to align blanket components and provide support against disruption loads—with a load path that equilibrates blanket, TF and PF loads through a base structure. Blanket piping services and auxiliary systems that interface with in-vessel components have played a major role in defining the overall device arrangement—concept details will be presented along with general arrangement features and preliminary results obtained from disruption analysis.

Final design of the generic upper port plug structure for ITER diagnostic systems

The generic upper port plug (GUPP) structure in ITER is a 6m long metal box which deploys diagnostic components into the vacuum vessel. This structure is commonly used for all the diagnostic upper ports. The final design of the GUPP structure, which has successfully passed the final design review in 2013, is described here.The diagnostic port plug is cantilevered to the vacuum vessel with a heavy payload at the front, so called the diagnostic first wall (DFW) and the diagnostic shield module (DSM). Most of electromagnetic (EM) load (∼80%) occurs in DFW/DSM. Therefore, the mounting design to transfer the EM load from DFW/DSM to the GUPP structure is challenging, which should also comply with thermal expansion and tolerance for assembly and manufacturing. Another key design parameter to be considered is the gap between the port plug and the vacuum vessel port. The gap should be large enough to accommodate the remote handling of the heavy port plug (max. 25t), the structural deflection due to external loads and machine assembly tolerance. At the same time, the gap should be minimized to stop the neutron streaming according to the ALARA (as low as reasonably achievable) principle. With these design constraints, the GUPP structure should also provide space for diagnostic integration as much as possible. This requirement has led to the single wall structure having the gun-drilled water channels inside the structure. Furthermore, intensive efforts have been made on the manufacturing study including material selection, manufacturing codes and French regulation related to nuclear equipment and safety.All these main design and manufacturing aspects are discussed in this paper, including requirements, interfaces, loads and structural assessment and maintenance.

Shutdown dose rate analysis of European test blanket modules shields in ITER Equatorial Port #16

‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5×106n/cm2s at port interspace, while the EPP gaps and BSM gaps are responsible for 5×107n/cm2s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121μSv/h averaged near the port closure flange was obtained, out of which, only 4μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range of 300–350μSv/h were observed in the interspace. Thus, although the SDDR are found above the project target the good performance of the TBM shields design was demonstrated. Two main reasons for high SDDR were identified in this work—the neutron streaming through the equatorial PP gaps and the neutron penetration through the blanket shield modules and vacuum vessel.

Multiphysics Engineering Analysis for ITER Diagnostic First Wall and Shield Module Design*

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

PPPL ST-FNSF Engineering Design Details

One of the goals of the PPPL Spherical Tokamak (ST) Fusion Nuclear Science Facility (FNSF) study was to generate a self-consistent conceptual design of an ST-FNSF device with sufficient physics and engineering details to evaluate the advantages and disadvantages of different designs and to assess various ST-FNSF missions. This included striving to achieve tritium self-sufficiency; the ability to provide shielding protection of vital components and to develop maintenance strategies that could be used to maintain the in-vessel components (divertors, breeding blankets, shield modules and services) and characterize design upgrade potentials to expanded mission evolutions.With the conceptual design of a 2.2 m ST pilot plant design already completed emphasis was placed on evaluating a range of ST machine sizes looking at a major radius of 1m and a mid-range device size between 1 m and 2.2 m.This paper will present an engineering summary of the design details developed from this study, expanding on earlier progress reports presented at earlier conferences that focused on a mid-size 1.7 m device. Further development has been made by physics in defining a Super-X divertor arrangement that provides an expanded divertor surface area and places all PF coils outside the TF coil inner bore, in regions that improve the device maintenance characteristics. Physics, engineering design and neutronics analysis for both the 1.7 m and 1 m device have been enhanced. The engineering results of the PPPL ST-FNSF study will be presented along with comments on possible future directions.

Neutronic analysis of the Diagnostic Equatorial Ports in ITER

The Diagnostic Port Plugs provide an infrastructure for integrating the diagnostics necessary to operate the machine, while withstanding the nuclear and mechanical loads. They have to provide neutron shielding in order to ensure accessibility to the Port Interspace (PI) region for maintenance operations. This report is about the study to mitigate the ITER Shutdown Dose Rate in Port Interspace. This is a strong design driver for the Equatorial Port Plugs (EPPs) requiring a neutron attenuation of 7 orders of magnitude while maintaining the weight limit. To achieve this challenging task, the design has evolved to an improved double labyrinth configuration with welded shims at the back of the EPP to reduce the gap with the Vacuum Vessel Port Extension. This proved to be very successful for this task. Neutron stoppers have been placed after the gaps necessary for the insertion of the Diagnostic Shielding Modules (DSMs) housed inside the EPP. Effort was also placed in the further development of the DSMs. The latest design relies on a stainless steel structure and lightweight efficient neutron absorbing materials inside.Studies of cross talk with lower ports and the influence of the streaming through the blanket area are presented to give the status of the radiation environment in the Port Interspace region and their contribution to the Shutdown Dose Rate (SDR).

Final design of the generic equatorial port plug structure for ITER diagnostic systems

The Diagnostic Generic Equatorial Port Plug (GEPP) is designed to be common to all equatorial port-based diagnostic systems. It is designed to survive throughout the lifetime of ITER for 20 years, 30,000 discharges, and 3000 disruptions. The EPP structure dimensions (without Diagnostic First Walls and Diagnostic Shield Modules) are L2.9×W1.9×H2.4m3. The length of the fully integrated EPP is 3174mm. The weight of the EPP structure is about 15t, whereas the total weight of the integrated EPP may be up to 45t. The EPP structure provides a flexible platform for a variety of diagnostics. The Diagnostic Shield Module assemblies, or drawers, allow a modular approach with respect to diagnostic integration and maintenance. In the nuclear phase of ITER operations, they will be remotely inserted into the EPP structure in the Hot Cell Facility. The port plug structure must also contribute to the nuclear shielding, or plugging, of the port and further contain circulated water to allow cooling during operation and heating during bake-out. The Final Design of the GEPP has been successfully passed in late 2013 and is now heading toward manufacturing. The final design of the GEPP includes interfaces, manufacturing, R&D, operation and maintenance, load cases and analysis of failure modes.

Design modification of ITER equatorial EC launcher for electron cyclotron heating and current drive optimization

The design modification of the ITER equatorial electron cyclotron (EC) launcher to enhance off-axis current drive performance, changing the steering capability of injection beams from toroidal to poloidal, is described. The modification requirement impacts the design of all 3millimeter wave beam paths transmitted through the port plug and blanket shield modules (BSMs) installed in front. The optimized configuration of the mm wave design consisting of both steering and fixed mirrors, beam ducts between the waveguide outlet and BSM openings are successfully developed. Assuming a 100% pure HE11 mode propagated at the waveguide output, a 99% transmission efficiency from the waveguide outlet to the BSM output is obtained. This modification ensures that mm wave beams from both middle and bottom beam row pass through the same BSM opening and will lead to the increase of neutron shielding potential. The configurations of BSMs are also modified maintaining the structural design concept of each module. The preliminary result of electromagnetic analysis shows that induced force and momentum torque on the modified BSMs are below the mechanical criteria of the BSMs’ support structure.

Development of ITER Equatorial EC Launcher Components Toward the Final Design

Development of an electron cyclotron equatorial launcher has proceeded with a series of prototype tests and design enhancements intended to improve reliability and functionality of the launcher. The design enhancements include adaptation of the launcher steering angles such that one of three beam rows of the launcher is necessarily flipped to perform counter current drive to conform to a new ITER physics requirement. Also, the top and bottom steering rows have been tilted at an angle of 5 deg so that the top and bottom beam rows can be accessed from on-axis to near midradius. Furthermore, the position of the fixed focusing mirror that forms a quasi-optical in-vessel millimeter-wave (mm-wave) transmission line is modified to increase the nuclear shielding capability. A high-power experiment of the mm-wave launching system mock-up fabricated based on the design confirmed a successful steering capability of 20 to 40 deg. It was measured that some stray radio-frequency power propagated in the beam duct and generated some heat on the duct. Prototype tests also include the fabrication of the blanket shield module and partial port plug mock-up and have shown no serious technological issue regarding the fabrication and cooling functionality.

Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail.

F4E studies for the electromagnetic analysis of ITER components

Fusion for Energy (F4E) is involved in a relevant number of activities in the area of electromagnetic analysis in support of ITER general design and EU in-kind procurement. In particular several ITER components (vacuum vessel, blanket shield modules and first wall panels, test blanket modules, ICRH antenna) are being analyzed from the electromagnetic point of view. In this paper we give an updated description of our main activities, highlighting the main assumptions, objectives, results and conclusions. The plasma instabilities we consider, typically disruptions and VDEs, can be both toroidally symmetric and asymmetric. This implies that, depending on the specific component and loading conditions, FE models we use span from a sector of 10 up to 360° of the ITER machine. The techniques for simulating the electromagnetic phenomena involved in a disruption and the postprocessing of the results to obtain the loads acting on the structures are described. Finally we summarize the typical loads applied to different components and give a critical view of the results.

Dynamic Amplification Factor of the ITER Diagnostic Upper Port Plug

The diagnostic upper port plug in ITER is a long metal box cantilevered to the vacuum vessel port with 42 × M52 studs and nuts. The plug structure has a heavy payload at the front, such as the diagnostic first wall and the diagnostic shield module to protect the diagnostic components from plasma and neutron fluxes. This kind of structural configuration is susceptible to a resonance with the transient external load. For the upper port plug, the design-driving load is electromagnetic (EM) forces due to plasma disruptions. In this paper, the dynamic amplification factor (DAF) of the structure is calculated for such EM loads. The bolted joint at the back flange of the plug structure is also considered together with the port extension of the vacuum vessel and its influence on the dynamic behavior is investigated. The analysis results show that the bolted joint reduces the DAF as well as the natural frequency of the structure.