ITER Port Research Articles

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module (TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development (R&D) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic R&D program for DEMO relevant technology development. In the HCCB concept Li2TiO3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept (case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept (case-2), the pebble bed is vertically (poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations. Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.

Electro-Magnetic Analysis of the ITER Upper Visible Infrared Wide Angle Viewing System

The Upper Visible Infrared Wide Angle Viewing System (UWAVS) is a diagnostic used in five upper ports of ITER. Each UWAVS provides visible and infrared views of various sections of the divertor. A single UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay mirrors to direct the in-vessel light out of the tokamak to the port cell camera sensors.For the in-vessel components, the transient electro-magnetic (EM) environment resulting from the ITER magnet operation and plasma events induces design driving Lorentz forces. As such, all in-vessel systems require detailed electro-magnetic finite element analysis (FEA) to derive the resulting time dependent Lorentz loads.ANSYS Maxwell software was used to perform transient electro-magnetic simulations of the UWAVS in ITER upper port 14. A 20 degree sector, cyclic symmetric model was employed and included, inner and outer vacuum vessel, blanket shield modules, diagnostic fist wall (DFW) and shield module (DSM), upper port plug structure, DSM shield blocks, and a detailed model of the UWAVS in-vessel assembly.The resulting data includes eddy current density and vector plots along with force and moment summation for various UWAVS components. Front end optical components are specifically reported as these components have significant EM loads.

Radiation In-Port Cross-Talks for ITER Port Diagnostics

This paper emphasizes the need of estimation of the mutual influence, called “cross-talk,” for neutronic analyses of neighboring diagnostics systems shared by the same ITER port. Using examples of several diagnostic systems inserted inside the ITER Equatorial and Upper Port Plugs, we have demonstrated this mutual influence. Cross-talk effects have been shown by examining the radiation environment inside the port plug in terms of neutron energy spectra and Shut-Down Dose Rate (SDDR) inside the Port Interspace (PI) area. In-port cross-talk was investigated for the diagnostic systems deployed in two Equatorial Port Plugs (EPP) #17 and #8, and for the components of Upper Port Plug (UPP) #3. One example of in-port cross-talks is a gamma shadow effect of the Tritium and Deposit Monitor (TDM) shield block, which affects the SDDR inside the PI of EPP#17. Where the gamma radiation originated from the dominant radioactive sources of the irradiated structures of Core-Imaging X-ray Spectrometer (CIXS) is blocked by the TDM shield. Another example is an influence of neutron streaming along the Fast Ion Loss Detector (FILD) channel on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) in EPP#8. For the example of UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), performed neutronic analysis identified excessive neutron streaming along the CXRS shutter, which must be reduced by further design iterations.

RAMI comparison of two China test blanket modules

Tritium breeding blanket module is one of the most important parts of fusion reactor. One of the main goals of ITER is to test and validate design concepts of tritium breeding blankets. China has proposed two test blanket module (TBM) concepts, including Helium cooled ceramic breeder test blanket module system (HCCB TBS) and the dual functional lithium-lead test blanket module system (DFLL TBS), which will be tested in one of the three equatorial ports of ITER. A detailed comparative study on the design schemes of these two TBSs concepts was performed in this paper. In this study, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was applied on those two test blanket systems. The modeling and analyzing process were done by the Reliability and Probabilistic Safety Assessment Software System RiskA. In addition, comparative study in the phases of functions, capabilities and efficiencies of tritium breeding, heat removal and coolant purification is also performed. It is found that HCCB TBS is a better choice currently to fulfill the engineering demand of ITER construction for its lower difficulty of technical implementation. However, DFLL TBS possesses a high performance prospect from the viewpoint of long-term ITER development due to its inherent advantages compared to HCCB TBS. Discussions of priorities for future research needed for these two TBS are also proposed in the paper.

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.

Rigid and adjustable fixation of the shielding modules into the ITER port plugs

During the final design review of Diagnostic Port Plugs, it has been highlighted that the current system of fixation, based on gaps, while it is not harmful for the port plug, it throws large uncertainties over the alignment of the optical systems placed inside the DSMs at the same time that the real mechanical behavior of the assembly is clearly unknown. Because the DSM is not rigidly fixed to the Port Plug structure, the DSM and its internal components suffer, during abnormal events like seismic or electromagnetic disruptions, an important level of amplification on the accelerations induced. This magnification is largely derived from a rattling effect that the DSM is vibrating inside the port plug. This effect produces an increase of the efforts acting on the anchoring components of the port plug at the same time that the final position of the DSM after an abnormal event is unknown and greatly influenced by the manufacturing and assembly tolerances. In order to solve all these issues, a new locking system has been developed based on a rigid configuration. Non-relevant modifications are required in the Port Plug structure or in the DSMs for the implementation of this new system because the interfaces are kept intact. Results based on this new configuration are available and they present an important attenuation of the induced accelerations on components inside the DSMs. In addition, the system permits the individual adjustment of each one of the interfacing component to finally achieve the correct aligned positions. Therefore, this new locking system has been assumed for all IO Diagnostic Port Plugs.

CAD-Based Shielding Analysis for ITER Port Diagnostics

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

Reliability of activation cross sections for estimation of shutdown dose rate in the ITER port cell and port interspace

This paper explores the quality of available activation cross section (XS) data for accurate Shutdown Dose Rate (SDDR) prediction in the ITER Port Cell and Port Interspace areas, where different maintenance activities are foreseen. For this purpose the EAF library (2007 and 2010 versions) has been investigated, as it is typically used by the ITER community. Based on both reports/papers on SDDR in ITER and own calculations, major nuclides contributing to the SDDR coming from the activation of i) relevant materials placed in ITER and ii) candidate materials for the bioshield plug as L2N and barite concretes, are identified. Then, relevant production pathways are obtained. EAF XS quality for all pathways is checked following the procedure used for validating and testing the successive EAF versions. Also, possible improvements from using the TENDL-2015 library are assessed by comparing EAF and TENDL XS with available differential experimental data from EXFOR. Results point out that most of the activation XS related to materials currently placed in ITER are reliable, and only a few need improvement. Also, many of the XS related to both L2N and barite concretes need further work for validation.

Engineering estimates of impurity fluxes on the ITER port plugs

Predictions of impurity fluxes are required for design analysis of the ITER optical diagnostics. In the present paper a simplified model is proposed for calculation of the neutral impurity fluxes on the recessed surfaces which are not in direct contact with plasma. The method is based on the Monte-Carlo simulation of the neutral particles transport in prescribed and fixed plasma background. The plasma parameters are projected from experimental observations, scalings and ITER modelling results. Blobs are approximated as stationary hot species. Results of 2D simulations with toroidally uniform wall and of the ‘2.5D model’ are presented. In this latter the 3D geometry of ports is implemented, but details of the incident ion flux distribution on the first wall panels are neglected. The calculated worst case gross deposition rate of Be in the middle of the port plug faces reaches almost 0.1 nm s−1. At the same time, the obtained Be erosion to deposition ratio at those locations is always larger than 5, indicating high probability of net erosion conditions there.

Upper wide-angle viewing system for ITER.

The Upper Wide Angle Viewing System (UWAVS) will be installed on five upper ports of ITER. This paper shows major requirements, gives an overview of the preliminary design with reasons for some design choices, examines self-emitted IR light from UWAVS optics and its effect on accuracy, and shows calculations of signal-to-noise ratios for the two-color temperature output as a function of integration time and divertor temperature. Accurate temperature output requires correction for vacuum window absorption vs. wavelength and for self-emitted IR, which requires good measurement of the temperature of the optical components. The anticipated signal-to-noise ratio using presently available IR cameras is adequate for the required 500 Hz frame rate.

Full scale test platform for European TBM systems integration and maintenance

This article deals with description and current status of a project of a non-nuclear, full size (1:1 scale) test platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER port cell #16. The facility called TBM platform reproduces the ITER port cell #16 and port interspace with all the relevant interfaces and mock-ups of the corresponding main components. Thanks to the modular design of the platform, it is possible to adapt or change completely the interfaces in the future if needed or required according to the updated configuration of TBSs. In the same way, based on customer requirements, it will be possible to adapt the interfaces and piping inside the mock-ups in order to represent also the other, non-EU configurations of TBM systems designed for port cells #02 and #18. Construction of this test platform is realized and funded within the scope of the SUSEN project.

Analysis of the impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field

The Test Blanket Module (TBM) will be used in the test port of ITER to demonstrate tritium self-sufficiency and the extraction of high grade heat for electricity production. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material was designed and fabricated to perform and validate relevant electromagnetic and thermal technologies of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) on the J-TEXT. Its size is one third of the CN HCCB-TBM. By using the finite element analysis technology, this paper analyzed the impacts on the equilibrium magnetic field over the plasma region after introducing the structure material RAFM steel. The distribution of toroidal field (TF) ripple and the magnitude of the error field with the mock-up at different positions were given. Simulation shows the distribution of the null field region formed by poloidal field (PF). The influence to tokamak discharge has been evaluated by drawing the magnetic field lines. Based on the results above, we have optimized and finished the installation of the mock-up to J-TEXT which meets the needs of the experiments and to ensure the normal discharge.

Design of the ITER VUV edge imaging spectrometer and R&D for the rotary motional vacuum feed-through

The ITER vacuum ultraviolet (VUV) spectrometer is aimed at monitoring impurity ion species in plasmas, and it consists of three systems, core survey, edge imaging, and divertor spectrometers depending on positions. In the VUV edge imaging spectrometer, a field mirror and a collimation mirror are used for imaging of the light onto the entrance slit of the VUV imaging spectrometer. To optimize the design of the field mirror box in the ITER port plug, the mirror box mock-up was developed to test installation and alignment. Especially, the mirror alignment using a special tilt stage was tested, and the motional feed-through with a double bellows was developed for the shutter operation. The preliminary qualification test for this rotary motional vacuum feed-through was performed to meet the functional requirement of ITER mechanical feed-through. Leak-tightness for over 15,000 times (larger than operation times) of 360 rotation and backward motion was tested for 1N.m loading condition to the axis rod of the feed-through. In the test, the vacuum inside of the feed-through was monitored, and the feed-through showed good rotation function and leak-tightness during test.

Thermal-hydraulics of LLCB TBM under different ITER operational conditions

Lead–lithium cooled ceramic breeder (LLCB) TBM is the Indian DEMO relevant blanket module, which will be tested in one-half of a designated ITER port from day 1 operation of ITER. LLCB TBM will be tested in various ITER operational phases. Different TBM may be tested corresponding to the four different phases of plasma operation. The paper present a review of thermal analyses based on the various heat loads on the LLCB TBM obtained from neutronics calculations to confirm heat removal and structural integrity under different ITER operational conditions. This includes the normal plasma operation, fusion power excursions, surface heat flux due to MARFES, disruption heat loads during thermal and current quench etc., in inductive and non-inductive plasma operations. Transient events, like plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux (high power densities) for short durations onto the first wall of TBM. Detailed thermal-hydraulic simulation studies have been performed using ANSYS. These analyses, therefore, will provide important information about the temperature distribution in different materials used in TBM and their temperature-time histories.

Design activities toward the achievement of the conceptual phase of the EU-TBM sets

Europe is developing two reference tritium breeder blanket concepts that will be tested in ITER under the form of Test Blanket Modules (TBMs): the helium–cooled lithium–lead (HCLL) which uses the liquid Pb–16Li as both breeder and neutron multiplier and the helium–cooled pebble-bed (HCPB) with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Both concepts are using the EUROFER reduced activation ferritic–martensitic steel as structural material and pressurized helium technology for heat extraction. In view of the conceptual design review, design activities of the TBM sets have been focused on specific areas. The convergence of TBMs design, developed in their pre-conceptual phase by different European Fusion laboratories, has been sought in order to optimize fabrication and maintenance in ITER port #16. The structural strength of the helium back manifold of both TBMs is now based on the HCLL stiffening rod concept proposed in 2010. Through iterations between the nuclear analyses and design, the radiation shield design reaches a good compromise between its shielding performances versus manufacturing. A shield-to-TBM attachment has been implemented which complies with constraints related to thermal expansion under operating conditions and mechanical strength to handle dynamic loads such as major disruptions and seismic event. A full set of thermomechanical, hydraulic and structural analysis of the TBM sets has been carried out. They cover the loads selected as design drivers in the conceptual phase along with exploration of operational domain in terms of helium cooling conditions versus RCC-MRx code criteria compliance. The analysis results show that thermal loads are the main design driver for most of TBM sub-components. Pressure loads are also a design driver for the TBM stiffening grids under in-TBM loss of coolant accident. From the analyses carried out for the shields, the most demanding scenario has been demonstrated to be the baking mode. For these internal pressure loads, the capacity checks carried out have demonstrated the feasibility and potential of the shield design proposed.

Final design of ITER port plug test facility

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment.The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs.The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented.

Structural damages prevention of the ITER vacuum vessel and ports by elasto-plastic analysis with regards to RCC-MR

Several types of damages have to be prevented in order to guarantee the structural integrity of a structure with regards to RCC-MR; the P-type damages which can result from the application to a structure of a steadily and regularly increasing loading or a constant loading and the S-type damages during operational loading conditions which can only result from repeated application of loadings associated to the progressive deformations and fatigue.Following RCC-MR, the S-type damages prevention has to be started only when the structural integrity is guaranteed against P-type damages. The verification of the last one on the ITER vacuum vessel and ports has been performed by limit analysis with elasto-(perfectly)plastic material behavior. It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. Some elasto-plastic analyses have been performed considering several cyclic loadings to evaluate also more realistic structural margins of the against S-type damages.

Shielding optimisation of the ITER ICH&CD antenna for shutdown dose rate

The Ion Cyclotron Heating and Current Drive (ICH&CD) system will reside in ITER equatorial port plugs 13 and 15. Shutdown dose rates (SDDR) within the port interspace are required to be less than 100μSv/h at 106s cooling. A significant contribution to the SDDR results from neutrons streaming down gaps around the port frame, and the mitigation of this streaming is the main subject of these analyses. An updated MCNP model of the antenna was created and integrated into an ITER reference model. Shielding plates were defined in the port gaps, and scoping studies conducted to assess their effectiveness in several configurations, based on which a front dog-leg arrangement was selected for high resolution 3-D activation analysis using MCR2S.It was concluded that the selected configuration reduced the SDDR from ∼500μSv/h to 220μSv/h but were still in excess of dose rate requirements. Approximately 30% of this was due to cross-talk from neighbouring ports. In addition, increased dose rates were observed in the port interspace along the lines of sight of the removable vacuum transmission lines. Design optimisation is continuing, however an integrated approach is needed with regard to ITER port plug design and the shielding of surrounding systems.

Construction of a test platform for Test Blanket Module (TBM) systems integration and maintenance in ITER Port Cell #16

This paper describes a project of a non-nuclear, 1:1 scale testing platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER Port Cell #16. This TBM platform is currently under construction in Centrum výzkumu Řež, Czech Republic. The facility is realized within the scope of the SUSEN project and its full operation is foreseen in the first half of 2016.

Dose rate analysis for the diagnostic generic equatorial port plug in the ITER port plug test facility

The port plug test facility (PPTF) in the ITER hot cell building (HCB) will be used to test activated equatorial port plugs (EPP). The design of the PPTF and the HCB walls adjacent to the PPTF aims to have the docking room, which is next to the test tank room containing the PPTF, having as low as reasonable dose rate of 25μSv/h. In this work, the R2Smesh approach, developed by KIT for the determination of shut-down dose rate distributions in and around activated components, was used to provide high resolution maps of dose rate distributions in the test tank room, the docking room and the service room, assuming an activated diagnostic generic equatorial port plug (GEPP) module in the PPTF. The results show that the shielding is not enough with the preliminary PPTF and rooms design. A design optimization was performed in order to improve the radiation shielding around the PPTF. The results show that with an optimized design dose rates less than 25μSv/h can be achieved in the docking room and the service room.