Remote Handling Requirements Research Articles

Design and thermal-hydraulic analysis for CFETR divertor Dome

The main function of China Fusion Engineering Test Reactor (CFETR) Divertor Dome is to isolate neutral particles and impurity particles as well as improve the ability of particle exclusion. For the CFETR Divertor Dome, a new integral hybrid divertor-blanket structure was proposed in previous study. The new Dome should also be compatible with the Remote handling (RH) requirements due to neutron activation of in-vessel components, thus special RH channels are designed on the front of the plasma-facing units (PFUs). Due to direct bombardment or sputtering of plasma, the Dome will suffer heat flux up to 10 MW/m2 and nuclear heat under 1.5 GW fusion power during steady-state operation, which puts forward critical and tricky requests for the design of cooling system. While for the RH channels, the situation is even direr due to its location and poor cooling conditions. In this paper, cooling system of the Dome is proposed, and cooling channels with special thermal transfer structure are designed in the PFUs to guarantee them with heat removal ability up to 10 MW/m2. Special PFUs and cooling system are designed to adapt to the RH, to shadow them from the direct bombardment of plasma and provide enough heat removal capability. Thermal-hydraulic analysis is carried out to verify the heat transfer ability of the Dome cooling system. The results will provide supports to the electromagnetic and structure analysis of Dome and promote the final design of CFETR water-cooled divertor.

Preliminary engineering assessment of alternative magnetic divertor configurations for EU-DEMO

• The present work deals with a preliminary engineering development of alternative magnetic divertor configurations for EU-DEMO. • Parametric CAD master models for each configuration have been developed, including PF and TF coils, VV, ports and in-vessel components. • The configurations have been analyzed in terms of geometrical envelope and ports layout, considering structural and Remote Handling (RH) requirements. One of the main challenges in the roadmap to the realization of fusion energy is to develop a heat and power exhaust system able to withstand the large loads expected in the divertor of a fusion power plant. The challenge of reducing the heat load on the divertor targets is addressed, within the mission 2 ‘Heat-exhaust systems’, through the investigation of divertor configurations alternative to the standard Single Null (SN), such as the Snowflake (SF), Double Null (DN), X and Super-X (SX) divertors. This paper focuses on a preliminary engineering assessment of the alternative configurations proposed for the EU DEMO reactor. Starting from the description of the optimized plasma shape developed for each configuration, the 3D geometrical description of the Magnet System and of the main Mechanical Structures (Vacuum Vessel and in-vessel components) is presented. Based on the 3D geometry, the compatibility of the location and dimensions of ports with Remote Maintenance needs is discussed and possible design optimizations are proposed both for the Magnets system and the mechanical structures design. Finally, the various configurations are compared with regard to the engineering and feasibility aspects.

Conceptual design of CFETR divertor supporting structures based on DOF and stability theories for the remote handling requirements

The divertor area of China Fusion Engineering Test Reactor (CFETR) vacuum vessel (VV) consists of 48 modular cassettes which must be replaced during operation of the CFETR. Considering the working environment for the divertor in Tokamak, the process for replacing is conducted by remote handling (RH) ways. In this paper, the work mainly focuses on the development of a new approach to the conceptual design of CFETR Divertor RH compatible structures. The approach is based on two theories: degree of freedom (DOF) and the stability.DOF theory requires the divertor should be constraint completely, and then its DOF is equal to zero. Besides that, the stability theory requires the divertor cannot be shaken due to the projection of the centre of mass and cannot be rotated due to the larger tilting moment. Then a new design scheme for the compatible structures is put forward based on theories to fulfill the working requirements. Finally, the simulation process in Delmia is carried on in order to validate the RH feasibility.

A Haptic Shared-Control Architecture for Guided Multi-Target Robotic Grasping.

Although robotic telemanipulation has always been a key technology for the nuclear industry, little advancement has been seen over the last decades. Despite complex remote handling requirements, simple mechanically linked master-slave manipulators still dominate the field. Nonetheless, there is a pressing need for more effective robotic solutions able to significantly speed up the decommissioning of legacy radioactive waste. This paper describes a novel haptic shared-control approach for assisting a human operator in the sort and segregation of different objects in a cluttered and unknown environment. A three-dimensional scan of the scene is used to generate a set of potential grasp candidates on the objects at hand. These grasp candidates are then used to generate guiding haptic cues, which assist the operator in approaching and grasping the objects. The haptic feedback is designed to be smooth and continuous as the user switches from a grasp candidate to the next one, or from one object to another one, avoiding any discontinuity or abrupt changes. To validate our approach, we carried out two human-subject studies, enrolling 15 participants. We registered an average improvement of 20.8%, 20.1%, and 32.5% in terms of completion time, linear trajectory, and perceived effectiveness, respectively, between the proposed approach and standard teleoperation.

Initial port integration concept for EC and NB systems in EU DEMO tokamak

The integration of the heating and current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in particular heat, electromagnetic (EM), and pressure loads in normal and off-normal conditions. A number of pre-conceptual design options for the vacuum vessel (VV) port and the port-plug are under assessment and need to be verified against all requirements and related criteria. The identification of the functional (or physics) requirements of the HCD systems remains an on-going process during the pre-conceptual design phase, hence some initial assumptions had to be made as a basis for development of the design of the VV ports and the HCD port plugs.The paper will provide an overview of present margins in the functional/physics requirements and the rationale behind the assumptions made in order to facilitate development of the pre-conceptual design options. Furthermore it will introduce the initial design concepts of the electron cyclotron (EC) Launchers and the neutral beam (NB) injectors integrated in equatorial ports. The NB duct design in DEMO and related issues such as transmission and re-ionization losses will be also addressed.

A study of methods to enhance infrared emissivity of Molybdenum surfaces

Single crystal Molybdenum is one of the most promising materials for the First Mirror (FM) of ITER optical diagnostics due to its high resistance to erosion under neutral atom bombardment. Other advantages are: low CTE, high thermal conductivity, and advantageous mechanical properties at elevated temperatures. The FMs are normally located in the front-end of ITER port plugs, being subject to volumetric heat loads up to ∼1 W/cm3 and higher. Active cooling of Mo mirrors by water or gas flow has limited applicability due to remote handling, integration and risk requirements. An athermal design with radiative cooling was developed for the H-alpha FM unit (FMU) with an all-Mo/TZM structure comprising a TZM housing, two Mo mirrors and the TZM cleaning electrodes. The main idea is to minimize mirror angular displacements and to keep the alignment at temperatures up to ∼350 °C. That requires a balance between thermal contact and radiative heat sinks. The most uniform temperature profile is obtained by weak thermal contact between FMU and support structure, but that leads to the highest FM temperature, since the normal as-milled Mo surface has a very low effective emissivity (ε < 0.1 at 200…400 °C). Well-known tools for surface blackening are: enhanced roughness, oxidation, V-grooving, and coatings, which should be analyzed for outgassing rate, thermal radiation and long-term stability in Hydrogen environment. In this study, a number of techniques to enhance Mo surface effective emissivity were tested: V-grooving, detonation spray coating, drilling and surface electro-erosion. Spectral emissivity of the test samples was measured by Bruker Vertex-70 infrared Fourier spectrometer and the effective values were derived for subsequent thermal analyses. Alumina coatings were found to be the most effective tool with ε ≈ 0.8…0.9. However, V-grooving and drilling also provide acceptable results (εeff ≈ 0.25…0.3) and do not require qualification for applicability to ITER. The trade-off between different techniques is discussed in detail.

Status of Heating and Current Drive Systems Planned for ITER

An auxiliary heating power of > 70 MW is envisaged at ITER in order to obtain the plasma temperatures and plasma profiles required to achieve $Q > 10$ for 400 s (inductive ELMy H-mode) and $Q = 5$ for a pulse duration of 3600 s (noninductive discharge), for which a potential upgrade may be necessary. In addition to providing the desired heating, the systems are also expected to drive current, tailor the plasma profile, and control the plasma instabilities. It is difficult to realize such a broad range of functionalities through a single system. As a result, three systems are planned during the first operational phase of ITER, which include the neutral beam (NB), electron cyclotron (EC), and ion cyclotron (IC) systems. While the NB systems are expected to deliver 33 MW of heating power to the plasma and drive current through it by the use of two NB injectors, the IC and the EC systems deliver 20 MW each. IC covers a wide variety of heating and current drive schemes and EC heats the electrons providing local heating and current drive, which can be steered across the plasma cross section. These systems, operating in a hostile radiative environment, become radioactive due to the neutron flux from ITER. Coupled to this are the features of long operating pulse lengths at high powers, reduced maintainability, and increased remote handling requirements. As a result, these devices are challenging to realize compared with their present operational counterparts worldwide. In order to help mitigate the risks involved with the manufacturing, setting up, and operation of these systems at ITER, extensive prototyping and research and development (R&D) activities are underway at various laboratories of the participating domestic agencies (DAs) for each of these systems. This paper provides a brief description of the requirements of each of the three heating systems for the various plasma scenarios foreseen for ITER operation, their functional and technical advantages, the various developments over the period of time, and the present status of the prototype and R&D activities underway to realize these systems and the overall development schedule.

Use of virtual reality for optimizing the life cycle of a fusion component

Efficient development of a complex system such as a fusion component needs a stringent integration of standard and new constraints. For example, compared to the previous fusion experimental devices, remote handling (RH) and safety requirements are in ITER key parameters which must be integrated since the earliest design. For optimizing such integration studies, CEA, IRFM decided in 2010 to implement the use of virtual reality (VR) tools during the life cycle (from design to operation) of a fusion component. This paper describes a first feedback of such use for fusion engineering purposes.After a short overview of the CEA, IRFM VR platform capabilities, three main uses will be described: design review, simulation of remote handling and hands-on operations, with man in the loop.The Design review mode was intensively used within the framework of a fruitful collaboration with ITER design Integration Team. This mode, fully compatible with CAD software, enables scale one data visualization with stereoscopic rendering. It improves the efficiency in detecting inconsistencies inside models and machine sub-system design optimization needs. Several accessibility cases of major Safety Important Components (SIC-1) were studied giving important requirements to the design at an early stage.CEA, IRFM, in close collaboration with expertise of CEA, LIST for VR simulation software, applies VR technologies for designing RH maintenance scenario for ITER Test Blanket System (TBS) and Ion cyclotron Resonance Heating (ICRH) Port Plugs. RH compatibility studies using VR pointed out major design drivers while helping to propose credible solution. VR platform is intensively used in the design of WEST (Tungsten (W) Environment Steady-state Tokamak) components and assembly studies, providing important information about the feasibility of assembly processes, optimization of physical mock-ups and ergonomic posture and gestures of operator.Finally, new perspectives, as the integration of safety constraints (dose calculation) will be described, demonstrating the powerful of VR tools at different stages of the component lifecycle.

Advances in neutronics and radiological protection of HiPER 4a

The HiPER project, phase 4a, is evolving. In this study we present the progress made in the field of neutronics and radiological protection for an integrated design of the facility. In the current model, we take into account the optical systems inside the target bay, as well as the remote handling requirements and related infrastructure, together with different shields. The last reference irradiation scenario, consisting of 20MJ of neutron yields, 5 yields per burst, one burst every week and 30 years of expected lifetime is considered for this study. We have performed a characterization of the dose rates behavior in the facility, both during operation and between bursts. The dose rates are computed for workers, regarding to maintenance and handling, and also for optical systems, regarding to damage. Furthermore, we have performed a waste management assessment of all the components inside the target bay. Results indicate that remote maintenance is mandatory in some areas. The small beam penetrations in the shields are responsible for some high doses in some specific locations. With regards to optics, the residual doses are as high as prompt doses. It is found that the whole target bay may be fully managed as a waste in 30 years by recycling and/or clearance, with no need for burial.

The European Supplied Components of the ITER Vacuum Pumping Systems

The ITER Vacuum Pumping systems are designed to pump hydrogen isotopes, including tritium, helium and impurities. The EU is responsible for the in-kind supply of 8 torus-, 2 cryostat-, 3 Heating Neutral Beam-(HNB) and 1 Diagnostic Neutral Beam-(DNB) cryopumps, their 14 Cold Valve Boxes (CVBs) and the cryolines between them as well as the Leak Detection and Localization (LD&L) systems. The design of these systems has progressed well in the past with the exception of LD&L, where the results of a significant R&D program are required to define the design. The torus and cryostat cryopumps incorporate an inlet vacuum valve capable of controlling the pumping speed and require to be sufficiently instrumented for performance and inventory control. In order to fully define the related Procurement Arrangements (PA) and minimize any technological, programmatic or cost risks, a 1:1 Pre-Production cryopump (PPC) has been planned to be built and tested as well as a series of necessary R&D activities will need to be performed. These activities will address all technological challenges, specify the instrumentation needs of these Vacuum Pumping systems and analyze their performance, incorporating also all safety provisions and remote handling requirements.

Analysis of the new architecture proposal for the CMM control system

While developing divertor remote handling maintenance systems at the Divertor Test Platform 2 facility, some risks and sensitivity points related to the Cassette Multifunctional Mover control system software were discovered and evaluated. The control system architecture has to simultaneously fulfill the demanding ITER remote handling requirements and to face new requirements being uncovered during the trials. Especially evolving non-functional requirements such as reliability and safety have an effect on the control system architecture as it is getting more mature. An evaluation of the implications from architectural decisions is necessary before implementation efforts, as an architecture left to develop without evaluation may lead to a dead end and therefore soaring development costs. After studying existing architecture analysis methods an analysis method was developed to gain confidence to carry out the proposed changes.

The ITER EC H&CD upper launcher: Analysis of remote handling compatibility

The present design of the ECH (Electron Cyclotron Heating) upper port launcher has been evaluated in light of the ITER remote handling (RH) requirements. Changes to the launcher design associated with the accessibility, maintainability and manageability of replaceable components are presented. Captive bolts were placed along the flange of the Blanket Shielding Module (BSM). A hinge mechanism was integrated to simplify the (dis-)mounting of the BSM and a frame with incorporated cooling and actuation lines was suggested for simplified mounting and replacement of the steerable mirrors. Rotating the upper port plug upside-down improves maintenance access and component handling. Tools are proposed for manipulation of the port plug and its sub-components. The RH compatibility analysis can improve a design. Early consideration of RH requirements and implementation of necessary features is therefore vital.

Proposal of an improved design of IFMIF Test Cell components for enhanced handling and reliability

An adequate design of components to be manipulated by remote handling is a key factor in the success of any activated facility, having a decisive impact on availability, prompt and safe maintenance, occupational exposures and flexibility of the facility. Such components should satisfy at least the basic remote handling requirements of simplicity, accessibility, modularity, standardization and assembling adequacy. Highly activated components in the IFMIF facility are found in the Test Cell, a pit closed by stepped shielding plugs. The Test Cell confines the Test Modules which contain the samples and experiments. The present reference design of the IFMIF Test Cell shows some drawbacks, in particular the jamming tendency of the shielding plugs, slow and complex access to the Backplate, a low lifetime component, and difficult positioning of the Test Modules. This paper summarises several modifications aiming at improving, under such remote handling requirements, the present reference design of the Test Cell shielding plugs and aspects of the geometrical structure of the Test Cell. A functional modularization of the present shielding plugs has been carried out and positioning guides for the Test Modules have been devised.

Combined application of Product Lifecycle and Software Configuration Management systems for ITER remote handling

The advantages of Product Lifecycle Management (PLM) systems are widely understood among the industry and hence a PLM system is already in use by International Thermonuclear Experimental Reactor (ITER) Organization (IO). However, with the increasing involvement of software in the development, the role of Software Configuration Management (SCM) systems have become equally important. The SCM systems can be useful to meet the higher demands on Safety Engineering (SE), Quality Assurance (QA), Validation and Verification (V&V) and Requirements Management (RM) of the developed software tools. In an experimental environment, such as ITER, the new remote handling requirements emerge frequently. This means the development of new tools or the modification of existing tools and the development of new remote handling procedures or the modification of existing remote handling procedures. PLM and SCM systems together can be of great advantage in the development and maintenance of such remote handling system. In this paper, we discuss how PLM and SCM systems can be integrated together and play their role during the development and maintenance of ITER remote handling system. We discuss the possibility to investigate such setup at DTP2 (Divertor Test Platform 2), which is the full scale mock-up facility to verify the ITER divertor remote handling and maintenance concepts.

Setting up and managing a remote maintenance operation for fusion

Trying to set up and manage a remote maintenance operation for a thermonuclear fusion reactor is a complex undertaking. There are many problems and challenges which need addressing. This paper tries to guide the reader through this process by composing a list of generic problems and by analysing possible solutions. The first challenge before setting up a remote maintenance operation for a fusion reactor is the systematic analysis of all the remote handling requirements. Based upon this the remote handling concept, including facility layout and equipment, can be defined. The following aspects have to be considered and incorporated into the remote handling concept: • Remote handling task development. • Remote handling task logistics and resource management. • Command, control and human–machine interface system. • Viewing and camera systems. • Virtual reality and Augmented Reality software. • Automatic path planning and collision avoidance. • Remote transfer of heavy loads. • Maintainability of RH equipment. • Reliability, redundant systems and safety. • Rationalisation and modularity in both hardware and software. • Recovery from failure modes. • Condition monitoring and fault detection/prediction. • Ability to deal with unforeseen problems. Oxford Technologies Ltd. has a proven track record in setting up and running the Remote Handling group at the JET Joint Undertaking in Culham, UK. Based on the unique experience gained at JET, Oxford Technologies Ltd. also developed the current design and remote handling concept of the ITER Hot Cell during a study in 2004. Examples of both the JET remote handling experience and the ITER Hot Cell design and layout are given throughout this paper.

Design progress of the ITER divertor cassette-to-vacuum vessel locking system

Recent progress of the ITER divertor cassette-to-vacuum vessel locking system design and analyses is presented. The detailed thermal and non-linear static and dynamic stress finite element analysis have been performed to validate all design modifications of the locking system. These modifications have been driven mainly by remote handling requirements and reliability of locking system at action of various operational thermal and electromagnetic loads applied on divertor cassette.

Radiation Shielding for Fusion Reactors

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel.

Remote maintenance concepts for net

Abstract NET will produce fusion reactions, which will result in the structure and internal components of the machine becoming radioactive. Consequently, NET will use remote handling techniques to perform maintenance. The strategy being followed to ensure that NET is fully compatible with the remote handling requirements is described together with a preliminary list of the remote handling equipment being developed.

Development of a first wall clamp for remote handling

RFX is equipped with a graphite first wall made up of 2016 tiles individually supported from the vacuum vessel. Due to the small dimensions of the ports, human access inside the vacuum vessel is prevented and a robot manipulator has been designed to perform remote operations on the graphite tiles. The clamping system that withstands the applied loads must accomplish the remote handling requirements. A bayonet clamp has been chosen because it is simple, safe and can be built with clearances and chamfers that, with a proper compliance of the manipulator wrist, allow the compensation of the positioning errors between the parts that have to be matched. To avoid the vacuum welding of the clamp, the surfaces in contact under vacuum are coated with an anti-welding coating. The paper first describes the main design choices, then deals with the load capability of the clamping system and reports the results of experimental and FE analyses; the last section gives the results of the tests that were performed on anti-welding coatings.