Divertor Cassette Research Articles

Radiation tolerance testing of piezoelectric motors for ITER (first results)

As a part of ITER diagnostics design and development, ultrasonic and stick-slip piezoelectric motors and several types of piezoelectric ceramics were tested under severe neutron and gamma ray irradiation (∼ 1019 per cm2 at E > 0.1 MeV) in a pool-type fission reactor. The fission neutron and gamma ray spectra were corrected to comply with conditions expected under ITER divertor cassettes. Taking into account possible neutron-induced amorphization of crystal structure, four types of piezoelectric ceramics were pre-selected. To restore the electronic structure of piezoelectric ceramic damaged by gamma rays, recovery polarization (i.e. heating while applying reverse bias) was suggested. The motors and piezoceramic samples were assessed for performance before and after irradiation as well as after recovery polarization. Recovery polarization was performed at 330 °C as it is expected close to the divertor cassettes during baking. Two most promising types of piezoelectric ceramics survived multiple cycles of depolarization/polarization at both 330 °C and 200 °C as appropriate for baking ITER divertor and vacuum vessel, respectively. All the experiments and procedures are described in detail.

Hydraulic Characterization of the Full Scale Mock-Up of the DEMO Divertor Outer Vertical Target

In the frame of the pre-conceptual design activities of the DEMO work package DIV-1 “Divertor Cassette Design and Integration” of the EUROfusion program, a mock-up of the divertor outer vertical target (OVT) was built, mainly in order to: (i) demonstrate the technical feasibility of manufacturing procedures; (ii) verify the hydraulic design and its capability to ensure a uniform and proper cooling for the plasma facing units (PFUs) with an acceptable pressure drop; and (iii) experimentally validate the computational fluid-dynamic (CFD) model developed by the University of Palermo. In this context, a research campaign was jointly carried out by the University of Palermo and ENEA to experimentally and theoretically assess the hydraulic performances of the OVT mock-up, paying particular attention to the coolant distribution among the PFUs and the total pressure drop across the inlet and outlet sections of the mock-up. The paper presents the results of the steady-state hydraulic experimental test campaign performed at ENEA Brasimone Research Center as well as the relevant numerical analyses performed at the Department of Engineering at the University of Palermo. The test facility, the experimental apparatus, the test matrix and the experimental results, as well as the theoretical model, its assumptions, and the analyses outcomes are herewith reported and critically discussed.

Power exhaust concepts and divertor designs for Japanese and European DEMO fusion reactors

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan–Europe broader approach DEMO design activity (BA DDA). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling (P rad tot/P heat ⩾ 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P sep/R p handling of 30–35 MW m−1 in order to maintain the radiation fraction in the main plasma at the ITER-level (f rad main = P rad main/P heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f rad main to ∼0.65 and handling the ITER-level P sep/R p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P sep = 250–300 MW and 150–200 MW, respectively. Both results showed that large divertor radiation fraction (P rad div/P sep ⩾ 0.8) was required to reduce both peak q target (⩽10 MW m−2) and T e,i div. In addition, the JA divertor performance with EU-reference P sep of 150 MW showed benefit of the closed geometry to reduce the peak q target and T e,i div near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C–200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.

Requirements and interface management for the divertor system design and integration in the Divertor Tokamak Test (DTT) facility

The construction of a Divertor Tokamak Test (DTT) facility is fully included in the European effort towards the realization of fusion energy with the aim of studying possible alternative solutions, with respect to the ITER research program, to tackle the problem of power exhaust. This means that the machine must be compatible with different possibilities for the divertor cassette design and the plasma scenario. The design of first “reference” divertor is largely based on the ITER-like divertor, but in the future other solutions are going to be tested, including liquid metal divertors. To provide a justification basis for the design of the reference cassette and guidelines for the future experimentations it is fundamental to define the requirements of this power exhaust system, in particular the heat flux compatibility and the interfaces with the surrounding interacting systems of the tokamak (the water cooling pipes, the divertor magnetic coils, the divertor diagnostics, etc.). This work shows how the divertor system can be included in a wider Requirement Management procedure that has been defined for DTT highlighting the relevant interface parameters. Additionally, a data analysis tool has been developed to provide an interactive visualization of requirements traceability to directly show the impact of changes in the divertor on the design and operations of the machine and viceversa. In this way, during future studies on alternative design divertors or plasma scenarios, the limiting constraints on both sides can be easily individuated and traced to their motivation, and possibly discussed.

The DTT device: Preliminary remote maintenance strategy

In this paper we show an overview of the preliminary strategy planned for remote maintenance of neutron–activated and contaminated components of DTT machine, in the 2019 reference configuration. The remote maintenance of such a complex machine has impact on different aspects of the DTT machine: layout of the tokamak hall, vacuum vessel and cryostat structures, in–vessel components. To date, the number and size of vacuum vessel ports as well as the segmentation and size of in–vessel mechanical components of DTT have been established by a compromise between operational and maintenance needs. An extensive multidisciplinary work has been done in deriving the requirements for the DTT remote maintenance strategy. Each vacuum vessel sector is divided into five ports: in the current configuration, the top and bottom divertor cassettes are expected to be removed, respectively, from four lower lateral ports and from four equatorial horizontal ports; the first wall modules at the inboard side are expected to be installed/removed from all the upper ports; the first wall modules at the outboard and upper sides are segmented such that they can be removed from four equatorial horizontal ports. The work describes the current strategies for divertor and first wall remote maintenance systems, as well as a first conceptual design of the remote maintenance equipment of DTT machine.

Design validation of in-vessel mirrors and beam dump for first plasma operations in ITER

First plasma operation in ITER will start after completing the assembly of the tokamak vessel and the installation of the main sub-systems, but prior to the installation of the blanket modules and the divertor cassettes. Utilization of temporary limiters and divertor replacement structures will provide a poloidal and toroidal reference side position to the plasma edge to protect the vacuum vessel and other components already installed during operations. An additional set of mirrors is required to reflect the power injected from the upper launcher towards the plasma resonance for EC-assisted breakdown of the plasma up to a beam dump needed to trap and absorb the power of the beams after crossing the plasma in order to reduce the stray radiation escaping back into the vacuum chamber down to less than 10% of the total power. The quasi-optical system has been designed, with shape and size of the mirrors compliant with the requirements provided by ITER for their installation, realization and plasma performances, resulting in two standard focussing mirrors and one grating mirror. The beam dump consists of a box with five metal plates, the first providing a spreading of the high incident power and the others coated with absorbing material with thickness distribution studied to gradually reduce the power during the multiple reflections inside the box, avoiding damages to the coating itself. This work focuses on the validation of the quasi-optical design of the mirrors and the assessment of the dump performances, based on a multi-bounces model developed ad-hoc for this purpose. The study includes a tolerance analysis for the beam dump to include the effect of uncertainties in the thickness of the absorbing coating and misalignments of the mirrors, to verify the performances of the dump also when operating in different conditions with respect to the nominal ones.

Manufacturing of lower divertor cassette of JT-60SA

First divertor cassette covering a 10° sector in the toroidal direction has been manufactured for JT-60SA. Modularized plasma facing components such as inner and outer targets, baffles, and domes are assembled into the divertor cassette. All plasma facing components are actively cooled. Carbon armor tiles bolted to water-cooled copper alloy heat sinks are used to remove heat loads of 0.3–1MW/m2 for 100 s and 10 MW/m2 for less than 5 s. The divertor cassette has two pipe joints between it and the inlet and outlet pipes on the vacuum vessel side. After removing the divertor cassette, the inner target can be replaced by cutting and welding the other two pipe joints with the divertor cassette. A remote pipe welding tool has been developed both to install the divertor cassette into the vacuum vessel and to connect the inner target with the divertor cassette. The space around the cooling pipes is so limited that they are cut and welded remotely from the inside for replacement. The laser welding method was used, and the focusing mirror inside the pipe was rotated to perform circumferential welding. As a countermeasure for welding groove misalignment, the aiming accuracy of the laser was improved by simultaneously controlling the rotation and lifting of the tool. Thereby, reliable welding is achieved even with large gaps of up to 0.5 mm and grooves with a maximum angular deviation of 0.5 ° between connecting pipes. The outer diameter of the cooling pipe is 59.8 mm, the wall thickness is 2.8 mm, and the material is SUS316L. After connecting the inner target, the other plasma facing components have been assembled into the first lower divertor cassette. When installing the divertor cassette towards the vacuum vessel, the laser welding tool described above is used to connect the cooling pipes between divertor cassette and vacuum vessel in JT-60SA, because the outer diameter and wall thickness of the cooling pipe are the same as the inner target.

Direct steam condensation at sub-atmospheric pressure: experimental tests and similitude analysis

This paper deals with experimental and theoretical analyses of the steam direct condensation at sub-atmospheric pressure. These operation conditions occur in the Vacuum Vessel Pressure Suppression System (VVPSS) of ITER Nuclear Fusion Reactor. This safety system permits to manage the Ingress of Coolant Event (ICE cat. IV) accident condition, postulated to occur in the ITER Vacuum Vessel (VV) as a consequence of the rupture of the shielding blanket cooling piping or the rupture of the divertor cassette piping.Experimental tests of steam direct condensation at sub-atmospheric pressures in a 1:22 scale facility were performed at the laboratory Guerrini of DICI-University of Pisa. Presently a full-scale facility is being building at DICI. This paper illustrates the similitude analysis elaborated to scale up the experimental results obtained in the reduced scale facility. The main variables which influence the steam direct condensation were analyzed and the relative scale laws determined. CFD numerical simulations of direct steam condensation in the actual geometry and in a vessel of equal volume permitted to verify some of the assumptions on which the scale laws are based. Scaling studies were first performed in order to calculate the transient of steam mass flow rate occurring during the ICE IV event in the scale apparatus, starting from the thermal hydraulic studies performed at ITER and successfully compared with the experimental results carried out in scaled apparatus. This confirmed the suitability of scale laws and in the same time the capability of the VVPSS to condense the injected steam at sub-atmospheric pressure, matching the safety goal to reduce the system pressurization.

DEMO Divertor preliminary safety assessment

European DEMOnstration plant is currently finalizing a pre-conceptual design phase. A carefully selected set of safety analyses are contextually being performed on DEMO systems in order to assess its integrated performance and capability to achieve expected targets while keeping in a safe operation domain. A key in-vessel system is the Divertor, in charge of exhausting a major part of the plasma ions thermal power in a region far from plasma core to control plasma pollution. To accomplish this mission a modular design approach was adopted based on two main sub-systems: the divertor cassette and plasma facing units (PFUs). Each cassette supports two vertical plasma facing units, one inboard and one outboard, intercepting the main part of the particle loads from the plasma. An assembly of 3 of such divertor cassettes and related PFUs is then foreseen for each of DEMO 16 sectors, hence resulting into 48 cassettes. Two independent heat transfer systems for cassettes and plasma facing units respectively provide heat sink function and connect the Divertor to DEMO Balance of Plant.An overview of the preliminary safety analyses performed on such system is here proposed. At first, a safety-perspective overview of current design is proposed. In particular, moving from the divertor system functions, specific component failure modes and related possible consequences, a set of relevant postulated initiating events are identified.Then, aiming at investigating whether accident consequences remain within accepted safe domain, two accidents from this set were selected to be analysed deterministically by means of MELCOR code. In particular, an in-vessel Loss of Coolant Accident initiated by a large rupture of the divertor PFU cooling pipes and a Loss of Flow Accident (LOFA) initiated by heat transfer system (HTS) pump seizure have been analysed. The nodalization includes divertor PFU and related HTS among other in-vessel components and safety devices (e.g., Vacuum Vessel Pressure Suppression System) devoted to Vacuum Vessel pressurization transient control or Fast Plasma Shutdown System. Parametric analyses were also performed to investigate possible impact of the design solutions (e.g., implementation of isolation valves in Divertor loop) on the accident consequences.

Thermal-hydraulic study of the DEMO divertor cassette body cooling circuit equipped with a liner and two reflector plates

In the framework of the Work Package DIV 1 – “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the steady-state thermal-hydraulic performances of the DEMO divertor cassette cooling system. The research activity has been focussed onto the most recent design of the Cassette Body (CB) cooling circuit, consistent with the DEMO baseline 2017 and equipped with a liner and two Reflector Plates (RPs), whose main functions are to protect the underlying vacuum pump hole from the radiation arising from plasma and shield the PFCs inlet distributors, respectively. The research campaign has been carried out following a theoretical-computational approach based on the finite volume method and adopting the commercial Computational Fluid-Dynamic (CFD) code ANSYS-CFX. The CB thermal-hydraulic performances have been assessed in terms of coolant and structure temperature, coolant overall total pressure drop and flow velocity distribution, mainly in order to check coolant aptitude to provide a uniform and effective cooling to CB, liner and RPs structures. Moreover, the margin against coolant saturation has been evaluated in order check whether any risk of its bulk vaporisation is prevented. The outcomes of the study have shown some criticalities, mainly in terms of coolant bulk vaporisation occurrence at the corner of the Inner Vertical Target (IVT) and uneven coolant flow distribution among RPs plasma-facing channels, that have suggested some design variations whose effectiveness has been numerically assessed. In particular, the solution proposed to contain and reduce the critical region of the IVT corner has been predicted to be particularly effective, while further studies are needed to improve coolant flow distribution among the RPs plasma-facing channels. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Shunt and Rogowski coil measurements on ASDEX Upgrade in support of DEMO detachment control

Detachment control in DEMO is a fundamental requirement to prevent damage to the plasma facing components. Thermo-currents flowing through the plasma facing components of the divertor cassette are driven by the thermoelectric voltage generated by the plasma temperature difference between the inner and outer target plates. Shunt and Rogowski coil measurements to measure thermo-currents for detachment control on ITER are planned. Thermo-current measurements on ASDEX Upgrade have been carried out in support of establishing designs considered suitable to measure thermo-currents in DEMO.

Development of the remote handling connector for ITER divertor diagnostic system

Diagnostics is a vital system for ITER to collect information during operation. Mineral insulated cables route the electrical and optical diagnostics sensor signals from the divertor area to the diagnostic hall. The diagnostic divertor cassettes are replaced during maintenance, which raises the requirement for a connector between divertor cassette and in-vessel wall. Since it needs to be handled remotely during the installation of the 16 diagnostic divertor cassettes, it is called Remote Handling Connector. Its operating space is limited and its environmental conditions are ultra-high vacuum, baking temperature 350 °C, irradiation and challenging electromagnetic forces. The preliminary design of the Remote Handling Connector is based on the conceptual design of Outboard and Inboard Configurations and is highly impacted by finding an appropriate balance between the external and internal space limitations and system requirements. The design of the system has been an iterative process including several conceptual and mechanical design phases, thermal, magnetic and structural load analysis, risk analysis, and remote handling assessment. Mock-ups were manufactured and tested at Divertor Test Platform 2 and Remote Handling Connector Platform at VTT Tampere Finland. The developed system was evaluated at a preliminary design review meeting organized at the end of 2019. The preliminary design phase provides a baseline for the final design of the Remote Handling Connector system.

Ceramic-Chromium Hall Sensors for Environments with High Temperatures and Neutron Radiation.

Ceramic-chromium Hall sensors represent a temperature and radiation resistant alternative to Hall sensors based on semiconductors. Demand for these sensors is presently motivated by the ITER and DEMO nuclear fusion projects. The developed ceramic-chromium Hall sensors were tested up to a temperature of 550 °C and a magnetic field of 14 T. The magnitude of the sensitivity of the tested sensor was 6.2 mV/A/T at 20 °C and 4.6 mV/A/T at 500 °C. The sensitivity was observed to be weakly dependent on a temperature above 240 °C with an average temperature coefficient of 0.014%/°C and independent of the magnetic field with a relative average deviation below the measurement accuracy of 0.086%. A simulation of a neutron-induced transmutation was performed to assess changes in the composition of the chromium. After 5.2 operational years of the DEMO fusion reactor, the transmuted fraction of the chromium sensitive layer was found to be 0.27% at the most exposed sensor location behind the divertor cassette with a neutron fluence of 6.08 × 1025 n/m2. The ceramic-chromium Hall sensors show the potential to be suitable magnetic sensors for environments with high temperatures and strong neutron radiation.

Heat pipe technology based divertor plasma facing component concept for European DEMO

Heat pipes (HP) are considered being used in the divertor target because of their high thermal conductivity and their capability to substantially enlarge the heat transfer area to the cooling circuit. Here, a divertor target concept based on a heat pipe design is introduced being able to dissipate heat fluxes of up to 20 MW/m2. It consists of a 230 mm long water-based heat pipe with a capillary structure that combines axial grooves with sintered porous material. The analysis of the integration of the HP in the divertor target indicates that design changes of the current divertor cassette design are required to accommodate such a solution.

Performance of first mirrors for divertor impurity monitor in ITER: compensation of misalignment and effect of beam divergence

We investigated feasibility of an alignment method using first mirrors of the divertor impurity monitor (DIM) in ITER based on ray tracing analysis. A misalignment could occur between the divertor cassette and measurement system. It is shown that the misalignment can be compensated to some extent by rotating the first mirror. The maximum compensable misalignment and the required accuracy in the alignment are discussed. It is likely that the surface roughness of the first mirrors gradually increases during ITER operation. We investigated the required quality of the first mirror as focusing on the beam divergence using the ray tracing model. Moreover, to check the performance, helium plasma irradiation was performed on molybdenum mirror samples, and the beam divergence was measured using a laser and a detector. Based on the simulations and experiments, it is shown that the helium effect does not harm the performance of the mirror when the mirror temperature is less than 800 K.

Simulation of EM loads acting on the plasma-facing units of ITER divertor cassette

Some aspects of electromagnetic (EM) simulation of the central divertor cassettes (CDC) in the ITER tokamak are described. The study is focused on EM loads anticipated on the plasma-facing units (PFU) of CDC. The paper is devoted to the EM simulations of the most dangerous scenarios associated with high mechanical stresses: fast downward vertical displacement events of Categories II and III with 36-ms linear current quench (FD VDE-II and FD VDE-III with 36-ms LCQ). The selected events have been analyzed using the original TORNADO code intended for simulation of transient EM processes in 3D solids using the finite element representation. Results of computations include evolutions of EM loads on PFUs in the form of integral EM forces and moments, their peak values and relevant time points.

Activation analysis of the European DEMO divertor with respect to the different breeding blanket segmentation

• Neutronic analysis of European DEMO fusion reactor single-null type divertor cassette. • Divertor neutron activation calculation with three different integrated breeding blanket modules (HCPB, WCLL MMS, WCLL SMS). • Neutron activation in divertor functional layers (radionuclide and reaction analysis). Neutron activation is an unavoidable process in large scale fusion devices operating with deuterium and tritium fuels. Neutron activation leads to the production of radioactive materials. Activated materials can generate heat and ionizing radiation in their environment. Consequently, values of activation characteristics need to be determined in order to ensure safety and performance of the fusion devices. In this work activity and decay heat inventories for the DEMOnstration power plant (DEMO) 2015 baseline model divertor were calculated. Operation scenario of 1898 days was considered as well as three different breeder blanket configurations of DEMO reactor: single-module segmentation water cooled lithium lead (WCLL SMS), multi-module segmentation water cooled lithium lead (WCLL MMS) and helium cooled pebble bed (HCPB). Divertor model is divided into 62 segments and 4 layers with different material compositions. Neutron transport and activation inventory calculations were performed respectively with MCNP and FISPACT codes. In addition to activation calculation results, divertor radionuclide analysis is also presented in this work.

On the numerical assessment of the thermal-hydraulic operating map of the DEMO Divertor Plasma Facing Components cooling circuit

Within the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the thermal-hydraulic behaviour of the DEMO divertor cassette cooling system, focussing the attention on the 2018 configuration of the Plasma Facing Components (PFCs) circuit consistent with the DEMO baseline 2017. The research campaign has been carried out following a theoretical-computational approach based on the finite volume method and adopting the commercial Computational Fluid-Dynamic (CFD) code ANSYS CFX. A steady-state CFD analysis has been carried out for the PFCs cooling circuit under nominal conditions and its thermal-hydraulic performances have been assessed in terms of coolant total pressure drop, flow velocity and Critical Heat Flux (CHF) margin distributions among the Plasma Facing Units (PFUs) channels, to check whether they comply with the corresponding limits. Results obtained have clearly predicted a total pressure drop of ≈1.7 MPa, slightly higher than the prescribed limit of 1.4 MPa, as well as a sufficient margin against CHF onset, higher than 1.4 except for three IVT PFU channels where a minimum value of 1.39 is predicted. Moreover, a parametric study has been performed in order to assess the operating map of the cooling circuit in the phase-space of coolant inlet temperature, pressure and mass flow rate, to be intended as that domain where the circuit thermal-hydraulic performances let it stay within the prescribed requirements. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Computational evaluation of N-16 measurements for a 14 MeV neutron irradiation of an ITER first wall component with water circuit

Abstract During ITER operations the water coolant flowing through components such as the first wall, blanket modules, divertor cassettes and vacuum vessel will become activated by high energy neutrons. Two key neutron-induced reactions will occur with oxygen in the water producing the radioactive isotopes N-16 and N-17, which have relatively short half-lives of a few seconds. These nuclides are transported in coolant loops and, unmitigated, their decay emissions will induce additional nuclear heat in components, potentially including superconducting magnets, and lead to an increase in the occupational dose for workers and sensitive equipment outside the biological shield. Variations in irradiation, water flow rate and cooling circuit parameters make it difficult to predict nuclear heating. A water activation experiment has recently been performed at the 14 MeV Frascati Neutron Generator to accurately measure N-16 and N-17 produced by irradiating an ITER first wall mock-up. This experiment aimed to validate the methodology for water activation assessment used for ITER and to provide scientific justification to reduce safety factors, which have a large impact on ITER component design and qualification. This paper provides a detailed description of neutronics calculations performed together with the GammaFlow code to model the temporal evolution of activated water, along with MCNP6.1 and FISPACT-II to calculate the detector response. The calculated reaction rates associated with nuclear data from ten libraries have been compared with measured data, although as many cross-sections originated from the same library effectively five nuclear data libraries have been compared.

EU-DEMO divertor: Cassette design and PFCs integration at pre-conceptual stage

Abstract The pre-conceptual design of the DEMO divertor cassette with a novelty, alternative path of the main cooling pipes inside cassette body is presented in this paper, focusing on cassette design and Plasma Facing Components (PFC) integration. The divertor cassette design is reviewed, considering recent updates in the DEMO configuration model as presented by the Programme Management Unit (PMU) in 2018. The new configuration requires the cooling pipes to be integrated inside the cassette body. The components affected by these changes and the impact on the divertor design are analyzed. The study focuses on a new integration system between cassette and cooling pipes. The paper describes the integration on the new cassette geometry and the divertor sub-systems. The design activities related to this system are discussed in detail in terms of CAD modeling and considerations with respect to manufacturing such as welding technologies and non-destructive testing.