Design and experimental activities in view of Water Detritiation–Isotopic Separation Systems combination in TRENTA facility

To support the research activities needed to characterize the performance of various components for the Water Detritiation System (WDS) and the Isotope Separation System (ISS) processes needed for the ITER design, an experimental facility called TRENTA, simulating for the ITER WDS and ISS protium separation column, has been commissioned at Tritium Laboratory Karlsruhe (TLK). The TRENTA facility has been conceived to allow operation in a closed loop with respect to tritium inventory and to allow investigations of key design and operation issues of combined CECE (Combined Electrolytic Catalytic Exchange) and CD (Cryogenic Distillation) processes in similar conditions as envisaged for the ITER WDS–ISS. Activities combining CECE and CD processes are on going at TLK. For the CD system, dedicated heat-exchangers have been designed and manufactured to make the combination possible. The system of heat-exchangers has to provide a double barrier to avoid tritium contamination of the helium stream. The design of the heat-exchangers for feeding of the CD column, equilibrator loop and condenser will be presented. In addition, the ongoing experimental activities for to the investigation of different CD fillings will be presented.     

Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose.Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min^?1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 10¹⁶ Bq kg^?1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water.The combination of the μCCR with the O₂ sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R&D activity on HTW processing with PERMCAT.     

The negative ion source test facility ELISE

The ITER neutral beam system is using inductively coupled radio frequency (RF) ion sources, that have demonstrated the required ITER parameters on (small) sources with extraction areas up to 200 cm². As a next step towards the full size ITER source IPP is presently constructing the test facility ELISE (“Extraction from a Large Ion Source Experiment”) operating with a “half-size” source which has approximately the width but only half the height of the ITER source. The modular driver concept is expected to allow a further extrapolation to the full size in one direction to be made. The main aim of this experiment is to demonstrate the production of a large uniform negative ion beam with ITER relevant parameters in stable conditions up to one hour.Plasma operation of the source is foreseen to be performed continuously for 1 h; extraction and acceleration of negative ions up to 60 kV is only possible in pulsed mode (10 s every 180 s) due to limitations of the existing IPP HV system. The design of the source and extraction system implements a high experimental flexibility and a good diagnostic access while still staying as close as possible to the ITER design. The main differences are the source operating in air and the use of a large gate valve between the source and the target chamber.ELISE is expected to start operation at the end of 2011 and is an important step for the development of the ITER NBI system; the experience gained early will support the design as well as the commissioning and operating phases of the PRIMA NBI test facilities and the ITER neutral beam system.     

Design progress on 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 these requirements, 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) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs.Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 × 10^?5 Pa at 100 °C containing a port plug. The heating system shall provide water at 240 °C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests.This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.     

The ITER full size plasma source device design

In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m² in D₂ (D^?) and 350 A/m² in H₂ (H^?) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m².The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.     

Negative ion test facility ELISE—Status and first results

The new test facility ELISE (Extraction from a Large Ion Source Experiment) has been designed and installed since November 2009 at IPP Garching to support the development of the radio frequency driven negative ion source for the Neutral Beam System on ITER. The test facility is now completely assembled; all auxiliary systems have been commissioned and are operational. First plasma and beam operation is starting in October 2012.The source is designed to deliver an ion beam of 20 A of D^? ions, operating at 0.3 Pa source pressure at an electron to ion current ratio below 1. Beam extraction is limited to 60 kV for 10 s every 3 minutes, while plasma operation of the source can be performed continuously for 1 hour. The ion source and extraction system have the same width as the ITER source, but only half the height, i.e. 1 × 1 m² source area with an extraction area of 0.1 m². The aperture pattern of the extraction system and the multi driver source concept stay as close as possible to the ITER design. Easy access to the source for diagnostic tools or modifications allows to analyze and optimize the source performance. Among other possibilities many different magnetic filter field configurations inside the source can be realized to enhance the negative ion extraction and to reduce the co-extraction of electrons. Beam power and profiles are measured by calorimetry and thermography on an inertially cooled target as well as by beam emission spectroscopy. Cs evaporation into the source is done via two dispenser ovens.     

Processing highly tritiated water desorbed from molecular sieve bed using PERMCAT

Tritium handling facilities use molecular sieve beds (MSB) to collect and recover tritiated water. After reaching the capacity limit of the MSB, the water is desorbed and decontaminated in a water detritiation system (WDS). In the case of highly tritiated water (HTW) absorbed into a MSB, an inherent safe option for processing is necessary due to the HTW specific properties. Ideally, HTW should be processed immediately in a continuous mode. With this in consideration, the water desorption process from a zeolite bed was developed and optimized in a dedicated non active facility. The results of this experiments were applied into the regeneration of a MSB previously loaded with HTW containing an activity of 1.9 × 10¹⁴ Bq kg^?1. The water was desorbed, by step increasing the temperature bed and fed by helium carrier gas into the PERMCAT for detritiation and tritium recovery. The processed water was collected in a dry MSB downstream of the PERMCAT. These initial studies successfully demonstrate the viability of the process. The obtained results of the preliminary study and the subsequent tests with tritium, will provide useful information for the design of tritium processes relying on MSB, such as the water processing foreseen for the test blanket modules in ITER.     

Development of water-cooled solid breeder test blanket module in JAEA

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m², 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.     

Korean high heat flux test facility by using electron beam system for ITER first wall semi-prototype

A Korean high heat flux test facility for the semi-prototype (SP) qualification of an ITER first wall (FW) will be constructed to evaluate the fabrication technologies required for the ITER FW, and the acceptance of these developed technologies will be established for the ITER FW manufacturing procedure. Korea participated in this qualification program, and is responsible for suitable arrangements for the heat flux test of our fabricated SPs. Qualification testing can be started provided that adequate quality and control measures are implemented and validated by the ITER Organization (IO). The controlling measures required for all heat flux tests shall be concrete and demonstrate the satisfaction of the IO test programs. Each country shall provide a test plan covering the quality and controlling measures in the high heat flux test facility to be implemented throughout the test program. Korean high heat flux testing for these ITER plasma facing materials will be performed by using a 60 kV electron beam and a power supply system of 300 kW, where the allowable target dimension is 70 cm × 50 cm in a vacuum chamber. In addition, this facility needs a cooling system for a high-temperature target and decontamination system for beryllium filtration.     

Effect of transient heating loads on beryllium

Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m². Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process.     

The performance test and analysis of ITER Main and Correction Busbar conductor

The first ITER Main Busbar (MBCN1) and Correction Busbar (CBCN1) conductor samples were manufactured in ASIPP and tested in the SULTAN facility. This paper introduces the sample manufacture, including strand, cabling, jacketing and sample preparation, and discusses the performance of MBCN1 and CBCN1 conductors. The testing results show that both samples have high T_cs, and meet the ITER requirement.Due to the ITER acceptance standard T_cs of MB conductor was changed to 6.7 K at 45.5 kA/3.9 T. The performance of MBCN1 conductor after cyclic load fits the ITER requirement, but the sample was only tested at 57 kA/2.75 T before cycling test. Using some hypothesis and equation to extrapolate the T_cs performance of MBCN1 conductor before cycling test, the result also fits the ITER requirement.For CBCN1 conductor, the central line of the central cooling spiral shifted about 1.3 mm during the cabling. The deviation causes an increase of the max self-field by about 0.005 T, which could not influence the CBCN1 conductor real T_cs performance at peak field.     

The ITER divertor pumping system,design evolution,simplification and performance

The ITER tokamak will be fuelled at a time averaged rate of up to 200 Pam³ s^?1 requiring neutralised gas in the divertor to be pumped to balance the fuelling and remove the fusion helium and other impurities in the exhaust. This is achieved on ITER using large bespoke cryo-sorption pumps. In this paper design evolution of the ITER divertor pumping system is outlined from the 1998 configuration to the current design. Details of the new, 6 direct pump, system design which will be used in the build of ITER are given. The operating modes of the new system configuration for different plasma scenarios are described and the performance of the new system is analysed and compared with previous baselines.     

Characterization and qualification of advanced insulators for fusion magnets

Intensive research over the past decades demonstrated that the mechanical material performance of epoxy based glass fiber reinforced plastics, which are normally used by industry as insulating materials in magnet technology, degrades dramatically upon irradiation to fast neutron fluences above 1 × 10²² m^?2 (E > 0.1 MeV). which have to be expected in large fusion devices like ITER. This triggered an insulation development program based on cyanate ester (CE) and blends of CE and epoxies, which are not affected up to twice this fluence level, and therefore appropriate for large fusion magnets like the ITER TF coils. Together with several suppliers resin mixtures with very low viscosity over many hours were developed, which renders them suitable for the impregnation of very large volumes. This paper reports on a qualification program carried out during the past few years to characterize suitable materials, i.e. various boron-free R-glass fiber reinforcements interleaved with polyimide foils embedded in CE/epoxy blends containing 40% of CE, a repair resin, a conductor insulation, and various polyimide/glass fiber bonded tapes. The mechanical properties were assessed at 77 K in tension and in the interlaminar shear mode under static and dynamic load conditions prior to and after reactor irradiation at ~340 K to neutron fluences of up to 2 × 10²² m^?2 (E > 0.1 MeV). i.e. twice the ITER design fluence. The results confirmed that a sustainable solution has become available for this critical magnet component of ITER.     

Feasibility study of an intense D–T fusion source: “The New Sorgentina”

The international community agrees on the importance to build a large facility devoted to test and validate materials to be used in harsh neutron environments. Such a facility, proposed by ENEA, reconsiders a previous study known as “Sorgentina” but takes into account new technological development so far attained. The “New Sorgentina” Fusion Source (NSFS) project is based upon an intense D–T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating targets. NSFS produces about 1 × 10¹³ n cm⁻² s⁻¹ over about 50 cm³. Larger volumes of lower neutron flux will be available (e.g. for TBM experiments) as well as collimated channels to study some features of the ITER neutron camera. The NSFS facility will overcome problems related to the ion source and accelerating system, by means of an upgraded version of the JET–PINI ion beams. NSFS has to be intended as an European facility that may be realized in a few years, once provided a preliminary technological program devoted to the operation of the ion source in continuous mode, target heat loading/removal, target and tritium handling, inventory as well as site licensing. In this contribution, the main characteristics of NSFS project will be presented.     

High heat flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

Fabrication and high heat flux test of large mockups for ITER first wall semi-prototype

The main topic of an ITER blanket first wall procurement is to qualify whether each party has the key technology needed for the fabrication and joining of first wall components. A semi-prototype qualification project will be released requiring that the single components of a full-scale first wall must be fabricated and successfully pass high heat flux tests using a hypervapotron cooling channel. In this work, various mockup types have been modeled and fabricated to develop the joining technology for a semi-prototype. The semi-prototype, which has three double-fingered panels, is a scaled-down component of a full-size first wall. The standard or slit mockups with a 80 mm × 80 mm single beryllium tile joined to a CuCrZr heat sink were fabricated to qualify our HIP (Hot Isostatic Pressing) technology for the joining of semi-prototype. These standard mockups were installed to perform a high heat flux test in the Korea heat load test facility (KoHLT). For a preliminary test of a semi-prototype, thermo-hydraulic mockups of 710 mm × 100 mm were designed and fabricated to verify the Cu/SS cooling performance, such as hypervapotron. For the high heat flux test in our KoHLT facility, the normal cycle is based on an expected heat flux of 300 s in accordance with the ITER qualification specifications. These tests will be performed to qualify the joining technologies, which is required for an ITER blanket first wall and a semi-prototype.     

Status of the IDTF high-heat-flux test facility

The ITER Divertor Test Facility (IDTF) was designed for the high heat flux tests of outer vertical targets, inner vertical targets and domes of the ITER divertor. This facility was created in the Efremov Institute under the Procurement Arrangement 1.7.P2D.RF (high heat flux tests of the plasma facing units of the ITER divertor).The heat flux is generated by an electron-beam system (EBS), 800 kW power and 60 kV maximum accelerating voltage. The component to be tested is mounted on a manipulator in the vacuum chamber capable of testing objects up to 2.5 m long and 1.5 m wide. The pressure in the vacuum chamber is about 3*10⁻³ Pa. The parameters of the cooling system and the water quality (deionized water) are similar to the cooling conditions of the ITER divertor. The integrated control system regulates all IDTF subsystems and data acquisition from all diagnostic devices, such as pyrometers, IR-cameras, video cameras, flow, pressure and temperature sensors.Started in 2008, the IDTF was commissioned in 2012 with the testing the outer vertical full-scale prototypes and the completion of the PA 1.7.P2D.RF task. This paper details the main characteristics of the IDTF.     

Tungsten and CFC degradation under combined high cycle transient and steady state heat loads

Within the ITER divertor lifetime millions of transient events are expected during H-mode operation due to edge localized modes (type I ELMs). These will deposit their energy on plasma facing materials that are pre-heated to various surface temperatures, depending on the steady state heat load (SSHL) at the respective location, leading to synergistic effects. An electron beam facility was used to simulate ELM-like heat loads with ITER relevant power densities (≈0.5 GW/m²) and pulse duration (0.5 ms). At the same time additional SSHL was applied to obtain different base temperatures. Experiments were performed on actively cooled pure tungsten and the carbon fiber composite (CFC) NB41, applying 10³–10⁶ pulses of 0.5 ms duration with a power density of 0.14–0.55 GW/m² and 0.55–0.68 GW/m² on tungsten and CFC, respectively. Surface temperatures were about 200 °C, 400 °C and 700 °C for tungsten and about 450 °C for CFC. Crack formation in tungsten was preceded by roughening due to plastic deformation. In case of T_surf ≈ 200 °C cracks propagated comparably fast (brittle material), while slow propagation and recrystallization around the crack edges indicated fatigue damage at higher temperatures. Compared to tungsten, CFC showed a higher damage threshold.     

Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m²) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles.The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares.In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time.