ITER Test Blanket Module Research Articles

Test blanket modules piping systems calculation method according to RCC-MRx code

The ITER Test Blanket Modules (TBMs) are installed and operated inside the vacuum vessel (VV) at equatorial ports located within port plugs (PP). The Pipe Forest (PF) is a network of pipes connecting the Ancillary Equipment Unit (AEU) to the TBMs. Depending of the TBM, the Pipe Forest includes various piping system such as Water Coolant (350°, 188 bars), Helium Coolant (500°, 100 bars) Lithium Lead and Tritium Extraction System. During operating states, the Pipe Forest main function is to withstand loads in normal, incidental and accidental situations according to RCC-MRx code criteria (including design rules for mechanical components for fusion installations). It includes severe thermomechanical loadings due to the thermal expansion added to seismic and vertical displacements events. In addition, the PF layout must satisfy constructability in constraint environments and maximize the space available for man access, which constitutes a real challenge for design.Since no existing RCC-MRx post-processing tool exists off the shelves, a specific methodology for pipe stress analysis is being developed by ITER, CEA and ASSYSTEM. This methodology starts with a post-processing calculation tool of stresses and reaction forces for looping back into the structural calculation. Moreover, a specific extraction module from 3D model to processing tool allows the conversion of CAD data to calculation entry files. Finally, a feedback loop is also re-integrated into the core CAD model through an import module, enabling the tracking of displacements and evaluate the impact of incorporating pipe movements into the analysis.This paper describes the pipe stress analysis methodology and its application for Pipe Forest system design life cycle. It covers the historical evolution of the routing including introduction of expansion loops. More recently, it incorporates the integrated calculation of the entire port cell, spanning from the back of the TBM sets, up to the connection pipes of the vertical shaft.

Assessment of the relevancy of ENEA Water Loop facility with respect to ITER WCLL TBS Water Cooling System by considering their thermal-hydraulic performances

The Water-Cooled Lead-Lithium (WCLL) is one of the two candidate concepts for the Breeding Blanket (BB) of DEMO. A Test Blanket Module (TBM) together with its Water Cooling System (WCS) is going to be installed and tested in the ITER reactor. The WCS acts as primary cooling circuit of the TBM module, and it is designed to reproduce the water thermodynamic conditions expected at the DEMO BB inlet.During last years, ENEA and the DIAEE of Sapienza University of Rome have carried out the conceptualization of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone. The W-HYDRA platform is composed by three individual facilities called: Water Loop, Steam, and LIFUS5/Mod4. Water Loop replicates the salient thermal-hydraulic features of the ITER WCLL WCS, and it is equipped with a test section placed inside a Vacuum Vessel (VV) to investigate mock-ups of the whole TBM or its individual parts.This paper assesses the relevancy of the WL facility with respect to ITER WCLL TBM System comparing their thermal-hydraulic performances during selected operational and accidental conditions. Two RELAP5/Mod3.3 models were developed, and the outcomes of the simulations showed a good agreement, thus demonstrating the effectiveness of WL facility to support the ITER TBM program.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

Experimental study and discrete element method analysis of impact of vibration on pebble packing

To develop tritium breeding technology for fusion reactors, Korea actively participated in the ITER Test Blanket Module (TBM) program, focusing on the development of a ceramic breeder-type TBM. The challenge was to introduce pebbles into the TBM without causing any damage to them. This study conducted experiments and Discrete Element Method (DEM) analyses to optimize the pebble-filling process for breeding blanket development. Even when facing obstacles such as grooved shapes on the walls of the pebble-filled space, these grooves did not significantly reduce the packing ratio. Vibrating the pebbles proved essential to increase the pebble packing ratio. Both experiments and DEM analyses confirmed an increase in pebble packing when vibration is applied. The study also investigated the impact of the timing of vibration application. It is better to vibrate after the pebble injection is done during injection to increase the pebble packing ratio. This approach increases the pebble packing ratio, but it also makes additional empty space. Consequently, maintaining continuous vibration from the initiation to the completion of the injection process can increase the packing ratio without the need to repeat the injection process to fill the empty space.

The CHIMERA facility development programme

The CHIMERA fusion technology facility will enable testing of large in-vessel component modules under reactor-like conditions of combined in-vacuum thermal power density and magnetic field. With an integral large superconducting magnet and a PWR-like water loop, CHIMERA is also ideally placed for experiments on liquid metal breeding blanket prototypes. Facility construction is underway at the UKAEA site in South Yorkshire. The superconducting magnet is fully wound and terminated, and the bespoke pulsed magnet power supply has been delivered. Even before construction and commissioning is complete, a parallel and supporting research and development programme is in progress and is reported here. A bespoke infrared heating system has been developed, capable of applying 0.5 MW/m2 to component surfaces up to the size of the ITER test blanket module first wall. The modules of this heater are highly specialised and designed to endure the high magnetic forces from the CHIMERA static and pulsed magnets. CHIMERA will feature a range of diagnostics, including load cells to measure static and pulsed magnetic forces, and induced current sensors, all of which have been tested to confirm acceptable operation in the pulsed magnetic field. Manufacture of the test mock-up to be used for facility commissioning is underway. Testing will be enhanced with simulation ‘digital twin’ capability, and a development project is now producing first virtual test results, informing commissioning of the CHIMERA device.

ITER Test Blanket Module—ALARA Investigations for Port Cell Pipe Forest Replacement

The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF.

CHIMERA Fusion Technology Facility: Testing and Virtual Qualification

The Combined Heating and Magnetic Research Apparatus (CHIMERA) fusion technology test facility is under construction. The facility will be uniquely capable of semi-integral testing of fusion materials and component modules up to the size of the ITER test blanket module box, under combined conditions of in-vacuum high heat flux, static and pulsed magnetic fields, and high-temperature/high-pressure water cooling. This paper reports the high-level capabilities of the CHIMERA baselined design and the planned program of testing and describes the proposed strategy for use of simulations for virtual testing, qualification, and in-situ monitoring. The first step in testing of a component mock-up is to take data from as-built geometry and other measurements and transmit them to an integrated computational model that can closely mimic the physical asset and form a digital replica. Not only can this digital replica be queried in advance of physical testing in the facility, allowing optimization of the test program, but combined with subsequent test data, it also can deliver much greater insight into experimental results than can be obtained using test data alone. The digital replica is used as the basis for a digital twin, which is live coupled to the running experiment, and is under development as a proposed key facet of fusion reactor surveillance in-service. Physical mock-ups for testing can be subjected to in-vacuum heat flux up to 0.5 MW/m2 over the entire surface while within a strong horizontal magnetic field. The central field can be up to 4 T with a peak in the test region of 5 T. The same component mock-ups can also be subjected to repeated magnetic field pulses with ramp rate 12 T/s, which can simulate loading conditions of a plasma disruption. Facility upgrades are underway to include a liquid metal circulation loop to allow the study of magnetohydrodynamics effects and to add a high-heat-flux system using a very high-power continuous-wave laser to achieve divertor-relevant heat fluxes of 20 MW/m2 over the area of a small-scale mock-up. Four examples are given to illustrate the physical testing program that is currently foreseen.

Design of a Prototypical Mock-Up for the Experimental Investigation of WCLL First-Wall Performances

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) and it is one of the two candidates adopted as driver BBs in DEMO. In this framework, at the ENEA research centre of Brasimone, the realization of the experimental platform, W-HYDRA, is envisaged. The platform is dedicated to the support of the development of WCLL BB and ITER TBM and the investigation of the DEMO balance of plants. One of the most important experimental infrastructures is the water-loop facility, the aim of which is to provide water at a high pressure and temperature (PWR conditions), with a sufficient mass-flow rate and power for the experimental testing of BB and TBM components. The facility will be equipped with a vacuum chamber and an electron beam gun for the reproduction of high surface heat flux on plasma-facing components. In the present work, the design of a prototypical mock-up (MU) of the WCLL BB first wall is described. The MU is used to investigate the thermal, hydraulic and structural behavior of the current first-wall design under relevant heat loads at the expected operational conditions. The delineation of the main experimental test’s features and the instrumentation needed is assessed in the paper. A preliminary CFD calculation on the prototypical MU and the computational results are also presented.

Overview of recent ITER TBM Port Plug R&D activities

The objective of ITER project is the demonstration of the scientific and technological feasibility of fusion energy. Tritium fuel is required for the fusion reaction in DEMOnstration reactors and future commercial reactors but it is not available in nature because of its instability. To investigate tritium breeding self-sufficiency, two equatorial ports in ITER are dedicated to operating test blanket modules (TBMs) using the TBM Port Plug (TBM-PP). The TBM-PP design should be investigated with the physical mock-ups to validate the leak tightness in the primary vacuum boundary and to ensure TBM replacement and refurbishment compatible with Remote Handling (RH) operations in the ITER Hot Cell Facility during the ITER lifetime. This paper summarizes the achievements on the two most relevant R&D topics as described herewith. Several TBM-PP components (i.e., back parts of the TBM-Frame, the Dummy-TBMs, the TBM-Shields, and the TBM feeding pipes) are part of the vacuum boundary implying that these components are Vacuum Quality Classification (VQC) 1A. To test and develop high vacuum sealing for the TBM-Frame flange and the TBM-Set or Dummy-TBM flange, the ITER Large Seal Test Rig (LSTR), and metallic gasket seals for the TBM application have been manufactured and installed in the ITER site. The helium leak tests were performed at different operating temperatures to investigate the vacuum performance of metallic gasket seals with VQC 1A. The TBM-PP design shall guarantee the feasibility of a rapid replacement and refurbishment compatible with ITER Remote Handling (RH) operations by using RH tools in the Hot Cell Facility during the ITER's lifetime. Therefore, an experimental program has been performed to demonstrate the feasibility of the critical RH refurbishment tasks including the insertion and removal of TBM-Set/Dummy-TBM and the RH bolting.

Status of Scoping Nuclear Analyses for the Evolving Design of ITER TBM Port Cells

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future fusion reactors. The concerned Test Blanket System (TBS) consists of a Test Blanket Module (TBM) residing inside the TBM–Port Plug (TBM-PP) and its associated ancillary systems in the Tokamak facility. In this paper, the results of a full suite of nuclear analyses concerning the shielding performance of the Pipe Forest (PF) and Bioshield Plug (BP), to reflect on the evolution of their designs, are discussed. On the BP side, the design of the peripheral part has been reviewed considering the ventilation openings and butterfly doors, to assure the design compliance with the Radiation Map (RadMap) requirements for the neutron flux in the Port Cell (PC), behind the BP. On the PF side, the pipes routing and maintenance corridor door have been redesigned, by taking into account results from previously concluded nuclear analyses. The neutronics model was developed from CAD and was used to perform transport simulations in two plasma modes: on and off. For plasma-on mode, the plasma neutron field in the Port Interspace (PI) as well as behind the BP was assessed and few shielding options were explored. The responses due to decay neutrons from 17N in activated cooling water were also considered. For the plasma-off mode, the focus was shifted to further refine the ShutDown Dose Rate (SDDR) maps, which is of importance for maintenance operations that are foreseen to take place at various stages of ITER operation, in particular following the FPO-1, FPO-2, and Short operation scenarios. In addition, detailed activation analyses were carried out to provide a provisional waste classification.

ITER Test Blanket Module—Engineering Investigations for Port Cell Pipe Forest Replacement

The ITER Test Blanket Modules (TBMs) are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each of them includes two TBMs. After each plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs by new ones during the ITER long-term shutdown (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace called pipe forest (PF). The TBS equipment in PC shall be designed so that occupational radiation exposure can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER Policy. To implement ALARA process requirements, a design shall consider careful integration assessment since early phase to address all engineering aspects of the replacement sequence. The design of the TBS equipment in PC shall optimize the reconfiguration operation by including human factor considerations and providing adequate space. Provisions for dose reduction measures (DRMs) considering protection means and minimization of occupancy times in radiation fields are also investigated by assessing remote technics to reduce operator exposure. This article describes engineering investigations for optimized PF replacement operations in various engineering fields. Results on the design of the TBS equipment in the TBM-PC are discussed. The PF design shall implement nuclear engineering aspects. Design options shall demonstrate constructability and maintainability meaning compliance with ALARA process and adequate construction codes (RCC-MRx, EN 13480). TBM-PP replacement sequence drives several elements of the TBM-PC component design and associated integration. Concepts are discussed with regard to nuclear engineering and operation aspects. Use of prototyping demo, including virtual reality (VR), will support the preliminary validation during the system engineering process in order to prepare the full-scale mock-up validation phase.

Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

• T inventory in the ITER EU Water-Cooled Lithium-Lead Test Blanket Module investigated. • T from vacuum vessel (gas, neutrals) and solute T in PbLi included in modeling. • Effect of neutron damaging on trap concentrations included in models. • T retention in the range of 0.75 g is expected over plasma operation period. • Comparison with the TMAP7 code performed. Tritium self-sufficiency is a key requirement for future fusion power plants. Therefore, it will be necessary to minimize tritium losses to the surrounding systems, such as the tritium breeding modules, plasma-facing components, cooling system, etc. These components and systems will act also as a tritium sink, as tritium will be retained in the metal walls in traps produced during manufacturing or induced by neutron irradiation. The design of the tritium breeding systems and plasma-facing components will therefore have a direct impact on the performance, operational safety and cost of any future power plant. Consequently, accurate modeling of tritium transport and retention in these systems is needed for future engineering designs of the tritium breeding systems and supporting safety requirements. In this work, tritium permeation and retention was modeled with the diffusion-transport code TESSIM-X for the current European WCLL (Water Cooled Lithium Lead) Test Blanket Module (TBM) for the ITER TBM Programme, with the inclusion of neutron-induced traps. Trap saturation is considered for dpa values of 0.25 and above, leading to an increase in trap concentrations relative to undamaged material of roughly a factor 3. The code was also benchmarked against the well-established TMAP7 code.

Maxwell force analysis of CN HCCB TBM based on preliminary design

The Chinese (CN) Helium Cooled Ceramic Breeder (HCCB) Test Blanket Module (TBM) will be tested in ITER according to the ITER TBM program. In order to well achieve the test goals and to meet the ITER technical requirements, the design of CN HCCB TBM is continuously updating since 2004 and it is currently in preliminary design (PD) phase. As an in-vessel component of ITER, TBM will be exposed to strong magnetic field. Consequently, large electromagnetic (EM) loads may generate on TBM structures and should be quantitatively evaluated to provide input for the structural integrity assessment of TBM. In this paper, the Maxwell force (MF), one of the EM loads, is numerically calculated and discussed. To improve the calculation efficiency, the finite element (FE) model is simplified by ignoring the cooling channels. Hence, the corrected B-H curve based on effective volume ratio of ferrite material is adopted to exclude the influence caused by absence of cooling channels. Current obtained results are comparable with those in previous work and agree well with analytical ones also, which indicates that the global FE model developed here is acceptable and the analytical equation derived in this paper can be used to approximately predict the MF on TBM.

Current status and future perspectives of EU ceramic breeder development

• EU reference tritium breeding pebbles are produced using a melt-based process. • Process scale-up is underway to ensure the supply for ITER. • Biphasic materials show enhanced mechanical strength and stability. • Design relevant data were summarised in a material property handbook. • Crucial neutron irradiation experiments will complement the material qualification Biphasic ceramic pebbles, consisting of Li 4 SiO 4 and Li 2 TiO 3 , are being developed as the EU reference tritium breeding material for ITER and DEMO. A modified melt-based process for the pebble fabrication was established and optimised with regard to the process stability and yield. In parallel, the properties of pebbles and pebble beds were evaluated in dedicated experiments. These include investigations into e.g. the long-term thermal stability of pebbles, the thermomechanical properties of pebble beds, as well as the compatibility between the ceramics and the structural material. The available data was recently summarised and issued as the Material Property Handbook on Advanced Ceramic Breeders. During the EU DEMO conceptual phase, two main issues are proposed to be pursued in view of ITER. Irradiation campaigns are planned to ensure that the advanced ceramic breeder material is adequately tested and an extensive process scale-up with an increased production rate is foreseen to secure the supply for the ITER test blanket modules.

Assessment of a bolted-flange connection for ITER Test Blanket Module

In ITER, the ‘Pipe Forest (PF)’ is a network of pipes that connects the Ancillary Equipment Unit to the Test Blanket Module (TBM) at the level of the equatorial port-plug of the reactor prototype. The goal of ITER's TBM program is to validate concepts to be adopted for the DEMO tritium Breeding Blankets. Different types of pipes are mounted on the port-plug among which those used for cooling or tritium processing. During plasma operation, the PF has to accommodate severe thermomechanical loads. It shall also provide protection from abnormal contamination and be designed to ease connection/disconnection operations. With numerical simulation, this work assesses the use of bolted flanges as an alternative to the welded solution to connect the PF on the TBMs inside the port cell environment. The proposed designs integrate metallic seals as well as an embedded cooling system in the flanges when necessary, to limit the temperature of the seals and the bolts and prevent irreversible damage. It is shown that bolted-flange junctions for the PF are a credible solution. Still, important stresses in the socket of the flanges are present but do not represent a problematic issue for dimensioning purposes, as they shall be accommodated with a small amount of plastic strain.

Development and demonstration of a semi-automatic system for the bulk production of lithium titanate (Li2TiO3) pebbles by solid state reaction process (SSRP)

Abstract Lithium titanate ( L i 2 T i O 3 ) has been selected as ceramic solid tritium breeding materials for the ITER Test Blanket Module (TBM) and also for DEMO reactor. Looking towards its future demand in large quantities, it is necessary to develop a system for bulk production of the lithium titanate pebbles. A semi-automatic system for the bulk production of lithium titanate (Li2TiO3) pebbles by Solid-State Reaction Process (SSRP) has been designed, developed and successfully demonstrated. Raw materials; viz. lithium carbonate ( L i 2 C O 3 ) and titanium dioxide ( T i O 2 ) were mixed in required stoichiometry and fed into the continuous rotary synthesizer. L i 2 T i O 3 powder was synthesized in a continuous rotary synthesizer at 10 kg/h production capacity. This synthesized powder was milled in air jet mill. Milled powder was fed into the extrusion-spheronization based automated granulation system and green, spherical L i 2 T i O 3 pebbles of were produced at the production rate of 25 kg/h. Green spherical pebbles were dried, classified and sintered It was found that the synthesized L i 2 T i O 3 was in monoclinic β- L i 2 T i O 3 phase according to JCPDS No. 33-0831 and the properties of the sintered Li2TiO3 pebbles were well within the desired values. The process and various process parameters are discussed in details in this paper.

Overview of recent ITER TBM Program activities

Abstract The ITER Test Blanket Module (TBM) Program has significantly evolved since 2018. The number of equatorial ports allocated for operating the Test Blanket Systems (TBSs) has been reduced from three to two. As consequence, four TBSs can be simultaneously installed and operated, versus six previously. Since the dedicated space in the various rooms of the Tokamak Complex has been kept unchanged, the existing space constraints have been substantially relaxed. The paper addresses the possible selection of the four TBSs for the initial configuration and design of the involved TBMs and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. It describes also the main progress of the design of the infrastructures needed for hosting the four TBSs (e.g., port plugs, port cell common components, common maintenance tools and equipment). The design of the TBS Connection Pipes (CPs) System has reached its final phase since, being captive, it needs to be installed before the First Plasma.

Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module

The Water Cooled Lithium Lead (WCLL) is one of the selected breeding blanket (BB) concepts to be investigated in the EUROfusion Breeding Blanket Project (WPBB), and it was also recently chosen as one of the mock-up for ITER Test Blanket Module (TBM) program. The program foresees the test of different BB mock-ups, called Test Blanket Modules, with all the related ancillary systems. A pre-conceptual design of the Water Cooling System (WCS) of the ITER WCLL-TBM was developed considering the same cooling function of the EU-DEMO WCLL-BB primary heat transfer system (PHTS), but matching different boundary conditions: a scaled source power and far lower heat sink temperatures. A complete thermal-hydraulic (TH) model of the WCS loop and TBM set was developed using a modified version of RELAP5/Mod3.3 system code to verify component sizing and to investigate the system behavior during steady-state and transient conditions. The full plasma power scenario was simulated and used as an initial condition for transient calculations. ITER Normal Operational State (NOS) was studied to evaluate the system response. Simulation results highlighted the need for an electric heater to keep the WCS system in stable operation. A sensitivity analysis was carried out to optimize the heater duty cycle.

A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas. The shield module is made of stainless steel plates and water channels. It is identified as an important component for radiation protection because of its radiation exposure control functionality. The radiation protection classification leads to more assurance of the component design. In order to validate and verify the design of the shield module, a neutronic laboratory-scale experiment is designed and executed. The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 1010 n s−1. The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials. The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene), respectively. The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 cross-section data. The paper covers neutronic design, irradiation and the outcome of the experiment in detail.

Evaluation of conservative and innovative manufacturing routes for gas cooled Test Blanket Module and Breeding Blanket First Walls

Different manufacturing routes are investigated at the KIT INR for the realization of First Walls (FW) for nuclear fusion components, such as the ITER Test Blanket Module (TBM) and DEMO Breeding Blankets (BB) for the Helium Cooled Pebble Bed (HCPB) Breeding concept. One conventional manufacturing route mainly basing of Electrical Discharge Machining (EDM) and forming was demonstrated successfully. However, the procurement costs are high. Therefore, options also to apply Additive Manufacturing (AM) as alternative were investigated. This paper compares the HCPB reference concept for FW fabrication to innovative concepts basing on AM.