Continual technology development for fusion has come to rely on the principle of "design by analysis" where advanced finite element analysis (FEA) or finite volume analysis provides insight on the performance of engineered systems. Extensive three-dimensional (3D) computations in fluid dynamics, heat transfer, neutronics, magneto-hydrodynamics and electro-magnetics are involved in an iterative design process for magnets, vacuum vessels and in-vessel components. Many difficulties arose in the integration of computer-assisted design (CAD) packages and the numeric models and results from different FEA codes. Over the last decade, engineers developed a vast array of specialized translators and interpolation programs to deal with geometry, mesh and load transfers between single-discipline codes, often with mixed outcomes. Now, several multiphysics codes that allow calculations on the same mesh and easy transfer of loads and other boundary conditions are emerging in the commercial market. These codes often have a robust library of physics models and solvers that address both steady state and transient phenomena and provide simultaneous solutions to heat transfer, fluid flow and structural mechanics problems. This article reviews three existing design tools, provides some examples of how the multiphysics codes are impacting practical engineering design, and identifies some important gaps that still exist today.

Detailed profiles of nuclear heating and radiation damage parameters were determined in ITER blanket modules at different poloidal locations. The results indicate that the nuclear parameters are sensitive to the configuration and material composition. Nuclear analysis was performed for several sections with cutouts in the back of the blanket modules for manifolds, inter-modular keys, and in-vessel coils to assess the impact on vacuum vessel shielding.

The prediction of electromagnetic loads on the ITER blanket modules during a plasma disruption is considered for two different blanket modules and different disruption events. The key features of the analysis procedure will be presented including the modeling the blanket modules (first wall and shield block), the plasma disruption current, and the integration of these models to produce a high-fidelity electromagnetic simulation that describes a plasma disruption event. The electromagnetic calculations are performed using the Opera-3d software. The loads due to the eddy currents were calculated for two outboard blanket modules and a number of selected disruption scenarios. Once these loads have been calculated they can also be exported for additional postprocessing to assess the mechanical loading effects.

The traditional design sequence for a fusion device starts from desired plasma performance and a coil set that optimizes formation of the required plasma equilibrium. Designers then reserve space for internal components, vessel, and cryostat. The ITER process has taught us that this traditional sequence should be revised. The discovery of convective transport in the plasma scrape-off layer has greatly increased heat flux to the first wall (FW), implying that power is flowing along field lines. Shaping of the FW's plasma-facing-surface and divertor components is now a critical design consideration and constraint. Plasma duration has increased to the point that active cooling of internal components is required. Ever more complex plasma diagnostics require complex openings to view the plasma and complex routing of items like cables behind the blanket. As we move toward the next generation of fusion machines, there is a need for many engineering diagnostics to monitor the operating state of actively cooled components. Internal coils for ELM, resistive wall mode, and plasma rotation control further complicate the region between blanket modules and vessel. Adding new components between the vessel and blanket removes material that is either shielding external components or breeding tritium in a reactor. Traditionally, such additional internal components are added during later design phases when space has been fixed. Using design by analysis during the conceptual design phase allows the space required for internal components to be more accurately defined so balanced trade-offs among magnets, vessel, and internal components are made. The result is a concept that is easier to integrate and does not disproportionately constrain later design phases for any one system.

Plasma facing components like the ITER blanket shield modules (BSMs) often consist of a first wall and a shield block that are mounted to a vacuum vessel wall. The ITER mount is comprised of an Inconel bolt, Ni-Al-bronze collar, insulating layers of alumina and a flexible Inconel cartridge that allows compliance to reduce stress. Here we describe our efforts to perform multiphysics simulations on the flexible mount using current densities calculated for ITER halo scenarios and off-normal disruption events over various size faults or breaks in the insulator layers. This effort also entailed a study of fault size and simulations of high current flow over small area faults that would likely lead to melting. In addition to ohmic heating and thermal analysis, melting and solidification physics were included using computational fluid dynamics to track the solid/liquid interface and the degree of melting near the fault. Temperature dependent conductivities were used for the solid and liquid phases of all the materials. We concluded that for the anticipated off-normal conditions in ITER, the flexible mount is very fault tolerant for both large and small area faults.

Plasma disruptions and edge localized modes can result in transient heat fluxes as high as 5 MW/m(2) on portions of a tokamak reactor first wall (FW). To accommodate these heat loads, the FW will likely use water-cooled hypervapotron heatsinks to enhance the heat transfer. In this article, we present the results of a computational fluid dynamics (CFD) study using 70 degrees C inlet water at 2.7 MPa to investigate the tooth height and backchannel depth of 50-mm-wide hypervapotrons with 6-mm-pitch and 3-mm side slots. We compare a popular design with 4-mm-high teeth and a 5-mm backchannel to a more optimal case with 2-mm-high teeth and a 3-mm backchannel under nominal heat loads (0.5 MW/m(2)) on a 100-mm-heated length and under single-phase flow conditions. Better heat transfer in the latter case and the smaller backchannel permit a factor of two reduction in the required mass flow while maintaining the same beryllium armor surface temperatures near 130 degrees C. The shallow teeth and smaller backchannel allow the 40 fingers in a typical panel to flow in parallel and simplify the water circuit. A comparison of the two hypervapotron designs during off-normal loading (5.0 MW/m(2)) and two-phase flow then follows. The design with 2-mm teeth has a 3.5% higher beryllium surface temperature of 648 degrees C and reduces the critical heat flux (CHF) by similar to 2%. Hypervapotron width also plays a role in heat transfer and CHF. CFD results for 36 and 70 mm wide hypervapotrons compared to the 50-mm case reveal similar thermal performance at low heat flux, but a reduction in CHF with increasing width. This study highlights the necessary compromise between design margin during transient events, effective heat transfer under nominal conditions, limitations on finger width, and the simplicity needed in the water circuit design.

The paper summarizes the proposed strategy for electro-magnetic (EM) load analysis of the components in large-scale electro-physical devices and describes an example of its application for the evaluation of forces and torques acting on the modular blanket of the tokamak ITER during plasma transients. It relies on the mutually complementing utilization of the newly developed method for operative and systematic estimates of the EM loads and related parameters over a wide range of conditions with traditional finite-element transient EM analyses on integrated models for select cases. The new method exploits possibilities to simplify the mathematical model determined by the particular considered problem and system. It assumes a reducing of the complex problem to a composition of specific local, global and composite ones, which makes this low-cost method operative, flexible and well supportive for the development through all its phases as a predictive mode analysis, also to search the advanced configurations of conducting components and for their design optimization. The flexibility and computational efficiency of a combination of the two methods ensure that the EM loads can be calculated in time to support the blanket design and analysis activities to the next milestones, the Preliminary Design Review scheduled for late 2011 and the Final Design Review a year later. The first results of detailed transient analyses on integrated models for select cases are discussed. (C) 2012 Elsevier B.V. All rights reserved.

A plasma current disruption is usually initiated by impurity influx that causes a rapid decrease in plasma thermal stored energy (thermal quench). Thermal quench occurs in 500–2000μs on a large device like ITER. Depending on the β value, the plasma may be either paramagnetic or diamagnetic. Thermal quench causes a large shift in paramagnetism (or diamagnetism) and a corresponding change in toroidal flux. The flux swing can be 1–2 Weber with the rate of change of the toroidal field between 25 and 150T/s for a device like ITER. The toroidal field shift induces poloidal current in the vessel and possibly in internal components. We have developed a method for simulating the thermal quench field shift that is compatible for use with the electromagnetic simulation codes. The method is based on a radially thin shell having the shape of the last closed flux surface with poloidal current driven to duplicate the toroidal field shift. The magnitude of the current and its time history are adjusted to duplicate the flux change during a disruption thermal quench. We will present the results of using this method to simulate the induced currents in a vacuum vessel having two shells.

Recent experimental data from the ITER critical heat flux (CHF) mock-ups was used to benchmark a 3D CFD code concerning subcooled boiling heat transfer for high heat flux removal. The predicted temperatures show good agreement with experimental measurements for a range of operating parameters and of cooling configurations. Specifically, it applies to a hypervapotron channel exposed to a 5MW/m2 surface heat load and cooled by velocity of 2m/s. Such flow geometry and operating condition seem necessary for ITER-enhanced heat flux first wall modules if an adequate design margin in CHF is needed. A detailed CFD and heat transfer analysis performed on a prototyped CAD model provided a higher confidence on the design and is deemed a desirable feature for continued design exploration and optimization processes. This is particularly crucial in regard to flow distribution among the FW fingers.

Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m(2). Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling.This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 degrees C for a wide range of heat fluxes from 3 MW/m(2) to 10 MW/m(2) and flow velocities from I m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin to dominate. Beyond this threshold, higher heat fluxes lead to the boiling crisis and eventual burnout. This predictive capability is essential in determining the critical heat flux margin for the design of complex 3d components.

Plasma disruptions and Edge Localized Modes (ELMS) may result in transient heat fluxes as high as 5 MW/m 2 on portions of the ITER first wall (FW). To accommodate these heat loads, roughly 50% of the first wall will have Enhanced Heat Flux (EHF) panels equipped with water-cooled hypervapotron heat sinks. Recent advances in computational fluid dynamics (CFD) enable designers to predict thermal performance even under transient two-phase flow conditions. The challenge is to design a heat sink that operates well under nominal 0.5 MW/m 2 conditions, but still has enough design margin to accommodate off-normal events. In this article, we present the results of a CFD study to investigate the tooth height and backchannel depth of 50-mm-wide hypervapotrons with 3-mm-pitch and 3-mm side slots as proposed for the fingers in the EHF FW panels. The typical EHF panel contains approximately 40 hypervapotron fingers connected to a common manifold. The water inlet temperature is 70°C at a pressure of 2.7 MPa and a mass flow rate of 0.435 kg/s per finger. The heated surface of the CuCrZr hypervapotron fingers are armored with 8-mm-thick beryllium tiles of various areas. The standard design with 4-mm-high teeth and a 5-mm-backchannel is compared to a more optimal case with 2-mm-high teeth and a 3-mm-backchannel under nominal heat loads and single-phase flow conditions. Better heat transfer in the latter case and the smaller backchannel permit a factor of two reduction in the required mass flow while maintaining the same beryllium armor surface temperatures near 130°C. The shallow teeth and smaller back channel allow the 40 fingers in a typical panel to flow in parallel and simplify the water circuit. The two hypervapotron designs are then compared during off-normal loading and two-phase flow. The design with 2-mm teeth has a 3.5% higher beryllium surface temperature of 648°C. This study highlights the necessary compromise between design margin during transient events, effective heat transfer under nominal conditions and the simplicity needed in the water circuit design.

Enhanced radial transport in the plasma and the effect of ELMS may increase the ITER first wall heat loads to as much as 4 to 5 MW/m2 over localized areas. One proposed heatsink that can handle these higher loads is a CuCrZr hypervapotron. One concept for a first wall panel consists of 20 hypervapotron channels, each measuring 1400 mm long and 48.5 mm wide. The nominal cooling conditions anticipated for each channel are 400 g/s of water at 3 MPa and 100°C. This will result in boiling over a portion of the total length. A two-phase thermalhydraulic analysis is required to predict accurately the thermal performance. Existing heat transfer correlations used for nucleate boiling are not appropriate here because the flow does not reach fully developed conditions in the multi-segmented channels. Our design-by-analysis approach used two commercial codes, Fluent and Star-CCM+, to perform computational fluid dynamics analyses with conjugate heat transfer. Both codes use the Rensselear (RPI) model for wall heat flux partitioning to model nucleate boiling as implemented in user-defined functions. We present a comparison between the two codes for this Eulerian multiphase problem that relies on temperature dependent materials properties. The analyses optimized the hypervapotron geometry, including teeth height and pitch, as well as the depth of the back channel to permit highly effective boiling heat transfer in the grooves between the teeth while ensuring that no boiling could occur at the back channel exit. The analysis used a representative heat flux profile with the peak heat flux of 5 MW/m2 limited to a 50 mm length. The maximum surface temperature of the heatsink is 415°C. The baseline design uses 2 mm for the teeth height, a 3 mm width and 6 mm pitch, and a back channel depth of 8 mm. The teeth are detached from the sidewall by a 2-mm-wide slot on both sides that aids in sweep-out and quenching of the vapor bubbles.

A nuclear analysis was performed for an alternate ITER first wall attachment scheme to estimate the nuclear heating in critical components. With a variety of components near the first wall without direct cooling, there is concern that the operating temperatures will be above engineering limits. Using DAG-MCNP5, a 1-D analysis framework was used to estimate the nuclear heating in a detailed 3-D model of the first wall and shield attachment mechanism. The parts of the stainless steel hinge at the base of system experience nuclear heating between 1.85 and 5.07 W/cm3. The copper yoke and Inconel-718 yoke pin experience a heat rate of 2.44 W/cm3. If stainless steel bolts are used, the bolt heads will experience 2.9 W/cm3, with only a small reduction from plugging the bolt access holes. If molybdenum bolts are used, the bolt head heating is increased to 3.4 W/cm3. The heating in the hinge parts and stainless steel bolts is expected to lead to temperatures higher than engineering limits. Due to a higher thermal conductivity and melting temperature, the Mo bolts are expected to operate at acceptable temperatures.

Enhanced radial transport in the plasma and the effect of ELMS may increase the ITER first wall heat loads to as much as 4 to 5 MW/m 2 over localized areas. One proposed heatsink that can handle these higher loads is a CuCrZr hypervapotron. One concept for a first wall panel consists of 20 hypervapotron channels, 1400 mm long and 48.5 mm wide. The nominal cooling conditions anticipated for each channel are 400 g/s of water at 3 MPa and 100degC. This will result in boiling over a portion of the total length, and two-phase thermalhydraulic analysis is required to predict accurately the thermal performance. Existing heat transfer correlations used for nucleate boiling are not appropriate here, because the flow does not reach fully developed conditions in the multi-segmented channels. Our design-by-analysis approach used two commercial codes, CFdesign and Fluent, to perform computational fluid dynamics analyses with conjugate heat transfer. The Fluent simulations use the Rensselaer (RPI) model for wall heat flux partitioning to model nucleate boiling as implemented in user defined functions. A more computationally expensive volume-of-fluid (VOF) multiphase model encompassing only several hypervapotron teeth provided a check on the results. We present a comparison between the two codes for this Eulerian multi-phase problem that relies on the steam tables for the fluid properties. The analyses optimized the hypervapotron geometry including teeth height and pitch and the depth of the back channel to permit highly effective boiling heat transfer in the grooves between teeth while ensuring that no boiling could occur at the back channel exit. The analysis used a representative heat flux profile with the peak heat flux of 5 MW/m 2 limited to a 50-mm-length. The surface temperature of the heatsink is kept well below 350degC. The baseline design uses 2 mm for the teeth height, a 3 mm width and 6 mm pitch, and a back channel depth of 8 mm. The teeth are detached from the sidewall by a 2-mm-wide slot on both sides that aids in sweep-out and quenching of the vapor bubbles.

We report here the fabrication processes used to manufacture US Party Team First Wall Qualification Mockups along with the detailed microstructural characterization and mechanical properties of the Be/CuCrZr/316L HIP bonds. A companion submission to this conference describes details of the PMTF heat flux testing and the performance of the first US FWQM.

To evaluate beryllium-to-copper joining techniques for potential use by US manufacturers in making first wall components for International Thermonuclear Experimental Reactor (ITER), we tested two mock-ups with S65C beryllium (Be) tiles Hot Isostatic Pressing (HIP) bonded to CuCrZr heat sinks. Under the aegis of the US ITER Project Office, Sandia prepared the mock-ups working with industrial vendors and performed high heat flux testing at Sandia's Plasma Material Test Facility (PMTF) to ascertain the robustness of the Be/Cu joints to 1000 thermal fatigue cycles at a heat flux level of 1.5 MW/m(2). Thermal stress analysis provided insight into choosing the heat flux and flow conditions required for accelerated fatigue testing at 1000 cycles and 1.5 MW/m2 that is comparable to the 12,000 cycles and 0.875 MW/m(2) required for the ITER First Wall Qualification Mock-ups.Each mock-up had three Be tiles, 35.5 mm square and 10 mm thick, bonded to a CuCrZr heat sink 134.5 mm x 36 mm x 25 nun with a single bored 12.7 mm (dia.) cooling channel. The bonding techniques included various interlayer metallizations and HIPping at 100 MPa pressure and temperature of 580 or 560 degrees C for 2 h. Each tile had a thermocouple (TC) in the center 1 mm below the Be/Cu interface. The test arrangement allowed for both mock-ups to be tested at the same time with alternate heating and cooling cycles of equal duration of 30 s. A total power of 12.7 kW was absorbed by the heated area of 4000 mm(2) during the on-cycle. The mock-up was cooled by water at 2.3 m/s (0.27 kg/s), I MPa and 20 C inlet temperature. These operating conditions did not permit the mock-ups to cool down to their initial temperature state during the off-cycle.Both mock-ups survived 1000 cycles with no significant changes. The temperature of the top surface on each reached 254 degrees C; while the center TCs reached 136 and 139 degrees C, respectively. Despite localized changes observed in the surface emissivity, the corrected temperature distributions on the surfaces varied by only a few degrees and did not change significantly during testing. We characterized the Be/Cu joint by ultrasonic testing before and after testing and sectioned the mock-ups for further evaluation. This article discusses the fabrication techniques, the results of the ultrasonic and thermal testing, and the time-dependent performance insights from computational fluid dynamics. (C) 2009 Elsevier B.V. All rights reserved.

Understanding the manner in which the First Wall Qualification Mockup (FWQM) responds structurally to simulated ITER conditions is important to the establishment of a reliable first wall. This paper provides a thermal and structural response analysis for the first round of qualification tests performed at Sandia National Laboratories. The results display the stresses and strains created in the FWQM as a result of the thermal expansion that occurred when subjected to cyclic heat flux under simulated ITER normal and MARFE conditions. From this structural response, further insight may be gained into the likelihood of fatigue failure of the Beryllium//CuCrZr interface once the first wall is in operation in ITER. While fully determining the reliability of this joint is beyond the scope of this study, some suggestions are made as to how this topic might be addressed with further research. Also investigated are the thermal patterns seen during testing that indicated slight variation from the intended test parameters. It is shown that these disparities from the ideal test parameters do not significantly affect the qualification of the FWQM.

Tritium retention and removal are critical issues for the success of ITER or any DT fusion reactor. The Tokamak Fusion Test Reactor, TFTR, is the first fusion facility to afford the opportunity to study the tritium retention and removal over an extended period. In TFTR, tritium accumulates on all surfaces with line of sight to the plasma by codeposition of tritium with carbon. Measurements of both deuterium and tritium retention fractions have yielded retention between 0.2 and 0.6 of the injected fuel in the torus. Tritium has been successfully removed from TFTR by glow discharge cleaning and by air purges. The in-vessel inventory was reduced by a factor of 2, facilitating machine maintenance. In TFTR, the amount of dust recovered from the TJTR vacuum vessel has varied from several grams to a few kilograms.

The International organization of ITER (International Thermonuclear Experimental Reactor) specifies a requirement of 3 mm in diameter for the largest permissible flaw in the joint of the beryllium (Be) armor tiles and the underlying heat sink made of a copper-chrome-zirconium (CuCrZr) alloy for the first wall (FW). We investigated the sensitivity of a new non-destructive process of detecting these flaws using a method in which we mapped the phase tag of the temperatures on the surface of a sample during thermal cycling with a sinusoidally varying water temperature. A method with hot-cold water test that we had pioneered during the 1990's for the development of a water-cooled mid-plane modular limiter for Tore Supra bad worked well with the high conductivity armor made or pyrolytic graphite brazed to copper tubes. The paper describes the experimental system. test samples and some experimental results.

The International Organization of ITER (International Thermonuclear Experimental Reactor) specifies a requirement of 3 mm in diameter for the largest permissible flaw in the joint of the beryllium (Be) armor tiles and the underlying heat sink made of a copper-chrome-zirconium (CuCrZr) alloy for the first wall (FW). We investigated the sensitivity of a new non-destructive process of detecting these flaws using a method in which we mapped the phase lag of the temperatures on the surface of a sample during thermal cycling with a sinusoidally varying water temperature. A method with hot-cold water test that we had pioneered during the 1990's for the development of a water-cooled mid-plane modular limiter for Tore Supra had worked well with the high conductivity armor made of pyrolytic graphite brazed to copper tubes. The paper describes the experimental system, test samples and some experimental results.