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An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices.

Aluminum (Al) is a convenient proxy for beryllium (Be) plasma material interaction studies since they have a number of physical and chemical similarities. Al samples were exposed at the lower outer strike point of an L-mode divertor plasma in DIII-D (conditions 7–11 × 1018 D-ions cm−2 s−1, Te = 12–47 eV). The gross erosion rate was directly measured using post-mortem ion beam analysis of small 1 mm-sized samples where local re-deposition was determined to be negligible. The gross erosion rate was also calculated using spectroscopic methods, but these rates greatly underestimate the direct (i.e. non-spectroscopic) measurement. The direct measured erosion yields were within the range of published D+ → Al ion beam sputtering yields. The ionizations per photon (S/XB) coefficients used in the spectroscopic analysis were determined in separate experiments using He plasmas at the PISCES-B linear plasma facility at UCSD. The measured S/XB coefficients were on average ∼6× higher than the theoretically calculated values.

ITER is under construction and will begin operation in 2020. This is the first 500 MWfusion class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a system code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively.

An accelerated fusion energy development program, a fast-track approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation,and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 κ = 2.3, BT= 5.4 T, IP = 6.6 MA, βN = 2.75, Pfus = 127 MW. The modest bootstrap fraction of fBS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Lastly,more » proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ~ 10 in ITER.« less

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which has the potential to be a high pe...

Comparisons have been made between heat flux measurements from Langmuir probes and embedded thermocouples in the divertor of DIII-D. Good agreement has been found near the outer strike point (OSP) during L-mode operation with Neutral Beam Injection (NBI) using a sheath power transmission factor (SPTF) of 7, predicted by collisionless 1-D sheath theory. Previous SPTF measurements taken from Langmuir probes and IR imagery on DIII-D demonstrated values below the theoretical limit. The Langmuir probe array has since been upgraded and an embedded thermocouple array has been utilized to measure heat flux. The SPTF has also been measured during a NBI heated H-mode shot. This shot demonstrated a SPTF greater than 7 neat the OSP, which is due to a larger scrape-off layer (SOL) current density during H-mode operation. These studies represent a significant advancement towards finding agreement between theoretical predictions of the SPTF at the divertor and experimental measurements from the divertor diagnostics.

Solid W is projected as the preferred plasma facing material. Unfortunately, W surfaces could suffer radiation damage under DT operation and will melt under Type-I edge localized modes and disruption events. A possible approach is the use of a low-Z sacrificial material, like Si deposited on the W-surface to withstand a few type-I ELMs and/or disruptions via the vapor shielding effect. Accordingly, sets of Si–W test buttons were fabricated and exposed in the DIII-D lower divertor. We found that when the Si–W buttons were exposed to a few DIII-D vertical displacement event disruptions, tungsten–silicide was formed which melts at 1414 °C. This clearly indicates that the Si–W combination cannot be used as a transient tolerance surface material, since the W surface can be damaged. Even when Si is used as a wall conditioning material the Si–W surface temperature should be operated at much lower than 1400 °C.

Experimental observation of net erosion of molybdenum being significantly reduced compared to gross erosion in the divertor of DIII-D is reported for well-controlled plasma conditions. For the first time, gross erosion rates were measured by both spectroscopic and non-spectroscopic methods. In one experiment a net erosion rate of 0.73 ± 0.03 nm/s was measured using ion beam analysis (IBA) of a 1 cm diameter Mo-coated sample. For a 1 mm diameter Mo sample exposed at the same time the net erosion rate was higher at 1.31 nm/s. For the small sample redeposition is expected to be negligible in comparison with the larger sample yielding a net to gross erosion estimate of 0.56 ± 12%. The gross rate was also measured spectroscopically (386 nm MoI line) giving 2.45 nm/s ± factor 2. The experiment was modeled with the REDEP/WBC erosion/redeposition code package coupled to the ITMC–DYN mixed-material code, with plasma conditions supplied by the OEDGE code using Langmuir probe data input. The code-calculated net/gross ratio is =0.46, in good agreement with experiment.

The net erosion of molybdenum by the divertor plasma in the DIII-D tokamak was determined from the reduction in thickness of a thin film test sample after a short exposure to well controlled plasma conditions. The spatial distribution of Mo deposited on adjacent carbon surfaces was also measured. Integration of the total quantity of Mo deposited within 2 cm of the source, gave only 19% of the amount lost from the film indicating that most of the Mo is transported to greater distances, in spite of the short pathlength for ionization of Mo in the divertor plasma. These measurements provide benchmark data for comparisons between gross and net erosion and between measurements and simulations of erosion and deposition, which are discussed in companion papers at this conference. Erosion and deposition of carbon, and deuterium retention were also examined.

Thermo-oxidation is controlled exposure in anoxygen-containing atmosphere at elevated temperatureand is being considered as a technique for the de-tritiation of carbon-based codeposits in ITER. In addi-tion, unplanned oxidation may also occur duringaccidental air ingress. The impact of thermo-oxidationon ITER diagnostic mirrors causes concerns. A dedi-cated study was performed in DIII-D, where molybde-num and copper mirrors were installed in the mainchamber, in the divertor, and at a location remote fromthe plasma and exposed for

A Fusion Nuclear Science Facility (FNSF) is a necessary complement to ITER, especially in the area of materials and components testing, needed for DEMO design development. FNSF-AT, which takes advantage of advanced tokamak (AT) physics should have neutron wall loading of 1-2 MW/m2, continuous operation for periods of up to two weeks, a duty factor goal of 0.3 per year and an accumulated fluence of 3-6 MW-yr/m2 (~30-60 dpa) in ten years to enable the qualification of structural, blanket and functional materials, components and corresponding ancillary equipment necessary for the design and licensing of a DEMO. Base blankets with a ferritic steel structure and selected tritium blanket materials will be tested and used for the demonstration of tritium sufficiency. Additional test ports at the outboard mid-plane will be reserved for test blankets with advanced designs or exotic materials, and electricity production for integrated high fluence testing in a DT fusion spectrum. FNSF-AT will be designed using conservative implementations of all elements of AT physics to produce 150-300 MW fusion power with modest energy gain (Q

A Fusion Nuclear Science Facility (FNSF) is necessary to make possible a DEMO of the Advanced Tokamak (AT) type after ITER. One candidate, Fusion Nuclear Science Facility-AT (FNSF-AT), should have ...

Prior heat pulse testing of plasma facing components (PFCs) has been completed in vacuum environments without the presence of background plasma. Edge localized modes (ELMs) will not be this kind of isolated event and one should know the effect of a plasma background during these transients. Heat-pulse experiments have been conducted in the PISCES-A device utilizing laser heating in a divertor-like plasma background. Initial results indicate that the erosion of PFCs is enhanced as compared to heat pulse or plasma only tests. To determine if the enhanced erosion effect is a phenomena only witnessed in the laboratory PISCES device, tungsten and graphite samples were exposed to plasmas in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Mass loss analysis indicates that materials that contain significant deuterium prior to experiencing a transient heating event will erode faster than those that have no or little retained deuterium.

Recent progress in understanding of disruptions and in developing methods to avoid disruption damage is presented. Nearly complete mitigation of conducted heat loads has been achieved with high-Z gas jet shutdown. The resulting local radiation heat flash melting in the main chamber might be a concern in ITER, especially with beryllium walls. During the current quench, significant vessel forces can occur due to halo currents I-halo; however, these are found to fall reliably below a boundary of (halo current fraction times halo current peaking factor)

The Porous Plug Injector (PPI) has proven to be an invaluable diagnostic for in situ characterization and quantification of erosion phenomena in DIII-D. Previous work has led to derivation of three primary figures of merit for chemical erosion (CE) in attached and cold divertor conditions: relative intensity of C+ chemical and physical sources, the CE yield (Ychem), and effective photon efficiencies for chemically eroded products. Application of these figures for accounting of observed absolutely calibrated CI and CII emission intensities is demonstrated to produce a self-consistent solution at the DIII-D targets. Reinterpretation of the CI (C0) spectral lineshape profile supports the relative roles of local chemical versus physical sputtering as previously determined for CII (C+). Comparison of calculated in situ Ychem to that measured ex situ suggests a tokamak-specific lower energy threshold for CE and has potentially major implications for prediction of tritium co-deposition near the divertor targets in ITER.

Liquid metal breeders such as Lithium or the eutectic Lead-Lithium alloy PbLi have the potential for attractive breeding blankets, especially if the liquid metal serves as breeder and coolant. However, cooling of first wall and blanket structure is a challenging task because the magnetic field degrades the heat transfer and can cause a really high pressure drop. To overcome these problems, dual coolant blankets with helium cooled FW/blanket structure and a self-cooled breeding zone had been proposed, with electrical insulation by ceramic-coatings or sandwich flow channel inserts. Such concepts are in principle simpler than helium cooled blankets, but the thermal efficiency is limited to ˜35 % as in any helium cooled blankets with steel structure. A much higher efficiency up to about 45 % became feasible when the sandwich insulator was replaced by flow channel inserts (FCI) made of a SiC composite. This FCI serves as thermal insulator too, allowing an exit temperature of ˜700° C, suitable for a BRA...

The US Fusion Nuclear Science and Technology program selected the Dual Coolant Lead Lithium (DCLL) concept as the primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is the structural material, helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and tritium extraction. The DCLL TBM has undergone major design changes since 2005. We present here the most recent thermo-mechanical analysis of the newly revised DCLL TBM. The analysis described here is aiming to verify the thermo-mechanical response of the DCLL TBM under relevant normal operating conditions as well as during a loss of coolant accident (LOCA). A full 3-dimensional solid model of the entire DCLL TBM structure was developed, which included FW, top and bottom lids, internal supporting ribs, manifolds, plena, and flexible frame-attachment supports. A coupled thermo-mechanical analysis was performed for both normal- and off-normal operating conditions. Thermal loads included surface heat load, volumetric heating, as well as detailed position- and location dependent heat transfer along all coolant channels. Structural loads incorporated helium coolant pressure loads, self-weight, as well as the weight of the PbLi. Maximum structure temperatures of nearly 560 o C along with a maximum resultant net displacement of more than 10 mm were mapped for normal operating conditions and a number of stress concentration locations were identified. The ITER SDC-IC-1300 criteria were applied to the LOCA analysis results. It is shown that the DCLL TBM exhibits admissible behavior regarding the ITER Design Criteria and that the most recent design modifications did not compromise the structural integrity.

The effects of D plasma pre-irradiation of vacuum plasma spraying tungsten (VPS-W), when exposed to He plasma exposures were investigated by using the linear plasma simulator, PISCES-A and the divertor materials evaluation system (DiMES) in DIII-D tokamak. Polished VPS-W samples with and without preexposure to 1 × 10 26 D/m 2 deuterium plasma were loaded into PISCES-A as well as DiMES for He exposure. Scanning electron microscopy (SEM) images show that significant surface morphology changes occur under tokamak condition. Mass loss of the samples is found to be relevant to poloidal loading position in DIII-D, indicating erosion induced by ELMing He plasma exposure. The effects of D pre-irradiation on retention are discussed by comparing results of nuclear reaction analysis (NRA) and thermal desorption spectroscopy (TDS). © 2011 Elsevier B.V. All rights reserved.

To move to a fusion DEMO power plant after ITER, a Fusion Nuclear Science Facility (FNSF) is needed in addition to ITER and research in operating tokamaks and those under construction. The FNSF wil...

The pre-conceptual design for the US DEMO inboard dual-coolant lead–lithium breeding blanket is introduced for the first time followed by the assessment for the most important magnetohydrodynamic issues for the blanket module itself and access ducts. The considered issues include: (i) the magnetohydrodynamic pressure drop, (ii) electric insulation in poloidal flows using the silicon carbide flow channel insert, and (iii) countercurrent flows in the access ducts.

Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (∼650–700 °C) from the RAF/M structure, which has a corrosion temperature limit of ∼480 °C. The RAF/M material must also operate at temperatures above 350 °C but less than 550 °C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.

The objective of the Fusion Development Facility (FDF) under consideration is to fill the gaps between ITER and current experiments and a fusion demonstration power plant (DEMO). FDF should carry forward advanced tokamak (AT) physics and enable development of fusion's energy applications. Near-term AT physics and nonsuperconducting magnet technology will be used to achieve steady state with burn, producing 100-250 MW fusion power with modest energy gain (Q

The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization of fusion reactors and enable development of fusion’s energy...

Erosion of the main chamber plasma-facing components is of concern for ITER. Plasma interaction with the outboard chamber wall is studied in DIII-D using Langmuir probes and optical diagnostics. Fast camera data shows that edge localized modes (ELMs) feature helical filamentary structures propagating towards the outboard wall. Upon reaching the wall, filaments result in regions of local intense plasma–material interaction (PMI) where peak incident particle and heat fluxes are up to two orders of magnitude higher than those between ELMs. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. A moderate change of the gap between the separatrix and the outer wall strongly affects PMI intensity at the wall. Material samples exposed near the outboard wall showed net carbon deposition in high-density discharges (near the Greenwald limit) and tendency towards net erosion in lower density discharges (∼0.45 of the Greenwald limit).

An improved, self-contained gas injection system for the divertor material evaluation system (DIMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH(4) are found to be 58 +/- 12 in an attached divertor (n(e) similar to 1.5 x 10(13)/cm(3), T(e) similar to 25 eV, T(surf)similar to 450 K), and 94 +/- 20 in a semi-detached cold divertor (n(e) similar to 6.0 x 10(13)/cm(3), T(e) similar to 2-3 eV, T(surf) similar to 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C(+) at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH(4) from the Pill, then compared with thatmore » seen in intrinsic sputtering. (C) 2009 Elsevier B.V. All rights reserved.« less

Testing of breeding blanket modules (TBMs) is one of the ITER goals foreseen from the very beginning of the ITER Project. Six half port TBMs and associated systems are expected to be tested simultaneously in three available Test Ports. This paper presents an initial assessment of the TBM and ITER interface requirements. Four areas of interface were identified. The first area is the port cell interface area, including components like the port plug frame, backside shield, dummy TBM and corresponding tools needed for the TBM maintenance and replacement. The second area is the hot cell, including the needed additional hardware for the service of TBMs, additional remote handling tools, and additional building space needed for the maintenance of the TBM ancillary equipment and the corresponding testing utilities and tools. The third area is the tokamak cooling water system (TCWS) with the need to accommodate six TBM heat transfer systems, each with a footprint of 57 m2. The fourth area of interface is the tritium plant. In all these areas modifications in the current ITER design are needed to accommodate the TMB testing. These changes must be incorporated in the new ITER baseline design which is now under preparation. The latest experiments on JET revealed unexpectedly high sensitivity of plasmas in H-mode of confinement to ripples of the magnetic field. The ferromagnetic test modules can create additional ripples. This new issue of interface between ITER and TBMs is also addressed.

We summarize here the results of ongoing studies for magnetohydrodynamic (MHD) flows and heat transfer in the eutectic alloy lead–lithium (PbLi), which is used as a breeder and coolant, for three US Dual-Coolant Lead–Lithium (DCLL) blanket scenarios: ITER H-H, ITER D-T, and DEMO. The paper focuses on the important blanket feasibility issues, such as the MHD pressure drop, heat leakages from PbLi into the cooling helium flows, and the temperature distributions in the insulating flow insert and at the material interface. Both ITER scenarios look acceptable, i.e., all material restrictions can be easily met. In the DEMO scenario, the flow insert operates in harsh conditions causing high temperature drop in the insert and high interface temperature between the hot PbLi and the ferritic structure that may exceed the allowable material limits.

In support of the ITER Test Blanket Module (TBM) program and coordinated by the Test Blanket Working Group, ITER party members have been focusing on the liquid metal blanket design concepts, most of which have been extensively explored. For the demonstration power reactor (DEMO) design, we will have to accommodate the neutron wall loading and first wall heat flux, breed and extract adequate tritium for the D-T fuel cycle and achieve high coolant outlet temperature for high power conversion efficiency. Most proposed liquid metal TBMs have the potential of achieving similar DEMO goals and requirements. Furthermore, all liquid metal TBMs are to satisfy ITER safety requirements and to be operated and tested within ITER operation scenarios. For the development of liquid metal TBM concepts, many R&D elements are common to a few designs such as the areas of Reduced Activation Ferritic/Martensitic Steel (RAFM, also abbreviated as FS in the following) or V-alloy fabrication, thermal fluid MHD, FS/PbLi, FS/Li and V-alloy/Li compatibility, irradiation effects on different materials, tritium extraction, SiC flow channel insert (FCI) development, etc. With a well-coordinated ITER TBM program, different parties’ R&D activities can supplement and complement each other via collaborations. This paper will present respective designs and R&D programs from seven ITER parties.

In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. Infrared cameras, visible and vacuum ultraviolet cameras, pressure gauges and residual gas analyzers, thermocouples, and erosion and deposition measurements by insertable probes, quartz microbalances, and a rather extensive review of Langmuir probes are discussed. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and infrared cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER.

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.

Solid breeder and liquid breeder blanket concepts are being developed for testing in ITER. In addition to fundamental investigations on the use of reduced activation ferritic/ferritic–martensitic steels (RAFMS) or V-alloys as the structural material, tritium breeder and neutron multiplier materials, and corresponding corrosion issues, there are three classes of functional materials being developed. Performance of these functional materials will significantly impact the performance of respective blanket concepts. Two classes of materials are related to the self-cooled liquid breeder design concepts. The first class is MHD coatings for the self-cooled Li breeder concepts. The second class is for flow channel inserts (FCI) for the dual coolant PbLi concepts. The third class of functional material is tritium permeation barrier material commonly needed for solid and liquid breeder blanket concepts. A description and the development status of these functional materials are described in this paper.

The US is proposing a prototype of a dual coolant liquid lead–lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

Migration of pre-characterized carbon dust in a tokamak environment was studied by introducing about 30 milligrams of dust flakes 5-10 {micro}m in diameter in the lower divertor of DIII-D using the DiMES sample holder. The dust was exposed to high power ELMing Hmode discharges in lower-single-null magnetic configuration with the strike points swept across the divertor floor. When the outer strike point (OSP) passed over the dust holder exposing it to high particle and heat fluxes, part of the dust was injected into the plasma. In about 0.1 sec following the OSP pass over the dust, 1-2% of the total dust carbon content (2-4 x 10{sup 19} carbon atoms, equivalent to a few million dust particles) penetrated the core plasma, raising the core carbon density by a factor of 2-3. When the OSP was inboard of the dust holder, the dust injection continued at a lower rate. Individual dust particles were observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction for deuteron flow to the outer divertor target, consistent with the ion drag force. The observed behavior of the dust is in qualitative agreement with modeling by the 3D DustT code.

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. The porous plug injector (PPI) releases methane into the plasma at a controlled rate through a porous graphite surface flush to a tile. In this way, the perturbation to the local plasma can be minimized, while also simulating the immediate environment of methane molecules released from a solid graphite surface. Photon efficiencies of CH 4 for measured local plasma conditions are reported. The contribution of chemical vs physical sputtering to the source of C + at the target can, in principle, be assessed through measurement of CII and CD/CH band emissions during release of CH 4 from the PPI, and due to intrinsic emission. These first results from this new experimental tool demonstrate the potential for the PPI to provide definitive results in future applications in DIII-D and indicate the improvements required to obtain firm quantitative conclusions.

Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper.

MHD flow and heat transfer have been analyzed for a front poloidal channel in the outboard module of a Dual Coolant Lithium Lead (DCLL) blanket, with a flow channel insert made of a silicon carbide composite. The US reference DCLL blanket module [C. Wong, S. Malang, M. Sawan, S. Smolentsev, S. Majumdar, B. Merrill, D.K. Sze, N. Morley, S. Sharafat, P. Fogarty, M. Dagher, P. Peterson, H. Zhao, S. Zinkle, M. Youssef, Assessment of liquid breeder first wall and blanket options for the DEMO design, in: Proceedings of the 16th ANS TOFE Meeting, Madison, September 14–16, 2004.] has been considered. Effectiveness of the insert as insulator was assessed via numerical simulations based on a 2D model for a fully developed flow and on a 3D model for heat transport. Parametric studies were performed at σ = 5–500 (Ωm) −1 and k = 2–20 W/m K. Parameters resulting in a reasonably low MHD pressure drop and almost no heat leakage from the breeder into the cooling helium flows have been identified.

An attractive blanket concept for the fusion reactor is the dual coolant Pb–17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb–17Li is circulated for power conversion and for tritium breeding. A SiC f/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb–17Li and as the thermal insulator to separate the high temperature Pb–17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb–17Li outlet temperature of 700 ◦ C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb–17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program. © 2005 Elsevier B.V. All rights reserved.

A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully independent TBM and (2) a dual-coolant helium-cooled ferritic steel structure with self-cooled LiPb breeding zone that uses a flow channel insert as MHD and thermal insulator. Example test module designs and configuration choices for each line of ITER TBM are shown and discussed in the paper. In addition, near-term R&D items for decision-making on testing of both solid breeder and dual-coolant PbLi liquid breeder blanket concepts in ITER are identified.

Far Scrape-Off Layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma contact with the low field side (LFS) main chamber wall. In H-mode between edge localized modes (ELMs) the plasma?wall contact is weaker than in L-mode. During ELM fluxes of particles and heat to the LFS wall increase transiently above the L-mode values. Depending on the discharge conditions, ELMs are responsible for 30?90% of the net ion flux to the outboard chamber wall. ELMs in high density discharges feature intermittent transport events similar to those observed in L-mode and attributed to blobs of dense hot plasma formed inside the separatrix and propagating radially outwards. Though the blobs decay with radius, some of them survive long enough to reach the outer wall and possibly cause sputtering. In lower density H-modes, ELMs can feature blobs of pedestal density propagating all the way to the outer wall.

Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation in the ITER Test Blanket Module (TBM) Program. A US strategy for ITER-TBM has evolved that emphasizes international collaboration. A study was initiated to select the two blanket options for the US ITER-TBM in light of new R and D results from the US and world programs over the past decade. The study is led by the Plasma Chamber community in partnership with the Materials, PFC, Safety, and physics communities. The study focuses on assessment of the critical feasibility issues for candidate blanket concepts and it is strongly coupled to R and D of modeling and experiments. Examples of issues are MHD insulators, SiC insert viability and compatibility with PbLi, tritium permeation, MHD effects on heat transfer, solid breeder 'temperature window' and thermomechanics, and chemistry control of molten salts. A dual coolant liquid breeder and a helium-cooled solid breeder blanketmore » concept have been selected for the US ITER-TBM.« less

As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant lead-eutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R&D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

The US Advanced Limiter-divertor Plasma-facing Systems (ALPS) program is developing the science of liquid metal surface divertors for near and long term tokamaks. These systems may help solve the d...

Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ~1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (~3 kg) over the blanket lifetime of ~3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, th...

Far scrape-off layer (SOL) plasma parameters in DIII-D depend strongly on the discharge density and confinement regime. In L-mode, cross-field transport increases with the average discharge density and elevates the far SOL density, thus increasing plasma-wall contact. Far SOL density near the low field side (LFS) of the main chamber wall also increases with decreasing plasma current and with decreasing outer wall gap. In H-mode, between edge localized modes (ELMs), plasma-wall contact is weaker than in L-mode. During ELMs plasma fluxes to the LFS wall increase to, or above the L-mode levels. A large fraction of the net cross-field fluxes is convected through the SOL by large amplitude intermittent transport events. In high-density L-mode and during ELMs in H-mode, intermittent events propagate all the way to the LFS wall and may cause sputtering.

The Advanced Power Extraction (APEX) project has been exploring concepts for power producing blankets that can enhance the potential of fusion energy. The pursued concepts cover both liquid and solid wall designs. The solid wall blanket branch of the project has been concentrating on the use of nano-composited ferritic (NCF) steel structure coupled with Flibe (Li 2 BeF 4 ) coolant. The present blanket is a solid wall design with an innovative coolant scheme, which allows part of the coolant to be re-circulated in order to enhance the outlet temperature, and thus improve the power cycle efficiency. The structure is 12YWT, an oxide dispersion strengthened (ODS) ferritic steel, which has a maximum operating temperature of 800 °C and a compatibility with Flibe up to 700 °C. Several methods for fabricating the complex geometry of the first wall (FW) and blanket are presented. Preliminary coolant routing is proposed with solutions offered for minimizing heat losses and simplifying assembly and maintenance. Even though this blanket is somewhat complicated, its forward looking aims, that of maximizing nuclear performance to achieve a high thermal conversion efficiency, are well worth striving for.

In this article, we explore some of the safety issues associated with two advanced ferritic steel (AFS) molten salt blanket designs from the Advanced Power Extraction (APEX) design study [M.A. Abdou, The APEX Team, On the exploration of innovative concepts for fusion chamber technology, Fus. Eng. Des. 54 (2000) 181]. In particular, we examine radiological inventories, decay heat, waste disposal ratings, and toxic chemical inventories of these design concepts. In addition, we predict the thermal response of these blanket designs during accident conditions, and the mobilization of the radiological inventories and site boundary dose from the release of this mobilized material during a worst-case confinement-boundary-bypass accident. The molten salts being proposed for these blanket concepts are Flibe and Flinabe, and the structural material is a nano-composite strengthened ferritic steel alloy called 12YWT. The estimated dose at the site boundary is less than the no-evacuation limit of 10 mSv for a ground level release during conservative weather conditions if plant isolation occurs within 5 days.

As an element in the U.S. Advanced Power Extraction (APEX) program, the solid first wall and blanket design team assessed innovative design configurations with the use of advanced nano-composite ferritic steel (AFS) as the structural material and FLiBe as the tritium breeder and coolant. The goal for the assessment is to search for designs that can have high volumetric power density and surface heat-flux handling capability, with assurance of fuel self-sufficiency, high thermal efficiency and passive safety for a tokamak power reactor. We selected the re-circulating flow configuration as our reference design. Based on the recommended material properties of AFS we found that the reference design can handle a maximum surface heat flux of 1 MW/m2, and a maximum neutron wall loading of 5.4 MW/m2, with a gross thermal efficiency of 47%, while meeting all the tritium breeding, structural design and passive safety requirements. This paper will cover the results of the following areas of assessment: material design properties, FW/blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal-hydraulics analysis including MHD effects, structural analysis; molten salt and helium closed cycle power conversion system; and safety and waste disposal of the re-circulating coolant first wall and blanket design.

Several experiments exposing a 5 cm 2 solid and liquid lithium to the divertor plasma of the DIII-D are described. The divertor plasma strikepoint cleans and conditions the initially solid lithium surface. The effective sputtering rate and transport of lithium was found to be acceptable. Lithium has a sputtering yield of solid lithium T e ∼20 eV. The sputtered lithium is ionized in a short distance from the divertor and promptly redeposited. Experiments and modeling show the sputtered lithium is well shielded by the divertor plasma. The behavior of the liquefied lithium was dominated by its macroscopic movement and injection into the plasma caused by J × B magnetohydrodynamic (MHD) forces. Plasma MHD events, such as edge localized modes (ELMs) and locked modes, are found to provide simultaneously the energy to melt the lithium and the transiently high scrape-off layer (SOL) currents to cause the J × B motion. The macroscopic removal of lithium from the small sample, comprising