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A flow channel insert (FCI) is the key element of the DCLL blanket concept. The FCI serves as electrical and thermal insulator to reduce the MHD pressure drop and to decouple the temperature-limited ferritic structure from the flowing hot lead-lithium (PbLi) alloy. The main focus of the paper is on numerical computations to simulate MHD flows in the first experiments on PbLi flows in a stainless steel rectangular duct with a foam-based silicon carbide (SiC) FCI. A single uninterrupted long-term (∼6500 h) test has recently been performed on a CVD coated FCI sample in the flowing PbLi in a magnetic field up to 1.5 T at the PbLi temperature of 300 °C using the MaPLE loop at UCLA. An unexpectedly high MHD pressure drop measured in this experiment suggests that a PbLi ingress into the FCI occurred in the course of the experiment, resulting in degradation of electroinsulating FCI properties. The ingress through the protective CVD layer was further confirmed by the post-experimental microscopic analysis of the FCI. The numerical modeling included 2D and 3D computations using HIMAG, COMSOL and a UCLA research code to address important flow features associated with the FCI finite length, fringing magnetic field, rounded FCI corners and also to predict changes in the MHD pressure drop in the unwanted event of a PbLi ingress. Two physical/mathematical models have been proposed and 3D and 2D computations performed to explain the experimental results. Although the computations do confirm that the SiC FCI can significantly reduce the MHD pressure drop, these first testing results that yet do not match the theoretical predictions, suggest that more work on the FCI development and testing is still needed, first of all to ensure that the FCI can withstand PbLi ingress in a long run.

Surface micro-engineering can enhance the thermo-mechanical performance of plasma facing components (PFCs). For example, castellation of a surface can reduce thermal stress due to high heat loads and thus provide higher thermo-mechanical resilience. Recently, fabrication of a variety of micro-sized refractory dendrites with reproducible geometric characteristics (e.g., density, length, height, and aspect ratio) has been demonstrated. In contrast to flat surfaces exposed to high heat loads, dendrites deform independently to minimize near-surface thermal stress, which results in improved thermo-mechanical performance. Thus, the use of dendrites offers a unique micro-engineering approach to enhance the performance of PFC structures. A brief overview of W, Re, and Mo dendritic structures is given along with micrographs that show dendrite-coated surfaces. The thermal responses of representative dendrite structures are analyzed as a function of aspect ratios and dendrite geometry. The heat-management capability of needle-like dendrites exposed to a surface energy of up to 1 MJ/m 2 is analyzed and compared to a flat surface. It is concluded that dendrite structures can significantly reduce thermal stress in the substrate when compared to flat surfaces. Implications of dendritic surfaces on sputter erosion rates are also discussed briefly.

Structural steels for nuclear applications have undergone rapid development during the past few decades, thanks to a combination of trial-and-error, mechanism-based optimization, and multiscale modeling approaches. Deformation mechanisms are shown to be intimately related to mechanical design via dominant plastic deformation modes. Because mechanical design rules are mostly based on failure modes associated with plastic strain damage accumulation, we present here the fundamental deformation mechanisms for Ferritic/Martensitic (F/M) steels, and delineate their operational range of temperature and stress. The connection between deformation mechanisms, failure modes, and mechanical design is shown through application of design rules. A specific example is given for the alloy F82H utilized in the design of a Test Blanket Module (TBM) in the International Thermonuclear Experimental Reactor (ITER), where several constitutive equations are developed for design-related mechanical properties.

The U.S. Dual Coolant Lead Lithium (DCLL) ITER Test Blanket Module (TBM) is under development for operation in the ITER reactor. The DCLL TBM must satisfy the Structural Design Criteria for ITER In-vessel Components (SDC-IC), which provides rules for the design evaluation and stress analyses of in-vessel mechanical components of ITER with the purpose of ensuring that required safety margins are maintained relative to the types of mechanical damage which might occur as a result of imposed loadings. Primary stresses on the blanket structure come from the pressurization of coolants, the weight of the blanket element, and any electromagnetic forces due to plasma disruptions events. Secondary stresses in the materials due to thermal stress resulting from temperature gradients also contribute to the stress state of the structure. The response to primary stresses will depend on the distribution of loads, the blanket support, as well as material thermo-physical properties, which depend on operating temperatures, loads, fabrication and heat treatment and changes caused by neutron irradiation effects. A detailed structural and thermal analysis of the DCLL TBM under typical loading conditions was performed. Highly stressed locations in the TBM were identified and the stress was broken down into membrane, bending, secondary, and peak stress for evaluating local stress intensities and equivalent stress in order to apply the SDC-IC design rules. Both lowand high temperature damage rules were evaluated to show lack of excessive deformation and negligible thermal creep.

A rectangular single channel low pressure drop helium-cooled refractory metal heat exchanger (HX) tube for divertor applications was designed and manufactured for testing in the SNL E-beam facility. A unique fabrication feature of the rectangular HX channel design is that all welds, brazes, and joints are located at or near the bottom of the rectangular channel, i.e., far from any heated surface. The HX tube concept uses a thin (~ 2mm) layer of open-cell refractory foam bonded underneath the heated surface to enhance heat transfer to the helium coolant. The helium coolant flows through a 2-mm-wide slot and then through the thin foam layer (~2 mm × 12 mm × 127 mm; H/W/L) from the inlet to the outlet plenum. This design minimizes the path of helium flow through foam to about 11 mm and thus the pressure drop through the porous media is more or less constant along the length of the channel. The concept is scalable for cooling large flat surfaces, such as a flat-plate divertor, without substantially increasing the coolant pressure losses. We present CFD analyses used to optimize the design for minimum pressure drop through the porous media and for highest uniformity of surface temperatures. A designfor-manufacturing concept for a single HX-channel was developed with the goal to minimize welds or joints near heated surfaces. Based on the advanced HX-channel design a number of HX-channels were manufactured using Mo as a surrogate material instead of tungsten.

A flat-plate He-cooled divertor would provide a flat surface facing the plasma, would minimize the number of otherwise complex sub-modules needed to cool large areas, and could greatly reduce the c...

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.

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.

As part of the U.S. ITER test blanket module development effort, several flow channel insert (FCI) concepts using a variety of porous SiC and SiC/SiC composites are being developed. Using porous SiC, prototypes of FCI segments as large as 0.12 m × 0.75 m × 0.015 m were fabricated and heat tested with a maximum ΔT of ~150°C across the FCI walls. In this paper, we report on two heat tests of the FCI prototypes. The first test used radiative heating of the inside of the FCI along with convective cooling of the outside of the FCI, which resulted in a temperature drop of about ~147°C across the FCI wall. The second test involved partial submersion of the FCI structure in liquid PbLi, resulting in an inner wall surface temperature of about 600°C and an outer wall temperature of about 450°C (ΔT ~ 150°C). Detailed thermomechanical analyses of the tests were conducted, and results of the simulations are discussed in the context of actual FCI operating conditions.

We are carrying out a multidisciplinary multi-institutional program to develop the scientific and technical basis for inertial fusion energy (IFE) based on laser drivers and direct-drive targets. The key components are developed as an integrated system, linking the science, technology, and final application of a 1000-MWe pure-fusion power plant. The science and technologies developed here are flexible enough to be applied to other size systems. The scientific justification for this work is a family of target designs (simulations) that show that direct drive has the potential to provide the high gains needed for a pure-fusion power plant. Two competing lasers are under development: the diode-pumped solid-state laser (DPPSL) and the electron-beam-pumped krypton fluoride (KrF) gas laser. This paper will present the current state of the art in the target designs and lasers, as well as the other IFE technologies required for energy, including final optics (grazing incidence and dielectrics), chambers, and target fabrication, injection, and tracking technologies. All of these are applicable to both laser systems and to other laser IFE-based concepts. However, in some of the higher performance target designs, the DPPSL will require more energy to reach the same yield as with the KrF laser.

The unprecedented demands faced by fusion structures primarily derive from severe time varying thermal-mechanical loading of complex, large scale, and highly interconnected heat transfer-energy conversion structures. This grand challenge is often much too narrowly couched in terms of the development of radiation damage resistant materials, while the enormously larger challenge is the creation of material systems and multifunctional structures. In addition, the fusion system designer is faced with the untenable situation that neither the fully functional materials, nor the requisite computational tools, nor experimental simulation facilities currently exist for reliable integrity and lifetime assessments of fusion reactor structures. Considering the absence of material information and design tools, neither the materials nor the fusion designer can follow standard design processes. The design process has to become actively materials-related while materials development must closely follow design process needs. This indispensible interaction between materials and design processes leads to a ‘concurrent materials-structure design’ path, which is necessary to meet the enormous materials-structural engineering challenges of fusion.

The dual-coolant lead-lithium, or DCLL, blanket concept is of strong interest in the US fusion technology program. In the DCLL blanket, the flow channel insert (FCI) is a critical component. FCIs must have low electrical and thermal conductivity and be compatible with lead–lithium eutectic alloy (Pb–17Li) at elevated temperatures. FCIs must retain structural integrity and desirable properties even under irradiation and large temperature gradients during operation. FCIs must not fail in such a way that Pb–17Li enters the FCI and changes its electrical or thermal conductivity significantly. Another important issue for the DCLL is the development of a suitable tritium extraction from the Pb–17Li to achieve low tritium partial pressure, thus facilitating decisive tritium control. In this paper, the state of DCLL development in the US is presented including recent design modifications and results from recent RD Pb–17Li material capability and infiltration studies; simulations of MHD Pb–17Li flow characteristics and of resultant temperature distributions; and the analysis of FCI stress states based on these thermal loads. In addition, tritium extraction from Pb–17Li based on a vacuum permeator concept is shown to have the potential to achieve the desired tritium control. A discussion of DCLL optimization and unresolved DCLL issues and future R&D needs is also presented.

The High Average Power Laser (HAPL) project is pursuing development of an IFE power reactor using a solid first wall chamber. Tungsten has been chosen as the primary candidate armor material protecting the low activation ferritic steel chamber wall structure. The tungsten armor is less than 1-mm thick and is applied by vacuum plasma spraying (VPS). The failure strength of the tungsten-armor is critical, which is measured using a state-of-the-art spallation technology developed at UCLA. A nano-second laser is used to propagate a compression/tension stress wave through the composite layered structure. The tensile strength in the coating is then related to the displacement velocity of the free surface of the tungsten coating. VPS tungsten coated steel samples were tested using the laser spallation technique and coating strengths were evaluated and are reported.

During Helium implantation or generation in finite geometries, space dependent parameters and features affect Helium transport through the material. Conventional kinetic rate-theory models assume strictly homogeneous field parameters and as such can not directly resolve space dependent phenomena of helium transport. The current work outlines a new approach to simulate space-dependent helium transport during irradiation in finite geometries. The model and the numerical code, called HEROS, are described and applied to simulate typical IFE relevant helium implantation conditions. A case study using the HAPL IFE reactor design is used to demonstrate the capabilities of the HEROS code. It is shown that the HEROS code is capable of simulating very complex transient and space dependent Helium transport in finite geometries, including the simultaneous transient production of defects and space- and time-dependent temperature and temperature gradients. Space dependent nucleation and growth of helium bubbles during implantation are modeled along with the impact of biased migration and coalescence of Helium bubbles.

Large-scale finite element modeling (FEM) of the US Dual Coolant Lead Lithium ITER Test Blanket Module including damage evolution is under development. A comprehensive rate-theory based radiation damage creep deformation code was integrated with the ABACUS FEM code. The advantage of this approach is that time-dependent in-reactor deformations and radiation damage can now be directly coupled with ‘material properties’ of FEM analyses. The coupled FEM–Creep damage model successfully simulated the simultaneous microstructure and stress evolution in small tensile test-bar structures. Applying the integrated Creep/FEM code to large structures is still computationally prohibitive. Instead, for thermo-structural analysis of the DCLL TBM structure the integrated FEM–creep damage model was used to develop true stress–strain behavior of F82H ferritic steel. Based on this integrated damage evolution-FEM approach it is proposed to use large-scale FEM analysis to identify and isolate critical stress areas for follow up analysis using detailed and fully integrated creep–FEM approach.

Tungsten is a material that has the highest melting point of 3422 -C among metals. Therefore, when deposited as a coating, it can protect the substrate surface from high heat flux. In this study, tungsten (W) sprayed coatings were formed on stainless steel substrates by gas tunnel type plasma spraying at a short spraying distance. The Vickers micro-hardness of the W coating was measured as a guide to its mechanical properties. The tungsten coating has a hardness of Hv=260–300 along its cross section, which is a little lower than the hardness usually quoted for pure tungsten. Regarding the microstructure, the coating contains some pores. The results of X-ray diffraction and energy dispersive X-ray spectroscopy show that the coating consists of pure tungsten. D 2005 Elsevier B.V. All rights reserved.

The high average power laser (HAPL) program aims at developing laser inertial fusion energy (Laser IFE) based on lasers, direct drive targets and a solid wall chamber. The preferred first wall configuration is based on tungsten and ferritic steel as armor and structural materials, respectively. A key concern is the survival of the first wall under the X-ray and ion energy deposition from the fusion micro-explosion. The HAPL design and R&D effort in the chamber and material area is focused toward understanding and resolving the key armor survival issues. This includes modeling and experimental testing of the armor thermo-mechanical behavior in facilities utilizing ion, X-rays and laser sources to simulate IFE conditions. Helium management is addressed by conducting implantation experiments along with modeling of He behavior in tungsten. This paper summarizes the HAPL chamber activities. The first wall/armor configuration and design analysis are described, key chamber issues are discussed, and the R&D to address them is highlighted.

The high average power laser (HAPL) program goal is to develop a laser inertial fusion reactor using a solid first wall (FW). The FW of the inertial fusion energy (IFE) chamber is exposed to high energy photon, particle, and neutron fluxes at frequency of several Hz. The feasibility of using a micro-engineered refractory metals, such as tungsten foam as a solid FW armor is investigated. Refractory foams are a new class of materials with very limited thermo-mechanical property databases. Elastic properties of tungsten foams are not readily available, particularly at high temperatures. To estimate high-temperature elastic properties of tungsten foams, a three-dimensional finite-element model of tungsten foam was developed. True stress strain curves of tungsten foams at elevated temperatures were developed as a function of characteristic foam properties and compared with measured values. The thermo-mechanical response of the tungsten-foam armored FW was analyzed using a detailed three-dimensional finite element model. The thermo-mechanical response of a tungsten-foam protected first wall to a typical IFE pulse is presented. It is shown that the W foam armor can be tailored to meet the thermo-mechanical stress requirements of an IFE solid wall design.

Ceramic foam or cellular ceramics are proposed as a new solid breeder material configuration. Such cellular breeder materials would have an open cell structure consisting of a network of three-dimensional interconnected ligaments. Ceramic breeder foams could address some of the challenges facing packed breeder beds and potentially enhance thermal performance, increase breeding ratio, and improve structural reliability. Foam densities are not limited to those of mono-sized pebble beds, thermal conductivities are higher compared with similarly dense pebble beds; and morphology changes are expected to be much smaller and slower than in pebble beds. Heat transfer between breeder and coolant walls can be enhanced in principal, by bonding the stand-alone breeder foam to the structure. Correlations of thermo-mechanical properties of ceramic foams are reviewed to highlight the potential advantages of a foam configuration for solid breeders.

This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential candidate FW armor materials.

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0–100 keV) and ions (0.1–4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

An engineering scaling process is applied to the solid breeder ITER TBM designs in accordance with the testing objectives of validating the design tools and the database, and evaluating blanket performance under prototypical operating conditions. The goal of scaling is to ensure that changes in structural response and performance caused by changes in size and operating conditions do not reduce the usefulness of the tests. Initially, constitutive equations are applied to lay out the basic operating and design parameters that dominate blanket phenomena. The suitability of these similarity criteria for the TBM design is then confirmed by comparing finite element predictions of prototype and scale model responses. The TBM design also takes into account the need to check the codes and data for future design use. Specifically, predictability of tritium production and nuclear heating rates in a complex geometry, tritium release and permeation characteristics under fusion environments belong to this category. We conclude that this engineering scaling design process has maximized the value of ITER testing.

Ceramic foam and cellular materials are being used in a wide variety of industries and are finding ever growing number of applications. Over the past decade advances in manufacturing of cellular materials have resulted in ceramics with highly uniform interconnected porosities ranging in size from a few μm to several mm. These relatively new ceramic foam materials have a unique set of thermo-mechanical properties, such as excellent thermal shock resistance and high surface to volume ratios. Based on new advances in processing ceramic foams, we suggest the development of ceramic foams or cellular ceramics for solid breeders in fusion reactor blankets. A cellular breeder material has a number of thermo-mechanical advantages over pebble beds, which can enhance blanket performance, improve operational stability, and reduce overall blanket costs.

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 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.

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.

Tin-lithium (Sn- Li) has been identified as a candidate liquid metal coolant for fusion power reactors. Sn-Li coolants offer a number of advantages compared with pure lithium. The vapor pressure of Sn 25Li (0.25 Li mol fraction) is a factor of ∼1000 lower than that of pure Li, which allows an increase in coolant temperatures by as much as 450 K. Experimental data of the stability of ceramic materials in Sn- Li is scarce. The thermodynamic stability of various oxides, carbides, and nitrides in Sn Li is estimated as a function of lithium composition and temperature at saturated solute concentrations by evaluating the Gibbs free energy of reaction. (Δ r G). At 773 K most of the studied nitrides, carbides, and some oxides were found to be stable (Δ r G > 0). However, oxides of Fe-based alloys, such as Cr 2 O 3 and Fe 2 O 3 were found to be unstable (Δ r G < 0) for all lithium compositions.

As an element in the US Advanced Power Extraction (APEX) program, we evaluated the design option of using advanced nanocomposite ferritic steel (AFS) as the structural material and Flibe as the tritium breeder and coolant. We selected the recirculating flow configuration as our reference design. Based on the 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 and structural design requirements. This paper covers the results of the following areas of evaluation: materials selection, first wall and 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 recirculating coolant design.

A high-power hollow-cathode plasma-gun was utilized to perform durability tests of advanced silicon carbide (SiC) foam materials under extreme thermal cyclic loading conditions. SiC-foam is a primary candidate for in-cylinder thermal regenerators in diesel engines. The SiC regenerator would be exposed to rapid thermal cycling by alternating hot (∼700°C) and cold (∼50°C) gases with cycling rates ranging between 3 and 50 Hz and with flow rates of 0.1–0.3 l/cycle at gas pressures between 1 and 2 atm. Simulation of these conditions outside a combustion engine would require an elaborate experimental setup capable of alternating flow between hot and cold gas reservoirs. Furthermore, a high-temperature rapid gas flow switching system would have to be developed. Instead, a high power plasma gun was used to deliver the required hot working gas at a flow rate of 60 l/min and at a pressure of ∼2 atm. Instead of an elaborate gas flow switching system an electrically driven open flathead engine was used to alternate between the cold and hot gases at frequencies of up to 5 Hz. The plasma gun was found to be a highly economical system for delivering hot argon gases at necessary flow rates of 0.25 l/cycle. A description of the experimental setup using the high-power plasma gun is given and representative results of the SiC-foam thermal cycling performance are reported.

The thermal barrier coatings (TBCs) by gas tunnel-type plasma spraying exhibited ceramic-composite features consisting of a host oxide matrix ceramic with an embedded second phase material.The densities of the composite TBC were found to be higher than those sprayed with 100 wt% ZrO2 or Al2O3.In the coatings produced with powder mixtures of 50 wt%, the embedded splats are found to have a relatively uniform thickness between 1 and 10mm and they exhibited clear and pore-free interfaces with the host material.The micro-composite coatings also exhibited thicknessdependent functionally gradient Vickers hardness values by the hardness measurements across the coating thickness.A one-dimensional series heat transfer model was developed to estimate upper and lower bounds of the transverse thermal resistance as a function of alumina–zirconia weight ratio.The model shows that the addition of higher thermally conducting Al2O3 can result in an increase in the transverse thermal resistivity of YSZ. r 2002 Elsevier Science Ltd. All rights reserved.

A novel concept for drastically improving the surface heat load capability of helium-cooled tungsten-alloy tubes is being developed for plasma facing components. The concept utilizes ultra-low density (90% porosity) W-foam, which is chemical-vapor-deposited inside a W-tube. The W-foam enhances the effective heat transfer coefficient inside the tube by significantly increasing the conduction path from the wall to the coolant fluid. A mockup of the W-tube/W-foam system has been constructed for testing at the helium loop and electron beam facility at Sandia National Laboratories, Albuquerque, NM. A finite element model (FEM) was constructed based on a 3-D solid model of the test section. The enhanced heat transfer coefficient was determined based on fundamental heat transfer principles through porous media. The porous tungsten heat exchanger tube exhibits a 3 fold improved surface heat load capability relative to a plain W-tube at temperatures above 1200°C.

A kinetic rate theory model, which includes the formation of cascade-induced clusters (CIIC), is presented. Comparison of the model to ion irradiation data on vanadium reveals the effects of helium generation and cascade-induced interstitial and vacancy clusters on microstructure evolution. The model is based on a simplification of hierarchical rate equations for the clustering of helium bubbles, immobile vacancy clusters, glissile interstitial clusters, sessile dislocation loops, as well as precipitates and grain boundaries. The model shows that the transport of helium to dislocations, bubbles and grain boundaries is strongly transient because of coupling between the nucleation and growth modes of bubble evolution. Helium agglomeration in vacancy clusters is shown to reduce the excess vacancy flux to grow matrix and precipitate-affixed bubbles. The direct formation of vacancy and interstitial clusters in cascades reduces the growth rate of bubbles, and leads to enhanced nucleation of matrix bubbles. In addition to the dislocation and production bias mechanisms, a new mechanism of `helium nucleation bias' is shown to exist under high helium generation rates.

Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W–5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. Systems study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kW h. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

A novel ZrO 2 }Al 2 O 3 thermal barrier composite coating was produced using a gas-tunnel-type plasma spraying torch. To enhance visualization of the microstructure features, image enhancement techniques were used during the microphotograph digitization processes. The unique microstructure features of this composite coating include a relatively even distribution of embedded thin ZrO 2 splats in a Al 2 O 3 matrix, parallel alignment of the ZrO 2 splats relative to the substrate surface, absence of porosity between the ZrO 2 splats and the Al 2 O 3 matrix, etc. This anisotropic composite coating combined with the large di!erence in thermal conductivity between ZrO 2 and Al 2 O 3 will alter the thermal behavior of the coating. ( 2000 Elsevier Science Ltd. All rights reserved.

Using an axial feed DC-plasmatron, mixtures of fine Ni and WC powder (1–3 μm) were plasma sprayed onto stainless steel substrates. A series of high-density Ni–WC coatings were produces with uniform distribution of WC particles. The small powder sizes coupled with the axial feed system resulted in high vaporization rates of the Ni and significant decomposition of the WC powders. The high vaporization rate of Ni followed by condensation contributes to the high density of the coatings, while the high WC decomposition rate results in low WC content and low hardness values of the coatings. Based on published data and simple energy balance equations, evaporation of as much as 20% of the Ni powder was estimated, and XRD analysis indicates decomposition of as much as 80% of the WC particles.

Tritium permeation, particularly plasma driven permeation (PDP) through plasma facing materials has been identified as a critical safety issue for future fusion power devices. While PDP through metals has been reported earlier, this work presents the first measurements of permeation rates through SiC/SiC composite samples exposed to a modified RF-discharge plasma device. The response of the SiC/SiC composite samples to impinging plasma radiation was also investigated using SEM. In contrast to metals, SiC/SiC composite showed no evidence of PDP of deuterium. The SEMs revealed attack only of directly exposed fibers and fiber-matrix interfaces. In lieu of these measurements, SiC/SiC composites offer promising characteristics as plasma facing materials for future fusion devices.

The PULSAR pulsed tokamak power plant design utilizes the outboard shield for thermal energy storage to maintain full 1000 MW(e) output during the dwell period of 200 s. Thermal energy resulting from direct nuclear heating is accumulated in the shield during the 7200 s fusion power production phase. The maximum shield temperature may be much higher than that for the blanket because radiation damage is significantly reduced. During the dwell period, thermal power discharged from the shield and coolant temperature are simultaneously regulated by controlling the coolant mass flow rate at the shield inlet. This is facilitated by throttled coolant bypass. Design concepts using helium and lithium coolant have been developed. Two-dimensional time-dependent thermal hydraulic calculations were performed to confirm performance capabilities required of the design concepts. The results indicate that the system design and performance can accommodate uncertainties in material limits or the length of the dwell period.

The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density: and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, IOCFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a “single-piece” FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs. to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10–20 MW/m2 with a shallow minimum COE at about 18 MW/m2. Even though operation at the lower end of the this range of wall loading (10–12 MW/m2) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m2) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key physics and engineering issues central to achieving reactors with the features of TITAN-I and TITAN-II have emerged.

The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density; and to determine the potential economic operational, safety, and environmental features of high mass-power-density (MPD) fusion-reactor systems. Parametric system studies have been used to find cost-optimized designs. The design window for compact RFP reactors includes the range of 10–20 MW/m2. The reactors are physically small, and a potential benefit of this “compactness” is improved economics. The TITAN study adopted 18 MW/m2 in order to assess the technical feasibility and physics limits for such high-MPD reactors. The TITAN-I design is a lithium self-cooled design with a vanadium-alloy (V-3Ti-1Si) structural material. The magnetic field topology of the RFP is favorable for liquid-metal cooling. The first wall and blanket consist of single pass poloidal-flow loops aligned with the dominant poloidal magnetic field. A unique feature of the TITAN-I design is the use of the integrated-blanket-coil (IBC) concept. The lithium coolant in the blanket circuit is also used as the electrical conductor of the toroidal-field and divertor coils. A “single-piece” FPC maintenance procedure is used, in which the first wall and blanket are removed and replaced by vertical lift of the components as a single unit. This unique approach permits the complete FPC to be made of a few factory-fabricated pieces, assembled on site into a single torus, and tested to full operational conditions before installation in the reactor vault. A low-activation, low-afterheat vanadium alloy is used as the structural material throughout the FPC in order to minimize the peak temperature during accidents and to permit near-surface disposal of waste. The safety analysis indicates that the liquid-metal-cooled TITAN-I design can be classified as passively safe, without reliance on any active safety systems. The results from the TITAN study support the technical feasibility, economic incentive, and operational attractiveness of compact, high-MPD RFP reactors. Many critical issues remain to be resolved, however. The physics of confinement scaling, plasma transport and the role of the conducting shell are already major efforts in RFP research. However, the TITAN study points to three other major issues. First, operating high-power-density fusion reactors with intensely radiating plasmas is crucial. Second, the physics of toroidal-field divertors in RFPs must be examined. Third current drive by magnetic-helicity injection must be verified. The key engineering issues for the TITAN I FPC have also been defined. Future research and development will be required to meet the physics and technology requirements that are necessary for the realization of the significant potential economic and operational benefits that are possible with TITAN-like RFP reactors.

Strong emphasis was given to safety engineering in the TITAN study. Instead of an add-on safety design and analysis task, the safety activity was incorporated into the process of design selection and integration at the beginning of the study. Low afterheat structural material, V3 Ti1 Si was used. Passive safety features like Ar cover gas, lithium drain tanks were used to minimize the likelihood and consequence of lithium fire. Plasma accident scenarios were evaluated and found that passive safety features can be incorporated into the design. Detailed safety analysis indicated that only the assurance of coolant-piping and vacuum-vessel integrity is necessary to protect the public. The TITAN-I design meets the definition of level 3 of safety assurance design “small-scale passive safety assurance”.

The operating conditions of a compact, high-neutron-wall-loading fusion reactor severely limit the choices for structural, shield, insulator, and breeder materials. In particular the response of plasma-facing materials to radiation, thermal and pressure stresses, and their compatibility with coolants are of primary concern. Material selection issues are investigated for the compact, high mass-power-density TITAN-I reactor design study. In this paper the major findings regarding material performance are discussed. The retention of mechanical strength at relatively high temperatures, low thermal stresses, and compatibility with liquid lithium make vanadium-base alloys a promising material for structural components. Based on limited data, the thermal creep behaviour of V3TiISi and V15Cr5Ti alloys is approximated using the modified minimum committment method. In addition, the effects of irradiation and helium generation are superimposed on the creep behavior of V3Ti1Si. Coolant compatibility issues are investigated. The liquid lithium compatibility of the two vanadium alloys, V15Cr5Ti and V3Ti1Si, are compared, and the latter was chosen as the primary structural-material candidate for the liquid-lithium-cooled TITAN-I reactor. Electrically insulating materials, capable of operating at high temperatures are necessary throughout the fusion reactor device. Electrical insulator-material issues of concern include irradiation induced swelling and conductivity. Both issues are investigated and operating temperatures for minimum swelling and dielectric breakdown strength are identified for spinel (MgAI 2 O 4 ).

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currentl...

Syntactic closed-cell SiC-foam is being developed as a potential material for Flow Channel Inserts (FCIs). Syntactic foams consist of hollow spheres embedded in a matrix material. Manufacturing of syntactic polymer-base foams is a well-established technology; however, fabrication of syntactic ceramic foams has not been widely demonstrated. For FCI applications the syntactic SiC-foam should consist of near-stochiometric SiC with a well controlled closed-cell size distribution. Here, we report on the development status of the syntactic foam for FCI applications.

Development of a robust transition joint between refractory-based PFC components and external cooling piping system is critical for the successful operation of future fusion power reactors. A helium cooled tungsten-based divertor and a steel-based cooling system joint is particularly challenging, due in part to the large mismatch of thermo-physical properties between the materials and because of the high operating temperatures. A strong metallurgical bond is required in such a duplex joint structure. Drawing on broad existing industrial experience, a joint was designed using Ta as a transition material between W and steel. Thermo-structural FEM analyses were performed to investigate the mechanical response of this joint during powerup, steady state, and shut down of divertor operation.