Alice Ying

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons why the blanket/FW will pace fusion development toward a DEMO. This paper summarizes the top technical issues and elucidates the primary challenges in developing the blanket/first wall and identifies the key R&D needs in non-fusion and fusion facilities on the path to DEMO.

This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load >10 MW/m2 and surface heat flux >2 MW/m2, (2) high power conversion efficiency (>40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid “bare” first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn–Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin (∼2 cm) or thick (∼40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma β and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma–liquid interactions including both plasma–liquid surface and liquid wall–bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W–5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at ∼1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.

The development of tritium breeder, neutron multiplier and flow channel insert materials for the breeding blanket of the DEMO reactor is reviewed. Present emphasis is on the ITER test blanket module (TBM); lithium metatitanate (Li2TiO3) and lithium orthosilicate (Li4SiO4) pebbles have been developed by leading TBM parties. Beryllium pebbles have been selected as the neutron multiplier. Good progress has been made in their fabrication; however, verification of the design by experiments is in the planning stage. Irradiation data are also limited, but the decrease in thermal conductivity of beryllium due to irradiation followed by swelling is a concern. Tests at ITER are regarded as a major milestone. For the DEMO reactor, improvement of the breeder has been attempted to obtain a higher lithium content, and Be12Ti and other beryllide intermetallic compounds that have superior chemical stability have been studied. LiPb eutectic has been considered as a DEMO blanket in the liquid breeder option and is used as a coolant to achieve a higher outlet temperature; a SiC flow channel insert is used to prevent magnetohydrodynamic pressure drop and corrosion. A significant technical gap between ITER TBM and DEMO is recognized, and the world fusion community is working on ITER TBM and DEMO blanket development in parallel.

The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R = 4.8 m, a = 1.2 m, IP = 7.9 MA, BT = 7.5 T, βN < 2.7, n/nGr = 0.9, fBS = 0.52, q95 = 6.0, H98 ∼1.0, and Q = 4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25 years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

Abstract The tritium aspects of the DT fuel cycle embody some of the most challenging feasibility and attractiveness issues in the development of fusion systems. The review and analyses in this paper provide important information to understand and quantify these challenges and to define the phase space of plasma physics and fusion technology parameters and features that must guide a serious R&amp;D in the world fusion program. We focus in particular on components, issues and R&amp;D necessary to satisfy three ‘principal requirements’: (1) achieving tritium self-sufficiency within the fusion system, (2) providing a tritium inventory for the initial start-up of a fusion facility, and (3) managing the safety and biological hazards of tritium. A primary conclusion is that the physics and technology state-of-the-art will not enable DEMO and future power plants to satisfy these principal requirements. We quantify goals and define specific areas and ideas for physics and technology R&amp;D to meet these requirements. A powerful fuel cycle dynamics model was developed to calculate time-dependent tritium inventories and flow rates in all parts and components of the fuel cycle for different ranges of parameters and physics and technology conditions. Dynamics modeling analyses show that the key parameters affecting tritium inventories, tritium start-up inventory, and tritium self-sufficiency are the tritium burn fraction in the plasma ( f b ), fueling efficiency ( η f ), processing time of plasma exhaust in the inner fuel cycle ( t p ), reactor availability factor (AF), reserve time ( t r ) which determines the reserve tritium inventory needed in the storage system in order to keep the plant operational for time t r in case of any malfunction of any part of the tritium processing system, and the doubling time ( t d ). Results show that η f f b &gt; 2% and processing time of 1–4 h are required to achieve tritium self-sufficiency with reasonable confidence. For η f f b = 2% and processing time of 4 h, the tritium start-up inventory required for a 3 GW fusion reactor is ∼11 kg, while it is &lt;5 kg if η f f b = 5% and the processing time is 1 h. To achieve these stringent requirements, a serious R&amp;D program in physics and technology is necessary. The EU-DEMO direct internal recycling concept that carries fuel directly from the plasma exhaust gas to the fueling systems without going through the isotope separation system reduces the overall processing time and tritium inventories and has positive effects on the required tritium breeding ratio (TBR R ). A significant finding is the strong dependence of tritium self-sufficiency on the reactor availability factor. Simulations show that tritium self-sufficiency is: impossible if AF &lt; 10% for any η f f b , possible if AF &gt; 30% and 1% ⩽ η f f b ⩽ 2%, and achievable with reasonable confidence if AF &gt; 50% and η f f b &gt; 2%. These results are of particular concern in light of the low availability factor predicted for the near-term plasma-based experimental facilities (e.g. FNSF, VNS, CTF), and can have repercussions on tritium economy in DEMO reactors as well, unless significant advancements in RAMI are made. There is a linear dependency between the tritium start-up inventory and the fusion power. The required tritium start-up inventory for a fusion facility of 100 MW fusion power is as small as 1 kg. Since fusion power plants will have large powers for better economics, it is important to maintain a ‘reserve’ tritium inventory in the tritium storage system to continue to fuel the plasma and avoid plant shutdown in case of malfunctions of some parts of the tritium processing lines. But our results show that a reserve time as short as 24 h leads to unacceptable reserve and start-up inventory requirements. Therefore, high reliability and fast maintainability of all components in the fuel cycle are necessary in order to avoid the need for storing reserve tritium inventory sufficient for continued fusion facility operation for more than a few hours. The physics aspects of plasma fueling, tritium burn fraction, and particle and power exhaust are highly interrelated and complex, and predictions for DEMO and power reactors are highly uncertain because of lack of experiments with burning plasma. Fueling by pellet injection on the high field side of tokamak has evolved to be the preferred method to fuel a burning plasma. Extrapolation from the DIII-D penetration scaling shows fueling efficiency expected in DEMO to be &lt;25%, but such extrapolations are highly uncertain. The fueling efficiency of gas in a reactor relevant regime is expected to be extremely poor and not very useful for getting tritium into the core plasma efficiently. Gas fueling will nonetheless be useful for feedback control of the divertor operating parameters. Extensive modeling has been carried out to predict burn fraction, fueling requirements, and fueling efficiency for ITER, DEMO, and beyond. The fueling rate required to operate Q = 10 ITER plasmas in order to provide the required core fueling, helium exhaust and radiative divertor plasma conditions for acceptable divertor power loads was calculated. If this fueling is performed with a 50–50 DT mix, the tritium burn fraction in ITER would be ∼0.36%, which is too low to satisfy the self-sufficiency conditions derived from the dynamics modeling for fusion reactors. Extrapolation to DEMO using this approach would also yield similarly low burn fraction. Extensive analysis presented shows that specific features of edge neutral dynamics in ITER and fusion reactors, which are different from present experiments, open possibilities for optimization of tritium fueling and thus to improve the burn fraction. Using only tritium in pellet fueling of the plasma core, and only deuterium for edge density, divertor power load and ELM control results in significant increase of the burn fraction to 1.8–3.6%. These estimates are performed with physics models whose results cannot be fully validated for ITER and DEMO plasma conditions since these cannot be achieved in present tokamak experiments. Thus, several uncertainties remain regarding particle transport and scenario requirements in ITER and DEMO. The safety standard requirements for protection of the public and release guidelines for tritium have been reviewed. General safety approaches including minimizing tritium inventories, reducing tritium permeation through materials, and decontaminating material for waste disposal have been suggested.

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

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs.

The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li4SiO4, and two forms of Li2TiO3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R&D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER.

Among the international fusion solid breeder blanket community, there exists steady progress on the experimental, phenomenological, and numerical characterizations of the pebble bed effective thermo physical and mechanical properties, and of thermomechanic state of the bed under prototypical operating conditions. This paper summarizes recent achievements in pebble bed thermomechanics that were carried out by members of the IEA Fusion Nuclear Technology Subtask I Solid Breeding Blanket. A major goal is on developing predictive capability while identifying a pre-conditioned equilibrium stress state that would warrant pebble bed integrity during operations. The paper reviews and synthesizes existing computational modeling approaches for pebble bed thermomechanics prediction, and differentiating points of convergence/divergence among existing approaches. The progress toward modeling benchmark is also discussed. These advancements have led to a framework to help navigate future research.

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 m, κ = 2.3, BT = 5.4 T, IP = 6.6 MA, βN = 2.75, Pfus = 127 MW. The modest bootstrap fraction of ƒBS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. 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.

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&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&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 blanket concept have been selected for the US ITER-TBM.

The thermomechanical interaction of solid breeder and beryllium pebble beds with structural material (BSMI) has been identified as a critical issue for solid breeder blanket designs. For example, the expansion of pebble beds restrained by blanket structure exerts stresses on the pebbles as well as the blanket structure wall, which might cause the pebbles to break and jeopardise the blanket operation. However, at elevated temperatures thermal creep will reduce these stresses and might compensate for stress build-up due to irradiation-induced swelling. A significant influence of irradiation on the pebble bed properties is expected. Computationally, the BSMI can be assessed in two ways: (i) by applying appropriate finite element codes combined with the description of modules for the pebble beds. As input, these modules require data on characteristic pebble bed properties determined in different standard-type tests; (ii) by numerical simulations based on a discrete numerical model. Here, the stress profiles are calculated while the effective modulus and bed thermal expansion coefficients are back estimated. In this paper, recent experimental results on thermomechanical pebble bed properties for ceramic breeder (metatitanate and orthosilicate) pebble beds and beryllium pebble beds are presented, including data on the moduli of deformation, thermal creep, inner friction angle, and thermal conductivity of deformed pebble beds. Furthermore, modelling results of the BSMI for simple geometries are reported based both on homogeneous and discrete models and are compared with experimental results.

An international study conducted by technical experts from Europe, Japan, Russia, and the United States has evaluated the technical issues and the required testing facilities for the development of fusion blanket/first-wall systems and has found that some of the key requirements for the engineering feasibility of blanket concepts cannot be established prior to extensive testing in the fusion environment. However, because of availability and low cost, testing in nonfusion facilities (e.g., fission reactors and laboratory experiments) serves a critical role in blanket research and development (R&D) and reduces the risks and costs of the more complex and expensive fusion experiments. A comprehensive analysis shows that the fusion testing requirements for meeting the goal of demonstrating a blanket system availability in DEMO > 50% are as follows: a 1 to 2 MW/m2 neutron wall load, a steady-state plasma operation, a > 10-m2 test area, and a fluence of > 6 MW·yr/m2. This testing fluence includes 1 to 3 MW·yr/m2 for concept performance verification and >4 to 6 MW·yr/m2 for component engineering development and reliability growth/demonstration. Reliability and availability analyses reveal critical concerns that need to be addressed in fusion power development. For a DEMO reactor availability goal of 50%, the blanket availability needs to be ∼80%. For a mean time to recover from a failure of ∼3 months, the mean time between failure (MTBF) for the entire blanket must be >1 yr. For a blanket that has 80 modules, the corresponding MTBF per module is 80 yr. These are very ambitious goals that require an aggressive fusion technology development program. A number of scenarios for fusion facilities were evaluated using a cost/benefit/risk analysis approach. Blanket tests in the International Thermonuclear Experimental Reactor (ITER) alone with a fluence of 1 MW·yr/m2 can address most of the needs for concept verification, but it cannot adequately address the blanket component reliability growth/demonstration testing requirements. An effective path to fusion DEMO is suggested. It involves two parallel facilities: (a) ITER to provide data on plasma performance, plasma support technology, and system integration and (b) a high-volume plasma-based neutron source (HVPNS) dedicated to testing, developing, and qualifying fusion nuclear components and material combinations for DEMO. For HVPNS to be attractive and cost effective, its capital cost must be significantly lower than ITER, and it should have low fusion power (∼150 MW). Exploratory studies indicate the presence of a design window with a highly driven plasma. A testing and development strategy that includes HVPNS would decisively reduce the high risk of initial DEMO operation with a poor blanket system availability and would make it possible – if operated parallel to the ITER basic performance phase – to meet the goal of DEMO operation by the year 2025. Such a scenario with HVPNS parallel to ITER provides substantial savings in the overall R&D cost toward DEMO compared with an ITER-alone strategy. The near-term cost burden is negligible in the context of an international fusion program with HVPNS and ITER sited in two different countries.

The long-range goal of the Z-Pinch IFE program is to produce an economically-attractive power plant using high-yield z-pinch-driven targets (~3GJ) with low rep-rate per chamber (~0.1 Hz). The present mainline choice for a Z-Pinch IFE power plant uses an LTD (Linear Transformer Driver) repetitive pulsed power driver, a Recyclable Transmission Line (RTL), a dynamic hohlraum z-pinch-driven target, and a thick-liquid wall chamber. The RTL connects the pulsed power driver directly to the z-pinch-driven target, and is made from frozen coolant or a material that is easily separable from the coolant (such as carbon steel). The RTL is destroyed by the fusion explosion, but the RTL materials are recycled, and a new RTL is inserted on each shot.A development path for Z-Pinch IFE has been created that complements and leverages the NNSA DP ICF program. Funding by a U.S. Congressional initiative of $4M for FY04 through NNSA DP is supporting assessment and initial research on (1) RTLs, (2) repetitive pulsed power drivers, (3) shock mitigation [because of the high yield targets], (4) planning for a proof-of-principle full RTL cycle demonstration [with a 1 MA, 1 MV, 100 ns, 0.1 Hz driver], (5) IFE target studies for multi-GJ yield targets, and (6) z-pinch IFE power plant engineering and technology development. Initial results from all areas of this research are discussed.

The use of lithium ceramic pebble beds has been considered in many blanket designs for the fusion reactors. Lithium ceramics have received a significant interest as tritium breeders for the fusion blankets during the last three decades. The thermal performance of the lithium ceramic pebble beds plays a key role for the fusion blankets. In order to study the heat transfer in the blanket, the effective thermal conductivity of the lithium ceramics pebble beds has to be well measured and characterized. The data of effective thermal conductivity of lithium ceramic pebble beds is important for the blanket design. Several studies have been dedicated to investigate the effective conductivity of the lithium ceramics pebble beds. The objective of this work is to review and compare the available data, presented by various studies, of effective conductivity of lithium ceramic pebble beds in order to address the current status of these data.

Tritium breeding materials, which have the ability to react with neutrons and produce tritium, are required to fuel fusion reactions. Tritium is produced inside the fusion blanket by neutron irradiation of lithium ceramics (tritium breeders). Using the lithium ceramics in form of pebbles-packed bed is a promising concept for fusion blankets, and worldwide efforts have been dedicated to its R&D. Thermal properties of lithium ceramic pebble beds have a significant impact upon the thermal performance of the fusion blanket. Specifically, the effective thermal conductivity of lithium ceramic pebble beds is important for the design and analysis of fusion blankets. The experimental apparatus of this study was designed and built based on the principles of the steady state and axial heat flow methods in order to conduct the required measurements. The objective of this study is to measure the effective thermal conductivity of a Li2TiO3 pebble bed as a function of the average bed temperature. The pebble bed has pebbles of 1.7–2.0 mm diameter and a packing fraction of 61%. Helium at atmospheric pressure was used as a filling gas. The experimental results showed that the effective thermal conductivity decreased from 1.40 to 0.94 W/m K with the increase of the average bed temperature from 50 to 500 °C. The results presented in this work will help to create a database of the effective thermal conductivity of Li2TiO3 pebble beds which can be used for the design and analysis of fusion blankets.

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 discrete element method (DEM) is used to study the thermal effects of pebble failure in an ensemble of lithium ceramic spheres. Some pebbles crushing in a large system is unavoidable and this study provides correlations between the extent of pebble failure and the reduction in effective thermal conductivity of the bed. In the model, we homogeneously induced failure and applied nuclear heating until dynamic and thermal steady-state. Conduction between pebbles and from pebbles to the boundary is the only mode of heat transfer presently modeled. The effective thermal conductivity was found to decrease rapidly as a function of the percent of failed pebbles in the bed. It was found that the dominant contributor to the reduction was the drop in inter-particle forces as pebbles fail; implying the extent of failure induced may not occur in real pebble beds. The results are meant to assist designers in the fusion energy community who are planning to use packed beds of ceramic pebbles. The evolution away from experimentally measured thermomechanical properties as pebbles fail is necessary for proper operation of fusion reactors.

In magnetohydrodynamic (MHD) flows turbulence reduction occurs due to the Joule dissipation. It results in heat transfer degradation. In open channel flows, heat transfer degradation is also caused by the turbulence redistribution near the free surface. Both effects can be significant in fusion applications with low-conductivity fluids such as molten salts. In the present study, the “K–ε” model equations for turbulent flows and the free surface boundary condition are adjusted with taking into account MHD effects. Different orientations of the magnetic field, perpendicular and parallel to the main flow, have been considered. The model coefficients have been tuned by a computer optimization using available experimental data for the friction factor. The effect of free surface heat transfer degradation due to the turbulence redistribution has been implemented through the variation of the turbulent Prandtl number. As an example, the model is used for the analysis of a turbulent MHD flow down an inclined chute with the heat flux applied to the free surface.

In this paper, the discrete element method (DEM) approach has been applied to study mechanical behaviors of ceramic breeder pebble beds. Directly simulating the contact state of each individual particle by the physically based interaction laws, the DEM numerical program is capable of predicting the mechanical behaviors of non-standard packing structures. The program can also provide the data to trace the evolution of contact characteristics and forces as deformation proceeds, as well as the particle movement when the pebble bed is subjected to external loadings. Our numerical simulations focus on predicting the mechanical behaviors of ceramic breeder pebble beds, which include typical fusion breeder materials in solid breeder blankets. Current numerical results clearly show that the packing density and the bed geometry can have an impact on the mechanical stiffness of the pebble beds. Statistical data show that the contact forces are highly related to the contact status of the pebbles.

The effective thermal conductivity (keff) of lithium meta-titanate (Li2TiO3) pebble beds is an important parameter for the design and analysis of IN LLCB TBM (Indian Lead Lithium Ceramic Breeder Test Blanket Module). The keff of Li2TiO3 pebble beds under stagnant helium gas have been determined numerically using different uniform packing structures and random close packing (RCP) structures. The uniform packing structures of Li2TiO3 pebble bed are modelled by using the simple cubic, body centered cubic and face centered cubic arrangement. The packing structure of the RCP bed of Li2TiO3 pebbles is generated with the discrete element method (DEM) code. keff of Li2TiO3 pebble beds with different packing fractions have been reported as function of temperature; keff of the RCP Li2TiO3 pebble bed is compared with reported experimental results from literature. The numerically determined keff of the Li2TiO3 pebble bed agrees reasonably well with the experimental data.

Numerical simulation of time-dependent magneto-convection flows in a liquid metal breeding blanket unit cell prototypical to the geometry of Water-Cooled Lithium-Lead (WCLL) breeding blanket is performed by COMSOL Multiphysics. Its modeling capability for fully coupled magnetohydrodynamic (MHD)/heat transfer flows in a complex geometry under fusion relevant conditions is tested with the demonstration of the main flow features and temperature profiles. Here, the magneto-convection flow in the bulk region is dominated by the effect of natural convection, leading to two columnar-like, counter-rotating circulations which occupy the whole plenum region of the breeding zone along the external magnetic field direction. High velocity flow jets with the magnitude of 1–2 orders greater than inlet velocity are formed near structural walls that are parallel to the applied magnetic field and break into unsteady sidewall vortices that are mainly confined in boundary layers. Due to the buoyancy effect, the hot fluid particles are stratified near the top structure wall where the cooling effect is the least. The calculated maximum temperature at the wall is lower than the structure’s temperature limit but still exceeds the corrosion temperature limit.

Considering the optimization of blanket performance, it is desired that the bed morphology and packing state during reactor operation are stable and predictable. Both experimental and numerical work are performed to explore the stability of pebble beds, in particular under pulsed loading conditions. Uniaxial compaction tests have been performed for both KIT’s Li4SiO4 and NFRI’s Li2TiO3 pebble beds at elevated temperatures (up to 750 °C) under cyclic loads (up to 6 MPa). The obtained data shows the stress-strain loop initially moves towards the larger strain and nearly saturates after a certain number of cyclic loading cycles. The characterized FEM CAP material models for a Li4SiO4 pebble bed with an edge-on configuration are used to simulate the thermomechanical behavior of pebble bed under ITER pulsed operations. Simulation results have shown the cyclic variation of temperature/stress/strain/gap and also the same saturation trend with experiments under cyclic loads. Therefore, it is feasible for pebble bed to maintain its packing stability during operation when disregarding pebbles’ breakage and irradiation.

AbstractThe effective thermal conductivity of the pebble beds is one the most important design parameters for pebble bed solid breeder blanket. In the framework of IEA Implementing Agreement on Solid Breeder Subtask Group, measurement of pebble bed thermal conductivity by the hot wire method were defined as one of tasks to provide comparative information on the effective thermal conductivity of candidate ceramic pebble beds for DEMO blanket designs and ITER breeding blanket design. The authors previously reported the preliminary result of the pebble bed thermal conductivity for Li2O, Be and Al2O3. This paper presents the result of Li2TiO3, Li2ZrO3 (1 mm diameter) from CEA, and Li4SiO4 (0.25 - 0.63 mm diameter) from FZK.Observation was compared to the correlations, SZB model and HM model. Contact area fraction was obtained by correlation fitting, of which the value is 4.9×10−3 for Li2TiO3, Li2ZrO3 (the same value as Li2O) and 1×10−6 for and Li4SiO4.

ABSTRACT ABSTRACT A general formula for the second-order projection method combined with the level set method is developed to simulate unsteady, incompressible multifluid flow with phase change. A subcell conception is introduced in a modified mass transfer model to accurately calculate the mass transfer across the interface. The third-order essentially nonoscillatory (ENO) scheme and second-order semi-implicit Crank-Nicholson scheme is employed to update the convective and diffusion terms, respectively. The projection method has second-order temporal accuracy for variable-density unsteady incompressible flows as well. The level set approach is employed to implicitly capture the interface for multiphase flows. A continuum surface force (CSF) tension model is used in the present cases. Phase change and dynamics associated with single bubble and multibubbles in two and three dimensions during nucleate boiling are studied numerically via the present modeling. The numerical results show that this method can handle complex deformation of the interface and account for the effect of liquid–vapor phase change. The authors would like to thank Prof. S. Osher for fruitful discussion. This work is supported by the U.S. Department of Energy under Grant DE-FG03-86ER52123.

The successful operation (with β ≤ 60%, classical ions and electrons with Te = 250 eV) of the gas dynamic trap device at the Budker Institute of Nuclear Physics in Novosibirsk, Russia, extrapolates to a 2 MW/m2 dynamic trap neutron source (DTNS), which burns only ~100 g of tritium per full-power year. The DTNS has no physics, engineering, or technology showstoppers; the extension of neutral beam lines to steady state can use demonstrated engineering; and it supports near-term tokamaks and volume neutron sources. The DTNS provides a neutron spectrum similar to that of ITER and satisfies the missions specified by the materials community to test fusion materials (listed as one of the top grand challenges for engineering in the 21st century by the U.S. National Academy of Engineering) and subcomponents (including tritium-breeding blankets) needed to construct DEMO. The DTNS could serve as the first fusion nuclear science facility (FNSF), called for by ReNeW (the Research Needs Workshop), and could provide the data necessary for licensing subsequent FSNFs.

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This paper reports on experimental findings on liquid metal (LM) free surface flows crossing complex magnetic fields. The experiments involve jet and film flows using GaInSn and are conducted at the UCLA MTOR facility. The goal of this study is to understand the magnetohydrodynamics (MHD) features associated with such a free surface flow in a fusion-relevant magnetic field environment, and determine what LM free surface flow option is most suitable for lithium divertor particle pumping and surface heat removal applications in a near-term experimental plasma device, such as NSTX. Experimental findings indicate that a steady transverse magnetic field, even with gradients typical of NSTX outer divertor conditions, stabilizes a LM jet flow—reducing turbulent disturbances and delaying jet breakup. Important insights into the MHD behavior of liquid metal films under NSTX-like environments are also presented. It is possible to establish an uphill liquid metal film flow on a conducting substrate, although the MHD drag experienced by the flow could be strong and cause the flow to pile-up under simulated NSTX magnetic field conditions. The magnetic field changes the turbulent film flow so that wave structures range from 2D column-type surface disturbances at regions of high magnetic field, to ordinary hydrodynamic turbulence wave structures at regions of low field strength at the outboard. Plans for future work are also presented.

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 demanding heat removal, particle removal, and erosion issues of fusion plasma/surface interactions. ALPS combines tokamak experiments, lab experiments, and modeling. We are designing both static and flowing liquid lithium divertors for the National Spherical Torus Experiment (NSTX) at Princeton. We are also studying tin, gallium, and tin-lithium systems. Results to date are extensive and generally encouraging, e.g., showing: 1) good tokamak performance with a liquid Li limiter, 2) high D pumping in Li and non-zero He/Li pumping, 3) well-characterized temperature-dependent liquid metal surface composition and sputter yield data, 4) predicted stable low-recycle improved-plasma NSTX-Li performance, 5) high temperature capability Sn or Ga potential with reduced ELM & disruption response concerns. In the MHD area, analysis predicts good NSTX static Li performance, with dynamic systems being evaluated.

A numerical approach using the discrete element method (DEM) has been applied to study the thermo-mechanical properties of ceramic breeder pebble beds. This numerical scheme is able to predict the inelastic behavior observed in a loading and unloading operation. In addition, it demonstrates that the average value of contact force increases linearly with overall pressure, but at a much faster rate, about 3.4 times the overall pressure increase rate. In this paper, the thermal creep properties of two different ceramic breeder pebble materials, Li4SiO4 and Li2O, are also examined by the current numerical code. The difference found in the properties of candidate materials is reflected numerically in the overall strain in the pebble bed when the stress magnitude becomes smaller.

An integrated, multi-physics, dynamic predictive tool to quantify tritium retention, removal, and permeation for HCCR Test Blanket System (TBS) is presented in this paper. The tool expands from detailed COMSOL component models developed previously at UCLA, into an integrated, system-level blanket model using MATLAB/Simulink. It aims at achieving self-consistent predictions in particular concerning dynamic tritium concentration built-up in the He coolant. The integration is achieved by implementing COMSOL component models in the discrete section of Simulink S-Functions. The model replicates HCCR TBS process flow diagram and preserves main tritium flow characteristics for both helium cooling and tritium extraction systems. Current results demonstrate importance of detailed component models as well as dynamic simulation for improved accuracy on answers to questions related to safety/licensing and designs.

Characterizing the thermo-physical properties of the ceramic breeder pebble bed is an integral step of developing breeder blankets for fusion energy applications. To that end, thermal conductivity is an important parameter to identify. In granular pebble bed materials, the thermal conductivity depends on the solid pebble material as well as any gas filling the interstitial void spaces, thus an effective thermal conductivity () of the bulk is used. A transient hot-wire apparatus is developed through a collaborative study between the Fusion Science and Technology Center at UCLA and the National Fusion Research Institute (NFRI) to measure the effective thermal conductivity of Korean-made Li2TiO3 pebble beds. In this study, current is pushed through a single strand of high purity platinum wire. The heat generated is conducted away by the surrounding pebble bed; the logarithmic change in temperature being used to calculate the rate of heat conductance. The apparatus is filled with roughly an atmosphere of helium and placed in a furnace to test the pebble bed under reactor relevant temperatures. Results and future improvements are presented.

The effective thermal conductivity (keff) of lithium meta-titanate (Li2TiO3) pebble beds under fusion relevant environments is an important property for the design of IN LLCB TBM (Indian Lead Lithium Ceramic Breeder Test Blanket Module). The transient hot wire technique was used to examine the thermal property of the Indian made Li2TiO3 material. The hot wire is used as both the heating element as well as for the temperature measurement. The keff of Li2TiO3 pebble bed has been investigated from room temperature to 800 °C. Experiments were performed on uncompressed Li2TiO3 pebble bed in stagnant helium gas filled at ambient pressure. A clear dependence of the keff on the temperature of the pebble bed was observed. The pebble bed has pebbles of 1 ± 0.15 mm diameter and packing fraction of 63%. The experimental results showed that the keff increased from 0.903 W/m°C to 1.204 W/m°C with the increase of bed temperature from 34.3 °C to 785.4 °C. The random close packing of poly dispersed Li2TiO3 pebble bed has been generated using discrete element method and then numerical modeling has been performed using finite element method to estimate keff. The numerically determined keff of the Li2TiO3 pebble bed agrees reasonably well with the obtained experimental data. The experimentally achieved keff results are also compared with the reported experimental results elsewhere and also with Zehner–Schlunder correlation.

In this paper, advancement on tritium transport simulations was demonstrated for a solid breeder blanket HCCR TBS, where multi-physics and detailed engineering descriptions are considered using a commercial simulation code. The physics involved includes compressible purge gas fluid flow, heat transfer, chemical reaction, isotope swamping effect, and tritium isotopes mass transport. The strategy adopted here is to develop numerical procedures and techniques that allow critical details of material, geometric and operational heterogeneity in a most complete engineering description of the TBS being incorporated into the simulation. Our application focuses on the transient assessment in view of ITER being pulsed operations. An immediate advantage is a more realistic predictive and design analysis tool accounting pulsed operations induced temperature variations which impact helium purge gas flow as well as Q2 composition concentration time and space evolutions in the breeding regions. This affords a more accurate prediction of tritium permeation into the He coolant by accounting correct temperature and partial pressure effects and realistic diffusion paths. The analysis also shows that by introducing by-pass line to accommodate ITER pulsed operations in the TES loop allows tritium extraction design being more cost effective.

A new high-performance cellular solid breeder is presented that has several times the thermal conductivity and is substantially denser compared with sphere-packed breeder beds. The cellular breeder is fabricated using a patented process of melt-infiltrating ceramic breeder material into an open-cell carbon foam. Following solidification the carbon foam is removed by oxidation. This process results in a near 90% dense robust freestanding breeder in a block configuration with an internal network of open interconnected micro-channels for tritium release. The network of interconnected micro-channels was investigated using X-ray tomography. Aside from increased density and thermal conductivity relative to pebble beds, high temperature sintering is eliminated and thermal durability is increased. Cellular breeder morphology, thermal conductivity, specific heat, porosity levels, high temperature mechanical properties, and deuterium charging-desorption rates are presented.

The purpose of this paper is to present our recent efforts on 3D MHD model development and our results based on the technique derived from induced-magnetic-field equations. Two important features are utilized in our numerical method to obtain convergent solutions. First, a penalty factor is introduced in order to force the local divergence free condition of the magnetic fields. The second is that we extend the insulating wall thickness to ensure that the induced magnetic field at its boundaries is null. These simulation results for lithium film free surface flows under NSTX outboard mid-plane magnetic field configurations have shown that 3D MHD effects from a surface normal field gradient cause return currents to interact with surface normal fields and produce unfavorable MHD forces. This leads to a substantial change in flow pattern and a reduction in flow velocity, with most of the flow spilling over one side of the chute. These critical phenomena can not be revealed by 2D models. Additionally, a design which overcomes these undesired flow characteristics is obtained.

We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall/blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel structures. The research described for these concepts includes both thermofluid MHD issues for the liquid metal coolant in the DCLL, and thermomechanical issues for ceramic breeder packed pebble beds in the solid breeder concept. Finally, future directions for ongoing research in these areas are described.

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.

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.

Cryogenic hydrogen adsorption using molecular sieve beds is considered to be one of the main candidate processes for recovery of produced tritium from purge gas in breeding blankets and it has been chosen for separating hydrogen isotopes in Tritium Extraction System (TES) of Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS). Various adsorbents and their performance have been studied for the cryogenic adsorption using small-scale experiments. However, large-scale experiments comparable to TBS-relevant scale are required to have sufficient confidence for component design and performance prediction of Cryogenic Molecular Sieve Bed (CMSB) for the TBS and beyond. To properly evaluate hydrogen adsorption performance of a large size CMSB, a series of experiments have been performed using PGLoop facility which is constructed and operated in the National Fusion Research Institute. The experimental conditions were set to include breeding-blankets-relevant parameters. As such, the effects of swamping ratio, total pressure, and flow rates on the performances of CMSB were studied in the range of hydrogen partial pressures from 100 to 700 Pa. While a slight reduction in hydrogen adsorption performance is observed in comparison to the small isotherm experiments, which can be attributed to scale-up effects, it shows that the experimental results agree reasonably well with existing literature data.

Available data and mathematical formulations concerning tritium transport in the FW/Divertor with tungsten and beryllium as plasma facing materials were implemented in the commercial code COMSOL Multiphysics. The goal is to develop a CAD-based multiphysics modeling capability so that FW/Divertor temperature and geometric features can be readily taken into consideration while tritium permeation to the primary coolant in a prototypical PFC can be more realistically addressed. This development began with the simulation of ion implantation experiments, validated against existing laboratory experimental results. Analysis shows that with ITER FW where Be is used as the plasma facing material, the low operating temperature, erosion, and the dwell time greatly hinder tritium bulk diffusion, permeation, and inventory accumulation. However, under DEMO high-temperature operating conditions, tritium can quickly diffuse through tungsten to structural material and reach a steady state inventory after a relatively short time. Additionally, its permeation to the coolant can be reduced when the Soret effect is considered. The findings and challenges of developing a 3-D predictive capability for tritium transport in a FW/Divertor PFC are discussed.

First-of-a-kind numerical simulation was performed to evaluate time dependent tritium transport properties for Korea’s HCCR (Helium-Cooled Ceramic Reflector) TBM (Test Blanket Module) design under ITER inductive operating conditions. The estimation of tritium inventories in various components of the HCCR submodule and its permeation amount into the helium coolant was obtained through three computational models involving: 1) a 3D FW standalone model where diffusion and permeation into FW He coolant through tritium ion implantation was studied, 2) a 2D Poloidal-Radial (P-R) mid-plane model where the effect of increased tritium concentration in the purge gas stream was accounted for, and 3) a 2D Toroidal-Radial (T-R) mid-plane model to study tritium concentration accumulation in the He coolant. The analysis shows that tritium inventory in the breeder reaches an equilibrium value in about 10 cycles, and is about 0.373 mg per submodule. Tritium inventory in the ferritic steel structure reaches its equilibrium value in less than 10 cycles, and has about 0.0012 mg per submodule at the end of the plasma burn. The amount of the tritium permeated into helium coolant is about 1.8% of the amount of tritium produced per cycle.

In this paper the problem of tritium transport in PbLi (Lead-Lithium) blankets has been studied and analyzed by means of our recently developed computational models. Several simulations are performed by incorporating the geometric configurations of the PbLi blankets including both DCLL (Dual Coolant Lead Lithium) and HCLL (Helium Cooled Lead Lithium) blankets. Tritium permeation loss percentage from the HCLL concept is about one order of magnitude higher than from the DCLL concept (~ 17%. vs. 1.2%). Sensitivity study also shows that the most relevant factors on tritium permeation are: 1) the level of tritium solubility in PbLi, 2) the gap velocity of the liquid metal in a DCLL blanket, 3) Hartmann number, and 4) the FCI (Flow Channel Insert) electrical conductivity.

Beryllium, in its pebble form, has been proposed in various blanket concepts to serve different purposes. Thermal property data for such a heterogeneous packed bed is needed, particularly data on the impact of compression forces on its magnitude and consequent temperature profile. The objectives of this work are to obtain and quantify experimental data on the effective thermal conductivity of a Be-He packed bed, on the interface heat conductance between Be and SiC, and on the effects of externally applied pressure on these effective thermal properties. The effective thermal conductivity of a Be-He pebble bed increases as the bed mean temperature increases. The values of effective thermal conductivity vary from 2.15 to 3.00 W/m.K for bed mean temperature ranges from 90 to 420 °C. Similar temperature effects are seen in the Be/SiC interface heat conductance, as the values of interface heat conductance range from 1140 to 2200 W/m2.K. In addition, effective thermal conductivity increases remarkably with the increase of applied pressure (by a factor of 2.53 at 2 MPa), while it remains higher than the initial value by ~0.3 W/m.K when external pressure is released (hysteresis effect).

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 use of fast moving liquid metal streams or "liquid walls" as a plasma contact surface is a very attractive option and has been looked upon with considerable interest over the past several years, both by the plasma physics and fusion engineering programs. Flowing liquid walls provide an ever replenishing contact surface to the plasma, leading to very effective particle pumping and surface heat flux removal. A key feasibility issue for flowing liquid metal plasma facing component (PFC) systems, pertains to their magnetohydrodynamic (MHD) behavior under the spatially varying magnetic field environment, typical of a fusion device. MHD forces hinder the development of a smooth and controllable liquid metal flow needed for PFC applications. The present study builds up on the ongoing research effort at UCLA, directed towards providing qualitative and quantitative data on liquid metal free surface flow behavior under fusion relevant magnetic fields.

A mathematical model for permeation of multi-components (H2, T2, HT) through a RAFM (Reduced activation ferritic/martensitic) membrane was described based on kinetic theory. Experimental conditions of tritium permeation for ARAA (Advanced Reduced Activation Alloy) material performed at INL were recreated in simulations for model validation. Both numerical simulations and experimental data indicated that the presence of hydrogen reduces tritium permeation rate significantly in low tritium partial pressure with 1000 ppm (0.1%) hydrogen-helium gas mixture at 1atm. Experimental behavior of tritium permeation flux dependence on tritium isotope partial pressure confirmed the kinetic theory. i.e., it still follows diffusion-controlled, square root dependence, with T2 partial pressures and a linear dependence HT pressure even though it is in a diffusion-controlled regime. In addition, the numerical model was validated with literature data for mono-isotope permeation through oxidized and clean MANET II (MArtensitic for NET) samples under surface-controlled and diffusion-controlled regimes. The simulation results agreed well with the experimental data, which indicated that the mono permeation rate through the oxidized sample is much lower (~2 orders) than clean sample and the permeation rate is proportional to p1 and p0.5 for oxidized and clean MANET II samples, respectively.

The use of flowing liquid metal streams or “liquid walls” as a plasma contact surface is a very attractive option and has received considerable attention over the past several years both in the plasma physics and fusion engineering programs. A key issue for the feasibility of flowing liquid metal plasma facing component (PFC) systems, lies in their magnetohydrodynamic (MHD) behavior. The spatially varying magnetic field environment, typical of a fusion device can lead to serious flow disrupting MHD forces that hinder the development of a smooth and controllable flow needed for PFC applications. The present study builds up on the ongoing research effort at UCLA, directed towards providing qualitative and quantitative data on liquid metal free surface flow behavior under fusion relevant magnetic fields, to aid in better understanding of flowing liquid metal PFC systems.

In this paper, an effective and accurate numerical model that involves a suggested mathematical formulation, viz., the stream functions (ψ and A)-velocity-magnetic induction formulation and a fourth-order compact difference algorithm is proposed for solving the two-dimensional (2D) steady incompressible full magnetohydrodynamic (MHD) flow equations. The stream functions-velocity-magnetic induction formulation of the 2D incompressible full MHD equations is able to circumvent the difficulty of handling the pressure variable in the primitive variable formulation or determining the vorticity values on the boundary in the stream function-vorticity formulation, and also ensure the divergence-free constraint condition of the magnetic field inherently. A test problem with the analytical solution, the well-studied lid-driven cavity problem in viscous fluid flow and the lid-driven MHD flow in a square cavity are performed to assess and verify the accuracy and the behavior of the method proposed currently. Numerical results for the present method are compared with the analytical solution and the other high-order accurate results. It is shown that the proposed stream function-velocity-magnetic induction compact difference method not only has the excellent performances in computational accuracy and efficiency, but also matches well with the divergence-free constraint of the magnetic field. Moreover, the benchmark solutions for the lid-driven cavity MHD flow in the presence of the aligned and transverse magnetic field for Reynolds number (Re) up to 5000 are provided for the wide range of magnetic Reynolds number (Rem) from 0.01 to 100 and Hartmann number (Ha) up to 4000.

A new 3D numerical model has been developed to simulate the thermal and mechanical characteristics of packed beds used in fusion reactor blankets and other applications. This method is based on an explicit numerical scheme which monitors the interaction of the particles contact by contact and their motion particle by particle. In this paper, a mathematical formulation as well as a model which predicts the packed bed thermomechanical states under imposed and induced loads are presented. The model is validated by comparing the numerical simulations with thermal expansion and uniaxial compression experiments of packed beds. The results of the calculated bed effective modulus compares reasonably well with the experimental data.

A SiC-based flow channel insert (FCI) is used as an electrical and thermal insulator in the Dual Coolant Lead Lithium (DCLL) blanket. To reduce the stress of the FCI structural material, the pressure equalization slot (PES) is implemented in the FCI wall. However, the PES affects the tritium transfer behavior and loss rate. Therefore it is important to examine the tritium loss rate and ensure it remains below an allowable limit. In the present study, we analyze tritium transport and quantify the tritium loss rate in a front duct of the DCLL-type outboard blanket where PbLi moves poloidally. Three types of poloidal ducts have been considered: one without the PES, one with the PES in the wall parallel to the magnetic field and one with the PES in the wall perpendicular to the magnetic field. Tritium concentration fields are obtained by solving a fully 3-D problem with appropriate boundary conditions at various interfaces. Results show a high tritium concentration at the location of reversed flow when a PES was located in the wall parallel to the field. Furthermore, when any PES was introduced, the PES changed the velocity profiles and thus changed the tritium concentrations in the core and gaps, which increases the tritium losses from 1.244% to 1.413% under the calculation conditions.

Understanding the behavior of a plasma chamber component in the fusion environment requires a simulation technique that is capable of integrating multi-disciplinary computational codes while appropriately treating geometric heterogeneity and complexity. Such a tool should be able to interpret phenomena from mutually dependent scientific disciplines and predict performance with sufficient accuracy and consistency. Integrated multi-physics simulation predictive capability (ISPC) relies upon advanced numerical simulation techniques and is being applied to ITER first wall/shield and Test Blanket Module (TBM) designs. In this paper, progress in ISPC development is described through the presentation of a number of integrated simulations. The simulations cover key physical phenomena encountered in a fusion plasma chamber system, including tritium permeation, fluid dynamics, and structure mechanics. Interface engines were developed in order to pass field data, such as surface deformation or nuclear heating rate, from the structural analysis to the thermo-fluid MHD analysis code for magnetohydrodynamic (MHD) velocity profile assessments, or from the neutronics analysis to the thermo-fluid analysis for temperature calculations, respectively. Near-term effort toward further ISPC development is discussed.

The need of a high performance CFD simulation to evaluate design accuracies involving subcooled boiling as a local high heat removal scheme makes the use of two-phase flow simulation challenging. This paper addresses the applicability of Bergles and Rosenow nucleate boiling model and Reynolds Averaged Navier Stokes (RANS) methods to ITER FW CFD/thermo-fluid design analysis, in which subcooled boiling in a hypervapotron is considered for the high heat flux removal. Initially, the subcooled flow boiling model adopted in the considered CFD code is evaluated by comparing the calculated wall temperatures with the estimated values that were derived from the empirical correlations. The geometry, volumetric flow rate, and the surface heat flux influence upon the heat transfer enhancement and the predicted wall temperature were then explored and discussed. In the subsequent thermo-fluid design analysis, a complete FW panel subjected to a local high heat flux of 5 MW/m2 was analyzed.

Abstract Numerical simulation of co-permeation of tritium and hydrogen from breeding zones to the coolant in the helium cooled pebble-bed blanket is performed in this paper. 3D multi-species convection–diffusion models integrated with thermal-fluid analysis in porous media are assessed and then used to estimate the associated tritium permeation for a solid breeder blanket module. Benchmark calculations give a reasonable agreement on the co-permeation rates with the experimental data. Simulation in a TBM unit show that purge gas flow can strongly affect tritium transport, increasing the purge flow velocity is an effective method to reduce tritium permeation to the coolant. In the case where hydrogen is added to the purge gas stream to promote tritium release, the co-permeation of H2, T2, and HT are taken into account in the permeation simulation, results show that permeation flux of T–T molecules is reduced due to the effect of co-permeation of hydrogen.

We apply coupled computational fluid dynamics and discrete element method (CFD-DEM) modeling tools with new numerical implementations of pebble fragmentation to study the combined effects of granular crushing and ensemble restructuring, granular fragment size, and initial packing for different breeder volume configurations. In typical solid breeder modules, heat removal from beds relies on maintaining pebble–pebble and pebble–wall contact integrity. However, contact is disrupted when an ensemble responds to individually crushed pebbles. Furthermore, restructuring of metastable packings after crushing events are, in part, dependent on gravity forces acting upon the pebbles. We investigate two representative pebble bed configurations under constant volumetric heat sources; modeling heat removed from beds via inter-particle conduction, purge gas convection, and contact between pebble beds and containers. In one configuration, heat is removed from at walls oriented parallel to the gravity vector (no gap formation possible); in the second, heat is removed at walls perpendicular to gravity, allowing for the possibility of gap formation between bed and wall. Judging beds on increase in maximum temperatures as a function of crushed pebble amount, we find that both pebble bed configurations to have advantageous features that manifest at different stages of pebble crushing. However, all configurations benefit from achieving high initial packing fractions.

The integrity of the electrically insulating coating at the coolant channel walls represents a feasibility issue for a self-cooled liquid metal blanket. The effects of cracks in the insulating layer on MHD pressure drop are investigated, for various crack locations, sizes and resistivities. It is shown that, once the crack resistance exceeds the Hartmann layer resistance, the MHD pressure drop increases significantly. Using the 2-D fully-developed MHD flow code, it has been found that the crack location has a large impact on the MHD pressure drop: for the same crack sizes and resistivities, the pressure gradient increases as the cracks move towards the central plane. The effect of the insulation on heat transfer is discussed. The heat transfer capability is reduced in the presence of the insulator coating.

Abstract A program of model development and experimentation has been undertaken to develop innovative methods to provide predictable and controllable thermal barrier regions for solid breeder blankets. In particular, particle beds have been studied because of their unique thermalhydraulic properties. It has been demonstrated that large variations in thermal conductance can be obtained in the thermal barrier region by external control over the gas pressure and composition in a metallic particle bed. By providing this “active” control mechanism, adjustments in the blanket temperature profiles can be made during operation to accommodate changes in power levels, time-dependent changes in material behavior, and design uncertainties. Data are presented for the effective thermal conductivity of several single-size and binary beds of aluminum, for a range of He and N2 gas pressures, and for a number of different porosities. Data for the wall conductance also are presented. The relative contribution to the temperature drop due to wall conductance is smallest in single-size beds with smaller particles. The variation of wall conductance with pressure is small or non-existent, suggesting that the wall region is dominated by conduction through the solid particle contact points, rather than gas in the Smolukovski zones.

Design windows on free surface flows in the APEX (advanced power extraction) study are derived from the viewpoints of the free surface heat transfer, the adaptation of liquid flows to the topological constraints, and temperature requirements for plasma operation and power conversion efficiency. Within these constraints, the temperature of the free liquid surface facing the plasma is the most critical parameter governing the amount of liquid that evaporates into the plasma chamber. Present analyses show that a 2 cm or a 40 cm thick lithium layer can be established throughout the ARIES-RS reactor using a velocity of 10 m s−1 while operating under the plasma compatible surface temperature. However, like solid metallic walls, the liquid lithium walls require the use of electrical insulators to overcome the MHD drag. As for Flibe free surface flows, the MHD effect caused by interaction with the mean flow is negligible, while a fairly uniform flow of 2 or 45 cm thick can be maintained throughout the reactor based on 3-D hydrodynamics calculations. However, being a low thermally conducting medium, the Flibe surface temperature highly depends on the extent of the turbulent convection. The heat transfer analyses based on the κ–ε model of the turbulence, including MHD effects and various boundary conditions, predict a range of temperatures that may be beyond the plasma compatible temperatures. If indeed the Flibe surface temperature is high relative to the plasma operation limit, further design adjustments will be required to accommodate this deficiency.

In this paper, recent numerical modeling and experiment work for predicting the effective thermal and mechanical properties of solid breeder blanket pebble bed materials is presented. The numerical modeling is based on the micro-mechanics displacement method in conjunction with an iterative process of successive releasing of the contact force of particles. Initial and final packed states for particle assembly and the contact forces for particles have been studied. In addition, a test article has been constructed to measure the thermal stress induced by the thermal expansion of the solid particle and to estimate the characteristic properties of particle materials. Corresponding experiments are carried out with aluminum and Li2ZrO3 pebble beds. Empirical correlations for the moduli of deformation are presented. These experiment data are compared with the numerical modeling results.

This paper presents the first results obtained in a facility constructed at the University of California, Los Angeles to study the interaction of ceramic breeder pebble beds with structural materials in conditions relevant to fusion energy power plant blanket operations. The experiments study the thermo-mechanical performance of lithium meta-titanate oxide (Li2TiO3) pebbles and silicon carbide clad constrained by a low thermal expansion alloy in vacuum and He atmosphere. The results show that large deformations due to induced thermal stresses are present during the first heat cycle but afterward are accommodated by a combination of pebble re-arrangement within the bed and thermally induced creep deformation. Initial results of a numerical simulation of the experiments using a finite element code that includes creep deformation is also presented. Planned operation of the UCLA thermo-mechanics test facility is summarized to conclude the paper.

Numerical simulation of tritium permeation from breeding zones to the coolant in the helium cooled pebble-bed blanket is performed in this paper. 2D and 3D convection–diffusion models are developed to account for the effects of purge stream convection. Incompressible transient Brinkman model with variable permeability is used in flow calculation, and transient diffusion and convection equations are simulated for the tritium permeation analysis. Tritium partial pressure, concentration and permeation flux are evaluated. The influence of convection on permeation is evaluated under different flow conditions.

AbstractTritium behavior in the breeder/coolant plays a crucial role in keeping the tritium loss under an allowable limit and realizing high tritium recovery efficiency. In this paper, progress toward the development of a comprehensive 3D predictive capability is discussed and presented. The sequence of transport processes leading to tritium release includes diffusion and convection through the PbLi, transfer across the liquid/solid interface, diffusion of atomic tritium through the structure, and dissolution-recombination at the solid/gas interface. Numerical simulation of the coupled individual physics phenomena of tritium transport is performed for DCLL/HCLL type breeder blankets under realistic reactor-like conditions in this paper. Tritium concentration and permeation are presented and the MHD effects are evaluated. Preliminary results shows that the MHD velocity profile has the significant effect in preventing tritium permeation due to the higher convection effects near the wall.

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

Beryllium was successfully bonded to a Reduced Activation Ferritic Martensitic (RAFM) steel with a maximum strength of 150 MPa in tension and 168 MPa in shear. These strengths were achieved using Hot Isostatic Pressing (HIP), at temperatures between 700 °C and 750 °C for 2 h and under a pressure of 103 MPa. To obtain these strengths, 10 μm of titanium and 20 μm of copper were deposited on the beryllium substrate prior to HIP bonding. The copper film acted a bonding aid to the RAFM steel, while the titanium acted as a diffusion barrier between the copper and the beryllium, suppressing the formation of brittle intermetallics that are known to compromise mechanical performance. Slow cooling from the peak HIP temperature along with an imposed hold time at 450 °C further enhanced the final mechanical strength of the bond.

In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

Pebble-scale models of the interactions inside packed beds are critical for determining alterations to thermophysical properties in the wake of changes to the packed bed due to cracking, sintering, or creep-deformation of the ceramic pebbles. Simultaneously, the helium purge gas flow through the pebble bed can change; while not specifically playing a role as coolant, it does have an impact on the thermal transport in the volumetrically heated bed. We present numerical tools that are capable of resolving pebble-scale interactions coupled to bed-scale thermofluid flow. The new computational techniques are used to show that maximum temperatures in pebble beds do not increase drastically in spite of the significant amount of cracking induced in our numerical model. Furthermore a complete flow field of helium moving through densely packed spheres is modeled with the lattice-Boltzmann method to reveal the strong effect of slow-moving helium gas on flattening temperature profiles in pebble beds with nuclear heating.

The discrete element method, as currently employed by members of the fusion community, is rooted on the assumption that each pebble is a perfectly elastic material that obeys Hertz's theory for normal interaction. This assumption impacts the magnitude of inter-particle forces predicted by the models. We scrutinize the Hertzian assumption with single-pebble crush experiments with carefully recorded force-displacement responses and compare them to the non-linear forces predicted by a Hertzian pebble with bulk properties reported in literature. We found each pebble generally has a non-linear force response but with varying levels of stiffness that qualitatively matched the curves from Hertz theory. Assuming Hertzian interaction, we backed-out an elastic modulus for each pebble. We define a softening coefficient, κ, as the ratio of the pebble's elastic modulus to the sintered bulk value from literature. After determining the κ value for every pebble in our batch, we discovered a probability distribution for different batches. The distribution is attributed to the varying micro-structure of each pebble. We incorporate the results into our DEM algorithms, distributing κ values at random to pebbles satisfying the probability curves of experiments. DEM simulations of pebble beds in oedometric compression are carried out to determine macroscopic responses of stress–strain, contact force distributions at maximum stress, and a prediction of pebbles crushing at that point. In all cases studied here, the pebble beds with modified Young's modulus had smaller overall contact forces and fewer predicted crushed pebbles.

The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have shown lower values as compared to those of experimentally observed data at 740 °C using an activation energy of 180 kJ/mol. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 °C may cause too much particle overlapping with a contact radius growth beyond 0.65 radius at a later time, when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition.

Experiments on thermomechanics interactions between clad and pebble beds have been performed with overstoichiometric lithium orthosilicate pebbles (pebble diameters between 0.25 and 0.63 mm) at temperatures of 700-800°C. The experimental results show that the thermal deformation of our pebble bed system is nonlinear and when the operating temperature is higher than 600°C, thermal creep deformation is generated. In this paper, constitutive equations of the elastic and creep deformation are derived from the experimental results. Incorporating the effective constitutive equations in finite element method (FEM), numerical investigations presenting the elastic and plastic deformation characteristics of pebble bed system are comparable to the experimental behaviors. In addition, discrete element method (DEM) is underdevelopment to derive constitutive equations for different pebble beds. The preliminary results of DEM show the stress distribution inside the pebble beds at steady or transient states, which helps us to identify the destructive region in a pebble bed system.

Fusion nuclear technology (FNT) research in the United States encompasses many activities and requires expertise and capabilities in many different disciplines. The US Enabling Technology program is divided into several task areas, with aspects of magnet fusion energy (MFE) fusion nuclear technology being addressed mainly in the Plasma Chamber, Neutronics, Safety, Materials, Tritium and Plasma Facing Component Programs. These various programs work together to address key FNT topics, including support for the ITER basic machine and the ITER Test Blanket Module, support for domestic plasma experiments, and development of DEMO relevant material and technological systems for blankets, shields, and plasma facing components. In addition, two inertial fusion energy (IFE) research programs conducting FNT-related research for IFE are also described. While it is difficult to describe all these activities in adequate detail, this paper gives an overview of critical FNT activities.

This paper presents design and analysis results regarding key aspects of the neutronics for a standard aspect ratio fusion nuclear science facility (FNSF). Optimization of the inboard design is based on maximizing tritium production and minimizing radiation damage to the ohmic heating coil (OHC). The calculations show that the outboard tritium breeding blanket alone is not enough for achieving tritium self-sufficiency in the FNSF. Options to enhance tritium production in FNSF include the use of either a thin, specifically designed tritium breeding blanket or the use of only a multiplier in the highly space-constrained inboard region. Trade-offs between breeding, shielding to protect the magnets, and materials for the inboard region shows that an inboard thickness of 50 cm is enough if a ceramic insulator is used. If an organic insulator is preferred, the inboard thickness has to be increased. Radiation-induced increase in the electrical resistivity of the copper has also been studied and its impact on the aforementioned optimization process for the blanket/shield thickness/material choice has been accounted for.

AbstractThis work has developed FEM models of ceramic breeder pebble beds and applied them to two categories of blanket design (edge-on and layer configurations) to predict the thermomechanical behavior of a pebble bed under ITER pulsed operating condition. To explore the pebble bed/structural wall separation phenomenon, a thermomechanical contact is considered using contact elements meshed along pebble/structure interface. The pebble bed/wall dynamic contact/separation process has been simulated, and the gap distance distribution and variation have been analyzed and presented. Pebble bed/wall separation occurs during the plasma-off period and varies with both location and time. A maximal radial gap of 0.64mm is found for an edge-on configuration after the 1st ITER cycle within the range of studied parameters. For the layer configuration, a poloidal gap of 1.99mm, larger than the pebble diameter, is found. The generated gap can cause the even large rearrangement of pebbles and result in a disturbed packing during further cycling. Consequently, a design solution is suggested to mitigate this situation.

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li2O, Li4SiO4 and Li2TiO3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2–3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation.

Breeding blankets are key components of fusion power plants and determine to a large degree their attractiveness. There is a large range of possible blanket concepts, characterized by breeding material, structural material, coolant, and geometrical arrangement. On one extreme side there are concepts requiring only a modest extrapolation of the present technology, but with limited attractiveness. On the other side there are concepts with the potential for very attractive plants, but involving a considerable development risk. Therefore, the selection of blanket concepts depends on the overall strategy for fusion power plant development. To ensure sufficient support over the long development time, it has to be shown in a credible way, that (a) there are feasible blanket concepts for fusion power plants which can be developed with a high confidence of success, and (b) the final product will be an attractive power plant which can be operated safely, with minimum impact on the environment, and at an acceptable cost of electricity. This requires an evaluation of the entire range of blanket concepts, starting from near term solutions and reaching up to very advanced designs with a potential for exceptionally high performance. Typical examples of candidate blanket concepts are compared in this study. The blanket development strategy developed in the EU is described as an example.

Numerical simulation of multi-phase flow is essential to evaluate the feasibility of a liquid protection scheme for the power plant chamber. The level set method is one of the best methods for computing and analyzing the motion of interface among the multi-phase flow. This paper presents a general formula for the second-order projection method combined with the level set method to simulate unsteady incompressible multi-phase flow with/out phase change flow encountered in fusion science and engineering. The third-order ENO scheme and second-order semi-implicit Crank–Nicholson scheme is used to update the convective and diffusion term. The numerical results show this method can handle the complex deformation of the interface and the effect of liquid–vapor phase change will be included in the future work.

(2019). Impact of Outer Fuel Cycle Tritium Transport on Initial Start-Up Inventory for Next Fusion Devices. Fusion Science and Technology: Vol. 75, Selected papers from the 23rd Topical Meeting on the Technology of Fusion Energy, pp. 1037-1045.

AbstractAbstractHCPB (Helium Cooled Pebbles Bed) Test Blanket Model (TBM) is based on solid breeder. Tritium has to be extracted using Helium purge gas flowing through breeder pebbles bed. He-stream must be cooled down and processed and all these steps are carried out in TES (Tritium Extraction System). A modeling work has been performed to study the behavior of the TES of the HCPB TBM and for Molecular Sieve 5A as adsorbent material. The result is the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of two main tritium gaseous species (H2 and HT).

Laminar heat transfer in self-cooled liquid metal blankets can be enhanced by increasing the aspect ratio of the ducts. To determine the potential benefits of elongated rectangular ducts, numerical simulations of MHD fully-developed flow and developing heat transfer were performed.Results show that as the aspect ratio increases (i.e., the ratio of the side wall to Hartmann wall length), the peak velocity and side layer flow quantity increase, which leads to enhancement of the average heat transfer coefficient along the side layer. The pressure gradient decreases with increasing elongation, providing an added benefit.However, results of the heat transfer analysis also indicate that the non-uniformity along the heated wall and the peak wall temperature both increase as the aspect ratio increases, due to smaller velocities in the corners and near the interface between the side layer and the core. The net benefit to reactor blanket design is therefore uncertain, because designs are usually constrained by the peak structure temperature. At fixed velocity, elongated ducts always have higher peak temperatures. However, the reduction in pressure gradient allows the designer to increase the average velocity, which improves thermal performance due to lower bulk temperature rise as well as higher wall heat transfer coefficient. Calculations show that peak temperatures can be reduced relative to the square duct case with lower pressure gradient by optimizing the velocity.Elongated ducts may suffer from larger pressure stresses due to geometric factors. Thermal stresses are also likely to increase, owing to the increased thermal gradients in the walls. Overall, it is difficult to guarantee that elongation will provide improved performance without a more detailed design analysis.

The use of lithium ceramic pebble beds in the design of blankets for fusion reactors makes the mechanical and thermal properties of ceramic pebble beds key issues to be investigated. In the last years measurements of the effective thermal conductivity of lithium oxide, lithium orthosilicate, lithium metatitanate and lithium metazirconate pebble beds were performed at the Research Centre of Karlsruhe (Germany), at University of California (USA) and at JAERI (Japan). The measurements performed at FZK and at UCLA were based on steady-state methods, at JAERI a transient method (Hot wire method) was used instead. In the present paper, the results from these experiments are compared and discussed.

The paper summarizes results of experimental and theoretical studies related to the flow of liquids with a free surface and poor electrical and thermal conductivity, such as molten salts, under conditions relevant to fusion energy systems. These results have been obtained over last several years when developing the liquid wall concept as a part of the APEX project [M.A. Abdou, The APEX TEAM, On the exploration of innovative concepts for fusion chamber technology, Fusion Eng. Des. 54 (2001) 181–247]. As a theoretical tool a modified K–ɛ model of turbulence coupled with the Navier–Stokes equations written in the thin-shear-layer approximation is used for studying wavy, turbulent flows in a spanwise magnetic field. The experimental part covers current results for supercritical flows in regimes transitional from “weak” to “strong” turbulence, which are expected to occur in the reference liquid wall flows. The paper also describes on-going work on novel schemes of heat transfer promotion and current directions for direct numerical simulation.

The successful design and development of a complex system, like the ITER test blanket module (TBM) warrants the need of extensive computer aided engineering (CAE) activities. In this light, a sophisticated numerical flow solver (‘SC/Tetra’ by CRADLE ® ), with a robust CAD interface, has been used to develop and evaluate helium coolant flow schemes for a solid breeder test blanket module design currently proposed by the US for testing in ITER. The traits of a particular cooling strategy for the TBM, namely the exit temperature of coolant, overall pressure drop, uniformity of temperature in the structure, robustness against transients, etc. can only be predicted by carrying out a complete three dimensional thermal-fluid analysis of the system in its entirety including all the structural and fluid components. The primary objective of this paper is to introduce the procedure for carrying out complex thermo-fluid analysis using the complete three dimensional CAD models of the TBM to evaluate the performance of TBM cooling schemes and to illustrate the way in which the results from these analyses can be useful towards a systematic design of an effective cooling solution for the test blanket module. © 2007 Elsevier B.V. All rights reserved.

A virtual test blanket module (VTBM) has been envisioned as a utility to aid in streamlining and optimizing the US ITER TBM design effort by providing an integrated multi-code, multi-physics modeling environment. Within this effort, an integrated simulation approach is being developed for TBM design calculations and performance evaluation. Particularly, integrated thermo-fluid/thermal-stress analysis is important for enabling TBM design and performance calculations. In this paper, procedures involved in transient coupled thermo-fluid/thermal-stress analysis are investigated. The established procedure is applied to study the impact of pulsed operational phenomenon on the thermal-stress response of the TBM first wall. A two-way coupling between the thermal strain and temperature field is also studied, in the context of a change in thermal conductivity of the beryllium pebble bed in a solid breeder blanket TBM due to thermal strain. The temperature field determines the thermal strain in beryllium, which in turn changes the temperature field. Iterative thermo-fluid/thermal strain calculations have been applied to both steady-state and pulsed operation conditions. All calculations have been carried out in three dimensions with representative MCAD models, including all the TBM components in their entirety.

DETROIT is an open-source vehicle-agnostic end-to-end framework for vehicular data collection, translation and sharing that facilitates the rapid development of automotive apps. With vehicles becoming increasingly connected, unlocking sheer amounts of data from the in-vehicle network (IVN) can accelerate the development of many useful apps. Unlike existing commercial and academic solutions that can only access a restricted set of standardized emission-related sensor data and lack feasible data accessibility by third-party developers, DETROIT offers a convenient interface to develop apps which can access a broad range of powertrain-related sensors and car-body events thanks to crowd-sourcing vehicular translation tables by fully automated CAN bus reverse-engineering. DETROIT is developed with the objectives of simplicity, scalability, privacy and liability. To the best of our knowledge, this is the first end-to-end framework consisting of a frontend, backend and a developer portal to cover vehicular data collection, translation and sharing with app developers. Besides an extensive framework benchmark to show the light resource overhead and feasibility of DETROIT, we also have evaluated it by reimplementing two existing mobility apps from academia. Developers have reported that DETROIT offers high sensor fidelity, enhanced application flexibility, as well as low implementation complexity.

The critical technical issues, evaluation and comparison of two inertial fusion enery (IFE) reactor design concepts developed in the Prometheus studies are presented. The objectives of this study were (1) to identify and characterize the critical issues and the R&D required to resolve them, and (2) to establish a sound basis for future IFE technical and programmatic decisions by evaluating and comparing the different design concepts. Each critical issue contains several key physics and engineering issues associated with the major reactor components and impacts key aspects of feasibility, safety, and economic potential of IFE reactors. Genetic critical issues center around: (1) demonstration of moderate gain at low driver energy, (2) feasibility of direct drive targets, (3) feasibility of indirect drive targets for heavy ions, (4) feasibility of indirect drive target for laers, (5) cost reduction strategies for heavy ion drivers, (6) domonstration og higher overall laser driver efficiency, (7) tritium self-sufficiency in IFE reactors, (8) cavity clearing at IFE pulse repetition rates, (9) performance, reliability and lifetime of final laser optics, (10) viability of liquid metal film for first wall protection, (11) fabricability, reliability and lifetime of SiC composite structures, (12) validation of radiation shielding requirements, design tools, and nuclear data, (13) reliability and lifetime of laser and heavy ion drivers, (14) demonstration of large-scale non-linear optical laser driver architecture, (15) demonstration of cost effective KrF amplifiers, and (16) demonstration of low cost, high volume target production techniques. Quantitative evaluation and comparison of the two design options have been made with special focus on physics feasibility, engineering feasibility, economics, safety and environment, and research and development (R&D) requirements. Two key conclusions are made based on the overall evaluation analysis: (1) The heavy-ion driven reactors appear to have an overall advantage over laser-driven reactors; and (2) However, the differences in scores are not large and future results of R&D could change the overall ranking of the two IFE concepts.

The mechanical behavior and properties of particle beds relevant to fusion ceramic breeder blanket designs are addressed through a combination of experimental and analytical studies. A series of uniaxial compression tests are performed in a INSTRON hydraulic-press test facility in which compressive loads are applied to Li2ZrO3 and aluminum packed beds. The experimental data shows that the mechanical properties such as the bed effective modulus are stress dependent and consistent with the analytical model predictions. Nevertheless, the experimental data shows that the bed effective modulus is approximately two orders of magnitude lower than that of the solid material.

The usefulness of the tritium-producing blanket in the International Thermonuclear Experimental Reactor (ITER) to the fusion research and development program can be maximized by selecting design parameters, features, and options that are reactor relevant without significantly increasing the risk in key areas such as device safety and operational reliability. For that reason, a helium-cooled solid breeder (SB) blanket is proposed since it combines the operation of the SB at high reactor-relevant temperatures with the operation of helium at moderate temperature and pressure to minimize risk. Results of the analysis done for this blanket concept indicate that it is very attractive. It can achieve a high tritium breeding ratio without breeding in the space-limited inboard region. It offers important safety features, including the use of inert gas with no chemical reaction or corrosion, low activation SB, and multiple containment of tritium. The concept provides great operational flexibility to accommodate changes in ITER operating parameters, such as power level, and to optimize the operating temperature of the structure. A novel and practical concept is proposed for the thermal resistance gap between the coolant and SB to allow their operating temperatures to be optimized.

Westinghouse Electric and the Electric Power Research Institute (EPRI) are engaged in a joint program to develop a 300 MVA generator with a superconducting field winding to demonstrate the increased efficiency, reliability and stability of such units as base load generators. Considerations such as conductor support, transient heating, electrical damping, magnetic shielding and dynamic stability demand creative structural and thermal designs. Unique structural designs are required to support the rotor damper shield and to limit transient torsional loads under fault operation. The rotor cooling system is designed to remove the shielding losses induced by fault operation as well as those encountered during steady state operation. So effective is the cooling system that the winding can undergo resistive transition over a large portion and still recover the superconducting state. The conceptual design of the 300 MVA rotor is reviewed. The structural design and calculated performance of the cooling system during steady state and fault operation are presented. The opportunities presented by new materials and manufacturing technology are summarized. Future developments, and model testing anticipated during the program are reviewed.

The design and operating characteristics of the ALICE (Advanced Liquid Ionized Condensation Experiment) facility at UCLA are here presented. The goal of this vapor condensation experiment is to rapidly generate an IFE prototypical post-shot vapor density in a control volume using characteristic liquid chamber material (flibe, Li2BeF4), and investigate the condensation rates for the proposed schemes. This experimental goal is achieved by: 1) a pulsed electrothermal plasma source that simulates the pellet explosion for rapid vapor generation and 2) an expansion chamber that represents the IFE liquid chamber. This paper reports also on the construction and operation of a furnace for flibe casting. Melting and handling procedures connected with the use of flibe are also discussed. The first flibe liner has been inserted in the plasma source. Results from the first low energy experiments are showed.

An assessment of the issues on using flibe for fusion applications has been made. It is concluded that sufficient tritium breeding can be achieved for a flibe blanket, especially if a few cm of Be is include in the blanket design. A key issue is the control of the transmutation products such as TF and F2. A REDOX (Reducing-Oxidation) reaction has to be demonstrated which is compatible to the blanket design. Also, MHD may have strong impact on heat transfer if the flow is perpendicular to the magnetic field. The issues associated with the REDOX reaction and the MHD issues have to be resolved by both experimental program and numerical solutions.

The use of thin, fast flowing liquid metal films as the divertor surface is a very attractive option for effective particle pumping and surface heat removal. The major "show-stoppers" for such a design concept are the magneto-hydrodynamic (MHD) effects associated with the flow, due to the presence of strong magnetic fields that vary spatially and temporally. This paper reports the preliminary experimental findings on the liquid metal free surface film flows under magnetic field conditions, similar to those experienced at the outboard divertor region of the NSTX machine at PPPL. The main goal of the study is to understand the MHD features associated with the flow, with emphasis on the variation of film thickness in the stream-wise and span-wise direction, as well as film surface stability. This study forms a part of a larger research effort at UCLA, directed towards providing qualitative and quantitative data on the liquid metal film flow behavior and identifying design constraints for the implementation of a liquid surface divertor module in NSTX.

The long-range goal of the Z-pinch IFE program is to produce an economically-attractive power plant using high-yield Z-pinch-driven targets (~3 GJ) with low rep-rate per chamber (~0.1 Hz). He present mainline choice for a Z-pinch IFE power plant uses an LTD (Linear Transformer Driver) repetitive pulsed power driver, a Recyclable Transmission Line (RTL), a dynamic hohlraum Z-pinch-driven target, and a thick-liquid wall chamber. The RTL connects the pulsed power driver directly to the Z-pinch-driven target, and is made from frozen coolant or a material that is easily separable from the coolant (such as low activation ferritic steel). The RTL is destroyed by the fusion explosion, but the RTL materials are recycled, and a new RTL is inserted on each shot. The RTL concept eliminates the problems of a final optic, high-speed target injection, and pointing and tracking N beams (N~100). Instead, the RTL concept must be shown to be feasible and economically attractive. Results of Z-pinch IFE studies over the last three years are discussed, including RTL experiments at the 10 MA level on Saturn, RTL structural studies, RTL manufacturing/cost studies, RTL activation analysis, power plant studies, high-yield IFE target studies, etc. Recent funding by a U.S. Congressional initiative of $4M for FY04 is supporting research on (1) RTLs, (2) repetitive pulsed power drivers, (3) shock mitigation [because of the high yield targets], (4) planning for a proof-of-principle full RTL cycle demonstration [with a 1 MA, 1 MV, 100 ns, 0.1 Hz driver], (5) IFE target studies for multi-GJ yield targets, and (6) Z-pinch IFE power plant engineering and technology development.

Copper has been investigated as a potential interlayer material for diffusion bonds between beryllium and Reduced Activation Ferritic/Martensitic (RAFM) steel. Utilizing Hot Isostatic Pressing (HIP), copper was directly bonded to a RAFM steel, F82H, at 650 °C, 700 °C, 750 °C, 800 °C and 850 °C, under 103 MPa for 2 h. Interdiffusion across the bonded interface was limited to 1 μm or less, even at the highest HIP’ing temperature. Through mechanical testing it was found that samples HIP’ed at 750 °C and above remain bonded up to 211 MPa under tensile loading, at which point ductile failure occurred in the bulk copper. As titanium will be used as a barrier layer to prevent the formation of brittle Be/Cu intermetallics, additional annealing studies were performed on copper samples coated with a titanium thin film to study Ti/Cu interdiffusion characteristics. Samples were heated to temperatures between 650 °C and 850 °C for 2 h in order to mimic the range of likely HIP temperatures. A correlation was drawn between HIP temperature and diffusion depth for use in determining the minimum Ti film thickness necessary to block diffusion in the Be/F82H joint.

This work introduces a unified cap model, implemented in ANSYS, to the field of ceramic breeder pebble bed thermomechanics. The developed continuum numerical model has been validated and applied to predict the temperature/stress-dependent creep phenomena and stress relaxation of Li <inf xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</inf> SiO <inf xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</inf> pebble bed, and the temperature/stress distributions of a DEMO-like breeder unit. In addition, considering the essential of experimental data for deriving the constitutive equations and for model benchmarking, we designed and conducted oedometric compression, creep and stress relaxation tests.

In the APEX study, one of the tasks focuses on the exploration and identification of the attractive options and issues for flowing liquid lithium walls in the NSTX device. In addition to constraints imposed by the machine, the operating conditions of the flowing liquid walls along the center stack and divertor areas are guided by MHD and heat removal requirements. In this paper, we present important MHD and heat removal issues and analysis for the proposed free surface lithium flows under NSTX conditions. It is shown that of all MHD effects, the one caused by the normal magnetic field is the most important. The flow over the center stack area is not affected by MHD interaction significantly, whereas flow over the inboard divertor undergoes strong MHD drag resulting in flow thickening by several times. The flow over the outboard divertor is essentially stopped. The analysis shows that a flow with an inlet velocity of 2 m/s and film thickness of about 4 mm can be established to provide surface temperature less than 400° C for the center stack under a projected NSTX total heating power of 10 MW operation.

Significant progress has been made on addressing critical issues for inertial fusion energy (IFE) chambers for heavy-ion, laser and Z-pinch drivers. A variety of chamber concepts are being investigated including drywall (currently favored for laser IFE), wetted-wall (applicable to both laser and ion drivers), and thick-liquid-wall (favored by heavy ion and z-pinch drivers). Recent progress and remaining challenges in developing IFE chambers are reviewed.

Several liquid-metal blanket design concepts employ rectangular ducts. If the duct walls can be accurately aligned with the magnetic field, substantial improvement in heat transfer may result due to the favorable characteristics of MHD (magnetohydrodynamic) side layers. A detailed analysis of the velocity and temperature profiles has been carried out under fully developed flow conditions for reactor-relevant Hartmann numbers in the range of 5000 and a wall conductance ration of 0.01. The results show that substantial changes in the wall temperatures can result from changes of only a few degrees in the magnetic field angle. An assessment of the possible sources of misalignment in an operating power-reactor blanket suggests that it will be very difficult to maintain sufficiently good alignment to fully exploit side layers as a means to enhance heat transfer.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

To evaluate beryllium-to-copper joining techniques for potential use by US manufacturers in making first wall components for International Thermonuclear Experimental Reactor (ITER), we tested two mock-ups with S65C beryllium (Be) tiles Hot Isostatic Pressing (HIP) bonded to CuCrZr heat sinks. Under the aegis of the US ITER Project Office, Sandia prepared the mock-ups working with industrial vendors and performed high heat flux testing at Sandia's Plasma Material Test Facility (PMTF) to ascertain the robustness of the Be/Cu joints to 1000 thermal fatigue cycles at a heat flux level of 1.5 MW/m2. Thermal stress analysis provided insight into choosing the heat flux and flow conditions required for accelerated fatigue testing at 1000 cycles and 1.5 MW/m2 that is comparable to the 12,000 cycles and 0.875 MW/m2 required for the ITER First Wall Qualification Mock-ups. Each mock-up had three Be tiles, 35.5 mm square and 10 mm thick, bonded to a CuCrZr heat sink 134.5 mm × 36 mm × 25 mm with a single bored 12.7 mm (dia.) cooling channel. The bonding techniques included various interlayer metallizations and HIPping at 100 MPa pressure and temperature of 580 or 560 °C for 2 h. Each tile had a thermocouple (TC) in the center 1 mm below the Be/Cu interface. The test arrangement allowed for both mock-ups to be tested at the same time with alternate heating and cooling cycles of equal duration of 30 s. A total power of 12.7 kW was absorbed by the heated area of 4000 mm2 during the on-cycle. The mock-up was cooled by water at 2.3 m/s (0.27 kg/s), 1 MPa and 20 °C inlet temperature. These operating conditions did not permit the mock-ups to cool down to their initial temperature state during the off-cycle. Both mock-ups survived 1000 cycles with no significant changes. The temperature of the top surface on each reached 254 °C; while the center TCs reached 136 and 139 °C, respectively. Despite localized changes observed in the surface emissivity, the corrected temperature distributions on the surfaces varied by only a few degrees and did not change significantly during testing. We characterized the Be/Cu joint by ultrasonic testing before and after testing and sectioned the mock-ups for further evaluation. This article discusses the fabrication techniques, the results of the ultrasonic and thermal testing, and the time-dependent performance insights from computational fluid dynamics.

Ceramic breeder pebble beds undergo complex thermally-induced stress build-up and relaxation processes during reactor operations due to the pebble bed thermal expansion and creep deformation. Understanding such processes can facilitate the evaluation of a solid breeder performance, including bed stress/strain equilibrium status, which will guide the design of stable blanket operation and assessment of lifetime. The efforts of this study cover both experimental testing and numerical modeling for this purpose. Measured stresses in pebble beds show a decreasing trend with thermal cycles, until ultimately reaching a saturated state. This stress relaxation is mainly caused by the combined effect of bed plastic rearrangement and accumulation of creep deformation under compressive stresses and high temperatures. As bed stress is reduced, the creep deformation becomes less significant and further cyclic operation would not alter the pebble bed mechanical state. To validate the thermally-induced stress and its variation with cycles, experiments of thermal stress measurement have been designed and conducted for pebble beds heated by both continuous and pulsed power sources. Moreover, the effects of mechanical pre-compaction were investigated with emphasis on understanding the relationship between the bed stress-state evolution and maintaining adequate levels of thermal contact between the pebbles and the coolant structure. The results of this study presents valuable data to serve as a basis for validation of the most recent pebble bed numerical models.

Analyses were performed of the effect of Harmann layers and side layers on heat transfer in laminar MHD flow in ducts and the dependence on the magnitude of the Hartmann number. Analytical and numerical results are presented for both fully developed and thermally developing cases. The presence of side layers in a rectangular duct usually increases the heat transfer coefficient on the side layer walls and decreases the heat transfer coefficient on the other two walls. For ducts with uniform thickness and conductivity on all walls, the studies show that a duct with higher conductance ratio gives higher average Nusselt number on the side wall. However, this behavior depends on the combination of Hartmann number and the conductance ratio. The heat generation inside the duct enhances the heat transfer coefficient.