D.-K. Sze

U.S. PLANS AND STRATEGY FOR ITER BLANKET TESTING

Abstract 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.

Configuration and engineering design of the ARIES-RS tokamak power plant

ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems.

Nuclear aspects of molten salt blankets

Abstract Nuclear aspects of candidate molten salts, namely a mixture of LiF and BeF2 (FLIBE) and a mixture of LiF, NaF, and BeF2 (FLINABE), were investigated for application as blanket coolants in tokamak fusion power plants. Tritium breeding, blanket energy multiplication, and neutron transmutation of these salts were assessed. Neutron activation of FLIBE and FLINABE was evaluated and site-boundary dose due to a worst-case loss of vacuum accident was estimated. Formation of F2, TF and T2 during power plant operation was analyzed and issues relevant to corrosion of structural materials due to the fluorine and fluoride species was assessed. Mechanism to control the corrosion of structural materials due to TF has been identified. Depletion of LiF, BeF2, and NaF in the salts was calculated and quantities of the make-up fluorides to be added into the salts were estimated.

Effect of impurities on the erosion behavior of beryllium under steady-state deuterium plasma bombardment

Abstract As a candidate plasma-facing material for ITER (International Thermonuclear Experimental Reactor), beryllium has been evaluated with respect to deuterium plasma erosion characteristics using the PISCES-B Mod facility at ion fluxes around 5 × 10 21 ions s −1 m 2 in the ion bombarding energy range from 100 to 300 eV. It is found that at elevated temperatures beryllium tends to be contaminated with trace amounts of plasma impurities such as carbon, hydrocarbon, oxygen and nitrogen even under the energetic deuterium plasma bombardment. Among these plasma impurities, carbon and hydrocarbon are observed to form thin films. This carbon deposition effect has made it difficult to interpret the weight loss data in evaluating the erosion yield. A first-order model has been developed to account for the dynamics of this carbon deposition process and it has been suggested that some surface chemistry effect plays an important role. This surface chemistry effect has been experimentally proved from the observation that carbon deposition can be avoided only at room temperature. Without impurity deposition, the beryllium surface after plasma bombardment is found to be cleaner in oxygen content and smoother in surface topography than the as-received material. Interestingly, however, even without carbon deposition, the erosion yield data still tend to agree with beryllium oxide data.

FLIBE-D2 permeation experiment and analysis

Abstract Experiment of D2 permeation through Ni facing with purified Flibe is being carried out under the Japan-US joint research project (JUPITER-II). The experiment is proceeding in the following phases; (i) fabrication and assembly of a dual-probe permeation apparatus, (ii) a single-probe Ni/D2 permeation experiment without Flibe, (iii) a dual-probe Ni/D2 permeation experiment without Flibe, (iv) Flibe chemical purification by HF/H2 gas bubbling, (v) physical purification by Flibe transport through a porous Ni filter, (vi) Ni/Flibe/D2 permeation experiment, and (vii) Ni/Flibe/HT permeation experiment. The present paper describes results of the single and dual Ni/D2 permeation experiments in detail.

Jupiter-ii flibe tritium/chemistry and safety experimental program

ABSTRACT The second Japan/US Program on Irradiation Tests for Fusion Research (JUPITER-II) began on April 1, 2001. Part of the collaborative research centers on studies of the molten salt 2LiF-BeF2 (also known as Flibe) for fusion applications. Flibe has been proposed as a self-cooled breeder in both magnetic and inertial fusion power plant designs over the last twenty years. The key feasibility issues associated with the use of Flibe are the corrosion of structural material by the molten salt, tritium control in the molten salt blanket system, and safe handling practices and releases from Flibe during an accidental spill. An overview of the experimental program to address the key feasibility issues is presented.

MIDTERM SUMMARY OF JAPAN-US FUSION COOPERATION PROGRAM TITAN

Abstract Japan-US cooperation program TITAN (Tritium, Irradiation and Thermofluid for America and Nippon) started in April 2007 as 6-year project. This is the summary report at the midterm of the project. Historical overview of the Japan-US cooperation programs and direction of the TITAN project in its second half are presented in addition to the technical highlights.

Safety implications of CLiFF liquid metal first walls

In this article we explore some of the safety issues associated with the Advanced Power Extraction (APEX) liquid metal first wall (FW) concept called the Convective Liquid Flow First Wall (CLiFF) design. In particular, we examine the chemical reactivity of and site boundary dose from three liquid metals being proposed for the CLiFF FW during a worst case confinement-boundary-bypass accident. The liquid metals considered are: lithium, tin, and gallium. The accident analyzed is a loss-of-vacuum accident (LOVA), during which air enters the CLiFF vacuum vessel (VV) from an adjoining room through a connecting diagnostic port or plasma-heating duct. The resulting lithium fire was analyzed and the energy released from this fire was found to be manageable for the CLiFF blanket design presently under consideration. The estimated dose at the site boundary is less than the no-evacuation limit of 10 mSv for ground level releases if plant isolation occurs within several hours.

Engineering options for the US Fusion Demo

Through its successful operation, the US Fusion Demo must be sufficiently convincing that a utility or independent power producer will choose to purchase one as its next electric generating plant. A fusion power plant which is limited to the use of currently-proven technologies is unlikely to be sufficiently attractive to a utility unless fuel shortages and regulatory restrictions are far more crippling to competing energy sources than currently anticipated. In that case, the task of choosing an appropriate set of engineering technologies today involves trade-offs between attractiveness and technical risk. The design space for an attractive tokamak fusion power core is not unlimited; previous studies have shown that advanced low-activation ferritic steel, vanadium alloy, or SiC/SiC composites are the only candidates we have for the primary in-vessel structural material. An assessment of engineering design options has been performed using these three materials and the associated in-vessel component designs which are compatible with them.

Heat flux limits on the plasma-facing components for a commercial fusion reactor

In this work, the heat flux limits of conventional plasma-facing components (PFC) were examined. The limits are based on maximum allowable temperature and stress levels in the structures. The substrate materials considered were V, SiC composite and HT-9. The use of Cu also was considered. However, low temperature limits, activation and very limited radiation damage lifetime, make the using of Cu in a commercial power plant unattractive. With selected heat transfer enhancement, the heat flux allowable is about 5.3 MW/m/sup 2/ for lithium-cooled V-alloy, 2.7 MW/m/sup 2/ for helium-cooled SiC composite, and 2.7 MW/m/sup 2/ for helium/water-cooled HT-9. Compared with the maximum heat flux attainable with Cu and cold water (13.4 MW/m/sup 2/), acceptable power plant materials place severe restrictions on heat removal. The thermal conductivity of SiC composite at 1000/spl deg/C and after irradiation is a factor of several lowered than the value we used. This indicates a need to examine the heat transfer problems associated with PFC, in terms of material development and enhancement in heat transfer. Physics regimes which can provide low peak and average heat flux should be pursued.