M.Z. Youssef

On the exploration of innovative concepts for fusion chamber technology

Abstract 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/m 2 and surface heat flux >2 MW/m 2 , (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.

Deuterium-tritium fuel self-sufficiency in fusion reactors

Conditions necessary to achieve deuterium-tritium fuel self-sufficiency in fusion reactors are derived through extensive modeling and calculations of the required and achievable tritium breeding ratios as functions of the many reactor parameters and candidate design concepts. It is found that the excess margin in the breeding potential is not sufficient to cover all present uncertainties. Thus, the goal of attaining fuel self-sufficiency significantly restricts the allowable parameter space and design concepts. For example, the required breeding ratio can be reduced by (A) attaining high tritium fractional burnup, >5%, in the plasma, (B) achieving very high reliability, >99%, and very short times, <1 day, to fix failures in the tritium processing system, and (C) ensuring that nonradioactive decay losses from all subsystems are extremely low, e.g., <0.1% for the plasma exhaust processing system. The uncertainties due to nuclear data and calculational methods are found to be significant, but they are substantially smaller than those due to uncertainties in system definition.

Analyses and Intercomparison for Phase I Fusion Integral Experiments at the FNS Facility

Phase I integral experiments of U.S./JAERI Collaborative Program on Fusion Breeder Neutronics which are carried out at the Fusion Neutronics Source (FNS) facility at JAERI ranged from D-T neutron source characterization experiments, tritium production rate (TPR) measurements in a reference Li/sub 2/0 assembly, first wall experiments with and without coolant simulation and beryllium neutron multiplier experiments in various configurations. Both U.S. and Japan have independently analyzed these experiments using their own data base and codes. Analytical predictions obtained by both countries are compared to measured values. Results of this intercomparison is presented in this paper.

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

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

Analysis of neutronics parameters measured in Phase-II experiments of the JAERI/US collaborative program on fusion blanket neutronics. Part I: Source characteristics and reaction rate distributions

Fusion blanket neutronics parameters measured in the Phase II assembly have been analyzed at both JAERI and the US. Both parties have analyzed the experiments independently by using different nuclear data and calculational methods based on 3-D Monte Carlo and 2-D Sn codes. This part includes the results of the analysis on the source characteristics in the assembly and the reaction rate distributions in the test zone consisting of Li2O with and without a beryllium multiplier. The source characterization has been made by measuring the neutron spectrum and various reaction rates. These reactions include 58Ni(n,2n), 58Ni(n,p), 27Al(n,α), 93Nb(n,2n), 197Au(n,2n), and 197Au(n,γ). The ratios of calculated to measured values are compared among both countries and the different nuclear data used. Considerable discrepancies have been observed for the 58Ni(n,2n), 58Ni(n,p) and 93Nb(n,2n) reactions depending on which nuclear data was used, while good agreement is seen for the reactions 197Au(n,2n) and 27Al(n,α). The distributions of these reaction rates in the test zone have also been analyzed to examine the prediction accuracy of neutronics parameters in a breeder zone. Using the recently measured cross sections at the FNS resulted in a significant reduction in the discrepancies for most reaction rates.

Phase III experimental results of JAERI/USDOE collaborative program on fusion neutronics

Abstract A pseudo-line D–T neutron source has been developed with new experimental techniques. This line source was applied in sophisticated neutronics experiments for an annular blanket arrangement simulating the tokamak geometry, as a new series in the JAERI/USDOE collaborative experimental program on fusion neutronics. The source characteristics of the present line source and the measurements for an annular assembly are described. The discussion on the experimental results focuses on the tritium production rate measured in an annular blanket and comparisons were made with the previous point source experiment, and also between the annular blankets with and without an armor reflector of graphite.

A Line D-T Neutron Source Facility for Annular Blanket Experiment: Phase III of the JAERI/USDOE Collaborative Program on Fusion Neutronics

AbstractA new experimental arrangement using pseudo-line D-T neutron source has been developed to investigate the neutronics performance in the fusion blanket and related components. The arrangement is simple in construction and gives well-simulated fusion reactor environment combined with the flexibility in testing various factors. In examining the deficiencies in nuclear data, calculation methods or modeling, it provides a powerful means either for basic benchmark experiments in clear geometry or design-supporting experiments with complexity of the recent blanket design. The features of the line source facility and the experimental scope are described.

Neutronics investigation of advanced self-cooled liquid blanket systems in the helical reactor

Neutronics investigations have been conducted in the design activity of the helical-type reactor Force Free Helical Reactor (FFHR2) adopting Flibe-cooled and Li-cooled advanced liquid blanket systems. In this study, comprehensive investigations and geometry modifications related to the tritium breeding ratios (TBRs), neutron shielding performance and neutron wall loading on the first walls in FFHR2 have been performed by improving the three-dimensional (3D) neutronics calculation system developed for non-axisymmetric helical designs. The total TBRs obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. However, it appeared that the most important neutronics issue in the present helical blanket configuration was suppression of neutron streaming through the divertor pumping areas and reflection from support structures for protection of poloidal and helical coils. Evaluation of neutron wall loading on the first walls indicated that the peaking factor would be moderated as low as 1.2 by the toroidal and helical effect of the helical-shaped plasma distribution in the helical reactor.

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

Abstract As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li2BeF4 and the low melting point molten salts such as LiBeF3 and LiNaBeF4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiCf/SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R&D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

Measurement and Analysis of Nuclear Heat Depositions in Structural Materials Induced by D-T Neutrons

AbstractNuclear heat deposition rates in ten different materials, Li2CO3, Graphite, Ti, Ni, Zr, Nb, Mo, Sn, Pb and W, subjected in D-T neutrons have been measured by a microcalorimetric technique in the frame work of JAERI/USDOE collaborative program on fusion neutronics. A great improvement in accuracy of experimental data was achieved by introducing a high sensitivity voltmeter and applying constant current on the thermal sensors. The measured heating rates were compared with calculations to verify the adequacy of the currently available data base relevant to the nuclear heating process. In general, calculations with data of JENDL-3 and ENDL-85 libraries gave excellent agreements with experiments for all materials except Zr. The calculation with the MBCCS suffered large discrepancy from measurement.