International Thermonuclear Engineering Reactor Research Articles

Fueling of ITER-Scale Fusion Plasmas

Fueling system functions for the International Thermonuclear Engineering Reactor (ITER) and similar scale devices are to provide hydrogenic fuel to maintain the plasma density profile for a specified fusion power, to replace the deuterium-tritium (D-T) ions consumed in the fusion reaction, to establish a density gradient for plasma particle (especially helium ash) flow to the edge, and also to supply hydrogenic edge fueling for increased scrape-off layer flow for optimum divertor operation. An additional function is to inject impurity gases at lower flow rates for divertor plasma radiative cooling, for wall conditioning, and for plasma discharge termination on demand. The burn fraction of ITER is about 1%, which is more than an order of magnitude lower than values typically assumed in fusion reactor studies. This low burn fraction results in large vacuum pumping and fuel processing systems to handle the larger D-T throughput. Gas and pellet fueling efficiency data from past tokamak experiments are reviewed; pellet fueling efficiency is significantly larger than that of gas injection. An overview of the current research and development status of gas and pellet fueling technology is presented.

A handy method for estimating radiation streaming through holes in shield assemblies

A ‘handy method’ has been developed for performing quick estimates of radiation streaming through cylindrical and rectangular holes in shield assemblies. Quick estimates are often needed during conceptual studies or detailed design of nuclear plants. For example, ITER (International Thermonuclear Engineering Reactor) has a large number of penetrations in the blanket and vacuum vessel and biological shield. This method estimates the levels of radiation that leaks through these holes as well as the magnitude of the fluxes along the hole wall. Results obtained using the ‘handy method’ are in good agreement with 3-D Monte Carlo calculations suggesting that the method provides a practical tool for streaming analysis.

Tritium Proof-of-Principle Pellet Injector—Phase II

As part of the International Thermonuclear Engineering Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. This repeating, single-stage, pneumatic injector, called the Tritium-Proof-of-Principle Phase II (TPOP-II) Pellet Injector, has a piston-driven mechanical extruder and is designed to extrude hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and DT mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and DT extrusions; integrate, test and evaluate the extruder in a repeating, single-stage light gas gun sized for the ITER application (pellet diameter {approximately} 7-8 mm); evaluate options for recycling propellant and extruder exhaust gas; evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory requiring secondary and room containment systems. In initial tests with deuterium feed at ORNL, up to thirteen pellets have been extruded at rates up to 1 Hz and accelerated to speeds of order 1.0-1.1 km/s using hydrogen propellant gas at a supply pressure of 65 bar. The pellets are typically 7.4 mm in diameter and up to 11 mm in length and are the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 11% density perturbation to ITER. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first wall tritium inventories by a process called isotopic fueling where tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge.

Effects of relativistic electrons on the calculated collective Thomson scattered spectra.

The effects of relativistic electrons on the spectral density function S(k,\ensuremath{\omega}) are evaluated in the collective Thomson scattering regime for ion diagnostics. An electromagnetic code that calculates the electron density fluctuation term of S(k,\ensuremath{\omega}) has been modified by treating the electrons with a relativistic Maxwellian susceptibility tensor. Results are presented for calculated scattered spectra representative of the International Thermonuclear Engineering Reactor plasma parameters. Minor effects of relativistic electrons can be seen in the S(k,\ensuremath{\omega}) spectra only in scattering regimes where plasma resonances are present. Relativistic effects in S(k,\ensuremath{\omega}) are negligible in other scattering regimes where plasma resonances are not observable in the spectra. Relativistic effects are more dramatic at higher fluctuation frequencies (\ensuremath{\sim}100 GHz), where the electron Bernstein resonances and electron plasma frequency resonances can be observed. Significant broadening of Bernstein resonance features toward lower frequencies and overlapping of resonance harmonic features at high electron temperatures were noted in the calculated scattered spectra.

AC losses and current distribution in 40 kA NbTi and Nb/sub 3/Sn conductors for NET/ITER

General working formulas usable for calculations of AC losses in superconducting multistage twisted cables have been developed. In situ contact resistances between subcables of different cabling levels as well as AC losses have been measured on full-size steel jacketed 40-kA NbTi and Nb/sub 3/Sn conductors developed for NET/ITER (International Thermonuclear Engineering Reactor) applications. In addition to the strand resistive barriers, stainless steel spacers have been used inside these cables to reduce the AC losses. Theory and measurements demonstrate the role played by the steel spacers in the reduction of the total AC losses of this kind of conductor. The authors give conditions for a good compromise between low AC losses and sufficient current transfer to ensure a correct current distribution among strands.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Diagnostics for ITER; A status report (invited)

The protocol agreement allowing the design of the International Thermonuclear Engineering Reactor (ITER) to proceed will be signed early this year. There is already a small U.S. Program evaluating diagnostic capability for this device and there will shortly be an international effort on diagnostics to ensure satisfactory integration of the diagnostic systems with all the complex systems that will be developed for this large ignited tokamak. The diagnostic studies, at least initially, will depend heavily on the output information provided by the conceptual design activity (CDA). ITER provides a very large plasma (R=6.0 m, a=2.15 m) with high average density (&amp;lt;ne≳∼1.0×1020 m−3), and temperatures (&amp;lt;Te≳∼&amp;lt;Ti≳≥10 keV). It is highly elongated and could have primarily noninductive current drive so that its pulse length can be very long (t≤2 weeks) to allow significant blanket testing in its technology phase. These parameters lead to considering new diagnostic methods for some measurements. A set of specifications for the diagnostic measurements were developed, and a corresponding set of diagnostic instruments was proposed during the CDA. The specifications for spatial and temporal measurement of the plasma parameters are very similar to those achieved on present day large tokamaks, where, on the whole, the access to the plasma is much more open and there is a negligible radiation background in which the diagnostics have to perform. The specifications require that many diagnostic signals participate in the control of complex fuelling, heating, and current drive schemes, so it is clear that the diagnostic design issues for these burning plasmas must be addressed early in ITER’s design life. This is particularly true because penetrations through the first wall are at a premium and only small holes can be tolerated because of radiation streaming. The needs for Research and Development, necessary to demonstrate the feasibility of diagnostic techniques for the ITER plasma parameters and the radiation environment were also addressed. In the area of radiation impact on diagnostic components, some priorities have been set and testing programs have been started. This report will address the present activities on diagnostics in the U.S. Program to address some of the problems left over from the CDA.

A High-Aspect-Ratio Design for the International Thermonuclear Engineering Reactor

Design features and performance parameters for HARD — the high-aspect-ratio (A = 4) International Thermonuclear Engineering Reactor (ITER) design variant developed by the U. S. ITER Team — are presented. The HARD design makes it possible for ITER to achieve both the ignition/extended-burn and the steady-state/technology-testing performance goals set forth in the ITER Terms of Reference. These performance capabilities are obtained in a device that is otherwise similar in concept, size and cost to the low-aspect-ratio (A = 2.8) ITER design defined during the ITER Conceptual Design Activity (CDA). HARD is based on the same physics and engineering guidelines as the CDA design and achieves the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) with inductively-driven plasmas as ITER CDA, but with much greater margin for inductive sustainment of the pulse duration. With non-inductive current drive, HARD operates at lower plasma current and higher plasma density and bootstrap current fraction than ITER CDA, is less constrained by beta limit and divertor considerations, and has increased peaking of the neutron wall load at the test module location. These factors give HARD a much better potential than ITER CDA to achieve the steady-state operation and 1 MWa/m2 technology-testing fluence goals of the ITER objectives.

Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating parameters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in thermal quench phase.

Low Temperature Irradiations in FFTF

The fusion materials program has little irradiation effects data at temperatures from 100 to 350 °C. Near-term machines such as the International Thermonuclear Engineering Reactor (ITER) will expose materials to neutron doses of 38 to 50 dpa at 150 °C or less.1 The data base for structural materials must be extended into this range. Also, lower temperatures are needed to investigate the lower bound for tritium release from solid breeder materials. A low temperature test vehicle is proposed for the Fast Flux Test Facility (FFTF), which will provide test temperatures of 100 to 350 °C. An 8.5–cm dia. by 100–cm test volume will be instrumented to collect temperature data and provide feedback for control. The spectrum and flux will provide accelerated damage accumulation for structural materials testing and the best available approximation of fusion reactor conditions for solid breeder materials testing. Breeder samples can be equipped with sweep gas flow lines for tritium recovery tests.