Controlled Thermonuclear Reactor Research Articles

Kinetic Theory of the Lorentz Gas

AbstractThe steady state kinetic equation for the Lorentz gas, where μ=ex · v/v, has been solved numerically. Analytic approximations have been developed for f(z, v, 0) in the limits of mean‐free‐path long and short compared with the system size. These have been shown to be good, and to provide a useful starting point for the numerical procedure employed, which iterates on f(x, v, 0). Their success suggests that counterpart approximations may work well for more complicated steady state kinetic problems of interest in the physics of the edge in proposed controlled fusion reactors, the diffusion of neutrons, the Brownian motion of heavy molecules, and the structure of shocks. It has also been demonstrated that the determination of f(z, v, 0) can be reduced to the solution of a one dimensional integral equation.

Strategy for Management of Solid Tritiated Wastes

Solid wastes are classified in terms of degassing rate and storage buildings constructed taking due account of the various contamination levels. New reprocessing methods should solve the problem in the future by elimination, de-commissioning or disposal. Solid waste originating in controlled-fusion reactors will prove a critical problem in the future because of the nature of the materials and the presence of highly gamma-irradiating activation products. Therefore, high-standard methods are required for reprocessing tritiated products to be placed in final sub-soil disposal sites. This paper deals with tritiated wastes originated in facilities, which contains neither activation nor fission products (except possibly in very small doses).

Macroscopic phenomena caused by high dose MeV energy He, Ne and Ar implantation

High dose implantation of reaction product MeV energy a particles into the construction elements of future controlled thermonuclear reactors might lead to serious problems. Among others the material suffers from swelling, large mechanical stresses develop in it and the bubbles into which the He gas accumulates weaken the material. When the interbubble material breaks, a layer will be separated and the specimen will show macroscopic surface deformations. An alternative process is when the implanted layer becomes mechanically unstable under the built up compressive mechanical stress and becomes rippled without separating from the bulk. Radiation damage, creep and sputtering enhanced by other particles coming from the active zone lead to further complications. Extensive research was devoted in our laboratory to explore and understand the above mentioned processes. 4He + ions from a 5 MeV Van de Graaff accelerator were applied to simulate α particles. Simultaneous radiation damage, creep and sputtering was simulated by heavier noble gas ions. A special technique was developed to perform quasi-simultaneous multiple energy implantation. A short insight into the achieved results will be given.

The low energy high current particle accelerator at the Colorado School of Mines

The General Ionex model 1545 particle accelerator recently acquired by the Colorado School of Mines Physics Department is described. Operating conditions such as beam energy, currents, ionic species, energy resolution, and vacuum are described. Measurements of radiative capture reactions of protons on D, 6Li, 7Li, and 11B between bombarding energies of 10 and 180 keV are reported. Application of these measurements to the diagnostics of advanced fuel controlled fusion reactors is suggested.

Investigation of helium and hydrogen deposition in an accelerator based neutron source

One of the major obstacles to be surmounted in the design of a controlled thermonuclear reactor (CTR) is the requirement of structural materials which can tolerate the intense radiation environment surrounding a fusion reactor core. The long lead time required for the development of suitable new alloys underscores the need for an intense source of energetic neutrons which can “simulate” the radiation environment associated with a fusion plasma. Among the various accelerator based neutron sources which have been proposed for such a simulation, the spallation neutron source (SNS) is considered to be very promising [1,2]. However, before the technical feasibility of SNS as a tool for simulation of radiation damage characteristics of a CTR can be established we have to compare various neutron damage parameters for samples irradiated in CTR and SNS. Several theoretical studies have been conducted for gas deposition characteristics of different designs of CTRs [3,4]. The aim of the present paper is to study helium and hydrogen deposition in samples irradiated in an SNS and compare the results with those of a CTR. As a reference design for an SNS we have chosen a proposed spallation neutron source called “EURAC” which is based on a 600 MeV beam of protons impinging on a liquid lead target. For this design of SNS we have carried out detailed evaluation of helium and hydrogen deposition in samples of Fe. We have also evaluated neutron and proton fluxes and their spectra using the high energy transport code HETC and studied the effect of variation in sample position on hydrogen deposition. The results of these calculations show that the helium and hydrogen deposition are very sensitive to the high energy part of the neutron spectrum in EURAC. It is also shown that more than half of the total helium production in the irradiation of a sample of Fe is due to neutrons above 40 MeV even though these neutrons constitute only about 3% of the total neutron flux. The results also show that most of the hydrogen deposition is due to the slowing down of secondary protons in the sample.

Primary vacuum pumps for the fusion reactor fuel cycle

The primary vacuum system for a tokamak-type controlled thermonuclear reactor with 1-GW thermal output is analyzed. The need for recovery, purification, and recycling of deuterium–tritium (D–T) fuel to the reactor leads to the following basic requirements for vacuum pumps at the reactor exhaust: effective pumping speed ∼250 m3 s−1; inlet pressure ∼0.1 Pa; outlet pressure ≥10 Pa (at full flow rate of ∼25 Pa m3 s−2); compatibility with tritium, nuclear radiation, static and dynamic magnetic fields, mechanical shock, and up-to-air accidents; and no contamination of pumped gases (primarily D–T and 4He) by vacuum pump media such as lubricants, operating fluids (vapors), or gases other than D–T and 4He. Based on present experience, the turbomolecular pump appears suitable. It would, however, require a substantial development effort both in technology and unit size to meet the requirements. There remains some doubt about its compatibility with mechanical shock, radiation, magnetic fields, and sudden up-to-air accidents, etc. Therefore, alternative pumping principles are proposed, all shown to work at least in laboratory size models: thermodynamic pumps for transport pumping to replace turbomolecular pumps and selective storage pumps for effluent processing in situ. The proposed systems require development efforts similar to the turbomolecular pump, but are shown to meet all requirements without the need for large protective valves. Selective pumps, pumping none of the hydrogen isotopes (but helium and impurities such as DTO, CO2, methane) or combinations of thermodynamic transport pumps, and regenerable volume getter pumps are shown to perform the major processing task, i.e., purification and recycling of DT, directly at the reactor exhaust.

The JET project: introduction and background

The Joint European Torus, JET, is an experiment, undertaken by 15 partners from 12 nations of Western Europe, to get information on the magnetic confinement of high-temperature plasma in conditions close to those needed for energy-producing controlled thermonuclear fusion reactors. Physically, JET is a very powerful toroidal-pinch electric discharge in a strong stabilizing magnetic field, a system known as a tokamak. The paper summarizes the main features of a tokamak and relates them to the papers in this symposium.

The scientific origins of controlled fusion technology

Summary This paper discusses the emergence of the concepts of fusion as an energy source and of the controlled fusion reactor. These concepts are shown to have arisen from the bringing together of several different branches of physics, notably nuclear physics, astrophysics, and gas discharge physics, in the period between the two world wars. By the late 1930s, enough information had become available for the possibility of a controlled fusion device to be explored, and a number of physicists seem to have recognized this independently. Such a possibility was the subject of discussion at Los Alamos during and immediately after the war, but the first detailed proposals to be recorded seem to be those of G. P. Thomson and P. C. Thonemann, both working in Britain in the middle to late 1940s. These proposals, which led to the setting up in Britain of the first coherent programme of controlled fusion research, are outlined and discussed.

Experimental devices in the osiris reactor to study effects of radiations on fusion reactor materials

Within the framework of the Technology Research Program on controlled fusion initiated by the European Communities, the Services des Piles de Saclay (SPS) of Commissariat à l'Energie Atomique (CEA) have been requested to perform some necessary experiments to study the irradiation behaviour of materials which are possible candidates for controlled fusion reactors. This paper describes the devices, generally adapted from a standard model “The COLIBRI”, which allow one to carry out, in the OSIRIS reactor, irradiations on the three great families of fusion reactor materials: - lithium containing materials of breeding blanket for in-situ tritium production, - protection materials, and - structural materials.

Residual stress in coated low-Z films of TiC and TiN: III. PVD coated films

Physical vapor deposition (PVD) is an attractive method of in-situ deposition for future controlled nuclear fusion reactors since it can prepare various low-Z ceramic films at much lower deposition temperatures. This paper is concerned with the properties of a titanium carbide coating on molybdenum by the activated reactive evaporation process (ARE), a technique known as PVD. Residual stress in films which might affect the adherence of the film were examined with particular reference to the deposition temperature. All films prepared in this study were found to have a compressive stress and its value decreased from 4.76 × 10 10 to 6.77 × 10 9 dyn/cm 2 as the deposition temperature was raised from 300 to 700°C. A discussion was given on the relatively high value of stress measured in PVD-TiC coatings. Problems in applying a PVD to the in-situ coating technique were also discussed.

Neutron-induced helium implantation in helium coolant pipes of fusion reactors

The neutron-induced implantation of energetic helium particles in the interior surface of the helium coolant pipes near the first wall of a controlled thermonuclear reactor has been studied analytically. A flux of energetic recoil helium particles as high as 1.0 × 10 10 He · cm −2 · s −1 has been calculated for the interior surface of 2-cm-diameter pipe made of Type 316 stainless steel near the first wall of a Tokamak fusion reactor with a power density of 1 MW/m 2. The calculated profile of the helium implantation rates indicates that about half the flux of energetic recoil helium particles will be implanted in the first 2.5-μm-deep layer of the surface. The rest will be implanted as deep as 40 μm. Since the calculated profile is significantly different from the one obtained from bombardment with a monoenergetic helium particle beam, use of multienergy beams or irradiation of helium-filled capsules in the core of a fast breeder reactor are recommended to assess the possible effects of the neutron-induced helium implantation.

Testing of low- Z-coated limiters in tokamak fusion devices

Coatings of materials with low atomic number Z are being developed for application both in experimental tokamak devices such as ISX and TFTR and for possible use on various components of large power-producing controlled thermonuclear reactors. A suitable coating has the potential of improving the plasma energy balance by lowering the atomic number of the impurities injected by plasma-wall interactions. This results in a decrease in plasma radiation losses and improved plasma stability. The coatings will be subjected to a severe environment and must be able to withstand physical erosion, chemical erosion, thermal shock, thermal fatigue and arcing without rapid degradation or loss of adhesion. Extensive testing on a laboratory scale was used to select the coatings most suitable for this environment. From this testing, which included pulsed electron beam heating, low energy ion bombardment and arcing, chemically vapor-deposited coatings of TiB 2 and TiC on Poco graphite substrates were selected and tested as limiters in ISX. Both limiter materials gave clean stable reproducible tokamak discharges on the first day of operation. After an exposure of 1 week, the TiC limiter showed only superficial damage with no coating failure. The TiB 2 limiter showed some small areas of coating failure. TiC-coated graphite limiters were also briefly tested in the tokamaks Alcator and PDX with favourable results.

Application of Generalized Variational Principles to Controlled Thermonuclear Reactor Neutronics Analysis

Application of Generalized Variational Principles to Controlled Thermonuclear Reactor Neutronics Analysis

Accidental Criticality of a Fusion-Fission Hybrid Blanket Design

Whenever a fissionable blanket is added to a controlled thermonuclear reactor, an investigation of the potential for accidental criticality is necessary to ensure the safety of this hybrid device. ...

Ｘ線光電子分光法による深さ方向の定量的状態分析とそのＳｉＣ表面への応用

An attempt of quantitative state analysis was made on the surface and the depth profile of inorganic compounds by X-ray photoelectron spectroscopy (XPS) which was combined by the sputter-etching with argon ions. A masking attachment was designed for an area of sample which is exposed to the non-uniform portion of the ion beam. Uniform sputter-etching could be attained, with the advantages on XPS observation of low background level and less impurity spectra from other origins than the sample. The photoelectron yields were examined for the quantitative analysis by XPS. The method established here was applied to analyze the surface and in-depth composition of SiC coatings onto carbon and molybdenum which are promising candidate materials as the first wall in a controlled thermonuclear reactor.

Research in microsphere fabrication and coolant loop technology for lithium oxide moving bed fusion power plant

A gravity-flow bed of lithium oxide microspheres has been proposed as the tritium breeding and blanket cooling medium for both magnetic confinement and inertial confinement controlled fusion reactors using the D-T cycle. The use of the lithium oxide moving bed could help eliminate MHD, corrosion, and materials compatibility problems associated with the use of liquid lithium. Adequate particle transport mechanisms which alleviate particle attrition and erosion seem attainable. Moving bed heat transfer coefficients an order of magnitude greater than for fluidized beds are indicat, ed. The fabrication of microspheres is complicated by the aggressive caustic nature of the lithium oxide but should be possible for commercial-scale production.

Dynamics and Control of a Packed Distillation Column for the Isotopic Enrichment of Plasma Exhausts from Controlled Thermonuclear Reactors

A control scheme is proposed based on our analysis of the disturbances expected during normal operation of the lead column in the fuel enrichment distillation cascade for a near-term tokamak fusion reactor fuel cycle. The primary objective of this control scheme is to minimize both the time and the amount that the atom percent protium in the bottoms product is above a setpoint level. As a secondary objective, distillate stream flow and composition fluctuations should be minimized to avoid downstream operational and control problems without requiring intermediate storage. A fixed material balance control scheme was found to be satisfactory for meeting the control requirements of this system. Because the concentration of protium in the bottoms product (the controlled variable) was relatively small, the distillate stream composition and the tritium/deuterium ratio in the bottoms stream proved to be essentially independent of the choice of controller parameters. This insensitivity permitted the controller parameters to be chosen solely on the basis of the primary objective and led to a high gain setting and low reset value for the controller. With the provision of a bottoms storage to dampen out the effect of oscillatory response, these controller settings minimized the overshoot and produced an more » averaged protium concentration in the bottoms very close to the setpoint level. « less

Some safety considerations in laser-controlled thermonuclear reactors

The potential safety problems which might be encountered in laser-driven controlled thermonuclear reactors (LCTR) have been examined and compared with previous studies of magnetically-confined systems. In several significant ways, including tritium hazards and some blanket cooling functions, laser-driven systems appear to differ only in quantitative ways from magnetic systems. Several aspects of safety of laser-driven systems are distinctive to them. These include cooling means designed to handle the pulsetile energy deposition from pellet explosions, and various equipment and containment penetrations associated with the lasers. Specify safety analyses have been performed for a loss-of-flow accident (LOFA) in a wetted-wall LCTR, and failure modes for laser penetrations are considered in detail to ascertain appropriate component placement and design criteria.

Helium gas bubble lattices in face-centred-cubic metals

ION irradiation of metals with helium or hydrogen isotopes commonly results in the formation of microscopic gas bubbles. At very high doses, radiation blistering and related phenomena can cause severe damage to surfaces1. Much research has been stimulated by the need to find materials with acceptable erosion and gas loading properties for use in future controlled thermonuclear reactors. Bubble formation is also important in fission-reactor fuel technology and other applications2,3. The underlying bubble structure of face-centred-cubic (f.c.c.) metals has remained largely unknown, although much has been done on depth profiling light gases implanted to high doses (see, for example, refs 3–5). Inert-gas bubbles lying on a space lattice have been reported previously only for the body-centred-cubic (b.c.c.) metal molybdenum6,7. There have been many observations, mainly for b.c.c. metals, of the similar ordering of voids produced by particle irradiation8,9. The void lattice spacing is typically 5–10 times that found for bubbles (∼5 nm). We report here on the bubbles produced in several f.c.c. metals by 30 keV helium-ion irradiation to a level just below the critical dose for blistering. A new result for this important class of metals is that the helium gas bubbles lie on a superlattice having an f.c.c. structure with principal axes aligned with those of the matrix.

Calculation of Radiological Dose due to Release of Tritium Oxides from Controlled Thermonuclear Reactors

A method has been developed for determining doses to the public resulting from releases of tritium as tritiated water vapor or as tritiated lithium compounds. This method has been included in a computer model. This model uses the Gaussian dispersion method to predict distribution of tritium species in the downwind environment. Movement of tritium into biological systems is determined by treating these systems as a series of interacting water compartments. Dispersion and uptake calculations are applied to two sample sites to predict health effects. Consequences predicted by the model are linear and can be scaled to any release quantity. For a continuous release of tritium at a rate of 10 Ci/day, the calculated dose would be 8 mrem/yr at the site boundary, with a dose commitment of 10 to 100 man-rem/yr within an 80-km radius. For an instantaneous release of 10/sup 8/ Ci, the calculated dose would be as high as 2200 rem at the site boundary, contributing a population dose of 0.6 to 2.6 x 10/sup 6/ man-rem within 80 km.