Fusion Reactor Environment Research Articles

SiC as a structural material in the plasma chamber of nuclear fusion reactors

Ceramic materials are necessary for several critical applications in fusion power reactors or as alternative materials for metallic materials. The leading candidate ceramic material for use in the plasma chamber of a fusion machine is SiC because of its low plasma contamination, low induced radioactivity, capability of high operation temperature and abundant inexpensive raw materials supply. Critical requirements with respect to thermal, mechanical, irradiation and environmental loading are given, and the properties of today's SiC materials compared with these requirements. Of special importance are the microstructure, activation and transmutation, erosion, swelling, thermal shock behaviour, strength and durability. The development of ceramics for heat engine applications has demonstrated the capability of engineering ceramics. First irradiation tests in fission reactors have indicated that there are no insurmountable problems for the use of these materials in a fusion reactor environment. However, more data have to be generated on present and newly developed types of SiC to confirm that SiC is a serious candidate for the desired application.

An Investigation of the Solubility of LiOH in Solid Li2O

The solubility of LiOH in Li2O was measured as a function of temperature and partial pressure of H2O. For these measurements, solid Li2O at temperatures of 973 to 1273 K was equilibrated with moisture in a helium carrier gas stream; monitoring the moisture content in the helium carrier gas allowed determination of LiOH solubility. The slope of the curve for the partial pressure of H2O(g) versus LiOH solubility was observed to be 0.6 and is interpreted as an indication of nonideality in the system. For a given partial pressure of H2O, the solubility of LiOH increased with increasing temperature. As a consequence of these measurements, the solubility of tritium (as LiOT) in a breeding blanket of solid Li2O is expected to be low under anticipated fusion reactor environments [e.g., 0.56 wppm tritium at 1000K for 1 Pa (10−5 atm) T2O].

Methods for forecasting performance limits of fusion reactor structural materials

For the purposes of both developing appropriate alloys for fusion reactors and ultimately designing and building such devices, it is necessary to determine the mechanical response to high energy neutron irradiation of candidate structural materials. The current and near-term facilities for providing the requisite testing environment are either limited by dissimilarity to fusion reactor environments (i.e., fission reactors), or irradiation volume, or both. This necessitates the development of both a testing technology commensurate with the available facilities — i.e., primarily small specimen technology — and a correlation methodology which recognizes the limitations of the anticipated data base and maximizes the use of that data. In this paper we (1) describe a general philosophy for such correlation methodologies, (2) provide a brief overview of small specimen test techniques and data characteristics, and (3) demonstrate with several examples how correlations can be developed and applied.

Helium effects on the creep and fatigue resistance of austenitic stainless steels at high temperatures

The effects of helium on the high temperature mechanical properties of austenitic stainless steel, also known as high temperature helium embrittlement, are reviewed. Special emphasis is given to such experimental data which simulate a fusion reactor environment best, i.e., to creep and low frequency fatigue experiments at high helium levels (≥ 100 appm He). Correlations between mechanical properties data and microstructural data are presented and compared to recently developed models for helium embrittlement. The presented data are critically assessed in the light of fusion reactor materials needs and some suggestions for future work are made.

Helium effects on the creep and fatigue resistance of austenitic stainless steels at high temperatures

The effects of helium on the high temperature mechanical properties of austenitic stainless steel, also known as high temperature helium embrittlement, are reviewed. Special emphasis is given to such experimental data which simulate a fusion reactor environment best, i.e., to creep and low frequency fatigue experiments at high helium levels (≥ 100 appm He). Correlations between mechanical properties data and microstructural data are presented and compared to recently developed models for helium embrittlement. The presented data are critically assessed in the light of fusion reactor materials needs and some suggestions for future work are made.

Partial pressures of H 2O above the diphasic [formula omitted] system

The temperature dependence of the partial pressure of H 2O(g) above the Li 2O(s)-LiOH(s, l) system was determined for temperatures between 300 and 617°C. The partial pressures were measured by means of a flowing gas technique combined with continuous monitoring of the concentration of water vapor in a helium carrier gas. For the reaction LiOH(s) = Li 2O(s) + H 2O(g) , second law heat and entropy of reaction values of ΔH o = 30.7 ± 0.6 kcal/mol and ΔS o = 29.5 ± 1.0 cal/mol.K were obtained. Above the melting point of LiOH (744 K), these values were ΔH o =19.9 ± 0.6 kcal/mol and ΔS o =14.8 ± 0.8 cal/mol.K . Current measurements yield ΔH m o = 5.4 ± 0.4 kcal/mol for the heat of melting of LiOH, which is in good agreement with the JANAF recommended value of 4.99 kcal/mol. The results of these measurements can be used to partially describe the behavior of a Li 2O solid breeding blanket in anticipated fusion reactor environments.

Ion impact desorption measurements of sputter-deposited copper on stainless steel

Hydrogen ion impact desorption of sputter deposited copper on type 304 stainless steel substrates has been measured using secondary ion mass spectrometry. Samples were prepared in a plasma simulator and then transferred to an analysis chamber, where each sample was subjected to hydrogen ion bombardment in the energy range 0.3–1 keV. By monitoring Cu + emission as a function of H ion fluence, it was possible to determine the cross section for desorption. The cross section was evaluated as a function of H ion energy; the maximum value was 1.1 × 10 −17 cm 2 at 1 keV. By monitoring the 63Cu + 65Cu + emission ratio as a function of H ion fluence, slight preferential sputtering of 63Cu was detected. Results from these experiments make it possible to predict the fate of sputter deposited material in a fusion reactor environment, such as whether it is more likely to remain on the first-wall or be released back into the plasma.

Vaporization behavior of lithium oxide: Effect of water vapor in helium carrier gas

The effect of water vapor in a helium carrier gas on the vaporization behavior of lithium oxide has been investigated in the temperature range 1023 to 1273 K. Based on the reaction Li 2O(s)+H 2O(g) → 2LiOH(g), the results of this study yield second and third law heats of reaction of 79.0 ± 3 and 82.1 ± 1 kcal/mol. Moisture significantly enhances the volatility of lithium oxide. The pronounced effect of water vapor on the volatilization of Li 2O (as LiOH) is important in understanding the behavior of a Li 2O solid breeding blanket in anticipated fusion reactor environments.

Low Energy Ion?Surface Interaction ? Atomic Physics in an Ordered Atom Environment

Ion-surface interactions have application within a number of areas of contemporary physics research. In particular, it is essential that an understanding of the interaction between an ion and a solid surface in a fusion reactor environment be achieved. A detailed understanding of the interaction is also essential for quantitative application of several of the modern methods of surface analysis currently available. It is unfortunate that the necessary level of understanding of the interac

Gamma Radiation Effects on Tritium Permeation Through Stainless Steel

Experiments were conducted in the gamma irradiation facility of the Advanced Test Reactor at the Idaho National Engineering Laboratory to determine whether gamma radiation has an appreciable effect on the normal permeation of tritium through stainless steel. Low concentrations of HT were allowed to diffuse through a 0.071-cm-thick tube of 316 stainless steel, heated between 590 and 733 K. Gamma irradiation intensities were varied from 1.3 to 155 C/kgXh (5 X 10/sup 3/ to 6 X 10/sup 5/ R/h). Ion chamber detectors were used to measure tritium concentrations on both sides of the tube. It was found that in the presence of excess H/sub 2/, the higher gamma irradiation intensity exhibited slightly higher permeation rates of tritium. When the walls of the permeation tube and the HT were highly oxidized, the permeation rates were much more scattered, and the gamma irradiation seemed to have no observable effect. It was concluded that the effect of gamma radiation on tritium permeation through stainless steel in a fusion reactor environment should be small. However, the relative ease with which tritium from HTO was seen to permeate the material raises questions regarding tritium management in breeder blankets.

The dependence of the high temperature mechanical properties of austenitic stainless steels on implanted helium

High temperature helium embrittlement effects on the creep properties of AISI 316 SS (solution annealed + aged) and DIN 1.4970 SS (solution annealed + cold worked + aged) have been investigated. The generation of helium due to (n. α) nuclear reactions in a fusion reactor environment has been simulated by homogeneous helium implantation at a cyclotron. The creep rupture tests with various applied tensile stresses have been carried out at 1023 K. (316 SS) and 1079 K (1.4970 SS). respectively, with four differently treatly sets of samples: (1) unimplanted controls; (2) after room temperature implantation of 100 appm He; (3) after implantation of 100 appm He at test temperature; (4) creep tested at high temperature during implantation (“in-beam”) with implantation rates of 10–100 appm He/b. In contrast to the ductile behaviour with transgranular failure of the unimplanted controls, all He-implanted samples showed brittle, intergranular early failure. The embrittlement effect was enhanced for the “in-beam” tested samples. The difference between the different treated sets of samples can be related to different bubble microstructures investigated by TEM. In addition, a comparison to reactor data for the DIN 1.4970 SS is presented.

Radiation effects in materials for fusion reactors

The 14-MeV neutrons produced in a fusion reactor result in different irradiation damage than the equivalent fluence in a fast breeder reactor, not only because of the higher defect generation rate, but because of the production of significant concentrations of helium and hydrogen. Although no fusion test reactor exists, the effects of combined displacement damage plus helium can be studied in mixed-spectrum fission reactors for alloys containing nickel (e.g., austenitic stainless steels). The presence of helium appears to modify vacancy and interstitial recombination such that microstructural development in alloys differs between the fusion and fission reactor environments. Since mechanical properties of alloys are related to the microstructure, the simultaneous production of helium and displacement damage impacts upon key design properties such as tensile, fatigue, creep, and crack growth. Through an understanding of the basic phenomena occurring during irradiation and the relationships between microstructure and properties, alloys can be tailored to minimize radiation-induced swelling and improve mechanical properties in fusion reactor service.

Formation of helium platelets in molybdenum

Over the past two decades, helium introduced into metals either by ion implantation or by the (n, α) reaction has been studied over a wide range of concentrations and temperatures1,2. Early studies were relevant to fission reactor technology3,4 but more recent emphasis has been on the role of helium in fusion reactor environments. One area of interest concerns the interaction of helium ions from the plasma with the first-wall surface while the influence of the helium from the (n, α) reaction on void nucleation, and on grain boundary bubble nucleation, is also under investigation5. One long-established property of insoluble helium and other inert gases in metals is their tendency to precipitate as bubbles. However, the inference that inert gas clusters always assume a three-dimensional form may not be justified. We present here experimental evidence showing that, at least in molybdenum, helium can initially aggregate in an unexpected planar form. Furthermore, we have found a bubble nucleation route involving the formation of several small bubbles from a single helium platelet.

Self-sustaining thin films as a means of reducing first wall erosion and plasma impurity influx

Neutral impurities ejected from Tokamak wall and limiter surfaces may travel several cm before being ionized very quickly upon entering the plasma edge. The influence of the unipolar sheath potential is exerted only within a very short distance of the surface and has no effect on neutral impurity atoms which are subsequently ionized by charge-exchange collisions or electron impact ionization. However, secondary ions emanating from the limiter surfaces with kinetic energies less than the sheath potential will have essentially zero probability of traveling more than a few Debye lengths before being redeposited. Similarly, secondary ions originating at the first wall are redeposited as a result of the deflection produced by the magnetic field. Impurity influx resulting from sputtering would therefore be substantially reduced for surfaces which produce a very high ion/neutral ratio when sputtered. It has been previously shown that the high secondary ion yield associated with the alkali metal potassium does not apply to the bulk metal but pertains to ionic compounds and thin (mono-layer) films. Two processes are discussed as a means of producing these films in a self-sustaining manner compatible with the fusion reactor environment.

Preliminary evaluation of beta-spodumene as a fusion reactor structural material

Beta-spodumene was investigated as a candidate material for use in fusion reactor environments. Properties which support the use of beta-spodumene include good thermal shock resistance, a very low coefficient of thermal expansion, a low-Z composition which would result in minimum impact on the plasma, and flexibility in fabrication processes. Specimens were irradiated in the Advanced Test Reactor (ATR) to a fluence of 5.3 × 10 22 n/m 2, E > 0.1 MeV, and 4.9 × 10 23 n/m 2 thermal fluence in order to obtain a preliminary evaluation of the impact of irradiation on the material. Preliminary data indicate that the mechanical properties of betaspodumene are little affected by irradiation. Gas production and release have also been investigated.

Flow and fracture of alloys in the fusion environment

The present paper examines both ductile and brittle fracture models of steels and assesses the impact of the fusion reactor environment on the fracture processes. In particular, the connections between plastic flow properties and fracture modes are reviewed for both ductile and brittle crack propagation. Highly radiation-hardened materials exhibit extreme flow localization resulting in channel fracture. Physical models for this phenomon are developed and an estimate for the associated fracture toughness is given. The impact of radiation-hardening and ductility loss on fatigue crack growth is examined. Next, models describing the chemical effects on fatigue and fracture are briefly discussed. Finally, fracture design criteria are proposed for first wall structures in fusion reactors.

Characteristics of irradiation creep in the first wall of a fusion reactor

A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from fission reactors are analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available.

Development of fatigue life criteria for experimental fusion reactor first-wall structures

An approach to the rational design of fusion reactor first-wall structures against fatigue crack growth is proposed. The approach is motivated by microstructural observations of fatigue crack growth enhancement in unirradiated materials due to volumetric damage ahead of a propagating crack. Examples are cited that illustrate the effect of mean stress on void nucleation and coalescence, which represent the dominant form of volumetric damage at low temperature, and of grain boundary sliding and creep cavitation, which are the dominant volumetric damage mechanisms at high temperature. The analogy is then drawn between these forms of fatigue crack growth enhancement and those promoted by irradiation exposure in the fusion reactor environment, such as helium embrittlement and atomic displacement. An enhanced strain range is suggested as a macroscopic measure of the reduction in fatigue life due to the higher fatigue crack growth rates. The enhanced strain range permits a separation of volumetric and cyclic effects, and assists in the assignment of rational design factors to each effect. A series of experiments are outlined which should provide the numerical values of the parameters for the enhanced strain range.

Development of mechanical property correlation methodology for fusion environments

It is suggested that a model-based analysis of the relationship between microstructure and properties and relatively simple properties, such as tensile data and fracture/failure parameters is one useful component in the overall effort to predict structural response in fusion reactor environments. The approach is applied to modeling yield strength changes in irradiated stainless steel with apparent success. Further analysis of ductility data may suggest significant changes in fracture mechanisms in highly irradiated stainless steel. Finally, a simple micromechanical model is used to estimate changes in fracture toughness; while there is not sufficient confirmatory data, the calculations indicate a substantial reduction in toughness due to irradiation.

The impact of hydrogen in a fusion reactor environment on titanium alloys

The impact of hydrogen in a fusion reactor on a titanium first wall and blanket structure has been investigated. Hydrogen is of concern in titanium because the combination of hydrogen isotopes in a fusion reactor and titanium's high affinity for hydrogen has the potential for producing hydrogen embrittlement and increasing the tritium inventory. The solubility of hydrogen in titanium depends upon a number of variables, chief of which is the type and amount of phase present, (alpha, beta, or alpha-beta). In this study the influence of alloying elements on phase stability and hydrogen solubility was investigated. The effects of three major sources of hydrogen were examined: (1) hydrogen produced via neutron transmutation reactions, (2) deuterium and tritium from the plasma chamber, and (3) tritium from the breeding material or coolant. Results indicate that under the conditions studied hydrogen from transmutation reactions does not raise the hydrogen concentration significantly above the level currently allowed in commercial alloys. However, significant amounts of deuterium and tritium can be absorbed in the first wall at the higher gas pressures encountered during the refueling cycle. The tritium concentrations anticipated with lithium or lithium oxide present, result in low tritium concentrations in titanium structures whereas tritium pressures anticipated for a lithium-lead breeding blanket may result in substantial tritium concentrations in a titanium structure.