R.J. Puigh

Irradiation creep and swelling of the fusion heats of PCA, HT9 and 9Cr-1Mo irradiated to high neutron fluence☆

Abstract The irradiation creep and swelling at 400–750°C of the fusion heats of PCA, HT9 and 9Cr-lMo were studied using pressurized tubes irradiated in the Fast Flux Test Facility (FFTF). At temperatures

Modeling, analysis and experiments for fusion nuclear technology

Selected issues in the development of fusion nuclear technology (FNT) have been studied. These relate to (1) near-term experiments, modeling, and analysis for several key FNT issues, and (2) FNT testing in future fusion facilities. A key concern for solid breeder blankets is to reduce the number of candidate materials and configurations for advanced experiments to emphasize those with the highest potential. Based on technical analysis, recommendations have been developed for reducing the size of the test matrix and for focusing the testing program on important areas of emphasis. The characteristics of an advanced liquid metal MHD experiment have also been studied. This facility is required in addition to existing facilities in order to address critical uncertainties in MHD fluid flow and heat transfer. In addition to experiments, successful development of FNT will require models for interpreting experimental data, for planning experiments, and for use as a design tool for fusion components. Modeling of liquid metal fluid flow is a particular area of need in which substantial progress is expected, and initial efforts are reported here. Preliminary results on the modeling of tritium transport and inventory in solid breeders are also summarized. Finally, the thermo-mechanical behavior of liquid-metal-cooled limiters is analyzed and the parameter space for feasible designs is explored. Because of the renewed strong interest in a fusion engineering facility, a critical review and analysis of the important FNT testing requirements have been performed. Several areas have been emphasized due to their strong impact on the design and cost of the test facility. These include (1) the length of the plasma burn and the mode of operation (pulsed vs. steady-state), and (2) the need for a tritium-producing blanket and its impact on the availability of the device.

Technical issues and requirements of experiments and facilities for fusion nuclear technology

The technical issues, development problems and required experiments and facilities for fusion nuclear technology have been investigated. The results have been used to develop a technical framework for a test plan that identifies the role, timing, characteristics and costs of major experiments and facilities. A major feature of this framework is the utilization of non-fusion facilities over the next 15 years, followed by testing in fusion devices beyond about the year 2000. Basic, separate effect and multiple interaction experiments in non-fusion facilities will provide property data, explore phenomena and provide input to theory and analytic modelling. Experiments in fusion facilities can proceed in two phases: (1) concept verification and (2) component reliability growth. Integrated testing imposes certain requirements on fusion testing device parameters; these requirements have been quantified. The nuclear subsystems addressed in the study are: (a) blanket and first wall; (b) tritium processing system; (c) plasma interactive components; and (d) radiation shield. The two generic classes of liquid and solid breeder blankets have significant engineering feasibility issues, and new experimental data must be obtained before selection of an attractive design concept. Liquid metal blanket issues are dominated by problems related to momentum, heat and mass transfer, which can be addressed in non-neutron test facilities. Solid breeder blanket issues are, however, dominated by the effects of radiation, including heating, transmutation and damage, which can be reasonably addressed in fission reactors. The tritium processing uncertainties are primarily related to the control and recovery systems, and most can be addressed in existing and planned non-neutron facilities. A dominant feature of plasma interactive components is the strong interrelation to both plasma physics and nuclear technology. Required facilities include thermomechanical test stands and confinement devices with sufficiently long plasma burn. The radiation shield poses no feasibility issues, but improved accuracy of predictions will reduce design conservatism and lower costs.

Miniature specimen tensile data for high energy neutron source experiments

A miniature tensile specimen technology has been developed in support of high energy neutron irradiated material testing. The work includes miniature specimen design and fabrication technique development, as well as both baseline and irradiated tensile testing. Baseline data obtained using miniature tensile specimens are in good agreement with published values obtained using larger specimen geometries. Tensile data from miniature specimens irradiated up to a high energy, E approx. 14 MeV, dose level of 1 10/sup 18/ n/cm/sup 2/ or 0.003 dpa, have been obtained for annealed and cold worked AISI 316 stainless steel.

The in-reactor deformation of the PCA alloy

This report documents the results from the irradiation of the PCA alloy in the Fast Flux Test Facility (FFTF). Both swelling and in-reactor creep data were obtained to a peak neutron dose corresponding to 80 dpa [16 × 1022n/cm2 (E > 0.1 MeV)]. These data are compared to the in-reactor deformation in cold worked 316 SS. The swelling data from this work for both PCA and 316 SS are also compared to swelling data on the 20% cold worked LSI alloy. This alloy is a titanium-stabilized austenitic steel developed for the liquid metal reactor program.

Thermal creep and stress-affected precipitation of 20% cold-worked 316 stainless steel

Abstract Measurements of the thermal creep of 20% cold-worked 316 stainless stpel have been performed for temperatures from 593 to 760°C, stress levels as high as 138 MPa and exposure times as long as 15,000 hours. The creep strains exhibit a complex behavior arising from the combined action of true creep and stress-affected precipitation of intermetallic phases. The latter process is suspected to be altered by neutron irradiation.

Titanium alloy tensile properties after neutron irradiation

Irradiated specimens from three duplex annealed titanium alloys were tested in uniaxial tension in air from room temperature to 550/sup 0/C. The EBR-II irradiation temperature was 550/sup 0/C; the maximum fluence was 5 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV), or 37 dpa, the highest neutron exposure received by any titanium alloy. Alloy 6242S increased in strength due to reactor exposure by approximately 20% at test temperatures from room to 400/sup 0/C. The strength of alloy 5621S was unaffected by reactor exposure, while Ti-6Al-4V displayed strength reductions at 550/sup 0/C. Largely ductility losses of 75 to 80% of unirradiated material total elongation were noted after reactor exposure in all three alloys; however, reasonable postirradiation ductility of 0.9 to 8.8% remained in all alloys.

FFTF as an irradiation test bed for fusion materials and components

Abstract The relatively large irradiation volume, instrumentation capabilities, and fast neutron flux associated with the Fast Flux Test Facility (FFTF) make this reactor a useful test bed for fusion materials and components irradiations. Significant fusion materials irradiations are presently being performed in the Materials Open Test Assembly (MOTA) in FFTF. The MOTA is providing a controlled temperature and high neutron flux environment for materials such as low activation alloys, copper alloys, ceramic insulators, and high heat-flux components. Conceptual designs utilizing the versatile MOTA irradiation vehicle have been developed to investigate irradiation effects on the mechanical and tritium breeding behaviors of solid breeder materials. More aggressive conceptual designs have also been developed to irradiate solid breeder blanket submodules in the FFTF. These specific component test designs are presented and their potential roles in the development of fusion technology discussed.

Fission Reactor Experiments for Solid Breeder Blankets

Solid breeder blanket testing needs have been characterized in the FINESSE study. These testing needs have unique characteristics, especially in comparison to liquid breeder blankets. First, there are a large number of potential breeder materials and material variables. Secondly, the influence of radiation on the primary uncertainties is large, but the influence of geometry is not. Finally, much of the important functional behavior of the solid breeder is not described by classical equations, but rather the controlling phenomena must be quantified by experiments. Therefore, test conditions must match fusion reactor conditions as closely as possible. These factors suggest that a reasonable number of moderate volume test sites in a neutron environment are needed. The suitability of fission reactors and the major solid breeder experiments are considered here.

Technical Requirements of Experiments and Facilities for Fusion Nuclear Technology

A process has been developed and applied to the technical planning of experiments and facilities for fusion nuclear technology. The process involves: (1) characterization of issues; (2) quantification of testing requirements; (3) evaluation of facilities; and (4) development of a test plant to identify the role, timing, characteristics, and costs of major experiments and facilities. The nuclear subsystems addressed are: (1) blanket, including the first wall; (b) radiation shield; (c) tritium processing system; and (d) plasma interactive components. Particular emphasis has been placed on the complex technical issues and development problems of the blanket.