Difference between revisions of "Fission Fragment Diffusion Through TRISO"
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The objective of this project is to assess the retention of activated fission products within graphite as a function of the microstructural, fission product, and environmental conditions examined. Overall project results should include a description of all experimental conditions examined, analytical methods employed, and resulting effects on transport and retention of the various species examined. | The objective of this project is to assess the retention of activated fission products within graphite as a function of the microstructural, fission product, and environmental conditions examined. Overall project results should include a description of all experimental conditions examined, analytical methods employed, and resulting effects on transport and retention of the various species examined. | ||
| + | = Irradiate and perform PAA on graphite films= | ||
| + | |||
| + | ==Fission Fragments of Interest== | ||
| + | |||
| + | ===silver=== | ||
| + | ===cesium=== | ||
| + | ===strontium=== | ||
| + | ===europium=== | ||
=References= | =References= | ||
Revision as of 20:24, 26 October 2016
Text from Appendix A
RADIOISOTOPE RETENTION IN GRAPHITE AND GRAPHITIC MATERIALS (RC-2) (FEDERAL POC – MADELINE FELTUS & TECHNICAL POC – PAUL DEMKOWICZ) (ELIGIBLE TO LEAD: UNIVERSITIES ONLY) (UP TO 3 YEARS AND $800,000)
Graphite is a primary core material across multiple types of advanced reactors (i.e., HTGR, FHR, and MSR) which have common material issues such as irradiation-induced material property changes, chemical reactivity, and material degradation. Fundamental studies determining the underlying mechanisms driving the material behavior as well as the impact from these effects on the core behavior is required for design and licensing can be completed for these advanced reactor concepts. A major issue of concern for MSR, FHR, and VHTR designs is the retention of activated fission products within graphite and graphitic materials such as the graphitic matrix composite used in TRISO particle fuel forms (pebbles or compacts). Radioactive material of fission product release from the fuel or from neutron reactions with molten coolant and fuel (lithium in FLiBe or FLiNaK in MSR designs) can be retained in carbon matrix, carbon-carbon composites, and graphite components. Research is needed on those graphite properties that are important for retention (and potential release) of these radioisotopes from a material possessing a graphitic crystal structure. Of particular interest is the chemisorption potential for various species, RSA efficiency, diffusion and intercalation efficiency, microstructure effects (grain size, BET, porosity distribution, source material), and at partial pressures of hydrogen (tritium and entrained water) over a range of high temperatures (500-1600C). This will assist in determining total inventory of retained products for accurate source term calculations required for licensing, determining the possibilities of tritium removal from carbon-based materials, and core component performance issues. The sorption/desorption isotherms of key fission products (including silver, cesium, strontium, and europium) in irradiated nuclear grade graphites for the high temperature reactors need to be determined. Research projects may use un-irradiated graphitic material and non-radioactive isotopes of the key fission products as surrogates to determine fission product retention behavior; however, comparison of parameters with the results from irradiated TRISO fuel forms and irradiated graphite experiments is encouraged. The objective of this project is to assess the retention of activated fission products within graphite as a function of the microstructural, fission product, and environmental conditions examined. Overall project results should include a description of all experimental conditions examined, analytical methods employed, and resulting effects on transport and retention of the various species examined.