Fuel matrix

Spent Fuel Treatment. Spent fuel assembhes from nuclear power reactors are highly radioactive because they contain fission products. Relatively few options are available for the treatment of spent fuel. The tubes and the fuel matrix provide considerable containment against attack and release of nucHdes. To minimi2e the volume of spent fuel that must be shipped or disposed of, consoHdation of rods in assembhes into compact bundles of fuel rods has been successfully tested. Alternatively, intact assembhes can be encased in metal containers. 

Fuel matrix - Fission products, bound in a ceramic matrix, may escape only by slow diffusion or melting of the matrix. 

The most significant use for HTPB propints is in ballistic missiles (Refs 12 13). The most unique usage is in the functioning of a laser by means of the chemical exhaust species-generated by a HTPB-Xmm Perchlorate fuel matrix (Ref 9)... 

Some petroleum products, especially those containing higher-molecular-weight compounds such as waxes, do not crystallize rapidly when cooled. Instead, they form a gel-like network throughout the fuel matrix. This network can begin forming at temperatures well above the pour point of a fuel and may render the product unpumpable. 

Further cooling of the fuel leads to wax crystal formation throughout the fuel matrix. The growing wax crystals develop into a larger latticelike network encompassing the bulk fuel volume. This latticelike network eventually causes the fuel to become highly viscous and to eventually gel into a semisolid mass. The lowest temperature at which fuel remains in the liquid state just prior to gellation is called the pour point. 

Dehazing refers to removal of water which has been dispersed throughout a fuel matrix. Haze created by water droplets can be removed by coalescence or by further dispersion through the action of a chemical dehazer. 

Less is known about the fate of the fission and neutron capture products that could result in the precipitation of unique alteration phases depending on the availability of these species in the fuel matrix. Burns et al. (1997) theorized that many of the U(VI) alteration phases may be capable of incorporating the long-lived radiotoxic isotopes, including 237Np, 99Tc, and 239Pu. In this chapter, we will discuss the evidence for Np incorporation into U(VI) phases and the behaviour of Pu in corroded spent nuclear fuel (SNF). 

Fission products, actinides, and lanthanides that are retained in the fuel matrix ... 

In SNF corrosion tests, there has been a tendency to use the release of more soluble species Tc, Cs, and Mo as markers for fuel corrosion (Finn et al. 2002). As none of these elements are present in the U02 matrix, this approach may not reveal the actual fuel matrix corrosion rate. Furthermore, short-term leaching tests may not expose possible diffusion-limited (tl/2) release of gap and grain boundary species and assume excessive rates of reaction based on initial fast release rates. The microstructure, radiation field, and composition will change over time, so that tests carried out on fuel today may not be relevant to fuel behaviour 300 to 1000 years from now, once the high p-,y-field has decayed. 

Indications from both microscopic analyses of metallic particles from corrosion tests and evidence from the Oklo natural reactors indicate that performance assessment calculations should not assume 99Tc is easily mobilized. It is entirely inappropriate to use "Tc release as a marker for fuel corrosion because Tc is not located in the fuel matrix. The TEM examinations of corroded e-particles have shown that Mo is preferentially leached from these phases, a behaviour that is similar to the one observed at Oklo. It is interesting to note that laboratory dissolution of billion-year old 4d-metallic particles for a chemical analysis required a corrosive mix of peroxide and acid (Hidaka Holliger 1998) similar to the experience at SNF reprocessing plants. It is doubtful that the oxidation potential at the surface of an aged fuel will be sufficient to move Tc(0) from the e-metal particles. 

Two master variables, pH and pe, can be used to define the limits of stability of the UO2 spent fuel matrix under repository conditions. The Pourbaix diagram depicted in Fig. 2 clearly indicates the stability space of UO2 and consequently the desired chemical conditions that ensure the correct performance of the waste matrix. 

Initially, Oz diffuses through the bentonite and granitic domains, controlling the redox state of the system. Once 02 is exhausted, granitic groundwater controls the redox state of the system. The results of these calculations performed with the PHREEQC geochemical code (Parkhust Appelo 1999) clearly indicate that there is a substantial variability in pH/pe space along the temporal and spatial evolution of the near field of a repositoiy. This has clear consequences for the subsequent interactions with the Fe canister material and finally with the spent fuel matrix. 

The fluids that have evolved as a result of the bentonite-groundwater interactions will contact the canister on their travel towards the spent fuel matrix. Most of the proposed canister materials in different countries have in common the presence of Fe in the system, either as cast iron (Sweden, Finland) or as stainless steel (France, Spain). While the bentonite-groundwater processes have... 

The spent fuel matrix is a ceramic material with a fascinating chemical composition and a large degree of phase heterogeneity. The physical state and chemical composition of spent fuel largely depends on the bum-up of the fuel once it is taken out of the reactor. In Fig. 6 we indicate the dependence of the chemical composition on the bum-up for a series of PWR fuels. However, the fact that remains constant is that U02 constitutes the major component of spent fuel, ranging within a total of 95-98% in weight (see Fig. 7). 

Oxidation of the fuel matrix and other redox sensitive Rn... 

Fig. 11. (a) Thermodynamic reaction pathway for the initial oxidative alteration of the spent fuel matrix at pH 8, calculated by using the PHREEQC code (adapted from Bruno etaL 1995). (b) Thermodynamic reaction pathway for the alteration of schoepite in granitic/bentonite groundwater at pH 8, calculated by using the PHREEQC code (adapted from Bruno et al. 1995 with permission). 

Bruno, J., Casas, I., Cera, E., Swing, R. C., Finch, R. C. Werme, L. O. 1995. The assessment of the long-term evolution of the spent nuclear fuel matrix by kinetic, thermodynamic and spectroscopic studies of uranium minerals. Materials Research Society Symposium Proceedings, 353, 633-639. 

Sebacic acid is used as part of the fuel matrix in polyamide based solid propints (Ref 3) and polyurethane-Amm perchlorate solid propints (Refs 3a, 4 5)... 

Finely divided oxidizers dispersed in fuel matrix. 

One of the key processes here is the dissolution of the spent nuclear fuel matrix in groundwater liberating radioactive fission products and actinides. Without this process no radioactivity will be released to the biosphere. 

Upon contact between the spent nuclear fuel and the groundwater, radiolysis of water will begin. From the point of view of a safety assessment it is relevant to define the worst-case, but still realistic, scenario. Hence, the maximum possible dissolution rate for the UOj fuel matrix (assuming congruent dissolution) must be defined. As shown above, oxidation of U(IV) to U(VI) is required for dissolution to occur. Consequently, the rate of dissolution can never exceed the rate of oxidation and the rate of oxidation can be used to estimate the maximum dissolution rate. It has also been shown that, for longer irradiation times, the only oxidant that must be taken into account is H2O2 and that the surface concentration of H2O2 approaches the steady-state concentration fairly rapidly. The concentration will never exceed the steady-state concentration and therefore we can use the steady-state approach to make a realistic estimate of the maximum dissolution rate. 

Most models assessing the long-term behaviour of Cs and Sr from fallout include processes relevant to LMM ionic species only. However, radionuclides may be associated with particles due to (a) release of fuel matrix or clusters, (b) condensation of volatiles on available particle surfaces after the release of (c) interactions with aerosol particulates during atmospheric transport. For volatile radionuclides (e.g., Cs, °Sr), all three mechanisms may be equally relevant, while the deposition of nuclides of refractory elements (e.g., " Ce, Nb, Zr) indicates the release of fuel particles. For fuel particles, depletion of volatiles relative to refractory elements would be expected to depend on the temperature reached during the releases, whereas the activity ratios for refractory elements should reflect the reactor fuel bum-up. 

The mineral elements can be held in the coal substance as organo-metallic salts, and also as a result of molecular adsorption and co-valent bonding. The mineral species dissolved in coal pore water, chiefly chlorides can also be considered as part of the inherent matter. The lignites and sub-bituminous coals can have a high fraction of the mineral elements, chiefly sodium, calcium and also aluminium and iron chemically combined in the fuel substance (9,10). The chemical reactivity and porosity of the fuel matrix decrease with the increase of coal age from lignite to bituminous rank. The loss of carboxyl, hydroxyl and quinone bonding sites in the fuel matrix results in a low "chemical" mineral matter content of bituminous coals. 

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