Influence of interface‐related parameters on some heat stable salts in the corrosion studies of carbon steel for post‐combustion CO
            2
            capture

ABSTRACTIn previous corrosion studies, carbon steels, especially the fine-grained steel TStE355, were identified as promising materials for heat-generating nuclear waste containers acting as a barrier in a rock-salt repository. In the present study detailed investigations have been performed on fine-grained steel to determine the influence of important parameters on its corrosion behaviour in disposal-relevant salt brines. These parameters are: brine composition (Mg Cl2-rich and NaCI-rich brines), temperature (90°C, 170°C), and salt impurities, such as H2S concentrations of 25 mg/I-200 mg/I salt brine.Under the conditions of the tests used here, carbon steel was subjected to general corrosion. Pitting and crevice corrosion or stress-corrosion cracking were not observed. The increase in temperature from 90°C to 170°C strongly enhanced the corrosion rate of the steel. In the MgCl2-rich brines, considerably higher rates (37-70 μm/a at 90°C, 200-300 °m/a at 170°C) were observed than in the NaCI-rich brine (5 μm/a at 90°C, 46 μm/a at 170°C). H2S concentrations in the MgCl2-rich Qbrine of up to 200 mg/l did not influence significantly the corrosion rate of the steel. The corrosion rates determined imply corrosion allowances that are technically acceptable for thick-walled containers. In view of these results, fine-grained steel continues to be considered as a promising material for long-lived HLW containers.

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Corrosion behavior of spent nuclear fuel in high pH solutions – Effect of hydrogen

MRS Proceedings ◽

10.1557/opl.2012.564 ◽

2012 ◽

Vol 1475 ◽

Cited By ~ 6

Author(s):

A. Loida ◽

R. Gens ◽

C. Bube ◽

K. Lemmens ◽

C. Cachoir ◽

...

Keyword(s):

Detection Limit ◽

Carbon Steel ◽

Nuclear Fuel ◽

Corrosion Behavior ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Detection Limits ◽

Type Solution ◽

Corrosion Studies ◽

Fractional Release

ABSTRACTIn accordance with the Belgian “supercontainer design”, spent nuclear fuel (SNF) will be encapsulated in carbon steel canisters, surrounded by a concrete overpack for disposal in poorly-indurated clay. After re-saturation of the barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Corrosion studies of SNF in ECW-type solution (Evolved Cement Water) and YCWCa-type solution (Young Cement Water with Ca) were performed under externally applied H2 overpressures over 426 days. Directly after H2 application, Tc concentrations decreased from >10-8 M to concentrations below detection limit. Based on the fractional release of selected fission products, low matrix dissolution rates of ~10-8/day were found in both experiments. U concentrations decreased finally to 1.5•10-9 M (YCWCa) and to 2.1•10-10 M (ECW), respectively. Am, Np and Pu concentrations were found throughout the experiments below their detection limits indicating an effective retention process.

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