Reexamining the Dissolution of Spent Fuel: A Comparison of Different Methods for Calculating Rates

Hanson, Brady D.; Stout, R.B.

doi:10.1557/proc-824-cc2.4

Cited by 9 publications

(7 citation statements)

References 7 publications

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“…5 is that it is not congruently released with uranium (FIAP Pu / FIAP U = 0.07 ± 0.02), similarly to previous studies on spent fuel dissolution [4,13,[32][33][34].…”

Section: Plutoniumsupporting

confidence: 79%

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Serrano‐Purroy¹,

Clarens

Glatz³

et al. 2009

Radiochimica Acta

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The dissolution behaviour of powdered commercial spent fuel (UO 2 with burn-up of 53 MW/d kg U) has been studied in a carbonate-containing solution ([HCO 3 − ] = 0.001 mol dm −3 ) by using a flow-through reactor specially designed for the use in a hot cell. This method allows studying spent fuel dissolution while avoiding the parallel process of secondary solid phase formation. The dissolution behaviour of U, Np, Pu, Sr and Cs was studied. The main trend of the results obtained in this work is that only neptunium releases congruently with uranium (FIAP Np /FIAP U = 1.21 ± 0.01) because both strontium and caesium have higher FIAP values (FIAP Sr /FIAP U = 2.3 ± 0.8; FIAP Cs /FIAP U = 5 ± 1) and plutonium lower (FIAP Pu /FIAP U = 0.07 ± 0.02). The FIAP value for uranium at the steady-state is 4(±2) × 10 −4 .

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“…5 is that it is not congruently released with uranium (FIAP Pu / FIAP U = 0.07 ± 0.02), similarly to previous studies on spent fuel dissolution [4,13,[32][33][34].…”

Section: Plutoniumsupporting

confidence: 79%

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Serrano‐Purroy¹,

Clarens

Glatz³

et al. 2009

Radiochimica Acta

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“…The fraction of inventory of an element i released into the aqueous phase, FIAPi, was calculated according to equation 1 [43,44]: As explained above, two complete replenishments were done at the beginning of the experiment to avoid a likely precipitation of U. Nevertheless, these values have important information particularly about the IRF.…”

Section: Solution Analysis Resultsmentioning

confidence: 99%

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

González-Robles

Serrano‐Purroy

Sureda

et al. 2015

Journal of Nuclear Materials

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The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd·kgU

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“…fractures, gap and grain boundaries, makes it difficult to know the initial surface area. Moreover as the experimental time goes by, the dissolution produces changes in the surface that are very difficult to anticipate [14]. Being aware of this problematic, the normalisation using the specific surface is still a useful tool to compare fuel samples with different morphology.…”

Section: Instant Release Fraction (%)mentioning

confidence: 99%

Instant release fraction corrosion studies of commercial UO 2 BWR spent nuclear fuel

Martínez-Torrents

Serrano‐Purroy

Sureda

et al. 2017

Journal of Nuclear Materials

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The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel.The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

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Reexamining the Dissolution of Spent Fuel: A Comparison of Different Methods for Calculating Rates

Cited by 9 publications

References 7 publications

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Instant release fraction corrosion studies of commercial UO 2 BWR spent nuclear fuel

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