Detailed characterisation of the rim microstructure in PWR fuels in the burn-up range 40–67 GWd/tM

Spino, J.; Vennix, K.; Coquerelle, M.

doi:10.1016/0022-3115(96)00374-1

Cited by 172 publications

(97 citation statements)

References 26 publications

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“…Recently, the tendency of increasing fuel burn-up (BU) produces 239 Pu by neutron capture of 238 U generating an external layer with a higher local BU, which presents higher porosity and fuel grain subdivision, resulting on the formation of the so-called rim zone also known as high burn-up structure (HBS) [9,10]. This layer is observed for BU's higher than 40 MWd·kgU -1 [9][10][11][12][13][14][15][16] and is a function of BU and the irradiation temperature, being the temperature threshold 1100ºC [13,15]. As a conservative approach for those SNF's, the radionuclides located in the HBS were also included in the IRF [6,8].…”

Section: Introductionmentioning

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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Section: Introductionmentioning

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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“…Investigations of the rim microstructure in PWR fuels in the burn-up range 40 to 67 GW dt HM -1 were performed by Spino et al [16] and tabulated data on burn-up, radial position a) Photo of the surface of an irradiated UO 2 pellet b) β-radiography of the pellet surface c) a-radiography of the pellet surface Kienzler et al [13] the following data were obtained for ~50 GWd t HM -1 fuel: a rim porosity of 3.8% for r/r 0 = 0.9, 17% for r/r 0 = 1, showing an average of 6.2% between this two values. Based on fit parameters for a pellet with r 0 = 4.6 mm, Fig.…”

Section: Resultsmentioning

confidence: 99%

Simulation of alpha dose for predicting radiolytic species at the surface of spent nuclear fuel pellets

Becker

Kienzler

2014

Open Chemistry

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In many countries, spent nuclear fuel is considered as a waste form to be disposed of in underground disposal. Under deep host rock conditions, a reducing environment prevails. In the case of water contact, long-term radionuclide release from the fuel depends on dissolution processes of the UO 2 matrix. The dissolution rate of irradiated UO 2 is controlled by oxidizing processes facilitated by dissolved species formed by alpharadiolysis of water in contact with spent nuclear fuel. To understand the effect of the radiation, the information of the dose rate at the surface of the fuel and its proximity is needed. α particles contribute strongly due to their high linear energy transfer. However, their dose rate and the energy deposition at the fuel surface are difficult to measure. Cylindrical fuel pellets as used in fuel rods show specific features, such as the rim zone, where a higher Pu concentration and a different porosity of the fuel matrix is present. The a particle dose rate was determined by simulations with the code MCNPX with focus on the rim zone of a pellet. As a result a 40% increased dose level in the rim zone exists in comparison to the center of a pellet. The potential dominant and inhomogeneous α-dose distribution is supposed to have a strong impact on radiolysis phenomena and in turn on an inhomogeneous dissolution of elements over the surface.

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“…This layer is likely mobile fission products such as Cs or I, which have been observed in larger pores in oxide fuels. 7,8 Fig. 1.…”

Section: Experimental Approach and Datamentioning

confidence: 97%

Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

Teague

Fromm

Tonks

et al. 2014

JOM

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Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

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Detailed characterisation of the rim microstructure in PWR fuels in the burn-up range 40–67 GWd/tM

Cited by 172 publications

References 26 publications

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

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

Simulation of alpha dose for predicting radiolytic species at the surface of spent nuclear fuel pellets

Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

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