Evaluation of the irradiation-averaged fission yield for burnup determination in spent fuel assays

In order to derive the burnup of spent nuclear fuel from the concentration of selected fission products (typically the Nd isotopes and 137Cs), their irradiation-averaged fission yields need to be known with sufficient accuracy, as they evolve with the changes in the actinide vector over the irradiation history. To obtain irradiation-averaged values, radiochemists often resort to robust generic methods – i.e., based on simple mathematical relations – that weight the fission yields according to the actinides contributing to fission, without performing core physics calculations. In order to assess the performance of those generic methods, a database of about 30 000 spent nuclear fuel inventories has been constructed from neutron transport and depletion simulations, covering a representative range of fuel enrichment, burnup, assembly designs and reactor types. When testing several existing methods for effective fission yield calculation, some inaccuracies were identified, originating from improper one-group cross-section parameters that do not accurately reflect resonance and self-shielding effects, and too crude approximations in the estimation of the actinide concentration evolution. Revised effective fission and absorption cross-section parameters are then proposed here, as a first improvement to the earlier burnup determination methods. As a second step, a novel method is proposed that reduces the error on their radiation-averaged fission yield values, and hence on burnup, while retaining a straightforward calculation scheme

Understanding the Corrosion Behaviour of Used Mixed Oxide (MOX) Fuels: Insights from Post-leaching Characterisation

The disposal of spent nuclear fuels (SNF) in a deep geological repository (DGR) is regarded as the best practical waste management option in many countries. The long-term safety of a DGR over an assessment time frame of up to one million years necessitates a comprehensive understanding of the corrosion behaviour of SNF once the waste canister is breached and groundwater comes into contact. Although various studies have addressed this topic in the last decades, some of the processes contributing to the (radiolytic) matrix corrosion of SNF in the generally reducing repository environment are not fully understood. Furthermore, only limited efforts were deployed to study the corrosion behaviour of irradiated MOX fuels. To examine the effects of environmental conditions on SNF corrosion, the SF-ALE project (Spent Fuel Autoclave Leaching Experiments) was started as a collaboration between the Belgian Nuclear Research Centre SCK CEN and the Forschungszentrum Jülich GmbH. Within SF-ALE, irradiated MOX fuel rod segments (burn-up between 29&nbsp;GWd/tHM and 52&nbsp;GWd/tHM) were leached in bicarbonate water at neutral pH and in synthetic cementitious water at pH 13.5 under reducing atmosphere (4 vol% H2 in Ar at 40 bars pressure) in order to assess the release of various radionuclides and fission gases over a timeframe of 3.5 years. Following the leaching phase, a post-leaching characterisation of the fuel rod segments was initiated. Scanning electron microscopy analyses revealed that the structure of the MOX fuel matrix was affected differently by the exposure to the varied environmental conditions. Furthermore, secondary phases and alteration products were observed on the fuel surfaces. This contribution introduces initial results of the post-leaching characterisation and their implications to the general understanding of the corrosion behaviour of SNF under repository relevant conditions.</p

First Phase of the Spent Fuel Autoclave Leaching Experiments (SF-ALE) at SCK-CEN

Deep geological disposal of high level nuclearwaste, such as spent nuclear fuel is being considered byseveral countries, including Belgium and Germany.Ultimately, such a scenario considers contact betweenunderground water and the fuel due to aging andcorrosion of the materials. The SF-ALE research programcontributes to a better understanding of the release ofradionuclides from the fuel as a function of thesurrounding environment. An improved autoclave setupwas developed, allowing to simultaneously investigate therelease of water-soluble and gaseous components fromthe fuel. Six experiments (three on irradiated UO2 andthree on irradiated MOX fuel samples) have recently beenstarted and will run over 1.5 years, followed by anextension of another 1.5 years. A description of the hotcell equipment, analytical techniques used and themicrostructural characterization of the irradiated fuelsamples will be presented

Understanding the corrosion behaviour of used mixed oxide fuels under repository relevant conditions: Initial results of autoclave leaching experiments

Abstract Disposal in a deep geological repository (DGR) based on a multibarrier concept is considered as the safest and most sustainable option for the management of spent nuclear fuels (SNF) in many countries. Demonstrating the long-term safety of a DGR for SNF over assessment time frames of up to one million years requires a profound understanding of the corrosion behaviour of SNF coming into contact with groundwater, when the waste canister is eventually breached. During the last decades, a number of studies addressed this topic, leading to a good phenomenological understanding of the long-term behaviour of SNF in a DGR. However, various processes contributing to the (radiolytic) matrix corrosion of SNF in the generally reducing repository environment are still not fully understood, and corrosion data on used mixed oxide (MOX) fuels are scarce to date. To investigate the effects of environmental conditions on the corrosion of SNF, the SF-ALE project (Spent Fuel Autoclave Leaching Experiments) was initiated as collaboration between the Belgian Nuclear Research Centre SCK&nbsp;CEN, Forschungszentrum J&uuml;lich GmbH, and the Belgian Agency for Radioactive Waste and Fissile Materials (NIRAS/ONDRAF). Within SF-ALE, the release of various radionuclides and fission gases from well-characterised, irradiated MOX fuel segments (burn-up between 29&nbsp;GWd/tHM and 52&nbsp;GWd/tHM) exposed to bicarbonate water as reference groundwater at neutral pH and a synthetic cementitious water at pH&nbsp;13.5 under reducing atmosphere (4&nbsp;vol% H2 in Ar at 40&nbsp;bars pressure) is investigated. The first project phase lasted for about two years and was completed in September 2020. The leaching solutions were renewed and a second leaching phase was started on the same MOX fuel segments, aiming for a total duration of about 3.5&nbsp;years. In this contribution, we report the results obtained during the first phase and describe the transition to phase two of the project.</p

Corrosion behaviour of irradiated mixed oxide (MOX) fuels: Effects of environmental conditions on radionuclide leaching

In the context of spent nuclear fuel (SNF) disposition, a deep geological repository (DGR) based on a multi barrier concept is considered in many countries as the safest and most sustainable disposal option. The licensing of such a DGR is challenging and demands amongst others, safety assessments considering time frames of up to one million years. Moreover, a profound understanding of the corrosion behaviour of SNF coming into contact with groundwater, when the waste canisters are eventually breached, is crucial. During the last decades, a number of studies addressed this topic, leading to a good phenomenological understanding of the long-term behaviour of SNF in a DGR. However, these studies addressed mainly uranium dioxide-based fuels, and various processes contributing to the (radiolytic) matrix corrosion of SNF in the generally reducing repository environment are still not fully understood.Since corrosion data on irradiated mixed oxide (MOX) fuels under reducing conditions are scarce to date, the SF‑ALE project (Spent Fuel Autoclave Leaching Experiments) was initiated. MOX fuel with a well-known irradiation history was characterised by optical microscopy, SEM and EPMA before leaching experiments were preformed. The latter aim at elucidating the impact of the environmental conditions on SNF corrosion, addressing the instant release of radionuclides as well as the (long-term) matrix corrosion. Three cladded fuel rod segments with burn-ups ranging between 29 GWd/tHM and 52 GWd/tHM were selected for leaching experiments and were exposed to bicarbonate water as reference groundwater at neutral pH, and a synthetic cementitious water (pH 13.5), to address repository conditions for disposal concepts with different engineered barrier systems. Since repository conditions are expected to be reducing as a consequence of hydrogen generation due to anoxic corrosion of metallic waste canisters, the autoclave leaching experiments, which lasted for about 3.5 years, were performed under a gas phase consisting of an argon/hydrogen mixture at a pressure of 40 bar. During the leaching experiments, both the leachates and the gas phases were regularly sampled in order to determine the release of more than 30 relevant radionuclides from the irradiated MOX by various analytical techniques.Results regarding the initial release fractions of the fission products caesium and iodine observed within the first two years of leaching were published recently [1]. In this contribution, we report results on the release behaviour of further selected fission products and actinides obtained throughout the whole duration of SF-ALE.References[1] Schreinemachers, C., Leinders, G., Mennecart, T. et al. "Caesium and iodine release from spent mixed oxide fuels under repository relevant conditions: Initial leaching results". MRS Advances (2022). DOI: 10.1557/s43580-022-00220-

Caesium and iodine release from spent mixed oxide fuels under repository relevant conditions: Initial leaching results

Autoclave leaching experiments are conducted on three well-characterised, irradiated, and cladded mixed oxide fuel-rod segments with burnups ranging from 29 GWd/tHM to 52 GWd/tHM to investigate the instant release fraction of fssion gases and long-lived fssion products and to assess the long-term fuel matrix corrosion. The segments are exposed to bicarbonate solutions as reference groundwater at neutral pH and a synthetic young cementitious water at pH 13.5 under reducing atmosphere (4 vol% H2 in Ar at 40 bar pressure), since 2018. The initial leaching results for the fssion products caesium and iodine as representative elements of the instant release fraction were found to depend on the leachate composition as well as on the fuel burnup

Leaching Experiments on UOX and MOX Spent Fuel: Results and Prospects of the SF-ALE Project at SCK CEN

To investigate the effects of the environmental conditions on the dissolution of irradiated nuclear fuels, the Spent Fuel Autoclave Leaching Experiments (SF-ALE) project is being carried out at SCK CEN, in collaboration with Forschungszentrum Jülich and the Belgian Agency for Radioactive Waste and Fissile Materials NIRAS/ONDRAF. A state-of-the-art spent fuel leaching setup was developed to examine the release of various radionuclides and fission gases from well-characterized UOX and MOX spent fuel segments, under different environmental conditions [1]. The first phase of the project included 733 contact days (approximately two years), and was completed in September 2020 by the complete renewal of the leaching solution. This also marked the start of the second phase of the project, in which continued leaching experiments will expand the total duration to 1279 contact days (approximately 3.5 years) on the same spent fuel segments. In the meantime, leaching data of the first phase have been acquired and processed. In the present contribution we will discuss and interpret the leaching results obtained during the first phase, and describe the transition to phase two of the project and its prospected outcomes