A model for release of fission products from a breached fuel plate under wet storage (original) (raw)
Related papers
Journal of Nuclear Materials, 2016
During reactor operation the fission gases Kr and Xe are formed within the UO 2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO 2 fuel itself are widely used as indicators for the release properties of 129 I, 137 Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H 2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.
CORROSION, 2018
Models for the corrosion of spent nuclear fuel (fission and actinide-doped uranium dioxide) provide the essential source term for the release of radionuclides from within a failed nuclear waste container in a deep geologic repository. Although redox conditions within a repository are expected to be anoxic, exposure of the fuel to groundwater will cause the generation of oxidants at the fuel surface, leading to its corrosion and the release of radionuclides. The influence of these oxidants will be partially mitigated by the anoxic corrosion of the inner walls of the steel container to produce the oxidant scavengers, Fe 2+ and H 2. This review summarizes the development of a finite element model developed to determine the influence of the various redox-controlling species (H 2 O 2 , Fe 2+ , H 2). Both one-dimensional and two-dimensional models are described, with the latter required to take into account the fractured geometry of the fuel.
2009 IEEE 13th International Multitopic Conference, 2009
Source term evaluation for a typical MTR type system has been carried out. For this purpose, a Matlab based program has been developed which used the Origen-2 code as a subprogram. The isotope inventory of the core of a typical MTR system has been done following a 180 days of full-power operation. The case of flow blockage accident with instantaneous release of radioactivity to containment air has been studied. The resulting source term for noble gases, for iodine, and for aerosols has been carried out for normal, emergency and for isolation states of the containment. The dependence of source term on values of exhaust rates has been studied and a typical trend of initial increase followed by approach to saturation value has been observed in all cases. The dependence of the containment retention factor on various radionuclides and on exhaust rates has also been studied in this work.
2010
The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged light water reactor fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and mixed oxide (MOX) fuels. This paper discusses the synthesis of these findings in the MELCOR severe accident code.
Limitations on Radionuclide Release From Partially Failed Containers
Materials Research Society Symposium Proceedings, 2008
In the long run, nuclear waste packages at the Yucca Mountain repository are likely to evolve into a combination of corroded materials mixed with relicts of intact Alloy-22 and other waste package materials. Different rates of corrosion, due to physical and chemical disturbances in the environment of the repository, will lead to different times of penetration between waste packages and at different locations on the same waste package. Radionuclides are released from waste packages by dissolution and transport in water. In this paper, we shed some light on the effect of residual heat release, and other physical processes that take place in the waste package during penetration times, on radionuclide release. We develop a flow-through conceptual model for a probable serious failure in which multiple penetrations allow water to flow through a partially failed waste container. This model demonstrates that evaporation at hotter protected areas creates a capillary pressure gradient that causes water to flow with its dissolved and suspended contents toward these relict protected areas, effectively preventing radionuclide release. We derive a dimensionless group to estimates the minimum size of the covered areas required to sequester radionuclides and prevent release, and explore the implication of the flow-through model on the Yucca Mountain repository performance.
Nuclear Fuel in a Reactor Accident
Science, 2012
Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.
Modelling of fission-product transport in the reactor coolant system
Annals of Nuclear Energy, 2013
The Phébus fission product (FP) programme studies the phenomenology of severe accidents in watercooled nuclear reactors. Five tests were performed in the frame of the programme covering fuel-rod degradation and FP behaviour released via the coolant system into the containment. To model FP transport and behaviour in the coolant system, numerous physical and chemical phenomena have to be taken into account. In the vapour phase, for example, FP speciation, vapour condensation and vapour/surface or vapour/aerosol reactions have to be considered. The aerosol phase has to be modelled with nucleation, growth and deposition processes. Finally, remobilisation phenomena like resuspension and revaporisation have to be taken into account for delayed release into the containment. Four Phébus FP tests (FTP0, FPT1, FPT2, FPT3) have been modelled with the ASTEC/SOPHAEROS code. Modelling shows an overall good estimation of retention for the main FPs (e.g., I, Cs, Mo). Furthermore, a strong connection is revealed in the gaseous phase chemistry between I, Cs, Cd and Mo which has a great impact on gaseous iodine release into the containment. The Phébus FP test modelling also exposes disagreement on FP retention when laminar gaseous flow is not well developed. Finally, probably the most significant shortcoming in modelling that Phébus-FP tests highlighted concerns vapour-phase iodine-chemistry modelling at low temperature. The study of this latter point is on going with the experimental programme ISTP/CHIP.
Journal of Nuclear Materials, 2017
The instant release of fission products from high burn-up UO 2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45 e63 GWd/t HM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride e bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H 2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burnup samples, fissures still provide possible preferential transport pathways.