Analysis of Reactor Fuel Rod Behavior

Uffelen, P. Van; Konings, R.J.M.; Vitanza, C.; Tulenko, J.S.

doi:10.1007/978-0-387-98149-9_13

Cited by 14 publications

(8 citation statements)

References 115 publications

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“…Later, the tetragonal phase becomes unstable, and the oxide changes to a monoclinic form. At this stage, the corrosion layer shows some porosity, consequently, only a portion of the oxide layer remains protective, and the corrosion is controlled by diffusion through the dense protective layer only [14].…”

Section: Corrosion Of Zirconium-based Alloys (Zy)mentioning

confidence: 99%

Evaluation of corrosion on the fuel performance of stainless steel cladding

Gomes

Abe

Silva

et al. 2016

EPJ Nuclear Sci. Technol.

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Abstract. In nuclear reactors, the use of stainless steel (SS) as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4) under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

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Section: Corrosion Of Zirconium-based Alloys (Zy)mentioning

confidence: 99%

Evaluation of corrosion on the fuel performance of stainless steel cladding

Gomes

Abe

Silva

et al. 2016

EPJ Nuclear Sci. Technol.

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“…RADAR [26] [27] is then explicitly time stepped forward (sub-cycled with respect to the larger time step of the implicit thermomechanical simulation), the RADAR models grid was kept much finer than the rest of the model, to reduce error, when linear projection was used to transfer the scaled volumetric power from RADAR to the rest of the model. The fuel burn-up serves as an important parameter for fission gas generation, fuel swelling and the changes in fuel thermal conductivity.…”

Section: Radar Modelmentioning

confidence: 99%

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

Mella

Wenman

2013

Journal of Nuclear Materials

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Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGR) are modelled in the ABAQUS finite element (FE) code. User supplied subroutines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilization of ABAQUS's well-developed pre-and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermomechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the viscoplastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. Keywords: Fuel Performance, ABAQUS, Finite Element, AGR Introduction Pellet-cladding interactionNumerous authors have investigated the phenomenon of PCI in the past [1][2][3][4][5][6][7][8][9][10]. These studies have included the AGR system. The AGR studies from the early 90s offer insight into the effective bonding forces between AGR cladding and fuel, albeit that these studies were for a very simplified geometry and material properties. Walker's models were two dimension r-T segments of a pellet and cladding in a coupled thermo-mechanical transient analysis. This analysis showed that the expected bond between pellet and clad will fail and cause the debonding of about 6% of the pellet-clad surface producing sufficient stress relief to prevent further debonding. Walker also shows that a failure of this size was sufficient to cause a two orders of magnitude reduction in the number of pellet cracks to propagate in the adjacent cladding. Most work in recent years has focussed on water cooled systems [2][3][11][12][13]. Marchal [11] showed through 3D modelling of PCI, in the CAST3M FE code, that the while pellet fracture did produce stress concentrations in the cladding there was also a significant stress remaining in the fuel fragment. These pellet fragments had sufficient remaining stresses that later interaction with the cladding through friction contact produced further radial cracking. Marchal's smeared crack model emphasises the non-linear behaviour of PCI cracking and that it is strongly dependent on fuel-cladding friction. *Manuscript Click here ...

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“…The pellet diameter is in the order of 1 cm; larger pellets are used for fuel rods in Boiling Water Reactors than in Pressurized Water Reactors. The pellet height is in the range of 1 to 2 cm for fuels of both reactor types [48]. During nuclear reactor operation the UO 2 fuel is subject to intensive transformations due to the fission reactions, the consecutive neutron capture and decay reactions as well as temperature induced transformations (e.g.…”

Section: Alteration Of Spent Nuclear Fuel and Radionuclide Release Unmentioning

confidence: 99%

“…Power ramping and other thermal excursions during reactor operation cause a coarsening of the grain size, extensive microfracturing and migration of fission gases and volatile fission products to grain boundaries, fractures, and the "gap" between the periphery of the fuel pellet and the surrounding cladding. Due to the configuration of the neutron energy spectrum in LWR, there is a higher density of epithermal neutron resonance absorption in 238 U nuclei at the radial outer edge ("rim") of the UO 2 fuel pellet, which results in a local enrichment in fissile Pu via Np decay and thus in higher local fission density [48,55]. The local BU at the rim of the UO 2 pellet can be 2-3 times higher than the average pellet BU, depending on the specific irradiation conditions.…”

Section: (Radio-)chemical and Mineralogical Properties Of Spent Nuclementioning

confidence: 99%

“…The fission gas release (FGR) occurs by diffusion to grain boundaries, grain growth accompanied by grain boundary sweeping, gas bubble interlinkage and intersection of gas bubbles by cracks in the SNF [48,103]. It is more corre-lated to the linear heat rating, which depends on the fuel temperature, than to the BU of the SNF [102,104].…”

Section: Fast/instant Release Of Radionuclides From Spent Nuclear Fuelmentioning

confidence: 99%

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Radionuclide behaviour in the near-field of a geological repository for spent nuclear fuel

et al. 2012

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UO 2 fuel corrosion / Radionuclide release / Instant release fraction / Hydrogen effect / Coprecipitation / SorptionSummary. Even though chemical processes related to the corrosion of spent nuclear fuel in a deep geological repository are of complex nature, knowledge on underlying mechanisms has very much improved over the last years. As a major result of numerous studies it turns out that alteration of irradiated fuel is significantly inhibited under the strongly reducing conditions induced by container corrosion and consecutive H 2 production. In contrast to earlier results, radiolysis driven fuel corrosion and oxidative dissolution appears to be less relevant for most repository concepts. The protective hydrogen effect on corrosion of irradiated fuel has been evidenced in many experiments. Still, open questions remain related to the exact mechanism and the impact of potentially interfering naturally occurring groundwater trace components. Container corrosion products are known to offer considerable reactive surface area in addition to engineered buffer and backfill material. In combination, waste form, container corrosion products and backfill material represent strong barriers for radionuclide retention and retardation and thus attenuate radionuclide release from the repository near-field.

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Analysis of Reactor Fuel Rod Behavior

Cited by 14 publications

References 115 publications

Evaluation of corrosion on the fuel performance of stainless steel cladding

Evaluation of corrosion on the fuel performance of stainless steel cladding

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

Radionuclide behaviour in the near-field of a geological repository for spent nuclear fuel

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