Model development for the estimation of fission product release under normal and accident conditions in a HTGR

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This article has attempted to estimate the radioactivity release from fuel materials during normal and transient conditions by coupling the TRISO fracture and the fission product (FP) diffusion. Two calculation models, named TRISO Fracture Analyzer (TRIFA) and DIFfusion Analyzer (DIFA), are developed. TRIFA is initially used to calculate the fraction of fractured fuel particles, thus determining the amount of fission gas release. The obtained particle fracture function is then used as input for the diffusion rate calculation. DIFA simulates with a single spherical fuel element, a pebble, irradiated under normal and accident conditions. It describes the diffusive transport of fission products by numerically solving the diffusion equation. The finite difference method is applied to obtain fission product release rates from a pebble to coolant. The model comparisons show that the new developed models are reliable, fast, and correspond with previous results of other models. As for HTR-10, the coupled models, TRIFA and DIFA, are applied to calculate the level of fission product release after accidents. The following conclusions can be drawn. First, the mitigation should be carried out until the maximum fuel temperature reaches under transient. Second, the mitigation should be intensively considered if the burn-up exceeds 5%FIMA (similar to 48 GWd/MTU) when transient happens. Additionally, it is found that there is the threshold burn-up where the rapid FP release occurs due to the numerous TRISOs fractured. Further investigations are needed to extend the use of the method developed in this work to the safety assessments for high-temperature gas-cooled reactors (HTGRs). This article will hopefully serve as a platform for designing the advanced TRISO that can minimize the activity release, and providing the rationale of development of the intensive accident mitigation system in future. (C) 2009 Elsevier B.V. All rights reserved.
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NUCLEAR ENGINEERING AND DESIGN, v.239, no.6, pp.1066 - 1075

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NE-Journal Papers(저널논문)
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