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A review of fission product release theory is presented in support of fuel-failure monitoring analysis for the characterization and location of defective fuel. This work is used to describe: (i) the development of the steady-state Visual_DETECT code for coolant activity analysis to characterize failures in the core and the amount of tramp uranium; (ii) a generalization of this model in the STAR code for prediction of the time-dependent release of iodine and noble gas fission products to the coolant during reactor start-up, steady-state, shutdown, and bundle-shifting manoeuvres; (iii) an extension of the model to account for the release of fission products that are delayed-neutron precursors for assessment of fuel-failure location; and (iv) a simplification of the steady-state model to assess the methodology proposed by WANO for a fuel reliability indicator for water-cooled reactors. [ABSTRACT FROM AUTHOR]

This paper documents the extension of the Fuel And Sheath modeling Tool (FAST) for modeling transient conditions and presents a proof of concept validation exercise. This validation compares the predictions of FAST and ELESTRES/ELOCA fuel performance codes against experimental measurements from a simulated loss of coolant test conducted at Chalk River Laboratories. The comparison includes in reactor measurements of fuel temperature and internal gas pressure along with post irradiation (hot-cell) measurements of cladding deformation and oxidation. The results of this comparison show that the single pellet version of the FAST code was insufficient to fully account for strong axial dependence along an element. This was addressed by modifying the FAST code geometry to allow an arbitrary number of pellets in the fuel-stack. This full element version demonstrated improved agreement with measured cladding deformation, including circumferential ridging effects. [ABSTRACT FROM AUTHOR]

The Fuel And Sheath modeling Tool (FAST) is a general purpose nuclear fuel performance code. FAST includes models for heat generation and transport, thermal expansion, elastic strain, densification, fission product swelling, cracked pellet, contact, grain growth, fission gas release, gas and coolant pressure and sheath creep. The equations are solved on a two-dimensional (radial-axial) geometry of a fuel pellet and sheath using the Comsol Multiphysics finite-element platform. This paper presents the FAST code for normal operating conditions and results of the proof-of-concept testing against the ELESIM and ELESTRES-IST fuel codes as well as experimental data from seven irradiated fuel elements. In these seven cases, all of the codes were found to under-predict the measured average sheath strains. However, the FAST code was found to under-predict the mid-pellet sheath strains by a smaller margin than the other two codes. A larger data set is required to assess relative accuracy of the codes. [ABSTRACT FROM AUTHOR]

A computer code (U W B 1 ) that is applicable for both light and heavy water reactor fuel types is being developed at the University of West Bohemia. CANDU fuel geometry was investigated in this paper. Fuel depletion with U W B 1 code was performed and compared to other depletion codes. A feasibility study on the use of burnable absorbers in CANDU to improve operating margins was recently published [1]. The use of slight enriched fuel (SEU) with various burnable absorbers in CANDU has also been considered for future fuel designs. U W B 1 is a fast-calculating depletion code. It is being tested and compared extensively to other six different calculation codes, which include both deterministic and probabilistic types. Although the work focuses primarily on the progress of the multiplication factor during fuel depletion, fuel inventory details were also considered. Effective cross sections as the derivation of reaction rates that directly influence Bateman burnup equations were also compared. Further improvement on the U W B 1 code is ongoing. The results showed that U W B 1 nuclear fuel depletion code is capable to calculate fuel depletion of CANDU fuel bundle with and without burnable absorbers. As expected, the accuracy of the U W B 1 code is lower than the other codes. This paper recommends that U W B 1 code be considered as a tool for supporting fuel design assessment, since discrepancies are comparable and the code is one to two orders of magnitude faster than other depletion codes. [ABSTRACT FROM AUTHOR]