Intra-nodal Study for the Mixed LEU-MOX Cores

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Title: Intra-nodal Study for the Mixed LEU-MOX Cores
Author: Jiang, Qunlei
Advisors: Dmitriy Y. Anistratov, Committee Co-Chair
Paul J. Turinsky, Committee Chair
Zhilin Li, Committee Member
Abstract: One favored method being considered for the disposal of surplus weapons grade plutonium (WGPu) is to burn the WGPu as mixed oxide (MOX) fuel in commercial existing Light Water Reactors (LWRs). Duke Power Company intends to irradiate MOX fuel assemblies in their four Westinghouse pressurized water reactors (PWR). The introduction of MOX fuel into LWRs poses several challenges for the reactor physics analysis. The difference in properties of uranium and plutonium induces neutron energy spectrum difference between the MOX and LEU assemblies, which creates a large thermal flux gradient at the interface between these assemblies. Current methods for predicting the intra-nodal flux distribution have difficulty to model this gradient. This study is focused on improving the fidelity of the core simulator utilized in FORMOSA-P to model mixed LEU-MOX cores. In particular, the nature of challenge in regard to accurately model the LEU-MOX interfaces due to both strong spatial variations of the thermal flux and energy spectra, the later impacting the two-group cross section values, will be assessed. The specific focus is on pin-wise power reconstruction; however, issues related to the nodal solution will also be assessed. To complete the work on pin-wise power reconstruction, there are three ways to improve the prediction accuracy, those being to improve the prediction accuracy of the intra-nodal flux shape, improve the prediction accuracy of the intra-nodal kappa-sigma-fission shape, and to introduce group power form factors. However, since the intra-nodal flux and kappa-sigma-fission both are predicted using results obtained from the nodal solution, the prediction accuracy of the nodal solution for mixed LEU-MOX cores enters. This study is completed by using HELIOS, a transport theory based lattice physics code, and NESTLE, a diffusion theory core simulator. The single assembly (SA) calculation is done by HELIOS to generate the homogenized cross sections, discontinuity factors (DFs) at corner points and surfaces, and power form factors using infinite-medium spectra. Homogenized cross sections and surface average DFs are generally used in diffusion calculations. Obviously, they are not that accurate because the effect from neighbors is ignored in SA calculations, hence introduce errors in diffusion calculations. The colorset (CS) calculation done by HELIOS is to generate the same information but now accounting for LEU-MOX interface effects, where a colorset denotes an LEU-MOX assembly infinite checkerboard loading. The intra-nodal flux distribution is obtained by a SA NESTLE calculation using the finite difference method with a very fine spatial mesh. The surface current boundary condition imposed is obtained from the CS HELIOS calculation, with the NESTLE calculation completed using SA HELIOS determined homogenized cross sections or CS HELIOS determined intra-nodal cross sections. The shape of intra-nodal cross section shows that the thermal group 'flat' cross section can not represent the interfacial effect. This assumption no longer works for mixed LEU-MOX cores. The group dependent 'flat' cross sections contribute errors to the , intra-nodal fluxes and powers. Comparing the cross sections, DFs, and power form factors from SA lattice physics, CS lattice physics, and diffusion theory calculations, we can evaluate the errors induced by the MOX and LEU spectrum interactions on these values. Contrasting SA lattice physics and diffusion theory (using intra-nodal cross sections) predictions, for the SA lattice physics predictions about a 1.5% relative error in the LEU fuel assembly and 2.2% relative error in the MOX fuel assembly are observed in the thermal group surface averaged DFs. The relative errors in the fast group ADF ratios are small. However about a 2.7% relative error is observed in the thermal group ADFs' ratios at BOL. Likewise, about a 3% relative error in the LEU fuel assembly and 4.4% relative error in the MOX fuel assembly are observed in the thermal group corner point DFs at BOL. Also about a 1% relative error is observed in in both the LEU and MOX fuel assemblies at a higher level burnup (20 GWD/THM). With regard to the intra-nodal cross sections, stronger spatial dependences are noted for the down-scattering, thermal transport, fast and thermal absorption, and fast and thermal fission cross sections. A steep spatial gradient is noted for the intra-nodal flux due to interfacial effects. The resulting flux shape presents a difficulty problem to accurately functionalize. The study on the effect of ADF's and cross sections shows that errors in the cross sections, i.e. the energy and spatial spectrum used for generating the cross sections, are the main error contributors to the , node averaged fluxes and intra-nodal fluxes. ADFs do not greatly effect the value and node averaged fluxes. The results of the study imply that the accuracy of the SA lattice physics calculation to obtain cross sections, DFs and form factors are poor for the mixed LEU-MOX core simulations, in other words, the rehomogenization of cross sections is necessary during the nodal (diffusion theory) calculation for mixed LEU-MOX core simulations. Also, the employment of intra-nodal cross section is effective in improving the accuracy of diffusion calculation.
Date: 2004-07-08
Degree: MS
Discipline: Nuclear Engineering
URI: http://www.lib.ncsu.edu/resolver/1840.16/285


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