PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

Lemehov, S.; Nakamura, Jinichi; Suzuki, Motoe

doi:10.13182/nt01-a3166

Cited by 13 publications

(9 citation statements)

References 5 publications

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“…Lemehov et al 4) show smaller burnup spikes and their height does not decrease with the progress of burnup for the MOX fuel of average pellet enrichment, 5.3% and of Pu-rich agglomerate enrichment, 25%, which is not consistent with those of the present and Kameyama et al's 3) studies.…”

Section: Local Power Density and Burnupcontrasting

confidence: 74%

“…Lemehov et al 4) made the fuel irradiation behavior model PULTON, which makes burnup calculations of the Pu-rich agglomerate regions using radial shape functions for neutron flux distributions. Chang 5) calculated the infinite multiplication factor (k-inf) of unirradiated MOX fuel using the MCNP code in which a limited number of Pu-rich agglomerates are explicitly represented.…”

Section: )mentioning

confidence: 99%

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Effect of Pu-Rich Agglomerate in MOX Fuel on a Lattice Calculation

K¹,

Yamamoto²,

Namekawa³

2007

J. Nucl. Sci. Technol.

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The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17 Â 17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small.

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Section: Local Power Density and Burnupcontrasting

confidence: 74%

Section: )mentioning

confidence: 99%

Effect of Pu-Rich Agglomerate in MOX Fuel on a Lattice Calculation

K¹,

Yamamoto²,

Namekawa³

2007

J. Nucl. Sci. Technol.

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“…A number of different strategies for coupling have been performed in the past (Lee et al, 2000;Lemehov et al, 2001;Hamilton et al, 2012;Hursin et al, 2013;Soba et al, 2013) with different types of data transfer and levels of fidelity. This paper demonstrates coupling of a high-fidelity core analysis program, DeCART (Joo et al, 2004), to BISON.…”

Section: Neutronics Couplingmentioning

confidence: 99%

Advanced multiphysics coupling for LWR fuel performance analysis

Hales

Tonks

Gleicher

et al. 2015

Annals of Nuclear Energy

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a b s t r a c tEven the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas.The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application.A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm.Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use lower length scale models such as those used in the mesoscale MARMOT code to compute average properties, e.g. swelling or thermal conductivity. These may then be used by an engineering-scale model. Examples of this type of multiscale, multiphysics modeling are shown.

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“…The original TUBRNP [6] model for power and burn-up calculations included in the TRAN-SURANUS fuel performance code [7], and later also in other codes like FRAPCON3 [8], extended the RADAR model by including higher Pu isotopes, and modifying the radial shape function that accounts for resonance absorption by 238 U. TUBRNP was originally validated for UO 2 fuel in light-water reactors (LWRs) with experimental data from fuel with burn-up values between 35 and 64 MWd/kgHM. Later on, the RAPID model [9] was developed for the COSMOS code [10][11][12], PLUTON [13] was developed for the FEMAXI code [14], while a specific burn-up model for the RTOP code [15] was developed for fuel rods in Russian-type WWER reactors. RAPID was validated purely on the basis of profiles calculated by HELIOS up to 150 MWd/kgHM, while the others were validated against experimental data up to 83 MWd/kgHM.…”

Section: Introductionmentioning

confidence: 99%