Integral Fuel Burnable Absorber Research Articles

Benchmarking of the WIMS9/PARCS/TRACE code system for neutronic calculations of the Westinghouse AP1000™ reactor

The objective of this paper is to assess the accuracy of the WIMS9/PARCS/TRACE code system for power density calculations of the Westinghouse AP1000™ nuclear reactor, as a representative of modern pressurized water reactors (Gen III+). The cross section libraries were generated using the lattice physics code WIMS9 (the commercial version of the legacy lattice code WIMSD). Nine different fuel assembly types were analyzed in WIMS9 to generate the two-group cross sections required by the PARCS core simulator. The nine fuel assembly types were identified based on the distribution of Pyrex discrete burnable absorber (Borosilicate glass) and integral fuel burnable absorber (IFBA) rods in each fuel assembly. The generated cross sections were passed to the coupled core simulator PARCS/TRACE which performed 3-D, full-core diffusion calculations from within the US NRC Symbolic Nuclear Analysis Package (SNAP) interface. The results which included: assembly power distribution, effective multiplication factor (keff), radial and axial power density, and whole core depletion were compared to reference Monte Carlo results and to a published reactor data available in the AP1000 Design Control Document (DCD). The results of the study show acceptable accuracy of the WIMS9/PARCS/TRACE code in predicting the power density of the AP1000 core and, hence, establish its adequacy in the evaluation of the neutronics parameters of modern PWRs of similar designs. The work reported here is new in that it uses, for the first time, the combination of WIMS9/PARCS/TRACE codes to perform neutronics calculation for the Westinghouse’ AP1000™ reactor, as a representative of modern PWRs, with its challenging core configuration.

Lattice codes pin power prediction comparison

Abstract Lattice codes are able to produce burnup dependent fuel assembly homogenized neutron cross sections and pin power form factors. These data are used by diffusion nodal codes to produce fuel assembly powers and intra assembly power distributions. 2D transport codes are usually used for this type of the calculation. Four lattice codes (DRAGON 4, FA2D, NEWT and PSG2/Serpent) were used to calculate pin power distribution for reflected single fuel assemblies during depletion. The NPP Krsko 16 × 16 fuel assemblies without and with IFBAs (Integral Fuel Burnable Absorber), for two U-235 enrichments 4.40% and 4.80%, as found in cycle 24 of the plant, were modeled. The heterogeneous depletion calculation was performed up to burnup of 60 GWd/tU. The prediction capabilities of the codes are compared for infinite multiplication factor, some group cross sections, and for 2D pin power form factors, in order to assess their applicability for preparing cross section data for diffusion nodal codes.

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants

The negative impact of a boron dilution accident on the safety of a current pressurized water reactor (PWR) initiated investigations with the aim of checking the feasibility of reduced boron concentration operation. In addition, reduction of the maximum boron concentration in a PWR is a practical and feasible means to substantially reduce the radiation dose to operators and to minimize corrosion damage. Four types of integral burnable absorbers have been considered: gadolinium, integral fuel burnable absorber (IFBA), erbia, and alumina boron carbide. Under consideration of four different kinds of fuel assemblies (FA), four core design candidates were developed by applying current PWR OPR-1000 technology and by keeping major engineering design constraints and the equivalent fuel enrichment level used in the reference core (REF) design. However, an optimal design was targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between the REF and the optimal core designs is presented here. One of the designs is suggested as the most promising and favorable low boron core (LBC) design in this framework. The proper combination of axial and radial enrichment zoning patterns plus a mixture of fresh FAs with depleted assemblies in an LBC design candidate with an IFBA-bearing FA at equilibrium cycle could bring a two times narrower axial offset variation than that of the REF design, maintain an acceptable power peaking factor ˜23% lower than the design limit, and achieve higher fuel burnup. It was observed that this optimal LBC design could comply with current OPR-1000 reactor acceptance criteria associated with smooth reactivity swing, more flattened power distribution, and desired limiting safety parameters despite an 18% loss of shutdown reactivity worth at beginning of cycle when compared to the REF design.

Analysis of Reactivity Coefficients of Hydride-Fueled PWR Cores

A thorough investigation was performed to understand the physics behind the calculated trends in the reactivity coefficients of hydride-fueled pressurized water reactor cores. A two-step procedure was developed for this purpose: ranking the contribution to a reactivity coefficient of each of the system’s constituents; investigating, for each of the more important constituents, the spectral reasons for their specific response to the temperature perturbation. This procedure was applied to understand (a) the difference in the beginning-of-life and end-of-life fuel temperature coefficient of reactivity (FTC) behavior of hydride- as compared to oxide-fueled cores; (b) the different effect integral fuel burnable absorber (IFBA) and erbium burnable poisons have on the FTC and coolant temperature coefficient of reactivity (CTC); (c) the effects on the FTC and CTC of inclusion of thorium hydride in the fuel; (d) the effects of plutonium on the FTC and CTC in hydride- as compared to oxide-fueled cores.It is found that the physics characteristics of hydride-fueled cores are fundamentally different from those of oxide-fueled cores as particularly manifested by the behavior of the FTC. In oxide-fueled cores the main phenomenon affecting the FTC is the well-known Doppler broadening of the fuel resonances. Hydride cores feature an additional unique phenomenon of spectral shift in the thermal energy range; it is the result of upscattering of the thermal neutrons due to the increase in the fuel hydrogen temperature. The interplay between the spectral shift and the shape of the low-energy cross sections of the fuel isotopes is responsible for the sometimes very different values of the calculated FTC for hydride- versus oxide-fueled cores and even for the same fuel type at different burnups. It is also concluded that fissile plutonium can have different effects on the FTC and, although the physics phenomena are quite different, on the CTC in hydride-fueled cores. If the plutonium is present in sufficiently large quantities, its effect can be negative, while if it is present in relatively small quantities, it is more likely to give a positive contribution. An additional finding is that the buildup of 135Xe makes the FTC less positive in hydride-fueled cores, while it has little effect on the FTC of oxide-fueled cores. Also concluded is that thorium-containing hydride fuel cores feature a smaller FTC than that of oxide-fueled cores. This is due to a harder neutron spectrum in the Th-containing hydride-fueled cores leading to a smaller spectral shift, combined with the buildup of 233U the contribution to the FTC of which in hydride-fueled cores is mostly negative.The insight gained through the analyses reported in this work facilitated the identification of an optimal, safe uranium-based hydride-fueled core design; it consists of U-ZrH1.6 fuel in which 25% (volumetric) of the ZrH1.6 is replaced by ThH2. Burnable poisons have to be used to compensate part of the excess reactivity; IFBA is the preferred choice.

A study on the optimization of integral fuel burnable absorbers using deterministic methods

A deterministic fuel management study was performed to optimize Integral Fuel Burnable Absorber (IFBA) designs for Pressurized Water Reactors (PWRs). Penn State's IFBA designs were compared to Westinghouse's generic IFBA designs of 64, 80, and 160 IFBA pins per assembly. It was found deterministically that the current generic Westinghouse designs of 64 and 80 IFBA may be improved upon, but the typical 160 IFBA design by Westinghouse was not sufficiently improved upon. The results indicate it is not necessary to use large numbers of IFBA pins per assembly nor is it necessary to use exceedingly high boron enrichment in the IFBA pins. This deterministic study was performed as part of a genetic algorithm optimization study which used IFBA fuel designs and CASMO-3. The Penn State IFBA pin placement designs with the IFBA loaded toward the center of the assembly and surrounding the control rod guide tubes proved to be very successful. The Penn State IFBA pin designs with 64 and 80 IFBA pins showed significant improvement over the generic vendor designs, which have IFBA placed towards the assembly periphery. The Penn State 64 and 80 IFBA designs reduced the maximum peak pin NP by as much as 5% over the vendor's IFBA configurations.