Efficient generation of one-group cross sections for coupled Monte Carlo depletion calculations

Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The onegroup cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.

Safety implications of reactivity variations in fast thorium ADSRs

Nuclear power generation offers a reliable, low-impact and large-scale alternative to fossil fuels. However, concerns exist over the safety and sustainability of this method of power production, and it remains unpopular with some governments and pressure groups throughout the world. Fast thorium fuelled accelerator-driven sub-critical reactors (ADSRs) offer a possible route to providing further re-assurance regarding these concerns on account of their properties of enhanced safety through sub-critical operation combined with reduced actinide waste production from the thorium fuel source. The appropriate sub-critical margin at which these reactors should operate is the subject of continued debate. Commercial interests favour a small sub-critical margin in order to minimise the size of the accelerator needed for a given power output, whilst enhanced safety would be better satisfied through larger sub-critical margins to further minimise the possibility of a criticality excursion. Against this background, this paper examines some of the issues affecting reactor safety inherent within thorium fuel sources resulting from the essential Th90232→Th90233→Pa91233→U92233 breeding chain. Differences in the decay half-lives and fission and capture cross-sections of 233Pa and 233U can result in significant changes in the reactivity of the fuel following changes in the reactor power. Reactor operation is represented using a homogeneous lumped fast reactor model that can simulate the evolution of actinides and reactivity variations to first-order accuracy. The reactivity of the fuel is shown to increase significantly following a loss of power to the accelerator. Where the sub-critical operating margins are small this can result in a criticality excursion unless some form of additional intervention is made, for example through the insertion of control rods. © 2012 Elsevier Ltd. All rights reserved.

Direct observation of the structural component of the metal-insulator phase transition and growth habits of epitaxially grown VO2 nanowires.

We have grown epitaxially orientation-controlled monoclinic VO2 nanowires without employing catalysts by a vapor-phase transport process. Electron microscopy results reveal that single crystalline VO2 nanowires having a [100] growth direction grow laterally on the basal c plane and out of the basal r and a planes of sapphire, exhibiting triangular and rectangular cross sections, respectively. In addition, we have directly observed the structural phase transition of single crystalline VO2 nanowires between the monoclinic and tetragonal phases which exhibit insulating and metallic properties, respectively, and clearly analyzed their corresponding relationships using in situ transmission electron microscopy.

Unsteady transition in an axial-flow turbine. Part 1. Measurements on the turbine rotor

Previously published measurements in a low-speed, single-stage, axial-flow turbine have been reanalyzed in the light of more recent understanding. The measurements include time-resolved hot-wire traverses and surface hot film gage measurements at the midspan of the rotor suction surface with three different rotor-stator spacings. This paper investigates the suction surface boundary layer transition process, using surface-distance time plots and boundary layer cross sections to demonstrate the unsteady and two-dimensional nature of the process.

Modeling of SFR cores with Serpent-DYN3D codes sequence

DYN3D reactor dynamics nodal diffusion code was originally developed for the analysis of Light Water Reactors. In this paper, we demonstrate the feasibility of using DYN3D for modeling of fast spectrum reactors. A homogenized cross sections data library was generated using continuous energy Monte-Carlo code Serpent which provides significant modeling flexibility compared with traditional deterministic lattice transport codes and tolerable execution time. A representative sodium cooled fast reactor core was modeled with the Serpent-DYN3D code sequence and the results were compared with those produced by ERANOS code and with a 3D full core Monte-Carlo solution. Very good agreement between the codes was observed for the core integral parameters and power distribution suggesting that the DYN3D code with cross section library generated using Serpent can be reliably used for the analysis of fast reactors. © 2012 Elsevier Ltd. All rights reserved.

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results. Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al.; 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module. Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions. © 2011 Published by Elsevier B.V.

Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX. © 2009 Elsevier B.V. All rights reserved.