Fast Reactor Analysis Research Articles

Few-Group Methods for Fast-Reactor Transients

The suitability of flux- and bilinearly weighted collapsing schemes for fast-reactor transiert analyses was studied. Comparisons of few-group transient results were made with exact'' 26-group and three-trial-function energy synthesis results. Normal flux weighting was not satisfactory even when up to 12 groups were used. Bilinear schemes with only 6 or 8 groups, using unperturbed fluxes and perturbed adjoints (or vice versa), were found to be satisfactory. When reactivities were less than one dollar, unperturbed fluxes and adjoints were sufficient to give satisfactory results. With bilinear group collapsing, it is necessary to consider discontinuities arising at interfaces between regions where the flux and adjoint spectra are different and to use modified continuity conditions. (auth)

Delayed Neutron Data for Fast Reactor Analysis

The delayed neutron precursor decay curves which result from a pulse of fissions of the isotopes 233 U, 235U, 238U, 232Th, 239Pu, 240Pu, 241Pu, and242Pu are consistently fit to a single set of six ...

Fast Reactor Fuel Interactions with Floor Material after a Hypothetical Core Meltdown

The hypothetical accident approach to analysis of fast reactors has been applied to the meltdown of an entire core and its interaction with containment floor materials of construction. The objectiv...

An Evaluation of the Number of Neutrons Per Fission for the Principal Plutonium, Uranium, and Thorium Isotopes

Experimental values (apart from thermal-neutron data) of the number of prompt neutrons per fission () for 232Th, 233U, 234U, 235U, 238U, 239Pu, 240Pu, and 241Pu over the energy range from 0 to 15 MeV up to February 1970 are evaluated to derive the best available data, primarily for fast reactor analysis. Limited comments are made on data published subsequently. Data are renormalized where necessary to the latest recommended value (3.756) for for 252Cf. Many of the existing lower energy data for 233U and 235U indicate a nonlinear energy dependence that would be of significance for fast reactor analysis, but a limited number of recent specific studies on 235U strongly support linear dependence, and the existence of low-energy structure is not established. The more limited low-energy measurements on 239Pu indicate little, if any, structure. At energies above several MeV the normally assumed linear energy dependence of is more applicable, but the occurrence of the (n, n′f) and (n, 2nf) reactions in addition to the (n, f) reaction introduces a nonlinear dependence that is generally small but may be of significance for some isotopes. Consideration of the effects of first-, second-, and third-chance fission [(n, f), (n, n′f), and (n, 2nf) reactions] gives a good fit to the single case, 235U, in which a test can be made and these effects are considered in the evaluation of other isotopes.

An Analysis of the Neutron Capture Cross Section of 238U Between 1 keV and 15 MeV

The best available values of the 238U neutron capture cross section in the range of a few keV to 15 MeV are derived by evaluation of the experimental techniques and reference cross sections and, where necessary, revising and rejecting experimental values. A principal objective is to obtain 238U capture cross sections which are consistent with fission cross sections previously selected by the author and which can, therefore, be used for consistent analysis of fast reactors. All available experiments are discussed and a detailed tabulation of the individual experimental data and selected best cross sections is given.

Space-Energy Flux Synthesis in Transport Theory

The efficacy of the overlapping group method in fast-reactor analysis is investigated and tested on an idealized fast-reactor configuration. A full transport-theory treatment is adopted and the ove...

Authors and Papers

Authors and Papers

Electronically-conducting, refractory ceramic electrodes for open cycle MHD power generation : D. B. Meadowcroft, Central Electricity Research Laboratories, Materials Division, Leatherhead, Surrey. Energy Conversion8, 185–190 (1968)

The postulated core disruptive accidents (CDAs) are regarded as particular difficulties in the safety analysis of liquid-metal fast reactors (LMFRs). In the CDAs, core debris may settle on the core-support structure and form conic bed mounds. Then debris bed can be levelled by the heat convection and vaporization of surrounding coolant sodium, which is named “self-leveling behavior”. The self-leveling behavior is a crucial issue in the safety analysis, due to its significant effect on the relocation of molten core and heat-removal capability of the debris bed. Considering its complicate multiphase mechanism, a comprehensive computational tool is needed to reasonably simulate transient particle behavior as well as thermal-hydraulic phenomenon of surrounding fluid phases. The SIMMER program is a successful computer code initially developed as an advanced tool for CDA analysis of LMFRs. It is a multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. Until now, the code has been successfully applied in numerical simulations for reproducing key thermal-hydraulic phenomena involved in CDAs as well as performing reactor safety assessment. However, strong interactions between massive solid particles as well as particle characteristics in multiphase flows were not taken into consideration in its fluid-dynamics models. To solve this problem, a new method is developed by combining the discrete element method (DEM) with the multi-fluid models of SIMMER code to reasonably simulate the particle behaviors as well as the thermal-hydraulic phenomena of multiphase fluid flows. In this paper, the coupling method is validated by performing numerical simulations on a series of experiments with cylindrical particle bed in 3D situation. Based on various experimental conditions, Reasonable agreement between simulation results and corresponding experimental data can demonstrate the applicability of the method in reproducing the self-leveling behavior of debris bed. Sensitivity analysis on some model parameters of DEM is also performed to assess their impacts in the simulation. It is expected that the present method can be used as a computational tool to estimate self-leveling process of debris beds as well as the following behaviors in real reactor environment.