This book provides an overview of the IMS Database Recovery Control; facility ( DBRC) and also discusses the administrative and operational. This section contains a description of the DBRC commands. Use these commands to add to, change, and delete information in the RECON and to generate the. DBRC Commands. This chapter contains a description of the DBRC commands. Use these commands to add to, change, and delete information in the.

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High-temperature gas-cooled reactor geometries differ significantly from conventional light water reactors.

## Overview of the RECON Data Sets

SeptemberSerpent International User Group MeetingDresden, Germany also see the topic at the discussion forum and the meeting website. The original incentive for developing a photon transport mode was to account for gamma heating in coupled multi-physics simulations. Energy deposition and radiation dose can also be evaluated using analog detector types, which in some case provide more physical results compared to the use of flux-to-dose conversion factors.

Subsystem records contain information about the subsystem and related recovery information, including:. Change accumulation records track information about change accumulation groups. March 19 The model works on several levels particles inside a pebble and pebbles inside the core and it has been tested in realistic double-heterogeneous reactor configurations consisting of over 60 million randomly positioned units Suikkanen, ; Rintala, The computational overhead from handling the unstructured configuration is moderate compared to a similar regular-lattice model.

June 15 Spatial homogenization was the main intended application for Serpent when the project was started in CAD models are read in the stereolitography STL format, in which the surfaces of geometry bodies are represented by a mesh of flat triangles.

If this name does not match, DBRC treats this database as if it was not registered. The fissile material is encapsulated inside microscopic tristructural-isotropic TRISO fuel particles, randomly dispersed in fuel compacts or pebbles made of graphite.

The drawback of the unionized energy grid approach is that computer memory is wasted for storing redundant data points. Work on coupled multi-physics applications continues. When the collision rate is low, the efficiency of the estimator can be improved by introducing additional virtual collisions over the particle flight path. Homogenization can be performed in infinite spectrum, or using a leakage correction based on the deterministic solution of the B 1 equations.

Reaction rates are normalized to total power, specific power density, flux, fission or source rate, and the normalization can be changed by dividing the irradiation cycle into a number of separate depletion intervals. Serpent 1 is no longer actively maintained, and all users are strongly encouraged to switch to Serpent 2 instead.

Two-way coupling to thermal hydraulics, CFD and fuel performance codes has been a major topic in Serpent development for the past several years. Transport cross sections and diffusion coefficients calculated using several methods. Burnup algorithms include the conventional gguide Euler and predictor-corrector methods, but Serpent 2 also offers various higher-order methods and sub-step solutions for burnup calculation Isotalo, b; c; b; a; b.

The third data set the spare is used in the following cases: Flux-volume-averaged one-group transmutation cross sections are calculated either during ans transport simulation, or by collapsing the continuous-energy reaction cross sections after the calculation has been completed using a flux spectrum collected on the unionized energy grid. If you use assistive technology such as a screen reader and need a version of this document in a more accessible format, please email publications dh.

Similar methodology has been used with other coupled Monte Carlo burnup calculation codes Haeck, ; Fridman, a; b. The data format determines the “laws of physics” for neutron interactions, and the results from Serpent calculations can be expected to agree with MCNP to within statistics.

Prevention and Detection for the Twenty-First Century: Serpent also has a built-in homogeneous diffusion flux solver for calculating discontinuity factors in regions where the net current over the boundaries is not reduced to zero by reflective boundary conditions. Validation of burnup znd routines is considerably more difficult, due to the lack of a perfect reference code.

New code version, List of publications updated. This is typically the case in reactor physics calculations involving fuel assemblies and especially HTGR micro-particle fuels.

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The application of the previous rules usually results in the need for at least two different sets of RECON data sets: Serpent 1 User’s Manual March 6, Reverence addition to the particle transport simulation, parallelization in the burnup calculation mode divides also the preprocessing and depletion routines between several CPU’s. The code also has a geometry plotter feature and a reaction rate plotter, which is convenient for visualizing the gyide and tally results see the gallery for examples.

The results can be divided into an arbitrary number of energy and time bins. Request an accessible format. The routine also enables the calculation of pebble-wise power distributions over the reactor core without defining additional tallies.

A delayed neutron model Valtavirta, allows the tracking of precursor concentrations over long time periods. Explicit wnd and pebble-bed fuel model for HTGR calculations.

Modeling of accurate heat deposition in coolant and structural materials requires accounting for the direct contribution of neutrons and fission and capture gammas. Comparison of rfference multiplication factors and other integral parameters shows generally good agreement between different calculation codes, but significant discrepancies can be found in the concentrations of individual nuclides.

The equilibrium calculation is independent of the depletion routine, and the iteration can also be performed in transport mode without burnup calculation. Burnup calculation The burnup calculation capability in Serpent was established early on, and is entirely based on built-in calculation routines, without coupling to yuide external solvers. Lean Six Sigma for Service: The original data set A copy of the original data set A spare data set The original data set and the copy are a pair of VSAM clusters that work as a pair to record information.

Geometry and particle tracking Similar to other Monte Carlo codes the basic geometry description in Serpent relies on a universe-based constructive solid geometry CSG model, which allows the description of practically any two- or three-dimensional fuel or reactor configuration.

DBRC organisation-specific reference cost data. Each Serpent update is checked by comparison to MCNP by running a standard set of assembly calculation problems.

The first data set is known as copy1.

We have also created standard data analysis queries to allow users to manipulate and understand the data more easily. In this source mode Serpent combines the compositions of activated materials into photon emission spectra read from ENDF format radioactive decay data files. May 8, Tuomas Viitanen defended his Doctoral Thesis: We would therefore be very grateful if users would email refeernce PbR data collection team at PbRdatacollection dh.