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Title: Code System for Reactor Physics and Fuel Cycle Simulation.

Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.
Authors:
Publication Date:
OSTI Identifier:
1268002
Report Number(s):
VSOP94; 004244MLTPL00
RSICC ID: C00670MNYWS
DOE Contract Number:
AC05-00OR22725
Resource Type:
Software
Software Revision:
00
Software Package Number:
004244
Software CPU:
MLTPL
Source Code Available:
Yes
Other Software Info:
Owner Installation: INST FIIR REAKTORENTWICKLUNG, JULICH, INST REACTOR DVLP Contributors: Juelich Research Center, Juelich, Federal Republic of Germany Through the NEADB, France The nuclear data of 160 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared from ENDF/B, BNL-325, and, in special cases, from other sources. Resonance cross section data are given as input. The data currently used are the ones published by J. J. Schmidt and ENDF/B IV. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group GATHER library by a typical HTR neutron energy spectrum generated by the GATHER code. Graphite scattering matrices are based on the Young phonon spectrum in graphite. The auxiliary codes DATA-2 and DATA-3 are specifically developed for HTRs. They prepare the input data for the nuclear part of the code from the basic geometric design figures for the fuel elements and reactor core, respectively. The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for the many core regions differing in temperature, burn-up, and fuel element lay-out. The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the epithermal GAM code uses modified cross-sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations. A two-dimensional neutron flux map is synthesized from a fast one-dimensional diffusion code in four energy groups by means of r-z iterations. This is used for the burnup calculation of up to 200 different compositions in the core. The basic scheme has been developed from the FEVER code. The build-up history of 40 fission product nuclides in these compositions is followed explicitly. The diffusion part of the program system will be repeated at many short burn-up stages, and the spectrum module will be reiterated at some larger time steps, when some significant change in the spectrum is expected. The fuel management and cost module performs the fuel shuffling and general evaluations of the reactor and fuel element life history. The fuel management simulates the currently known shuffling and out of pile routes for various reactors. It has further been extended to include the typical features of the pebble bed reactor as burnup dependent optional reloading of elements, separated treatment of different fuel streams, and recycling in new fuel element types according to a consistent mass balance and timing. Optionally, four different types of data files can be set up with characteristic data of the reactor life. These are used for more detailed investigations and display programs. The restart option allows the study of special phases of the reactor life, e.g., changes of the fueling scheme, of the burn-up, of the power output, of the coolant temperature, and of the corresponding KPD code. Two dimensional thermal hydraulics studies for operating and emergency conditions can be performed with the recently developed TIK-THERMIX code. A 2-dimensional display of incore characteristics is given by the LSD being coupled with TIK. LSD also given the averaged temperature of the different spectrum zones in the core. KEYWORDS: NEUTRON; ENDF FORMAT; FLUX or DOSE PLOTTING
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE
Country of Publication:
United States

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