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RELAP5/MOD3 is a "best estimate" system code
suitable for the analysis of all transients and postulated accidents
in Light Water Reactor (LWR) systems, including both large- and
small-break loss-of-coolant accidents (LOCAs) as well as the full
range of operational transients. The one dimensional
RELAP5/MOD3 code is based on a non-homogeneous and non-equilibrium
model for the two-phase system that is solved by a fast,
partially implicit numerical scheme to permit economical calculation
of system transients. As a part of the severe accident
core meltdown code SCDAP/RELAP5, RELAP5/MOD3 describes the
thermal hydraulic aspects of both LWR systems and of the degrading
core. RELAP5 and the coupled version SCDAP/RELAP5 are being developed
at INEEL.
In addition, RELAP5 can be used to solve many plant
thermal-hydraulic problems. Modeling of single and two-phase
systems can be accomplished using many types of valves, heat exchangers and pumps.
Using the best estimate (BE) code, RELAP5, for probabilistic safety analysis (PSA) has
be performed with much success.
The code includes many generic models allowing to simulate general thermo hydraulic
systems. The models include pumps, valves, pipes, heat releasing or
absorbing structures, reactor point kinetics, electric heaters, jet
pumps, turbines, separators, accumulators, and control system logic
elements. RELAP5 represents the aggregate accumulation of
experience in modeling reactor core behavior during accidents,
two-phase flow processes, and LWR systems. The code development has benefited
from extensive application and validation against a lot of
experimental programs.
- ACGRACE Analysis
Code, GRaphing,
Advanced Computation
and Exploration of data
The SCDAP/RELAP5 code package is composed of:
- The Severe Core Damage Analysis Package (SCDAP) originally
designed at INEL to simulated the core behavior under simplified
thermal hydraulic conditions (water & steam, single pipe). It
has been verified against several in-pile and out-of pile test ([PBF],
CORA).
- The RELAP5 thermal
hydraulic code used for analyses of the reactor coolant systems
under normal operation as well as under design basis accidents
(i.e. LOCA). The code includes also a point-kinetics neutronic
model to model reactivity effects of the fluid composition and
density.
This coupled code package is used for detailed accident analyses
including severe accidents for existing as well as future power
plants.
For another post-processing the tool XMGR5
is available on request. The XMGR5 is a tool for 2-D
graphics including a versatile interface to various data types. (For
more information see [XMGR5]
)
Note: Parts taken
from
IRS
LWR
Safety
Research
and
from FZK/IRS
LWR
Safety
Research.
If you would like to know more about RELAP5, please contact RelapAdministrator@thcentral.com
.
Other Useful Links:
- A
COMPARISON OF THE PARET/ANL AND RELAP5/MOD3 CODES
FOR THE ANALYSIS OF IAEA BENCHMARK TRANSIENTS The
PARET/ANL and RELAP5/MOD3 codes are used to analyze the series of
benchmark transients specified for the IAEA Research Reactor Core
Conversion Guidebook (IAEA-TECDOC-643 , Vol. 3). The computed
results for these loss-of-flow and reactivity insertion transients
with scram are in excellent agreement and agree well with the
earlier results reported the guidebook. Attempts to also compare
RELAP5/MOD3 with the SPERT series of experiments are in progress.
Download PDF of paper.
- Program
of RELAP5 Users Group Program
of the International RELAP5 Users Group (IRUG); 19 and 20 October
2000
- Assessment
of the RELAP5-3D RELAP5-3D is the
latest version of the RELAP series of computer programs that are
used for the thermal-hydraulic safety analysis of water-moderated
nuclear reactors. In this version of the code, a three-dimensional
simulation of thermal-hydraulic and neutronic phenomena occurring in
a reactor can be simulated. Recently, a number of
papers have appeared in the literature in which attention has been
devoted to the RELAP5 code's predictions of low-pressure flow
boiling experiments. These studies have been motivated by the need
to use the RELAP5 code for the thermal-hydraulic safety analysis of
research reactors and advanced reactors with passive safety
features. The primary objective of the present study is to
study some of the experimental data on low-pressure subcooled flow
boiling available in the literature, and to utilize them as
validation cases, for thermal-hydraulic computer codes. Moreover,
this study was also conducted to investigate the suitability of the
SRL model for low-pressure subcooled flow boiling.
- RSICC
CODE PACKAGE PSR-423 - RELAP5
RELAP5 was developed to describe the behavior of a
light water reactor (LWR) subjected to postulated transients such as
loss of coolant from large or small pipe breaks, pump failures, etc.
RELAP5 calculates fluid conditions such as velocities, pressures,
densities, qualities, temperatures; thermal conditions such as
surface temperatures, temperature distributions, heat fluxes; pump
conditions; trip conditions; reactor power and reactivity from point
reactor kinetics; and control system variables. In addition to
reactor applications, the program can be applied to transient
analysis of other thermal-hydraulic systems with water as the fluid.
This package contains RELAP5/MOD1/029 for CDC computers and
RELAP5/MOD1/025 for VAX or IBM mainframe computers.
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