The RELAP5/MOD3.2 R/T engineering quality plant simulation code has been developed based on the system thermal-hydraulics code RELAP5 and the nodal spatial kinetics code NESTLE. Preliminary validation of the code as an engineering simulator was performed using a series of NEA PWR control rod ejection and withdrawal problems. Real time performance for simulator applications was demonstrated for ...

Recent advances in the computational speed of engineering workstations have enabled the development of a real-time version of the RELAP5 [Beelman, 1996] nuclear plant simulation code with Laboratory Discretionary Research and Development (LDRD) funding. In addition, the INEL is also funding the development of an enhanced real-time version of the existing three-dimensional nodal neutron kinetics...

The LSTF experiment simulating the SGTR accident at the Mihama Unit-2 reactor is analyzed using the RELAP5/MOD3.3 code. In the accident, and thus in the experiment, the ECC water was injected not only into the cold legs but into the upper plenum. Overall transients during the experiment such as pressures and fluid temperatures are simulated well by the code. The cold-leg fluid temperatures are ...

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Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very f...

The modern direction in the study of safety nuclear power plants is to ensure maximum level detail process modeling with a satisfactory computational resources. One approaches such task solving coupled use special software required levels detail, for example, systemic thermohydraulic codes hydrodynamics codes.
 This article describes developed coupling module between system code RELAP5/Mod...

The work completed had the following objectives: i) To compute pulse width (full width at half maximum-FWHM) of the power curves for the Three Mile Island Unit 1 Pressurized Water Reactor (TMI-1 PWR) core model in the event of a super-prompt-critical rod ejection accident (REA) at Hot Zero Power (HZP) in the central fuel assembly (Rod 7A) at both EOC and BOC for various control rod worths and d...

The nuclear community has coupled several three-dimensional Computational Fluid Dynamics (CFD) solvers with one-dimensional system thermal–hydraulic (STH) codes. This work proposes to replace the CFD solver by a reduced order model (ROM) reduce computational cost. code RELAP5-MOD3.3 and ROM of finite volume OpenFOAM are partitioned domain decomposition coupling algorithm using an implicit schem...