Conference Proceedings Paper
Simulate-3K Linkage with Reactor Systems Codes
NURETH-14 - 2011 September 25-30
Jerry Judd (Studsvik Scandpower, Inc.)
Gerardo Grandi (Studsvik Scandpower, Inc)
The comparisons demonstrate the applicability of S3K, either standalone or coupled with a system analysis code, to properly model system response during accident scenarios.
SIMULATE-3K is Studsvik Scandpower’s best-estimate three-dimensional core kinetics code. SIMULATE-3K has been coupled to several best-estimate reactor systems codes including, RELAP5- 3D, RELAP5-3.3, TRACE V5.0, and RETRAN-3D. The coupled codes can be applied to existing reactors and to advanced reactor designs.
The S3K linkage to each of the systems codes is a direct, explicit coupling of the two codes on a synchronous time-step basis. The coupling provides an execution method for the S3K three- dimensional neutronic model using the Nuclear Steam Supply System (NSSS) boundary conditions calculated by the systems code. Also, it allows the S3K calculated total core power and core power distributions to drive the system model core.
Detailed calculations from the component codes result in a methodology for analyzing limiting transients such as steam line breaks, rod drops/ejections, and ATWS scenarios. These transient events require detailed three- dimensional core data and information about the behavior of NSSS components. A coupled analysis of these transients is important because the core behavior is closely tied to the NSSS system. For example, to capture the timing and characteristics of the important thermal-hydraulic phenomena and/or operations events, such as valve closures, safety injection, or control system interactions, requires a detailed plant model.
The Peach Bottom 2 turbine trip transient is used to assess the accuracy of the coupled code calculations. Comparisons of the important plant parameters to results from RELAP5-3D, RELAP5- 3.3, and TRACE V5.0 calculations are shown and discussed.
The MSLB benchmark is also used to demonstrate the capabilities of the coupled code systems. Comparisons of the calculated reactor power to the reference data are shown can discussed.
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