Chang-Lung Hsieh

LAPUR5.2 BWR Stability Analysis in Chinshan Nuclear Power Plant

Boiling water reactors have the unique coupling mechanisms between neutronic and two-phase flow thermal-hydraulic behaviors and may induce instability by unstable power/flow oscillations. At each core reload design, it is important to employ decay ratio for the purpose of analyzing system stability and determining its operating boundary. Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such analysis. Comparisons made with vendor’s STAIF results indicated good agreements in decay ratios for Chinshan Nuclear Power Plant Unit 2 Cycle 21 reload design. This research focuses on the parametric sensitivity effect on the variation of decay ratio for different power/flow operating points. Based on the result of sensitivity studies, we presented fractional changes of decay ratios by varying certain important parameters under different power/flow points. It is concluded that density reactivity coefficient, gap conductance and recirculation loop gain on high operating power/flow points have larger fractional change of decay ratio.© 2006 ASME

Comparative Study in the Stability Analysis Code of LAPUR5.2 and LAPUR6.0 for the Kuosheng NPP

Under certain conditions, boiling water reactors (BWRs) would be susceptible to couple neutron-thermalhydraulic instability. It is important to predict such potential problems as early as possible and prevent the core instability from happening. In each BWR reload core design, fuel vendors are required to provide instability boundaries on power/flow map to assure safety operation of the nuclear reactor. In Taiwan, a LAPUR5.2 methodology had been adapted to build up the remarkable analysis mode for different types BWRs to verify vendor’s results. However, with upgrading nuclear safety technology, most of boiling water reactors has been adopting partial length fuel assemblies to reduce two-phase pressure drop and void fraction, to improve reactor stability. The question is that LAPUR5.2 methodology cannot precisely analysis stability characteristics from the variation of flow area in fuel assemblies. From the reasons of upgrading stability analysis, a LAPUR6.0 methodology had built to do the related researches. This research was based on a comparison study between LAPUR5.2 and LAPUR6.0 to realize the major differences and their effects on stability characteristics. According to the comparison results for Kuosheng Nuclear Power Plant Unit 2 Cycle 21 reload design, it shows that LAPUR6.0 can completely present pressure drop, void fraction and density reactivity coefficient from the changing of flow area and fuel spacers.Copyright