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.
Both in-phase (core wide mode) instability and out-of-phase (regional mode) instability are of great concerns in BWR stability issues. Normally, decay ratios for regional mode oscillations are much less than those under core wide conditions. However, under certain observation mode, the regional mode instability has the phenomenon of power increasing in one half of the core and at the same time, it decrease in the other half, so it looks like that the average power remains essentially constant. This research presents a study of fractional change of decay ratio to evaluate parametric effects of regional mode instability on reload core design power/flow stability boundary for the Chinshan Nuclear Power Plant Unit 2 Cycle 21 (BWR4). Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such out-of-phase stability analysis. Many important parameters, such as system pressure, core flow rates, moderator void fraction, fuel physical and geometrical properties, have strong influences on regional mode stability. Current investigations have shown that at some operation points along the stability boundary, certain parameters present more sensitive characteristics.
A method for direct determination of resonance phase shifts in a ͑001͒ CdTe/InSb thin-film system is developed using soft x-ray three-wave resonance diffraction. At the ͑002͒ Bragg peaks of CdTe and InSb, two inversion-symmetry related three-wave diffractions are systematically identified according to crystal symmetry and the resonance phase shifts versus photon energies are measured without turning the thin film upside down. The momentum-transfer selectivity at ͑002͒ reflections facilitates the quantitative determination of the phase shifts near the Cd L 3 , Te L 3 , and Sb L 2 edges.
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.
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