Tube inlet orifice design of a once-through steam generator for flow stabilization

A numerical study is conducted on the secondary side screw-type tube inlet orifice design of a once-through steam generator. An orifice length criterion for flow stabilization is derived by introducing the hydraulic resistance ratio of the orifice and the subcooled region to the two-phase and superheated regions. Various tube plugging conditions and power levels are considered, and the secondary coolant pressure at the tube outlet is adjusted to maintain a constant thermal power. Comprehensive numerical solutions are acquired to evaluate the minimum orifice length under various operating conditions. The results obtained show that a constant thermal power is maintained by properly adjusting the secondary coolant outlet pressure with a variation of the superheat degree and secondary coolant pressure drop when the steam generator operates at high power level. The steam generator performance is analyzed according to the tube plugging condition in terms of the degree of superheat, secondary side pressure drop, temperature distribution, and quality distribution. The secondary side outlet pressure curve for the constant thermal power operation is obtained, and the required minimum orifice length to suppress the flow oscillation below the allowable level is evaluated. The lowest power level results in the highest minimum orifice length, and non-plugging condition provides a limiting case for the orifice length criterion except near the 100 % power level.

     

Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet 1, 2, 3, 4, 5, 6, 7 and 8. This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.     

Fully-coupled transient simulation for RCS and SPCS of an integral reactor plant

Design features of SMART such as a built-in once-through steam generator (OTSG) and a close interaction between the feedwater flow rate and steam pressure controls leads to the necessity of fully-coupled transient analysis tools of the reactor coolant system (RCS) and the steam and power conversion system (SPCS) for the purpose of a plant control system development. A fully-coupled transient simulation tool, MMS/SMART, was developed to test the capability of the plant control system for the normal load-following event and the anticipated abnormal events. The MMS/SMART was composed of several interacting MMS modules with numerical data, each of which represented a component of the SMART plant and a control logic. The RCS and the SPCS with their control logics were modeled using default modules such as a pipe, pump and tank. The developed MMS/SMART was validated by using the scaled-down experimental data and the analysis result from the TASS/SMR code. A simulation result for the 100–50–100% load-following operation with a 25%/min rate shows that the feedwater flow rate and the steam pressure are controlled well as expected, except for small-amplitudes of steam pressure fluctuation at the lower power operating region. The loss of turbine load event was also simulated and the result shows that the plant can be operated stably with the steam bypass control system.     

Parallel process system and its application to steam generator structural analysis

A large-scale analysis to evaluate complex material and structural behaviors is one of interesting topic in diverse engineering and scientific fields. Also, the utilization of massively parallel processors has been a recent trend of high performance computing. The objective of this paper is to introduce a parallel process system which consists of general purpose finite element analysis solver as well as parallelized PC cluster. The later was constructed using eight processing elements and the former was developed adopting both hierarchical domain decomposition method and balancing domain decomposition method. Then, to verify the efficiency of the established system, it was applied for structural analysis of steam generator in nuclear power plant. Since the prototypal evaluation results agreed well to the corresponding reference solutions it is believed that, after reinforcement of PC cluster by increasing number of processing elements, the promising parallel process system can be utilized as a useful tool for advanced structural integrity evaluation.     

Experimental and numerical investigations on brittle failure probability and ductile resistance property

The scatter of measured fracture toughness data and the transferability of toughness between different crack configurations and loading conditions are major obstacles in applications of fracture mechanics. To address these issues, worldwide, significant efforts have been devoted to establishment of the local approach. The purpose of this work is to further investigate both brittle rupture and ductile rupture of typical nuclear materials by using miniature specimens. Systematic finite element analyses as well as corresponding fracture toughness tests are performed with respect to compact tension and pre-cracked V-notched specimens. Subsequently, assessment of brittle failure probabilities and estimation of fracture resistance curves are carried out. Promising results have been observed through comparisons between experimental and numerically predicted data, which enables one to determine practical safety margins of nuclear components containing a defect.