• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1071-1082
• /
• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

• Nuclear Engineering and Technology
• /
• v.52 no.6
• /
• pp.1110-1119
• /
• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.235-241
• /
• 1995

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor Power Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems.

• Nuclear Engineering and Technology
• /
• v.48 no.2
• /
• pp.545-553
• /
• 2016

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

• Nuclear Engineering and Technology
• /
• v.48 no.6
• /
• pp.1338-1348
• /
• 2016

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at $500^{\circ}C$ to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.988-1003
• /
• 2021

Existing elastic analysis methods cannot be adhered to in order to assess the structural integrity of a reactor vessel and internals for a beyond design basis earthquake. Elasto-plastic analysis methods are required, and the factors that affect the elasto-plastic behavior of reactor materials should be taken into account. In this study, a material behavior model was developed that considers the irradiation embrittlement effect, which affects the elasto-plastic behavior of the reactor material. This was used to perform the elasto-plastic time history analyses of the reactor vessel and its internals for beyond design basis earthquake. For this investigation, appropriate beyond design basis earthquakes and reliable finite element models were used. Based on the analysis results, consideration was given to the load reduction effect and the margin change. These were transferred to the internals due to the plastic deformation of the reactor vessel.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.37 no.12
• /
• pp.1175-1183
• /
• 2013

Even if some CFD software developers and its users think that a state-of-the-art CFD software can be used to reasonably solve at least single-phase nuclear reactor safety problems, there remain limitations and uncertainties in the calculation result. From a regulatory perspective, the Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of commercial CFD software for nuclear reactor safety problems. In this study, to examine the prediction performance of commercial CFD software with the porous model in the analysis of the scale-down APR (Advanced Power Reactor Plus) internal flow, a simulation was conducted with the on-board numerical models in ANSYS CFX R.14 and FLUENT R.14. It was concluded that depending on the CFD software, the internal flow distribution of the scale-down APR was locally somewhat different. Although there was a limitation in estimating the prediction performance of the commercial CFD software owing to the limited amount of measured data, CFX R.14 showed more reasonable prediction results in comparison with FLUENT R.14. Meanwhile, owing to the difference in discretization methodology, FLUENT R.14 required more computational memory than CFX R.14 for the same grid system. Therefore, the CFD software suitable to the available computational resource should be selected for massively parallel computations.

• Nuclear Engineering and Technology
• /
• v.51 no.8
• /
• pp.1970-1977
• /
• 2019

Understanding the irradiation hardening effect of structural steels under various irradiation conditions plays an important role in developing advanced nuclear systems. Such being the case, a crystal plasticity model for body-centered cubic (BCC) crystal based on the density of dislocations and irradiation defects is summarized and numerically implemented in this paper. Based on this model, nano-indentation hardness of Chinese A508-3 steels with ion irradiation is calculated. Very good agreement is observed between simulation and experimental data of several different irradiation doses subjected to various operating temperatures, from which, it can be concluded that indentation hardness increases with increasing irradiation dose at both room temperature and high temperature. Consequently, the validity of this model has been proved properly, and furthermore, the model established in this paper could guide the study of irradiation hardening effect and temperature effect to some extent.

• Journal of the Korean Society of Safety
• /
• v.23 no.5
• /
• pp.54-60
• /
• 2008

The objective of this study is to obtain the optimal process condition and the maximum decomposition efficiency by measuring the decomposition efficiency, electricity consumption, and voltage in accordance with the change of the process variables such as the frequency, maintaining time period, concentration, electrode material, thickness of the electrode, the number of windings of the electrode, and added materials etc. of the harmful atmospheric contamination gases such as NO, $NO_2$, and $SO_2$ etc. with the plasma which is generated by the discharging of the specially designed and manufactured $TiO_2$ catalysis reactor and SPCP reactor. The decomposition efficiency of the NO, the standard samples, is obtained with the plasma which is being generated by the discharge of the combination effect of the $TiO_2$ catalysis reactor and SPCP reactor with the variation of those process variables such as the frequency of the high voltage generator($5{\sim}50kHz$), maintaining time of the harmful gases($1{\sim}10.5sec$), initial concentration($100{\sim}1,000ppm$), the material of the electrode(W, Cu, Al), the thickness of the electrode(1, 2, 3mm), the number of the windings of the electrode(7, 9, 11turns), basic gases($N_2$, $O_2$, air), and the simulated gas($CO_2$) and the resulting substances are analyzed by utilizing FT-IR & GC.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1651-1664
• /
• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.