[en] During the flow reversal from a downward flow to an upward flow, flow stagnation occurs and it can induce deterioration in heat transfer and threaten the fuel integrity. Therefore, it is important for the downward inertial flow to be kept for a sufficiently long time by using active or passive system. In this paper, a performance evaluation is carried out on a passive type core cooling system such as a gravity core cooling system (GCCS) and an active type safety residual heat removal system (SRHRS) for a research reactor with 15 MW. The performance evaluation of a passive and active core cooling system for a research reactor with 15 MW has been carried out by using RELAP5/MOD3.3/P4. According to the analysis results on the failure of two PCS pumps, it is found that both the passive and active core cooling system have an adequate safety function for the research reactor. However, the passive core cooling system has not enough the safety margin for the CHFR and coolant temperature at the hot fuel assembly. So the active core cooling system shall be considered to enhance the safety margin for a research reactor with 15 MW

[en] The procedures may also be used as an aid for assessing and documenting the results of tests. The commissioning procedures should include information that specifies several items. Those are mainly (1) all the activities and performance parameters that are to be measured under specified steady state and transient conditions, (2) the requirements on performance, together with clearly stated acceptance criteria. The final phase of stage C commissioning is reactor performance test, which is to prove the integrated performance (neutron power calibration, Control Absorber Rod drop time, I and C functioning, Rod worth, Core heat removal with natural mechanism) and the safety of the research reactor at full power with fuel loaded. Commissioning procedure was developed to show the safety of the research reactor. Both indirect and direct indicators were selected to show that the safety is ensured: 1) indirect parameters which imply success of safety functions: power, flow, opening valves, system response as-designed; 2) direct parameters which shows no failure of safety functions: no meaningful increase in level of neutron in the cooling system. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters. A preliminary comparison to conservative estimation has shown that the nuclear reactor safety of the research reactor will be assured by verifying that the reactor power and the PCS flow rate are conservative

[en] The SMART (system-integrated modular advanced reactor) is a 330 MWt advanced integral PWR designed by KAERI for seawater desalination and electricity generation. Unlike the loop type commercial reactors, the nuclear steam supply system (NSSS) of the SMART adopts the design concept of containing most of the reactor coolant system (RCS) components, such as a core, four reactor coolant pumps (RCPs), eight steam generators (SGs), and a pressurizer in a single leak-tight reactor pressure vessel. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs). The thermal hydraulic analysis of the increased heat removal transients by the secondary system which are expected to occur with moderate frequency is performed in the SMART. And the results of the transient analysis are reviewed to ensure that the values of relevant system parameters are satisfied with the acceptance criteria

[en] Following the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (NRC) established the sets of requirements addressing their objective to improve the quality of operational information for dealing with emergency events in nuclear power plants. The Emergency Operating Guidelines (EOG) should be presented to provide technical information to prepare reactor-specific Emergency Operating Procedures (EOP) which cover operation during emergency events. EOG must provide guidance for both classes of emergencies. Thus, when a reactor trip occurs or should occur, the operators can refer to guidance which will provide a safe response whether or not a symptom set is identified: EOG written to treat specific symptoms are called event-based recovery guidelines (ERG); the EOG which provides guidance for undiagnosed events for which a reactor trip is required is called the Symptom-based Recovery Guidelines (SRGs). For helping to design and develop a set of EOG/EOP, a stencil in a sheet was proposed as an easy and intuitive tool to gather information of a research reactor related to safety/safety-related functions. With concept of safety functions, the stencil will be able to give a strong but very easy, straightforward, systematic, comprehensive tool to analyze the architecture of the reactor, to cover the whole SSC information

[en] The SMART (System-integrated Modular Advanced ReacTor) has a passive safety system such as a Passive Residual Heat Removal System (PRHRS). When a secondary system can't operate because of the malfunction of a feedwater system or a loss of off-site power (LOOP), the core decay heat is removed using the Condensation Heat Exchanger (CHX) in the PRHRS with natural convection. The CHX has 500 tubes submerged in an emergency cooling tank. So the heat removal performance can be affected by the material and diameter of the CHX tube. In this paper, the heat transfer performance for the diameter and material of the CHX tube was evaluated using the TASS/SMR-S (Transient And Setpoint Simulation/System-integrated Modular Reactor-Safety) code with the POSTECH CHX heat transfer test. The TASS/SMR-S code has several heat transfer models for the PRHRS CHX. The role of these models is to calculate the heat transfer coefficient in the CHX by the relevant correlations for all of the heat transfer modes

[en] When some accidents or events are occurred in the SMART, the secondary system is used to remove the core decay heat for the long time such as a feedwater system. But if the feedwater system can't remove the residual core heat because of its malfunction, the core decay heat is removed using the Passive Residual Heat Removal System (PRHRS). The PRHRS is passive type safety system adopted to enhance the safety of the SMART. It can fundamentally eliminate the uncertainty of operator action. TASS/SMR-S (Transient And Setpoint Simulation/ System-integrated Modular Reactor-Safety) code has various heat transfer models reflecting the design features of the SMART. One of the heat transfer models is the PRHRS condensation heat transfer model. The role of this model is to calculate the heat transfer coefficient in the heat exchanger (H/X) tube side using the relevant heat transfer correlations for all of the heat transfer modes. In this paper, the validation of the condensation heat transfer model was carried out using the POSTECH H/X heat transfer test

[en] In this study, the validation of the core heat transfer model in the TASS/SMR-S code on steady conditions was performed with the Bennett's heated tube tests and Thermal Hydraulic Test Facility (THTF) experiment. According to the selected test conditions, the CHF (critical heat flux) points and the axial temperature profiles at a fuel rod surface were calculated and compared with the test data. From the results of the calculation, the CHF position and the surface temperature of a fuel rod at the post CHF region were predicted more conservatively than expected by the test results. However, at low pressure, low mass flux and low heat flux conditions such as in the 3.07.9O test, model improvement is needed for a conservative calculation

[en] The SMART (System-integrated Modular Advanced ReacTor) which is a 330 MWt advanced integral PWR was developed by the KAERI (Korea Atomic Energy Institute) for electricity generation and seawater desalination. A thermal hydraulic evaluation and analysis of the SMART is performed by the TASS /SMR-S (Transient And Setpoint Simulation/System integrated Modular Reactor-Safety). The TASS/SMR-S code has various models reflecting the design features of the SMART such as the drift flux model, the core models (core power and core heat transfer model), the component models, and the specific models. One of the core models is the core heat transfer model. The role of this model is to calculate the heat flux and radial temperature profiles at a fuel rod surface using the relevant heat transfer correlations for all of the heat transfer modes. Also it is modeled to meet the requirements of the 10 CFR 50 appendix K EM model for the CHF (Critical Heat Flux) and post CHF conditions. In this paper, the validation of the core heat transfer model was carried out using the Bennett's heated tube tests

[en] In this paper, to apply Digital Radiography (DR) to industrial NDT (Nondestructive Testing), we have simulated absorbed energy and calculated absorbed dose, exposure dose on scintillator coupled CMOS substrate using Monte Carlo method by MCNPX Code for high energy X-ray and gamma source such as Co-60, Ir-192, Se-75. Therefore, in comparison with previous experimental result in the variation of MTF (Modulation Transfer Function) value, estimated life time of each scintillator coupled CMOS image sensor according to exposure time.

[en] In this paper, to apply Digital Radiography (DR) to industrial NDT (Nondestructive Testing), we have simulated absorbed energy and calculated absorbed dose, exposure dose on scintillator coupled CMOS substrate using Monte Carlo method by MCNPX Code for high energy X-ray and gamma source such as Co-60, Ir-192, Se-75. Therefore, in comparison with previous experimental result in the variation of MTF (Modulation Transfer Function) value, estimated life time of each scintillator coupled CMOS image sensor according to exposure time.