Hwan-Yeol Kim

Development of a Computer Code, ONCESG, for the Thermal-Hydraulic Design of a Once-Through Steam Generator

Development of the conceptual design of a 300 MWt integral reactor, SMART (System-integrated Modular Advanced ReacTor), for utilization in nuclear cogeneration plants has been completed at the Korea Atomic Energy Research Institute (KAERI). The major primary components of the SMART such as modular helical steam generators, main circulation pumps and a self regulating pressurizer are integrated into a reactor vessel. It is a common practice to employ a once-through steam generator in integral reactor designs because of its advantages in compactness and simplicity of the flow path arrangements. In this study, a thermal hydraulic design and performance analysis computer code for a once-through steam generator using helically coiled tubes, ONCESG, is developed. To benchmark the developed physical models and computer code, once-through steam generators developed by other designers are simulated and ONCESG calculated results are compared with the design data. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by ONCESG showed general agreements with the published data and it is demonstrated that the ONCESG code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

Validation of RCS depressurization strategy and core coolability map for independent scenarios of SBLOCA, SBO, and TLOFW

Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coolant Accident (LOCA), Station Black Out (SBO), Total Loss-Of-Feed-Water (TLOFW) was performed and effectiveness of Reactor Coolant System (RCS) depressurization strategies of OPR1000 was analyzed using MELCOR 1.8.6 code. Required injection flow rate has been derived using Core Exit Temperature (CET) information obtained from MELCOR calculation and a simple model and corresponding coolability map have been suggested to assist effective operator action. The depressurization strategies using secondary Atmospheric Dump Valve (ADV) for SBLOCA, pressurizer Safety Depressurization System (SDS) for SBO and TLOFW were introduced in 5 min since the initiation of Severe Accident Management Guidance (SAMG). Respective mitigation strategy employed leads to significant delay of the reactor pressure vessel failure and RCS pressure at Reactor Pressure Vessel (RPV) failure was lower than the SAMG target pressure of 2.86 MPa. Thus, possibility of High Pressure Melt Ejection (HPME) and impair of containment building is expected to avoid effectively. Using CET information obtained from MELCOR calculation, a simple model and a coolability map for the required injection flow rate were developed for recovery of core coolability. It is suggested that the coolability map based on MELCOR calculation results may provide decisive and intuitive information to operators for more effective safety management.

Thermal-Hydraulic R&Ds for the APR+ Developments in Korea

This paper briefly introduces recent progress in thermal-hydraulic R&Ds, which is mainly being performed at KAERI, for the APR+ (Advanced Power Reactor plus) development. The main R&D items for the APR+ reactor are associated directly with recent efforts to introduce new safety concepts in the APR+ standard design developments, which is currently in progress in the Republic of Korea. The R&D activities reported here mainly cover the thermal-hydraulic and severe accident areas and are being performed in experimental and/or analytical ways. They include: (1) advancement and optimization of safety injection system, (2) incorporation of passive safety features, such as advanced Fluidic Device (FD+) and passive auxiliary feedwater system (PAFS), and (3) incorporation of severe accident mitigation features.Copyright

Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

Severe accident analysis for Korean OPR1000 with MELCOR 1.8.6 was performed by adapting a mitigation strategy under different entry conditions of Severe Accident Management Guidance (SAMG). The analysis was focused on the effectiveness of the mitigation strategy and its adverse effects. Four core exit temperatures (CETs) were selected as SAMG entry conditions, and Small Break Loss of Coolant Accident (SBLOCA), Station Blackout (SBO), and Total Loss of Feed Water (TLOFW) were selected as postulated scenarios that may propagate into severe accidents. In order to delay reactor pressure vessel (RPV) failure, entering the SAMG when the CET reached 923 K, 923 K, and 753 K resulted in the best results for SBLOCA, SBO, and TLOFW scenarios, respectively. This implies that using event-based diagnosis for severe accidents may be more beneficial than using symptom-based diagnosis. There is no significant difference among selected SAMG entry conditions in light of the operators available action time before the RPV failure. Potential vulnerability of the RPV due to hydrogen generation was analyzed to investigate the foreseeable adverse effects that act against the accident mitigation strategies. For the SBLOCA cases, mitigation cases generated more hydrogen than the base case. However, the amount of hydrogen generated was similar between the base and mitigation cases for SBO and TLOFW. Hydrogen concentrations of containment were less than 5% before RPV failure for most cases.

Development of safety injection flow map associated with target depressurization for effective severe accident management of OPR1000

ABSTRACT If any severe accident occurs, application of the Severe Accident Management Guidance (SAMG) is initiated by the Technical Support Center (TSC). In order to provide advisory information to the TSC, required safety injection flow rate for maintaining the coolability of the reactor core has been suggested in terms of the depressurization pressure. In this study, mechanistic development of the safety injection flow map was performed by post-processing the core exit temperature (CET) data from MELCOR simulation. In addition, effect of oxidation during the core degradation was incorporated by including simulation data of core water level decrease rate. Using the CET increase rate and core water level decrease rate, safety injection flow maps required for removing the decay and oxidation heat and finally for maintaining the coolability of the reactor core were developed. Three initiating events of small break loss of coolant accidents without safety injection, station black out, and total loss of feed water were considered. Reactor coolant system depressurization pressure targeting the suggested injection flow achievable with one or two high pressure safety injections was included in the map. This study contributes on improving the current SAMG by providing more practical and mechanistic information to manage the severe accidents.

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

ABSTRACT During a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.

Air Clearing Oscillation Produced by APR1400 Prototype Sparger

KAERI has performed a series of experiments to investigate the performance of APR1400 prototype sparger in view of the dynamic load oscillation with the variation of test conditions such as discharged air mass, submergence of the sparger, valve opening time, and pool temperature during the air-clearing phase. The air mass and pool temperature are in the range of 0.8 ∼ 1.5 kg and 20 ∼ 90°C, respectively. The valve opening time can be adjusted in the range of 0.6 ∼ 1.7 sec according to the POSRV operating conditions. The maximum positive pressure amplitude, which is observed at the bottom of the quench tank, is increased with the submergence depth and maximum header pressure of sparger. The valve opening time has a considerable effect on the maximum amplitude. As the opening time decreases, the maximum amplitudes at the tank wall are increased. Air mass and pool temperature, however, have a weak effect on the maximum amplitude. Oscillation frequency is decreased with air mass in the range of 2.5 ∼ 4.5Hz.Copyright