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국민안전처가 개발 주체가 되어 재난 피해자를 위한 재난심리 치료를 위해 기능성 게임을 제안하고 운영할 수 있을까? 재난 회복력을 위한 기능성 게임 활용 가능성에 대한 연구 사례는 찾아보기 힘들다. 이러한 가능성을 논의하기 위해 재난 회복력과 기능성 게임과 관련된 주제어 검색으로 나타난 선행 연구들이 그러한 가능성을 언급하고 있는지 또는 관련 선행 연구들이...

Many countries have taken an interest in small and medium sized nuclear power plants. SMART, which was developed by Korea Atomic Energy Research Institute (KAERI), is a small sized integral type pressurized water reactor with a rated thermal power of 330 MW. In order to analyze thermal hydraulic characteristics of the SMART design, the TASS/SMR code has been developed. The code was validated using the results of basic and separate effect tests including small scale experiments for the SMART special components. To enhance an analysis capability of the TASS/SMR code for an integral type PWR, the KAERI has constructed the VISTA-ITL facility, and several integral effect tests have been performed at the VISTA-ITL facility. The TASS/SMR code is validated using the results of a loss of coolant flow transient, which is one of the integral effect tests performed at the VISTA-ITL. According to the evaluation results, the code predicts well the overall thermal hydraulic behaviors including the system pressure, fluid temperature, and mass flow rate. The main coolant pump model is important in order to simulate well the primary coolant flow behavior at an early transient.

The benefits of research reactor technology are always essentially accompanied by nuclear safety, which is ensured by a series of supporting analysis such as safety analysis. In order to extend the applicability of the Safety and Performance Analyzing Code (SPACE code) to safety analysis of typical open-tank-in-pool type research reactors using plate-type fuels, a Phenomena Identification and Ranking process was performed for most of the relevant internal event scenarios of a reference research reactor, i.e., the KiJang Research Reactor, designed and under development by the Korea Atomic Energy Research Institute (KAERI). The Structures, Systems, and Components (SSCs) important to nuclear safety of the reactor were included in the PIR processes. A summary of important results from the PIRT study on the research reactors is presented with a proposal on a preliminary set of data bases for a validation and verification matrix in addition to research issues.

Thermal–hydraulic behavior for a complete blockage of a single fuel channel in a generic 10 MW research reactor is studied by using the system analysis code RELAP5/MOD3.3 which is widely used in the nuclear industry. Fuel assembly geometry is lumped into a 4 channel model to model high and average power cases which are spatially discretized. Various axial power shapes coming from different control rods positions are considered in the analysis, where the minimum wall subcooled margin is found to exist for case with highest peaking for an average powered channel blockage transient. Vapor generation is observed from first and second highest peaking cases where cyclic variation of vapor inventory inside a blocked channel resulted in oscillatory behavior of the fuel temperature. Effect of a presence of an oxide layer is also tested which showed a slight increase in structure temperatures and vapor generation. Point kinetics model is utilized in the analysis code to observe the effect of reactivity feedback and consequences from different application ranges are compared. Analysis shows a consideration of assembly wise feedback results in increased feedback effect and decreased boiling which deviate from single channel wise feedback case. This calls for a detailed multi-dimensional simulation with neutronics and thermal–hydraulics simultaneously considered. Analyses results show that the consideration of feedback improves the outcome in terms of fuel temperature, and its integrity is conserved for all test cases.

SMART (system-integrated modular advanced reactor), which is a 330 MWt advanced integral nuclear power plant, was developed by the Korea Atomic Energy Institute (KAERI) for generating electricity and desalinating seawater. To enhance its safety, various design concepts were adopted, such as containing most of the reactor coolant system (RCS) components and a passive residual heat removal system (PRHRS). A thermal hydraulic evaluation and analysis of SMART is performed by TASS/SMR-S (transient and setpoint simulation/system-integrated modular reactor safety). The TASS/SMR-S code has various models reflecting the design features for SMART such as the core models (core power and core heat transfer models), the component model,s and the condensation heat exchanger (CHX) model. In this paper, the validation of the CHX model in the TASS/SMR-S code was performed with the POSTECH (Pohang University of Science & Technology) CHX test to evaluate the conservative prediction capability of the heat transfer coefficien...

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.

The transients and setpoint simulation/system-integrated modular reactor (TASS/SMR) code has been used to identify the safety margin of a 65-MWt advanced integral reactor and to evaluate its design performance. Although, the code has been verified by using simplified and analytical problems as well as a reliable system code, its validation has not been fully established. This paper deals with a validation of the TASS/SMR code by using two kinds of separate effect tests related to heat transfer at a helically coiled steam generator. The heat transfer experiments were performed by using a full-scale prototype of the steam generator cassette of the advanced integral reactor and a scaled-down steam generator cassette. Analytical results show that the TASS/SMR code predicts the thermal hydraulic parameters, including the system pressure and fluid temperature at the primary and secondary sides of the steam generator cassette, and the heat transfer rate through the steam generator cassette well. The validation results in this study show that the TASS/SMR code is applicable for heat transfer calculations related to the helically coiled steam generator of the advanced integral reactor.

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.

A series of evaluations for an inadvertent control rod withdrawal event has been performed to identify the appropriateness of the reactor protection system of a 65 MWt advanced integral reactor by using the Transients And Setpoint Simulation/System-integrated Modular Reactor (TASS/SMR) code. Evaluation results show that, if considering only a high core power trip function as the reactor protection system for the limitation of a core power increase, the occurrence of an inadvertent control rod withdrawal event at the lower core power levels under a high speed mode of a main coolant pump (MCP) results in a significant deterioration in the safety margin. Specially, a violation of a safety limit is possible under the MCP middle and low speed modes due to an unsuitable setpoint of the high core power trip function. To enhance the safety margin, a variable overpower trip (VOPT) function having multiple ceilings according to the MCP speed mode has been implemented into the reactor protection system for the 65 MWt advanced integral reactor. Reassessments by considering the VOPT function with multiple ceilings show that a safety enhancement in the case of an inadvertent control rod withdrawal event is obtained by a timely initiation of a reactor trip.

Overpressure protection analysis of KAERI's advanced integral reactor, which has been developed to verify the performance of the System integrated Modular Advanced ReacTor (SMART), has been performed using the Transients And Setpoint Simulation/Small and Medium Reactor (TASS/SMR) code. In the analysis, the loss of feed-water and the regulating bank withdrawal events on behalf of the decrease in the heat removal by the secondary system and the reactivity and power distribution anomalies are selected as the initiating events for the analysis because the highest peak pressures of the primary system occur during these events. Conservative assumptions and the various initial/boundary conditions have been applied to the overpressure protection analysis for the advanced integral reactor. Although the pressurization of the primary system occurs due to an unbalance between the power generation in the core and the heat removal through the steam generator, the peak pressures in the cases of using the loss of feed-water and the regulating bank withdrawal event as an initiating event are well below the acceptance criteria of 18.7 MPa, due to the reactor protection system and three pilot operated safety relief valves installed in the advanced integral reactor.

Performance evaluation of KAERI’s advanced integral reactor against an anticipated transient without scram has been carried out with the transients and setpoint simulation/small and medium reactor code, by considering a decrease in the heat removal by the secondary system, a loss of offsite power and an inadvertent control rod withdrawal event as an initiating event. In a decrease in the heat transfer by the secondary system and a loss of offsite power, the reactor coolant system pressures can be maintained below 110% of the design pressure during the transition period due to the effect of the large negative moderator temperature coefficient. On the other hand, in an inadvertent control rod withdrawal event, the pressure of the reactor coolant system increases up to the ASME service level C stress limit due to a high reactivity insertion into a reactor core by the adoption of a boron free core concept. Therefore, a hardware installation against an anticipated transient without scram is essential to mitigate the consequences resulting from an inadvertent control rod withdrawal event. A diverse protection system, which is an independent and diverse reactor shutdown system that is initiated by the signals of a high core power or a high pressurizer pressure, is adopted in the advanced integral reactor. According to the reassessment results by considering the diverse protection system for a reactor shutdown, the diverse protection system is helpful in mitigating the consequences of an anticipated transient without scram.

For an electricity generation and seawater desalination, a 330 MW System-integrated Modular Advanced ReacTor (SMART) was developed by KAERI. The safety level of the SMART is enhanced when compared to that of the typical commercial reactors, with the aid of an elimination of a large break loss of coolant accident by placing the major components of the primary system in a reactor vessel and the adoption of a new technology and a passive design concept into the safety system. However, the events related to reactivity and power distribution anomalies have been evaluated as vulnerable points when compared to the other initiating events in the SMART, since the reactivity worth of the control rods (CR) banks is quite large due to the boron free core concept. Especially, safety margins, i.e., minimum departure from nucleate boiling ratio (MDNBR), are significantly threatened during the CR banks withdrawal event. Therefore, MDNBR enhancement methodology for the CR banks withdrawal event should be considered to further enhance the safety level of the SMART design. Two methodologies have been suggested to enhance the MDNBR during the CR banks withdrawal event: the application of a DNBR trip function into a core protection system and a turbine trip delay methodology. Sensitivity studies are performed to evaluate the two MDNBR enhancement methodologies and show that the suggested methodologies could enhance the MDNBR during the CR banks withdrawal event of the SMART.

A series of experiments have been performed to understand the pool-boiling critical heat flux (CHF) behavior on small plates, varying the inclination angle and size of the heated surface under near atmospheric pressure: the first-phase experiment to clarify the CHF behavior at near the horizontal downward-facing position and the second-phase experiment to find out the general CHF behavior for overall inclination angles. The first- and second-phase experiment were performed for the inclination angles from −90° (horizontally downward position) to −40° using two plate-type test sections (20×200 mm and 25×200 mm) submerged in a pool of saturated water and for overall inclination angles from −90° to 90° using two plate-type test sections (30×150 mm and 40×150 mm) submerged in a slightly subcooled water pool, respectively. The CHF generally decreases as its inclination approaches to −90°, but there is a transition angle, at which the rate of decrease in the CHF suddenly changes. The measured CHF is lower for the wider test section due to the increased difficulty of bubble escape and this size effect increases as the inclination angle approaches to −90°.

The purpose of the study was to elucidate the mechanism underlying the enhancement of radiosensitivity to 60Co gamma-irradiation in human hepatoma cell line HepG2 pretreated with gliotoxin. Enhancement of radiotherapy by gliotoxin was investigated in vitro with human hepatoma HepG2 cell line. Apoptosis related proteins were evaluated by Western blotting. Annexin V/PI and reactive oxygen species (ROS) were quantified by Flow Cytometric (FACS) analysis. Gliotoxin (200 ng/ml) combined with radiation (4 Gy) treated cells induced apoptosis. Cells treated with gliotoxin (200 ng/ml) prior to irradiation at 4 Gy induced the expression of bax and nitric oxide (NO). The gliotoxin-irradiated cells also increased caspase-3 activation and ROS. Gadd45a, p38, and nuclear factor kappa B (NFkappaB) activated in irradiated cells was inhibited by Gliotoxin. Specific inhibitors of p38 kinase, SB203580, significantly inhibited NFkappaB activation and increased the cytotoxicity effect in cells exposed to gliotoxin combined with irradiation. However, SB203580 did not suppress the activation of Gadd45a in irradiated cells. Gliotoxin inhibited anti-apoptotic signal pathway involving the activation of Gadd45a-p38-NFkappaB mediated survival pathway that prevent radiation-induced cell death. Therefore, gliotoxin, blocking inflammation pathway and enhancing irradiation-induced apoptosis, is a promising agent to increase the radiotherapy of tumor cells.