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The elimination of active systems and their replacement with passive systems are emphasized to improve a system reliability in advanced nuclear power plants such as SMART. For accident situations, a condensate heat exchanger having vertical tubes is an ultimate heat sink to remove residual heat generated in the core. To adequately analyze the system behaviors, a realistic condensation model with pure steam or a steam-air mixture in the condensate heat exchanger is important for the various thermal hydraulic conditions. Improved correlations are proposed to analyze the thermal hydraulic behaviors for wide pressure conditions of up to 6 MPa. For a validation of the condensation heat transfer for the condensate heat exchanger tube side, simulations of the UCB (University of California at Berkeley) and POSTECH (Pohang University of Science and Technology) condensation experiment are carried out, and the results are compared with the experimental data. For pure steam conditions, the correlations predict well or under-predict slightly the UCB experimental results and predict the POSTECH experimental results within 25%. The improved correlation predicts that the condensation heat transfer coefficient decreases as the system pressure increases as shown the POSTECH experiment.

A pressurizer is a major component used to control a system pressure in a primary coolant system of a pressurized water reactor. To identify the system pressure characteristics on nitrogen gas-water and steam-water pressurizers for an integral pressurizer water reactor, an experimental study is carried out on the integral test facility, VISTA-ITL, which is a scaled down facility of the SMART plant. According to the experimental results, the pressurization rate and peak pressure are affected by the pressurizer type. The pressurization and depressurization rate with the steam pressurizer is smaller than the gas pressurizer since water has its inherent characteristic having latent heat while a nitrogen has ideal gas properties. The safety valve in both the gas and steam pressurizers is not opened during the transient. It is one of the characteristics of the integral reactor SMART design with a large pressurizer volume. The cooldown rate of the primary coolant system during the later part transient depends on the decay heat level regardless of the pressurizer type.

A 330 MWt integral type nuclear power plant, SMART, was developed for electricity generation and seawater desalination. Advanced design concepts were adopted such as an integral arrangement of the major components, and a passive residual heat removal system (PRHRS) to enhance the safety capability. The TASS/SMR code was developed using various thermal-hydraulic models reflecting the design features of SMART, such as the condensate heat exchanger in the passive residual heat removal system. The development and validation of the condensate heat exchanger model were performed using POSTECH and IIT heat transfer test results. The TASS/SMR code predicted well or slightly under-predicted the heat transfer coefficient at the condensate heat exchanger shell side compared with the experimental data. The heat transfer correlation considering the tube bundle effect improved the prediction of the heat transfer for a vertical tube bundle geometry.

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 system-integrated modular advanced reactor (SMART), which was developed by Korean Atomic Energy Institute (KAERI), has unique characteristics because of its integral type geometry. A series of design basis accident tests have been conducted using the SMART integral test loop facility to assess the performance of the safety systems in SMART. Using these tests data, system analysis codes have been validated. In this paper, validation with the MARS and TASS/SMR codes was performed for the test scenario of pressurizer safety valve line break conducted at the SMART-ITL (Integral Test Loop), which is an integral test facility for SMART. The two codes gave a reasonable prediction for the primary pressure and break flow. However, the water level in the reactor pressure vessel was under-predicted slightly.

Small reactors are a suitable option for nuclear system deployment in developing countries or non-electrical applications for various facilities. SMART is one of the small integral type reactors to apply flexibly local power demands or sea water desalination. A thermal hydraulic analysis code, TASS/SMR, having SMART specific models, was developed to simulate thermal hydraulic phenomena of the SMART plant. The improved TASS/SMR code predicts well the system behaviors under two-phase conditions compared with the OSU-MASLWR experimental results. A small break LOCA simulation of the SMART plant is improved a void distribution, a break flow, and a collapsed water level in the core.

SMART is a small sized integral type pressurized water reactor and its design minimizes the potential for an occurrence of transients and accidents. To achieve these purposes, engineered safety features enhance the accident resistance. A large volume of semi-passive pressurizer can accommodate a wide range of pressure transients. Low core power density increases the thermal margin, and negative fuel and moderator temperature coefficients yield beneficial effects on a core self-stabilization, and limit the reactor power during accidents. Considering the enhanced engineered safety features and design characteristics, anticipated transient without scram scenarios are adopted to evaluate the strength of the SMART design. The peak pressure does not exceed 116% of the nominal pressure and the fuel can maintain its integrity for the transient. The intrinsic SMART design characteristics having passive safety features and a large primary coolant inventory can maintain a stable condition safely although the reactor protection and diverse protection systems are not available for the anticipated transient without scram transients.

A flow mixing header assembly (FMHA) is installed in a system-integrated modular advanced reactor (SMART) to enhance the thermal mixing capability and create a uniform core flow distribution under both normal operation and accident conditions. In this study, the thermal mixing characteristics of the FMHA are investigated for various steam generator conditions using a commercial CFD code. Simulations include investigations for the effects of FMHA discharge flow rate differences, turbulence models, and steam generator conditions. The results of the analysis show that the FMHA works effectively for thermal mixing in various conditions and makes the temperature difference at the core inlet decrease noticeably. We verified that the mixing capability of the FMHA is excellent and satisfies the design requirement in all simulation cases tested here.

SMART was developed for electricity generation and seawater desalination and adopted a passive system to enhance its safety. This system could passively remove decay heat from the reactor core to the emergency cooldown tank (ECT) through the heat exchanger. A natural circulation flow was established as water covered the tube bundle inside the emergency cooldown tank. Heat transfer tests for the upward straight tube bundle in the emergency cooldown tank were performed to find the characteristics of the passive system design under natural circulation conditions. The heat transfer coefficient at the tube bundle was affected by the cooling water temperature, and the radial location of the tube. However, it has nearly a similar value at the bottom region regardless of the tube location. The average heat transfer coefficient for the tube bundle was slightly higher than that for the single tube owing to the turbulence effect among the tube bundles.

A safety evaluation of SMART adopting a fully passive safety system was performed for a small-break loss of coolant accident (SBLOCA) with various break locations and sizes using the MARS code. It is necessary for SMART, adopting a fully passive safety system composed of passive safety injection system (PSIS), automatic depressurization system (ADS), and passive residual heat removal system (PRHRS), to satisfy the passive safety performance requirements, i.e., the capability to maintain safe shutdown conditions for a minimum of 72 h without AC power supply or operator action in the case of a design basis accident. A number of SBLOCAs of different locations and sizes were analyzed using the MARS code. The results of the break spectrum analyses showed that the collapsed liquid level inside the core support barrel is maintained high enough above the top of the core owing to the sufficient passive safety injection flow from the core makeup tank (CMT) and safety injection tank (SIT). Therefore, the core is not uncovered for 72 h after the break without AC power supply or operator action, resulting in a continuous decrease of fuel cladding temperature throughout the transient.

A helically coiled once-through steam generator has been used widely during the past several decades for small nuclear power reactors. The heat transfer characteristics and dryout conditions are important to optimal design a helically coiled steam generator. Various experiments with the helically coiled tubes are performed to investigate the heat transfer characteristics and occurrence condition of a dryout. For the investigated experimental conditions, Steiner–Taborek’s correlation is predicted reasonably based on the experimental results. The pressure effect is important for the boiling heat transfer correlation. A boiling heat transfer tends to increase with a pressure increase. However, it is not affected by the pressure change at a low power and low mass flow rate. Dryout occurs at high quality test conditions investigated because a liquid film on the wall exists owing to a centrifugal force of the helical coil.

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...

Small reactors can apply to local power demands or remote areas. SMART, which can produce 90 MWe of electricity and 40,000 tons/day of sea-water desalination for a 100,000 population city, is a promising advanced integral type small reactor. The thermal hydraulic analysis code with a drift flux model, TASS/SMR, was developed for a conservative simulation of a small break loss of coolant accident in SMART. Taking into account SMART-specific inherent characteristics, the code adopts the Chexal–Lellouche correlation for a drift flux model. The capability of TASS/SMR code is validated using the results of the experimental data. The code predicts conservatively the void distribution compared with the experimental data. TASS/SMR calculation predicts reasonably major phenomena for the SBLOCA and is more conservative than or nearly the same as the results of the best estimated realistic system code, MARS. TASS/SMR code can be used for a SBLOCA analysis of SMART.

A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident.

SMART is a 330 MWt advanced integral PWR. An experimental validation for the SMART-specific thermal hydraulic phenomena is performed to generate background information required for SMART standard design approval. Series of fundamental experiments are carried out. The two-phase heat transfer at the helically coiled tube is affected linearly by a heat flux and system pressure. However, it has nearly the same value at the high quality region regardless of the pressure change. The condensate heat transfer depends on the tube diameter and the mass flux at the PRHRS heat exchanger. The condensate heat transfer increases at a high steam flow, low pressure and large diameter. The calculation results of MARS and TASS/SMR codes are predicted to have similar values because the two codes incorporate the same heat transfer correlations for a helically coiled tube and condensate heat exchanger.

TASS/SMR-S is thermal-hydraulic design code for safety analysis of SMART plant. SBLOCA is considered as the design basis accident in the SMART. The capability of TASS/SMR-S for SBLOCA analysis was assessed using LOFT L3-7 as TMI action plan. Because TASS/SMR-S has been developed only for SMART, which is an integral reactor with helical steam generator heat exchanger, the steam generator model of TASS/SMR-S cannot be used directly for LOFT experiment that involves the use of U-tube steam generator. Therefore, the general heat structure model of TASS/SMR-S was used for modeling the SG heat exchanger. Nevertheless, TASS/SMR-S predicted important thermal-hydraulic parameters such as system pressure, fluid temperatures, and cladding temperature of reactor within reasonable error ranges. Further, TASS/SMR-S was applied to simulate SBLOCA in the SMART plant. In this simulation, thermal hydraulic parameters similar to those predicted in LOFT L3-7 were predicted.

To identify the flow characteristics of the SMART reactor, a flow distribution model test and a numerical simulation are performed in KAERI. Among several part of the SMART reactor, the fuel assemblies are simulated using simulators because of the complexity. The geometries of the core in the SMART reactor and simulator are different, but some similarities are maintained such as the ratio of pressure drop in the vertical and cross directions. There are cross flow holes in each core simulator to reproduce the cross flow of SMART fuel assemblies. To know the flow characteristics of the cross flow, numerical analysis is performed. As the cross flow area is decreased, the pressure drop between inlet and outlet is decreased. Also, when the flow imbalance between two core simulators is constant, the cross flow area does not significantly affect the cross flow.

An advanced integral pressurized water reactor (PWR) of a small size (∼330 MWt) is being developed by the Korea Atomic Energy Research Institute (KAERI). The purposes of the reactor are a sea water desalination and an electricity generation. To enhance its safety, many advanced design concepts are introduced such as a passive residual removal system and a low power density core. For the safety validation of the designed reactor, a system analysis code named TASS/SMR, was developed. TASS/SMR code uses a one dimensional node/path modeling for the thermal hydraulic calculation and point kinetics for the core power calculation. The code also has specific models for the developed integral reactor, such as a helical tube heat transfer model and a passive residual heat transfer model. One of the important models for the safety or performance calculation is the core heat transfer model. The core heat transfer model of TASS/SMR was developed to meet the requirements of the 10 CFR 50 appendix K EM model as well as the realistic models. The developed model was validated with experimental data. The results show that the model predicts the heat transfer phenomena in the reactor core with a reasonable conservatism.

An experiment on a heat transfer and a natural-circulation performance for an integral-type reactor have been carried out using the VISTA facility, and the calculated results using the TASS/SMR and...

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.

Advanced integral-type pressurized water reactor with a maximum thermal power of 65 MW is under development at the Korea Atomic Energy Research Institute (KAERI). This 65 MW integral reactor incorporates a number of innovative design features. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the TASS/SMR code and a sensitivity analysis has been performed to evaluate a passive cooldown capability for various system conditions such as natural and forced circulation conditions for the reactor coolant system or the passive residual heat removal system, and number of active PRHRS trains. The reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation and the 65 MW integral reactor can cool the coolant to the SCS entry condition in the primary system for all the possible operational conditions.

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.

An integral type reactor, which is an innovative design to achieve a high degree of safety, is currently being developed at the Korea Atomic Energy Research Institute. A feedwater pipe break accident is one of the most important accidents regarding the safety of an integral type reactor. A best estimated calculation, a conservative calculation, and a parameter study for a feedwater pipe break have been carried out. The sensitivity analysis in this paper performed is to establish the parameters which greatly affect the feedwater pipe break accident. A power level, an initial system pressure, a moderator reactivity coefficient and a break size are the major parameters which maximize a system pressure. The important function that must operate following a feedwater pipe break accident is an opening of the pilot operated safety relief valves, and an initiation of the passive residual heat removal system. The integral reactor safety systems function properly and thus secure the reactor to a safe condition with ...

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.

An investigation of the thermal hydraulic characteristics and the natural circulation performance in the passive residual heat removal system (PRHRS) for an integral type reactor have been carried out using the VISTA facility and the calculated results using the MARS code, which is a best estimate system analysis code have been compared with the experimental results. The VISTA facility consists of the primary, secondary, and the PRHRS circuits, to simulate the SMART design verification program. The experimental results show that the fluid is well stabilized in the PRHRS loop and the PRHRS heat exchanger accomplishes well its functions in removing the transferred heat from the primary side in the steam generator as long as the heat exchanger is submerged in the water in the emergency cooldown tank (ECT). The decay heat and the sensible heat can be sufficiently removed from the primary loop with the operation of the PRHRS. The MARS code predicts reasonably well the characteristics of the natural circulation in the PRHRS. From the calculation results, most of the heat transferred from the primary system is removed at the PRHRS heat exchanger by a condensation heat transfer.

The system-integrated modular advanced reactor (SMART) new phase (SMART-P) with a rated thermal power of 65.5 MW is currently being developed at the Korea Atomic Energy Research Institute. It is an innovative design to achieve a high degree of safety by adopting inherent safety-improving features and passive safety systems. Realistic and conservative calculations and a parameter study for a steam-line pipe break have been carried out by means of the TASS/SMR code. A set of transients for the whole system of SMART-P is investigated from the point of view of fuel integrity. The results of the analyses show that the most conservative initial conditions are thermal design flow, high system pressure, high coolant temperature, and high core power. It is also assumed that off-site power is unavailable and the steam section pipe guillotine break with the least reactive control rod assembly stuck out in the fully withdrawn position is a limiting case under the most moderator density reactivity condition. The SMART-P safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters such as the critical heat flux ratio and the pressure. Natural circulation is well established in the primary and passive residual heat removal systems and is enough to ensure a stable plant shutdown condition after a reactor trips.

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.

An investigation of the thermal hydraulic characteristics in the passive residual heat removal system of the System integrated Modular Advanced ReacTor-P (SMART-P) has been carried out using the MARS code, which is a best estimate system analysis code. The SMART-P is designed to cool the system during accidental conditions by a natural convection. The dominant heat transfer in the steam generator is a boiling mode under a forced convection condition, and it is a single-phase liquid and a boiling heat transfer under a natural convection condition. Most of the heat is removed in the heat exchanger of the passive residual heat removal system by a condensation heat transfer. The passive residual heat removal system can remove the energy from the primary side as long as the heat exchanger is submerged in the refueling water tank. The mass flow is stable under a natural circulation condition though it oscillates periodically with a small amplitude. The parameter study is performed by considering the effects of an effective height between the steam generator and the heat exchanger, a hydraulic resistance, an initial pressure, a non-condensable gas fraction in the compensating tank, and a valve actuation time, which are useful for the design of the passive residual heat removal system. The mass flow in the passive residual heat removal system has been affected by the height between the steam generator and the heat exchanger, and the hydraulic resistance of the loop.

SMART is an integral type reactor of 330 MW, which enhances its safety by adopting inherent safety design features. Thermal hydraulic characteristics of transients in heat removal by a secondary system for the SMART have been carried out by means of the TASS/SMR and MATRA codes. The primary, secondary, and passive residual heat removal systems RHRS of the SMART were modeled properly. Then, a set of transients for the whole system was investigated. The results of the analyses using the conservative initial and boundary conditions showed that the safety features of the SMART design carried out their functions well and there was a strong moderator temperature coefficient due to the soluble boron free reactor affected by the transient behavior. The natural circulation was well established in the primary and passive residual heat removal systems during the transients and was enough to ensure a stable plant shutdown condition after a reactor trip.

Vortex shedding behind a square cylinder in an oscillatory incoming flow is studied at Reynolds numbers of 80, 100 and 200 by numerical solutions. The forcing frequencies investigated for the oscillatory incoming flow are ff/fso=0.4∼4.0 and the forcing amplitudes normalized by the mean incoming velocity are 0.1, 0.2 and 0.4, where fso is the vortex shedding frequency under steady incoming flow and ff is the forcing frequency of the oscillatory incoming flow. In the lock-on region, the flow is in a periodic state and there is a strong regularity between the drag and lift forces. The different phase diagram between the drag and lift forces is brought on by the different lift mode. The phase difference between the incoming velocity and the drag is nearly the same magnitude of 65∼70 degrees in the lock-on region. The time-averaged mean recirculation region in the oscillatory incoming flow is smaller than that for the steady one and is inversely proportional to the forcing amplitude or the Reynolds number. The time-averaged streamwise mean velocity in the oscillatory incoming flow recovers more quickly than that in the steady incoming flow.

SMART (System-integrated Modular Advanced ReacTor) is an integral light water reactor developed by KAERI for both electricity generation and various applications. SMART can produce 90MWe of electricity and 40,000 tons/day of sea-water desalination, which is deemed suitable for a city population of 100,000. The SMART project aims at standard design approval. The experimental validation for SMART specific thermal hydraulic phenomena is to generate background information required for SMART licensing. The objective of the experimental validation is to demonstrate the safety and performance of the SMART-specific system and component designs, and to provide a database for the validation of the design tools, such as a system analysis code. From the experimental results, it is proved that SMART was designed properly and the data are used to approve the validation of the system analysis code. Also, the experimental results were submitted to a regulatory body as technical background information of the SMART standard design, and they enhance the licensing possibility of this design.

This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties.