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[en] The present paper performs analytical evaluations for the potential distortions caused by the scaled models using RELAP5/MOD3 computer codes. By use of scaling analysis, two scaled models with same volumetric ratio are constructed for Korean Next Generation Reactor (KNGR), which is an advanced light water reactor. The scaling methodology adopted in this paper preserves two-phase natural circulation similarities between prototype and scaled models. One scaled model is at full height with reduced flow area. The other model is at reduced height with reduced flow area. By using appropriate scale factors the RELAP5/MOD3 input models are developed. Then, the transient responses of two ideal scaled models are simulated for Small Break Loss of Coolant Accident (SBLOCA) by using RELAP5/MOD3 computer code. The transient responses of two scaled models are compared with those of the prototype. The results indicate that qualitative and quantitative similarities are well preserved for both models during SBLOCA with different break sizes
[en] Highlights: • TASS/SMR is 1-dimensional system analysis code for the integral reactor, SMART. • Capability of TASS/SMR for subcooled boiling prediction is validated using KIT and FRIGG experimental data. • TASS/SMR code predicts well void distribution along the height for subcooled boiling flow. - Abstract: SMART, which was an advanced integral type small modular PWR, was developed by KAERI (Kim et al., 2014). To analyze the thermal hydraulic phenomena including behaviors at the SMART specific components, the TASS/SMR code which can predict a heat transfer for various thermal hydraulic conditions, has been developed. Information of the void distribution in a subcooled boiling flow is important in predicting the inception of a two-phase flow and an onset of the critical heat flux condition. The TASS/SMR code adopts an energy partitioning method and a critical enthalpy correlation determining a point of net vapor generation for subcooling conditions. A range of the subcooling degree investigated is 1.5–50.6 K to validate the method for a subcooled boiling flow prediction. The TASS/SMR code predicts well the void distribution along the height for the subcooled boiling flow conditions compared with the experimental data. The predicted location of the onset of void generation is simulated well at most investigated conditions and delayed slightly at the very high subcooling condition.
[en] Highlights: • SMART plant and TASS/SMR code have been developed by KAERI. • Capability of TASS/SMR code is validated using POSTECH and ITT experimental results. • The code predicts well heat transfer for a vertical tube bundle under natural circulation. • The correlation considering the tube bundle effect improves the prediction of the heat transfer. - Abstract: 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.
[en] Highlights: • VIATA-ITL constructed to simulate the SMART plant, which is an integral type PWR. • TASS/SMR code is developed to analyze thermal hydraulic phenomena of the SMART plant. • TASS/SMR code predicts well the results of the VISTA-ITL loss of coolant flow test. - Abstract: 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.
[en] The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs
[en] A set of EOPs will be constructed based the fundamental (or critical) safety functions of a nuclear reactor: the safety functions, whatever the reactor is, are the same: 1) controlling reactivity (or power), 2) cooling the radioactive materials, i.e., fuels and confining the radioactive material. The SMART with unique design features and operating characteristics of integral reactors has most of design basis events in common with conventional commercial loop-type PWR plants. This paper presents a preliminary strategy to develop Emergency Operating Guidelines of SMART covering safety functions, defense-in-depth, and sorts of accidents handled by the accident management. The references of EOPs were studied for a wide range of nuclear reactors with the Korean regulatory requirements and a set of safety functions for SMART reactor were identified.
[en] A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)
[en] Pre-test analysis using a 1-D moduleof MARS 2.0 code has been performed for the KNGR (Korean Next Generation Reactor) DVI (Direct Vessel Injection) performance test facility which is a full height and 1/24.3 volume scaled separate effects test facility focusing on the identification of multi-dimensional thermal-hydraulic phenomena in the downcomer during the reflood conditions of a large break LOCA. From the steady state analyses for various test cases at the late reflood condition, the degree of major thermal-hydraulic phenomena such as ECC bypass, ECC penetration, steam condensation, and water level sweep-out are quantified. The MARS code analysis results showed that: (a) multi-dimensional flow and temperature behavior occurred in the downcomer region as expected, (b) the proximity of ECC injection to the break caused more ECC bypass and less steam condensation efficiency, (c) increasing the steam flow rate resulted in more ECC bypass and less steam condensation, and (d) the high velocity of steam flow swept-out the water in the downcomer just below the cold leg nozzle. These results are comparable with those observed in the previous tests such as UPTF and CCTF
[en] Highlights: • Heat transfer characteristics and dryout for helically coiled tube are performed. • A boiling heat transfer tends to increase with a pressure increase. • Dryout occurs at high quality test conditions investigated. • Steiner–Taborek’s correlation is predicted well based on the experimental results. - Abstract: 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
[en] Various best estimate small break LOCA analyses are performed for Korean Next Generation Reactor using CEFLASH-4AS/REM and RELAP5/MOD3 computer codes. The KNGR is an Advanced Light Water Reactor adopting the advanced design feature of DVI configuration for ECCS. The study is performed to evaluate the best estimate performance of KNGR ECCS and to investigate the real physical phenomena expected to occur during the small break LOCA. The analysis results are qualitatively in good agreement until the loop seal clearing, and the general trends are similar after that in the sense that a DVI line break case results in a more rapid system depressurization and larger system inventory loss. However, the two phase mixture level transients estimated from the RELAP5/MOD3 results are very different from those of CEFLASH-4AS/REM. After the loop seal clearing, both of the broken and intact downcomer mixture levels drop to the cold leg elevation. Thus, the steam from the cold legs is more easily vented and the inner vessel mixture level is maintained well above the core throughout the transient. However, details of transient behavior are different depending on nodal scheme of downcomer node and break flow model which have primary importance in predicting the DVI performance during the small break LOCA. The comparative analysis for the performance between the DVI and CLI configuration show that the DVI is beneficial for cold leg break in system inventory and core water level, but, since it also requires the analysis of DVI line break which shows the worst core water level, it does not have significant advantage in terms of small break LOCA performance. However, to identify the real physical phenomena expected to occur during the SBLOCA with DVI configuration of ECCS, Further detailed analysis using a three-dimensional code is recommended. (Author). 10 refs., 6 tabs., 96 figs