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To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

An experimental simulation for a loss of coolant accident with a 4-in cold leg top-slot break was performed using ATLAS, an integral effect test facility at the Korea Atomic Energy Research Institute (KAERI), South Korea. The aims of this test are to resolve a safety issue and to validate the thermal hydraulic safety analysis codes for the activity of the 4th ATLAS-Domestic Standard Problem. The related safety issue for this test is that a loop seal clearing and loop seal reformation of APR1400 can lead to a core temperature excursion because of the characteristics of the loop seal geometry. In the experimental results, the core heater temperature slowly increased after loop seal reformation. This increase is attributed to the accumulated steam at the upper head. The accumulated steam increased the pressure and saturated the temperature in the core. The core heater temperature then increased due to the increased saturated temperature. This temperature increase, therefore, was not a core temperature excursion. Sensitivity analysis results using Multidimensional Analysis of Reactor Safety KINS Standard (MARS-KS), which is a best-estimate thermal hydraulic system analysis code, are also presented in this study. In the 4th ATLAS-Domestic Standard Problem, 15 participants performed calculations and sensitivity analyses using various thermal hydraulic safety analysis codes (MARS-KS, SPACE, RELAP5, and TRACE). From this exercise, several parameters that affect the calculation results are brought out. The sensitivity analysis results using MARS-KS with some of these parameters are also presented. The MARS-KS calculation results are similar to the experimental data. The effects of critical flow model, break line modeling, and fine node modeling are also discussed.

The advanced thermal-hydraulic test loop for accident simulation (ATLAS) was developed and operated at The Korea Atomic Energy Research Institute. The ATLAS is operated to simulate accidents in pressurized water reactors (PWRs). A domestic standard problem (DSP) using the ATLAS was proposed for transferring the database of the integral effect test to Korean nuclear researchers and developing the safety analysis methodology of PWRs. The fourth DSP (DSP-04) exercise was performed during 2015–2017 with 15 participants (13 organizations), that are universities, government, and nuclear industries. In DSP-04, a top-slot break at the cold leg was chosen as the target scenario to resolve an issue about the effect of loop seal clearing and the reformation on the peak cladding temperature during the cold leg top-slot break LOCA for APR1400. The participants performed a code calculation for the experimental simulation and sensitivity studies for the enhancement of the code. This paper includes brief information about the experimental and major code assessment results.

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR +) and is designed to completely replace a conventional, active auxiliary feedwater system. With the aim of validating the cooling and operational performance of the PAFS, a separate effect test facility, the PAFS Condensing heat removal Assessment Loop (PASCAL), was constructed by simulating a single passive condensation heat exchanger (PCHX) tube submerged in the passive condensation cooling tank (PCCT) according to the volumetric scaling methodology. During heat removal of the PAFS, the pool water in the PCCT plays a role in the ultimate heat sink of a decay heat. In this study, the effect of the PCCT water level on the cooling performance of the PAFS was experimentally investigated with the PASCAL facility. Quasi-steady-state and PCCT level decrease test cases were sequentially performed by varying the steam generator heater power from 300 to 750 kW to investigate the thermal-hydraulic behavior during the decrease of the PCCT water level. From the experimental results, it was found that the decrease of the PCCT water level enhanced evaporative heat transfer at the outer wall of the PCHX tube by reducing the degree of subcooling around the PCHX. That induced an increase of the heat removal rate by the PCHX during the transient. Thus, it can be concluded that the current design of the PCHX in the PAFS has sufficient capacity to cool down the decay heat during the whole transient of the PCCT water level decrease. [ABSTRACT FROM AUTHOR]

This study concerns the development of an integrated calculation methodology with which to continually and consistently analyze the progression of an accident from the design-basis accident phase via core uncovery to the severe accident phase. The depletion rate of reactor coolant inventory was experimentally investigated after the safety injection failure during a large-break loss-of-coolant accident utilizing the Seoul National University Integral Test Facility (SNUF), which is scaled down to 1/6.4 in length and 1/178 in area from the APR1400 [Advanced Power Reactor 1400 MW(electric)]. The experimental results showed that the core coolant inventory decreased five times faster before than after the extinction of sweepout in the reactor downcomer, which is induced by the incoming steam from the intact cold legs. The sweepout occurred on top of the spillover from the downcomer region and expedited depletion of the core coolant inventory. The test result was simulated with the MAAP4 severe accident analysis code. The calculation results of the original MAAP4 deviated from the test data in terms of coolant inventory distribution in the test vessel. After the calculation algorithm of coolant level distribution was improved by including the subroutine of pseudo pressure buildup, which accounts for the differential pressure between the core and downcomer in MAAP4, the core melt progression was delayed by hundreds of seconds, and the code prediction was in reasonable agreement with the overall behavior of the SNUF experiment. [ABSTRACT FROM AUTHOR]