Won Pil Baek

Simulation Capability of the ATLAS Facility for Major Design-Basis Accidents

A thermal-hydraulic integral-effect test facility [advanced thermal-hydraulic test loop for accident simulation (ATLAS)] is being constructed at the Korea Atomic Energy Research Institute. The ATLAS is a one-half-reduced-height and 1/288-volume-scaled test facility based on the design features of the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. The simulation capability of the ATLAS for major design-basis accidents (DBAs), including a large-break loss-of-coolant accident and direct vessel injection line-break and main-steam-line-break accidents, is evaluated by the best-estimate system code MARS with the same control logics, transient scenarios, and nodalization scheme. The validity of the applied scaling law and the thermal-hydraulic similarity between the ATLAS and the APR1400 for the major DBAs are assessed. It is confirmed that the ATLAS can maintain an overall similarity with the reference plant APR1400 for the major DBAs considered in the study. However, depending on the accident scenarios, there are some inconsistencies in certain thermal-hydraulic parameters, such as cladding temperature, subcooling at the lower plenum of the core, break flow rate, core and downcomer water level, and secondary pressure. The causes of the inconsistencies are carefully investigated by considering the detailed design features of the ATLAS. It is found that the inconsistencies are mainly due to the reduced power effect and the increased stored energy in the structure. The similarity analysis was successful in obtaining a greater insight into the unique design features of the ATLAS and would be used for developing optimized experimental procedures and control logics.

Hybrid accident simulation methodology using artificial neural networks for nuclear power plants

A hybrid accident simulation methodology for nuclear power plants is proposed to enhance the capabilities of compact simulator by introducing artificial neural networks. Two neural networks are trained with the target values obtained from the analyses of detailed computer codes and trained results are combined with the compact simulator to perform the following roles: (i) compensation for inaccuracies of a compact simulator occurring from simplified governing equation and reduced number of physical control volumes, and (ii) prediction of the critical parameter usually calculated from the sophisticated computer code: the autoassociative neural network improves the computational results of the compact simulator up to the accuracy level of detailed best estimate computer code, while the backpropagation neural network predicts the minimum departure from nucleate boiling ratio (DNBR). Simulations are carried out to verify the applicability of the proposed methodology for the loss of flow accidents and the results show that the neural networks can be used as a complementary tool to improve the results of a compact simulator.

Effects of Break Size on Direct Vessel Injection Line Break Accidents of the ATLAS

Abstract The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break among the analyzed small-break loss-of-coolant accidents in safety analysis. The first-ever integral effects test database for various DVI line break sizes from 5% to 100% was established with the Korea Atomic Energy Research Institute’s Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) test facility. This database enhances our physical understanding of the major thermal-hydraulic behaviors of the APR1400 during DVI line break accidents, and it can also be used to examine the prediction capabilities and identify any deficiencies in the existing best-estimate safety analysis codes. Effects of the break size were experimentally investigated, and the best-estimated MARS code was assessed against the experimental database. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than the data just before a loop seal clearing occurs, and it also produced a more rapid decrease in the downcomer water level than the data. These disagreements are the expected consequence of uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effects test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology on the DVI line break accidents of the APR1400.

Assessment of a tube-based bundle CHF prediction method using a subchannel code

Abstract At the conceptual design stage for advanced water-cooled reactors (AWCRs), a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. A basic idea for general CHF prediction is to utilize the tube-based CHF models covering wide applicable ranges with the help of supplementary terms for bundle effects. In this study, feasibility assessments have been performed with the bundle CHF data relevant to pressurized water reactors. A subchannel analysis is adopted in order to be able to consider the geometrical variations properly even in untested fuel bundle geometries. As a result, a CHF look-up table method (i.e., the use of a round tube CHF table with appropriate bundle effect factors) turns out to be a promising way to fulfill the needs in many aspects among some selected correlations and theoretical models. Though improvements of the supplementary factors, especially for the cold wall and the bundle heated length effects, are desirable to make better predictions, the tube-based bundle CHF prediction method clearly shows a potential as a general CHF predictor.

An integral equation model for critical heat flux at subcooled and low quality flow boiling

A new theoretical model of critical heat flux (CHF) is developed for the flow boiling condition from bubble-detached to low quality range. The CHF condition is postulated to occur when the superheated liquid layer on the heated wall, which is formed under the bubbly layer from the point of the onset of significant void generation, is depleted due to the evaporation along the heated length. The model shows a very promising agreement with the uniformly heated round tube data for both water and refrigerants by simply applying well-known constitutive relationships without any tuning constant for the CHF data. The significance of the proposed model in unifying the existing models is also discussed.

A SUMMARY OF 50 th OECD/NEA/CSNI INTERNATIONAL STANDARD PROBLEM EXERCISE (ISP-50)

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: “blind” and “open” phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.

An Integral Effect Test on the Reflood Period of a Large-Break LOCA for the APR1400 Using ATLAS

A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), has been constructed at the Korea Atomic Energy Research Institute. It is a 1/2-reduced-height and 1/288-volume-scaled test facility based on the design features of APR1400, an evolutionary pressurized water reactor developed by the Korean industry. ATLAS was used to perform a set of integral effect tests on the reflood period of a large-break loss-of-coolant accident (LBLOCA) after intensive performance tests had been conducted to verify ATLAS’s operational performance and controllability for major thermal-hydraulic components. The present LB-CL-09 test is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period that can be used to provide reliable data to help validate the LBLOCA analysis methodology for APR1400. The main objective of the present test is to identify the major thermal-hydraulic characteristics such as the direct emergency core coolant (ECC) bypass, downcomer boiling, and core cooling behavior during the reflood phase of an LBLOCA for APR1400 under conditions where the downcomer region interacts with the reactor core region and the heat could be transferred through the steam generator. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results. The decay heat and the ECC flow rate from the safety injection tank were simulated from the start of the reflood period. The ECC flow rate from the safety injection pump was 0.32 kg/s. The system pressure was fixed at ~0.1 MPa, and the initial outer-wall temperature was determined to be 205°C. The experimental results showed the typical thermal-hydraulic trends expected to occur during the reflood phase of the LBLOCA scenario.