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Session Overview
8.08: Safety
Tuesday, 17/Mar/2020:
3:30pm - 5:00pm

Session Chair: Maria Lorduy Alos, Universitat Politècnica de València, Spain
Location: B-1049

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Simulation of a SBLOCA Scenario in Different Technology Facilities

Maria Lorduy, Sergio Gallardo, Gumersindo Verdu

Universitat Politècnica de València

Over the last decades, many tests have been conducted in integral test facilities (ITF), which reproduce the behavior of their reference nuclear power plants under accidental scenarios. However, scaling is still an important concern due to the distortions that result from assumptions and simplifications in the scaling methods to design of facilities. Faced with this problem, counterpart tests contribute to address scaling issues and provide confidence in extrapolating to a full-scale plant.

This work presents the analysis and simulation with the thermal-hydraulic code TRACE5 of a counterpart test between the LSTF and ATLAS facilities. The experiment at issue reproduces a 1% small break loss-of-coolant accident (SBLOCA) in a cold leg. Furthermore, the high-pressure injection system failure and no inflow of non-condensable gas from the accumulators are assumed. In this scenario, the first accident management (AM) action is the manually controlled depressurization of the secondary system, followed by the auxiliary feedwater injection in both steam generators. As the cooling effect is significant, the injection of the accumulators and the LPI system are activated when the primary pressure is reduced to 4.51 MPa and 1.21 MPa, respectively.

The main objective is to investigate the thermal-hydraulic phenomena during a cold leg SBLOCA such as the loop seal clearing, the core heat-up, and the effectiveness of the AM actions. This study is completed with the test simulations and sensitive analysis of the initial and boundary conditions of the Test A5.1 in ATLAS that are determined by applying scaling ratios from the three-level scaling methodology to the reference conditions in the Test SB-CL-32 in LSTF.

The comparison between the experimental and simulation results shows the TRACE5 code capability to reproduce the main thermal-hydraulic phenomena during the test, therefore, they are a contribution to its assessment and validation.

Evaluation of Passive Containment Cooling System with Phase Change Material

Sung Gil Shin, Jai Oan Cho, Jeong Ik Lee


The necessity of the Passive Containment Cooling Systems (PCCSs) has been emphasized since the Fukushima Daiichi nuclear power plant accident. In response, the KAIST research team proposed a new PCCS concept using a phase change material (PCM) as a heat removal source. To apply PCM-based PCCS in the containment building, heat transfer performance of PCM-based PCCS should be first evaluated if the system can meet the safety system design criteria. In this study, PCM-based PCCS is modeled with a containment thermal hydraulic code and accident analysis is conducted to evaluate if the PCM-based PCCS has satisfactory heat transfer performance. Domestic and international design requirements related to PCCS are used as design criteria of PCM-based PCCS. The accident analysis is performed for an SBLOCA condition, during which the energy released inside the containment building is the greatest among many design basis accidents (DBAs). Maximum pressure and temperature (Max PT) analysis is performed for the case of double-ended suction leg slot break (DESLSB) and double-ended discharge leg slot break (DEDLSB). As a result of performing the max PT analysis, the maximum temperature and pressure in the case of DEDLSB is larger than that of DESLSB, so the design basis accident of PCM-based PCCS is determined with DEDLSB case. A long-term cooling analysis is conducted for DEDLSB. The utilized containment thermal hydraulic code is CAP (nuclear containment analysis pack) version 2.21. CAP is a lumped-parameter (LP) code developed by Korean industrial consortium for the analysis of thermal hydraulic behavior in the containment.

Research on the Heat Transfer of Passive Residual Heat Removal Heat Exchanger

Chuanxin Peng, Zan Yuanfeng, Zhuo Wenbin

Nuclear Power Institute of China

The heat transfer characteristics of the passive residual heat removal heat exchanger were studied in this paper. The heat transfer characteristics and related parameters of heat exchanger under different inlet steam condition(temperature, pressure, flowrate) and different water temperature and levels in the containment tank were obtained. And the flow heat transfer model of the primary and secondary side of the passive residual heat removal heat exchanger were established. The calculation results show that the prediction error was less than 20 percent.

Maximum heat transfer rate expected on porous oxide surfaces, for corrosion-resistance, considering a BA-added environment in NPP

Dong Hoon Kam1, Hyoung Suk Yu1, Yong Hoon Jeong1, Yang Jeong Park1, Jung Woo Kim1, Sung Oh Cho1, Yacine Addad2


Zr-based alloy, a cladding material, in NPPs undergoes rapid oxidation at high temperature steam condition. The reaction is an exothermic phenomenon, that surface temperature continuously increases, which induces further accelerated oxidation. As a byproduct, hydrogen is generated during the process, and without proper remedies, there is increased possibility of hydrogen explosion inside the CTMT. As experienced in Fukushima accident in Japan, appropriate and timely mitigation systems are required to keep the integrity. To mitigate or prevent the hydrogen explosion in NPPs, several ideas and corresponding approaches have been carried out so far. One of method is improving the present cladding material, that roles as a hydrogen source during the progression, to prevent, or at least mitigate, the damage. In this regard, an anodizing method has been proposed by the KAIST group to preform intact oxide layers on the cladding surface for the ATF cladding concept. By controlling the production conditions, stable and intact oxide layers can be made on the surface, and porous structures can also be generated to enhance thermal margin altogether. Mitigation of the oxidation has been confirmed in preceding studies, and in this study, CHF values considering real environments in NPPs have been summarized. When boric acid is added in the coolant, the boiling time effect diminishes beyond certain periods of time, and bare and anodized surfaces show different trends, respectively; noticeable effect is observable only on the bare surface. With SEM images and EDS results, accelerated oxidation phenomenon has been confirmed on the bare surface with the boric acid-added coolant. Stable oxide layers on the surface for the anodized ones prevent the additional oxidation in the experimental range.

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