Conference Agenda

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Session Overview
6.02: Severe accident analysis
Tuesday, 17/Mar/2020:
3:30pm - 5:00pm

Session Chair: Mirco Grosse, Karlsruhe Institute of Technology, Germany
Session Chair: Tunc Aldemir, Ohio State University, United States of America
Location: B-1048

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Severe Accident Mitigation System and Analysis for Korean Small Integral Reactor

Rae-Joon Park, Jaehyun Ham, Sang Ho Kim, Jong-Hwa Park


The Korean small integral reactor of SMART (System-integrated Modular Advanced ReacTor)100 has been developed, which contains core, reactor coolant pumps, steam generators and pressurizer within a single reactor pressure vessel. For the safety enhancement, SMART has the design characteristics of adopting the inherent and passive safety, simplified safety system, and advanced man-machine interface. The nominal thermal power of SMART is 365 MWth. The severe accident mitigation technology to improve the SMART safety include 1) vessel depressurization using ADS (Automatic Depressurization System) to prevent DCH (Direct Containment Heating) in case of a reactor vessel failure 2) reactor cavity flooding using the cavity flooding system with the IRWST (In-containment Refueling Water Storage Tank) for IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) to prevent the reactor vessel failure 3) hydrogen control system to remove hazards from hydrogen combustion considering the amount of hydrogen to be generated by 100% fuel cladding oxidation. The severe accident progression for the representative accident sequences including SBO and SBLOCA was analyzed by using MELCOR computer code. The severe accident issues for SMART, such as IVR-ERVC, steam explosion, DCH, MCCI (Molten Core Concrete Interaction), hydrogen control, were analyzed using many computer codes. This paper is focused on severe accident mitigation systems, severe accident analysis method, and some results on hydrogen analysis using MELCOR computer code.

Experimental and modelling results of the QUENCH-18 experiment on air ingress and aerosol release

Juri Stuckert1, Martin Steinbrueck1, Jarmo Kalilainen2, Terttaliisa Lind2, Jonathan Birchley3

1Karlsruhe Institute of Technology (KIT); 2Paul Scherrer Institut (PSI); 3Consultant

The primary aims of the QUENCH-18 bundle test were to examine the oxidation of M5® claddings in air/steam mixture following a limited pre-oxidation in steam, and to achieve a long period of oxygen and steam starvations to promote interaction with the nitrogen. Additionally, the QUENCH-18 experiment investigated the effects of the presence of two Ag-In-Cd control rods, and two pressured unheated rod simulators (60 bar, He). The low-pressurized twenty heater rods (2.3 bar, similar to the system pressure) were Kr-filled. In a first transient, the bundle was heated in an atmosphere of flowing argon and superheated steam by electrical power increase to the peak cladding temperature of 1400 K. During this heat-up, claddings of the two pressurized rods were burst at temperature of 1045 K. During the pre-oxidation stage a maximum cladding oxide layer thickness of about 100 µm was achieved. In the air ingress stage, the steam and argon flows were reduced, and air was injected. The first Ag-In-Cd aerosol release was registered at 1350 K and was dominated by Cd bearing aerosols. Later in the transient, a significant release of Ag was observed. A strong temperature escalation started in the middle of the air ingress stage. Later a period of almost complete steam consumption and partial consumption of the nitrogen occurred. Following this, the temperatures continued to increase and stabilized at melting temperature of Zr bearing materials. Almost immediately after the start of reflood there was a temperature excursion, leading to maximum measured temperatures of about 2450 K. A significant quantity of hydrogen was generated during the reflood (238 g). Nitrogen release (>54 g) due to re-oxidation of nitrides was also registered. The experiment exhibited a multiplicity of phenomena for which the data will be invaluable for code assessment and for indicating the direction of model improvements.

Severe Accident Modeling and 3D Graphic Visualization

Vladimir Nosatov1, Wei Huiming2, Oussama Ashy1

1WSC, Inc.; 2CNPSC

Results of using the MELCOR code for real-time simulation of severe accidents are presented. Recent work includes severe accident model implementation in the Seabrook Full Scope Simulator (FSS) and Barakah FSS with the purpose to extend the simulator’s range of regimes and scenarios available for training and emergency drills. Details of physical models, numerical solutions, interfaces with external models, as well as, the approach used for embedding the model on the existing FSS will be provided. As an extension to the FSS, assisting in result analysis and decision making; examples of visualization developed for China Nuclear and Radiation Safety Center (NSC) using a CPR1000 full scope simulator are presented in cooperation with WSC’s Joint Venture, China Nuclear Power Simulation Corporation (CNPSC).

Uncertainty Quantification Using the MAAP5 Code of Lower Head Failure Time in a Severe Station Blackout Accident of an Advanced Boiling Water Reactor

Jun Wei Chiou, Min Lee

National Tsing Hua University

In this study, the integral severe accident analysis code, MAAP5, is used to quantify the uncertainty in the prediction of failure time of reactor vessel of an Advanced Boiling Water Reactor (ABWR).

MAAP5 employs modularized phenomenological models to mimic the phenomena involved in the progression of core melt accident. These models were validated against the results of separate effects experiments. It can be expected that the predicted results of MAAP5 have large uncertainties and these uncertainties would definitely play a role in delineating the mitigating measures during a severe accident.

In the present study, the uncertainty in the predicted vessel lower head failure time of ABWR in an accident initiated by station blackout is explored. Sensitivity studies are used to identify the important model parameters that affect the target output parameters. The uncertainty of input parameters are propagated through code calculations and the input combinations of 29 input parameters are generated using the technique of Latin Hypercube Sampling. In the present study, 3000 MAAP5 calculations were made.

The vessel failure modes considered including the ejection of CRD (control rods drive) tubes, the attack of vessel wall by overlying metal and the heatup of drain line. The results show that the vessel fails due to the ejection of CRD tube in 822 calculations and due to attacking of metal layer in 2176 calculations. There are only two calculations that vessel fails due to heatup of drain line. When vessel fails due to the ejection of CRD tube, the vessel failure time is between 4.9 hour and 7.3 hour . The vessel failure time is between 4.1 and 7.6 hour when it fails due to the attack of overlying metal layer. Results reveal that only ECREPF is rather critical in the failure mode of ejection of CRD by Pearson's correlation coefficient method.

Assessment of the Possibility of Outlet Feeder and End-Fitting Oxidation of CANDU-6 During a Large Break LOCA with ECCS Failure for CANDU-6 Severe Accident Modeling

Bo W. Rhee1, Y.M. Song1, J.H. Bae1, T.W Kim2

1KAERI; 2Nuclear Engineering Services & Solutions

In the current DBA analysis of CANDU-6 plant using CATHENA code, the steel-water reaction are considered in neither the end-fittings nor the outlet feeder near the channel exit even for the severe accident conditions such as the large break LOCA with simultaneous failure of emergency core cooling system. However the issue addressed by S. Nihjanwan[1] that outlet feeder oxidation and consequent hydrogen production may not be negligible in the severe accident or even in the so called, “double failure accident scenarios” in CANDU-6 has not been taken seriously in the CANDU safety community.

Meanwhile a new integrated severe accident code for CANDU-6 called CAISER is under development at KAERI. So far the core degradation phenomena in a calandria tank including fuel melting, fuel rod slumping, melting and relocation of main components such as fuel channel has been modelled mechanistically and tested for preliminary validation. And the above-mentioned issue of the possibility of end-fitting and outlet feeder oxidation during the large break loss of coolant accident with a simultaneous failure of emergency core cooling system (LBLOCA + LOECC) is thought necessary to be investigated, and the necessity of including the relevant model in CAISER needs to be confirmed. In this study. the possibility of steel oxidation in CANDU-6 during LBLOCA with ECCS impairment was carried out by investigating the safety analysis results of Wolsong-1 Refurbishment Project and is described the result of the expected ranges of the temperatures of the end-fittings and outlet feeder during these accidents, and the possibility of oxidation of the outlet feeders and end-fittings is discussed.


[1]Y.M. Song, B.W. Rhee, J.H. Bae, Sunil Nijhawan, Hydrogen Source Term in CANDUs to include Oxidation of Steel, Transactions of the KNS Spring Meeting, May 17-18, 2018, Jeju, Korea.

Improvement of molten core coolability by an efficient spreading: VULCANO VEU9 Experiment and THEMA sensitivity calculations

Anne Boulin1, Laurence Buffe1, Jean-François Haquet1, Viviane Bouyer1, Arthur Denoix1, Hiroshi Goda2, Satoru Kamohara2


For additional safety improvement to mitigate the hypothetic severe accident, MHI has been conducting an R&D project to develop a new sacrificial material that will cover the RV cavity floor to improve molten core coolability by effectively spreading the molten core on the cavity floor to form a thin layer. This is achieved by lowering viscosity of molten corium-sacrificial material mixture around liquidus-line temperature as well as lowering its liquidus-line temperature compared to that of pure molten corium. In the framework of this project, MHI asks CEA, on one hand, to set-up a spreading experiment (VULCANO VE-U9) on the sacrificial material they defined and on another hand to perform sensitivity calculations with the THEMA code in order to define the corium inlet operating range representative of the end of a corium spreading.

This paper will describe, in a first part, the experimental set-up including different systems: the melting and pouring device, the corium transfer system and the spreading test section itself. The spreading area is an angular sector which is separated into two identical channels: one channel is covered with fused zirconia slabs and the other with sacrificial concrete slabs.The instrumentation provides measurements on corium flowrate for each channel and its progression, corium and substrate temperatures.

In the second part, the main THEMA models used for the VULCANO VE-U9 spreading test pre-calculations are presented. The balance equations are averaged over the melt thickness and a shallow water condition is assumed. For a 50 kg poured corium mass, it can be deduced from the sensitivity study for the corium inlet conditions (2000K, 1l/s) that the laminar spreading length is significant (~1,5m), the thickness is thin and no substrate ablation is observed.

The VULCANO VE-U9 experimental test is planned on October 2019.

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