Kwang Soon Ha

Validation of RCS depressurization strategy and core coolability map for independent scenarios of SBLOCA, SBO, and TLOFW

Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coolant Accident (LOCA), Station Black Out (SBO), Total Loss-Of-Feed-Water (TLOFW) was performed and effectiveness of Reactor Coolant System (RCS) depressurization strategies of OPR1000 was analyzed using MELCOR 1.8.6 code. Required injection flow rate has been derived using Core Exit Temperature (CET) information obtained from MELCOR calculation and a simple model and corresponding coolability map have been suggested to assist effective operator action. The depressurization strategies using secondary Atmospheric Dump Valve (ADV) for SBLOCA, pressurizer Safety Depressurization System (SDS) for SBO and TLOFW were introduced in 5 min since the initiation of Severe Accident Management Guidance (SAMG). Respective mitigation strategy employed leads to significant delay of the reactor pressure vessel failure and RCS pressure at Reactor Pressure Vessel (RPV) failure was lower than the SAMG target pressure of 2.86 MPa. Thus, possibility of High Pressure Melt Ejection (HPME) and impair of containment building is expected to avoid effectively. Using CET information obtained from MELCOR calculation, a simple model and a coolability map for the required injection flow rate were developed for recovery of core coolability. It is suggested that the coolability map based on MELCOR calculation results may provide decisive and intuitive information to operators for more effective safety management.

THERMAL HYDRAULIC ISSUES OF CONTAINMENT FILTERED VENTING SYSTEM FOR A LONG OPERATING TIME

This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

NUMERICAL INVESTIGATION OF THE SPREADING AND HEAT TRANSFER CHARACTERISTICS OF EX-VESSEL CORE MELT

The flow and heat transfer characteristics of the ex-vessel core melt (corium) were investigated using a commercial CFD code along with the experimental data on the spreading of corium available in the literature (VULCANO VE-U7 test). In the numerical simulation of the unsteady two-phase flow, the volume-of-fluid model was applied for the spreading and interfacial surface formation of corium with the surrounding air. The effects of the key parameters were evaluated for the corium spreading, including the radiation, decay heat, temperature-dependent viscosity and initial temperature of corium. The results showed a reasonable trend of corium progression influenced by the changes in the radiation, decay heat, temperature-dependent viscosity and initial temperature of corium. The modeling of the viscosity appropriate for corium and the radiative heat transfer was critical, since the front progression and temperature profiles were strongly dependent on the models. Further development is required for the code to consider the formation of crust on the surfaces of corium and the interaction with the substrate.

The Hydrogen Issue in the Initial Operation of a Filtered Containment Venting System

Abstract This study evaluated the hydrogen issue in the initial operation of a filtered containment venting system (FCVS). We calculated the volumetric concentration of hydrogen, steam, and air in the postulated FCVS connected with the OPR 1000, as a target nuclear power plant, under a station blackout using the MELCOR computer code (version 1.8.6). A large amount of steam and a flammable mixture generated during a severe accident are immediately released from the containment building to the FCVS when the pressure in the containment building approaches a set value. The constituent ratio of the flammable mixture of hydrogen, steam, and air can change due to the different thermal-hydraulic conditions between those due to a severe accident in the containment building and the initial condition in the FCVS. The volumetric concentration of hydrogen was 6% in the containment building just before the operation of the FCVS. It increased up to 9% in the FCVS vessel during the early operation, and steam condensation occurred simultaneously. The atmospheric condition including steam, hydrogen, and air in the FCVS can enter the combustion zone in the Shapiro diagram.

Numerical Simulation on the Spreading and Heat Transfer of Ex-Vessel Core Melt in a Channel

In the unlikely of nuclear reactor meltdown, the leaked core melt or corium must be contained in a device called core-catcher so that the corium can be cooled and stabilized. The ex-vessel behavior of corium involves complex physical and chemical mechanisms of flow propagation, heat transfer, and reactions with sacrificial substrates. In this study, the detailed characteristics of corium flow and heat transfer were investigated by using a commercial CFD code for VULCANO VE-U7 test reported in the literature. The volume-of-fluid (VOF) model was used to predict the interfacial surface formation of corium and the surrounding air, and the discrete ordinate model was adopted to calculate radiation between corium and the surroundings. It was found that cooling via radiation through the top surface of corium had a dominant effect on the temperature and viscosity profiles at the front of the corium flow.

A Core Catcher Concept and First Experimental Results

In a postulated core melt accident, if a molten core is released outside a reactor vessel despite taking mitigation actions, the core debris would relocate in the reactor cavity region and attack the concrete wall and basemat of the reactor cavity. This will potentially result in inevitable concrete decompositions and possible radiological releases. To prevent direct contact of the melt and basemat concrete of the cavity, a core catcher concept is suggested, which can passively arrest and stabilize the molten core material inside the reactor cavity. The core catcher system includes a retention device for the molten core material, a cooling water storage tank, and a compressed gas tank. Upon ablation of the sacrificial layer on top of the retention device while molten core material is discharged, a mixture of water and gas is injected from below. It is expected that a simultaneous injection of water and gas could prevent a possible steam explosion/spike. It could also suppress the rapid release of steam which might result in fast over-pressurization of the containment. A test facility for the core catcher using a thermite reaction technique for the generation of the melt was designed and constructed at KAERI. The first series of tests were performed by using a mixture of Al, Fe2 O3 , and CaO as a stimulant. As a first try, only water was injected from the bottom of the melt through five water injection nozzles when the melt front reached the water injection nozzles. In this paper, the core catcher concept and the related provisions are suggested. A description of the test facility for the core catcher, the thermite composition, and the methods of experiment is included. The first experimental results with only water injected from the bottom of the melt are discussed.Copyright

An Evaluation of a Direct Corium Cooling Method for the Ex-Vessel Melt Retention

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation and in the suppression of a steam explosion. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-inches large break loss of coolant accident without safe injection. The corium spreading regime was estimated by an asymptotic calculation. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water injection system was an effective corium cooling method in the ex-vessel core catcher to preclude a possible steam explosion and to suppress the quick release of steam.© 2010 ASME