Study of Refractory Material Installation in the Reactor Cavity

During a severe accident with a vessel failure, corium relocates from the vessel into the reactor cavity (PWR) or pedestal (BWR) and accumulates on top of the cavity floor to form a corium pool. This corium pool is hot enough to cause a Molten Corium-Concrete Interaction (MCCI) that can ablate the concrete structure even if water is present on top of the corium. MCCI will also produce steam and other gases that increase containment pressure as well as generate combustible gases (Hydrogen and Carbon Monoxide). Current MAAP5* calculations with conservative assumptions have shown that the ablation depth in a basemat constructed of siliceous concrete can be larger than the depth of liner, even if the reactor cavity is flooded by water. To retain the melt in the containment and to cool the corium pool before the erosion reaches the liner plate, several approaches are being considered. One of these approaches is the installation of a protective layer on top of the concrete floor to retard MCCI. The purpose of this paper is to study the performance of different protective materials under postulated severe accident conditions. The candidates for the protective materials are refractory materials and limestone/limestone-common-sand (LCS) concrete. The refractory material was chosen based on the thermal performance and dissolution rate of the refractory material calculated by analytical calculations and also by MAAP5. Adding the refractory protective material protects the underlying concrete basemat from melting temporarily, so that water ingression into the surface of the corium is not initially affected by addition of the concrete material. *MAAP5 is an integrated severe accident code owned by the Electric Power Research Institute and developed by Fauske and Associates, LLC.

Experimental Research on Steam Condensation on Cold Surface: Facility Design

The passive containment cooling system (PCCS) of the 3rd generation APWR utilizes natural phenomena to transfer the heat released from the reactor to the environment during postulated designed basic accidents. Steam condensation on the inner surface of the containment shell is one of the most dominate mechanism to keep the ambient conditions within the design limits. Extensive experiment and theoretical research shows condensation is a complex process, gas pressure, film temperature and velocity of the gas have impact on the heat transfer coefficient. To span the expected range of conditions and provide proper model for evaluating the condensation heat transfer process, SCOPE test facility was designed by State Nuclear Power Technology Research & Development Centre (SNPTRD) in various conditions anticipated the operating range of CAP1400 in accident conditions. Pressurized test section with a rectangular flowing channel was used, with one of the walls cooled to maintain low temperature for condensing, supplying systems was designed for different pressures, gas temperatures, velocities and coolant water temperatures. Facility components, test section structure, supplying systems and measurement technology were described in this paper, also results of some pre-tests was introduce to show property of the facility.

Experimental Measurements and CFD Evaluation of the Onset Flow Boiling at Low Pressure and High Subcooling Flow of R-11 Within Vertical Annulus

The knowledge of the onset of subcooled boiling in forced convective flow at high liquid velocity and subcooling is of importance in thermal hydraulic studies. Measurements were performed under various conditions of mass flux, heat flux, and inlet subcooling, which enabled to study the influence of different boundary conditions on the development of local flow parameters. Also, some measurements have been compared to the predictions by the three-dimensional two-fluid model of subcooled boiling flow carried out with the computer code ANSYS-CFX-13. A computational method based on theoretical studies of steady state two phase forced convection along a test section loop was released. The calculation model covers a wide range of two phase flow conditions. It predicts the heat transfer rates and transitions points such as the Onset of Critical Heat Flux.

Experimental Study of Moderator Circulation in CANDU6 Calandria Tank

As a CANDU6 reactor has a high pressure primary cooling system and an independently cooled moderator system, the moderator in the calandria would act as a supplementary heat sink during a loss of coolant accident (LOCA) if the primary cooling and emergency coolant injection systems fail to remove the decay heat from the fuel. For the safety concern it is required to predict the 3-dimensional velocity and temperature distribution of moderator fluid to confirm the effectiveness of moderator heat sink. Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions.

Analysis of Heat Transfer and Flow Characteristics of AP1000 Passive Residual Heat Removal Heat Exchanger

In this paper, the computational fluid dynamics (CFD) method is applied to the thermal-hydraulic analysis, while the porous media model is used to simplify AP1000 passive residual heat removal heat exchanger tube. The temperature as well as flow distribution in the secondary side of the heat exchanger are obtained, aiming at analysis of natural circulation ability. It can be noted that the fluid in the secondary side of heat exchanger moves driven by the effect of thermal buoyancy, forming the natural cycle, which takes away heat in tube bundle region. The heat transfer in water tank is mainly enhanced by vortex and turbulent flow, caused by the large resistance of tube bundle region as well as large temperature difference. This phenomenon is obvious especially for the recirculation of flow near the tube bundle. The enduring change of flow rate and direction enhance the heat transfer. Besides, the big temperature difference helps to increase the driving effect of natural circulation. Consequently, the heat transfer of the tank is enhanced by above mechanism. The results of this study contribute to the capacity analysis of passive residual heat removal of natural circulation system, providing valuable information for safe operation of AP1000.

Simulation of Gas Stratification Build-Up in the Containment Under Severe Accident Conditions

The presence of hydrogen stratification in a NPP containment in the case of a severe accident is a source of concern, as pockets of the gas in high concentration could lead to a deflagration or detonation risk, which might challenge the containment structural integrity. These issues, as well as the capability of various computer codes to predict the evolution of a representative accident, are addressed in the coordinated projects ERCOSAM of the 7th EURATOM FWP and the project SAMARA sponsored by ROSATOM. The projects aim to establish whether in a test sequence representative of a severe accident in a LWR hydrogen stratification can be established during the initial transient following a loss of coolant accident (LOCA) and whether and how this stratification can be broken down by the operation of Severe Accident Management systems (SAMs): sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Experiments with helium (as simulant of hydrogen) have been performed at “small scale” in TOSQAN (IRSN, Saclay), and “medium scale” in the MISTRA (CEA, Saclay), PANDA (PSI, Villigen) and SPOT ((JSC “Afrikantov OKBM”, Nizhny Novgorod) facilities. The present paper presents the analysis of the initial transient of some tests in the PANDA, TOSQAN and SPOT facilities using the GOTHIC computer code. The work therefore addresses the capability of the code and a relatively coarse mesh to simulate the pressurisation and build-up of steam and helium stratification for conditions representative of a postulated severe accident scenario, properly scaled to the various facilities. The prediction of the pressurisation is excellent, and the position of the gas concentration stratification front at the end of the steam and helium releases is generally well captured.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Study on DNB-Type Critical Heat Flux With Chopper-Cosine Axial Heat Flux Distributions

Departure from nucleate boiling (DNB) type critical heat flux (CHF) is one of most important thermal criteria for nuclear reactor design. Concerning on the typical chopper-cosine heat flux profile at reactor core, it is of great significance to predict the CHF under non-uniform heating conditions for reactor design and the performance promotion of reactor system. Some correction factors are proposed for the prediction of CHF with non-uniform axial power shapes. In this study, a mechanistic DNB-type CHF model has been developed on the basis of liquid sublayer dryout mechanism. The non-uniform axial heat flux is taken into account of upstream memory effect on boiling crisis in this model. The predictions of present model and Tong’s non-uniform heat flux shape factor method are compared with the experimental results in the vertical tube with chopper-cosine axial heat flux distributions. The comparison results show the present model has fairly good prediction capability for DNB-type CHF under non-uniform heating condition.

Cold Water Injection and Rod Ejection Analysis of Annular Fueled PWRs by a Hybrid Lumped Parameter Model

Annular fuels are being studied to increase the power of advanced third-generation reactors by 50%. This paper aimed to analyze transient scenarios through a hybrid lumped parameter-finite difference model in a pressurized water reactor with annular fuel. The model used in this work is more detailed than the double lumped parameter one, but still simple enough to model some transients in PWR fuels, as rod ejection accident and cold water insertion accident. The heat transfer equations are solved by the numerical semi-implicit Crank-Nicolson method together with point kinetics equations with six groups of delayed neutrons and a lumped parameter model for the reactor coolant. The model takes into account in an approximate way the hot spot by using a composed peaking factor equal to 2.5. The reactivity feedback is taken into account by considering the Doppler effect of fuel temperature, and also moderator temperature variation. The results were compared with solid fuel performance and showed that the annular fuel reached considerable lower fuel temperature profiles even for 150% power, as compared to 100% power for solid fuel, thus showing that this kind of fuel has a better safety performance for the transients analyzed. The rod ejection accident showed that feedback effects can lead the reactor to a new safe steady state condition.

Effect of Thermal Stratification on Full-Cone Spray Performance in Reactor Containment for a Scaled Scenario

The results of an experimental study on the nuclear reactor containment spray system are presented. Depending on the initial conditions, the spray nozzle configuration and flow rates, the spray may cause higher hydrogen concentration during depressurization due to steam condensation, or it may erode the hydrogen stratification by enhanced mixing. To investigate these phenomena, the tests are performed using a full-cone spray nozzle in PANDA facility at Paul Scherrer Institut, Switzerland. Temporal evolution and spatial distribution of the fluid temperature and the fluid concentrations are measured using thermocouples and mass spectrometers. Two tests are performed with initial vessel wall temperatures of 105°C and 135°C, which create condensing and non-condensing environments respectively. The different initial conditions lead to different density stratifications. The effect of these different density stratification on the flow patterns and mixing of gases in the vessels due to the action of the spray is revealed by these tests.