MARS-KS Calculations for ATLAS 6-Inch Cold Leg Break (SBLOCA) Test

An integral effect test on the SBLOCA (Small-Break Loss of Coolant Accident) aiming at 6-inch cold leg bottom break, SB-CL-09, was conducted with the ATLAS on November, 13, 2009 by KAERI. In this study, the calculations using MARS-KS V1.2 code were conducted for 6-inch cold leg break test of ATLAS (SB-CL-09) to assess MARS-KS code capability to simulate the transient thermal-hydraulic behavior for SBLOCA. The steady state was determined by conducting a null transient calculation and the errors between the calculated and measured values are acceptable for almost primary/secondary system parameters. The sequence of events except the location of loop seal clearing (LSC) and SIT injection time was predicted relatively well. The predicted pressurizer pressure agrees relatively well with the experimental data and the predicted break flow and mass are in good agreement with experiment. In MARS-KS calculation, the decrease of core collapsed water level is predicted well in blowdown phase, but just before LSC, water level is higher than experiment. However, the sudden decrease and increase of water level at the LSC are predicted qualitatively. After LSC, there is another water level dip at SIT injection time which is not in experiment. It is considered that this phenomenon is caused by rapid depressurization of downcomer due to significant condensation rate of vapor in downcomer when SIT water flows in it. For the downcomer water level, before the SIT injection time, water level is predicted well, however, it is significantly over-predicted at SIT injection time after SIT water flows in downcomer. Predicted cladding temperature generally agrees well with the experiment, while there is peak at SIT injection time in calculation which is not in experiment. The loop seals of 1A, 2B intermediate leg are cleared around 400 seconds in experiment, while only that of 1A is cleared in MARS-KS calculation at the same time. In conclusions, MARS-KS code has good capabilities to simulate cold leg break SBLOCA, however, there are some discrepancies in quantitatively predicting the steam generator secondary pressure, core collapsed liquid level, downcomer liquid level, and so on. Therefore, MARS-KS code including interfacial condensation model needs to be improved to predict more accurate results.

Pressure Drop and Flow Pulsation in Tight-Lattice Rod Bundle

In order to ensure the compactness and high-power density of a nuclear power reactor, the research on tight-lattice fuel bundle, with a narrow gap distance between fuels, has been highlighted. Recently, KAERI (Korea Atomic Energy Research Institute) has been developing dual-cooled annular fuel to increase a significant amount of the reactor power in OPR1000 (Optimized Power Reactor), a PWR (Pressurized Water Reactor) optimized in the Republic of Korea. The dual-cooled annular fuel is configured to allow a coolant flow through the inner channel as well as the outer channel. To introduce the dual-cooled annular fuel to OPR1000 is aiming at increasing the reactor power by 20% and reducing the fuel-pellet temperature by 30%, as compared to the cylindrical solid fuel, without a change in reactor components. In such a case, due to larger outer diameter of a dual-cooled annular fuel, the dual-cooled annular fuel assembly exhibits a smaller P/D (Pitch-to-Diameter ratio) than the conventional cylindrical solid fuel assembly. In other words, the dual-cooled annular fuel array becomes the tight-lattice fuel bundle configuration, and such a change in P/D can significantly affect the thermal-hydraulic characteristics in nuclear reactor core. In this paper, the pressure drop and flow pulsation in tight-lattice rod bundle were investigated. As the test sections, the tight-lattice rod bundle of P/D = 1.08 was prepared with the regular rod bundle of P/D = 1.35. The friction factors in P/D = 1.08 appeared smaller than those in P/D = 1.35. For P/D = 1.08, the twist-vane spacer grid became the larger pressure loss coefficients than the plain spacer grid. In P/D = 1.08, the flow pulsation, quasi-periodic oscillating flow motion, was visualized successfully by PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques. The peak frequency and power spectral density of flow pulsation increased with increasing the Reynolds number. Our belief is that this work can contribute to a design of nuclear reactor with tight-lattice fuel bundle for compactness and power-uprate and a further understanding of the coolant mixing phenomena in a nuclear fuel assembly.

Validation Experiment for the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) With PASCAL Facility

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.

Scaling Evaluation of the Two-Phase Natural Circulation

This paper presented the scaling evaluation of the two-phase natural circulation process between an assumed nuclear power plant and three test facilities with full pressure simulation and three different height scales, which were 1:2, 1:3 and 1:4. The Hierarchical Two-Tiered Scaling (H2TS) Methodology was adopted. By top-down scaling analysis, several characteristic time ratios were obtained, and then the calculation method of the scaling distortion were investigated. It has been found that the dominant processes in two-phase natural circulation can be well preserved no matter what the height scale is.

Experiment of Adiabatic Two-Phase Flow in an Annulus Under Low-Frequency Vibration

In order to investigate the possible effect of seismic vibration on two-phase flow dynamics and thermal-hydraulics of a nuclear reactor, experimental tests of adiabatic air-water two-phase flow under low-frequency vibration were carried out in this study. An eccentric cam vibration module operated at low motor speed (up to 390rpm) was attached to an annulus test section which was scaled down from a prototypic BWR fuel assembly sub-channel. The inner and outer diameters of the annulus are 19.1mm and 38.1mm, respectively. The two-phase flow operating conditions cover the ranges of 0.03≤<jg> ≤1.46m/s and 0.25≤<jf>≤1.00m/s and the vibration displacement ranges from ±0.8mm to ±22.2mm. Steady-state area-averaged instantaneous and time-averaged void fraction was recorded and analyzed in stationary and vibration experiments. A neural network flow regime identification technique and fast Fourier transformation (FFT) analysis were introduced to analyze the flow regimes and void signals under stationary and vibration conditions. Experimental results reveal possible changes in flow regimes under specific flow and vibration conditions. In addition, the instantaneous void fraction signals were affected and shown by FFT analysis. Possible reasons for the changes include the applied high acceleration and/or induced resonance at certain ports under the specific flow and vibration conditions.

Experimental Investigation of Friction Reduction in Transport of Fluids Using Surfactants

This paper presents the results of the influence of surfactants in reducing friction while driving fluid in pipes. Experimental research was conducted with water-surfactant mixtures which were tested three types of these additives: anionic, cationic and nonionic surfactants. Was designed and built an experimental facility in which the test area was acrylic pipe with an inner diameter of D = 19 mm and a length of 300 D. The concentrations of surfactants in the mixtures were 150, 250, 500, 750 and 1000 ppm, added according to a pilot program that took into account the amount and type of additive added in different liquid mass fluxes. Pressure losses were compared against those obtained when flow is the same water flows through the installation. The results obtained show a reduction of up to 43.9% of the friction which is achieved with a Re = 11243 and surfactant concentrations of 250 ppm (cetyl trimethyl chloride ammonium), to which was added as a stabilizer for the micro structure of the surfactant, sodium salicylate, which applies only to the cationic type surfactants. The results are promising but left to study such issues as: the injection and recovery of surfactant, more efficient mixing, the mechanisms that lead to a reduction of friction and the effect of temperature among others.

Investigations on Slug Flow in a Horizontal Pipe Using Stereoscopic Particle Image Velocimetry and CFD Simulation With Volume of Fluid Method

Investigations on gas-liquid flows in horizontal pipes are of immanent importance for Reactor Safety Research. In case of a breakage of the main cooling circuit of a Pressurized Water Reactor (PWR), the pressure losses of the gas-liquid flow significantly govern the loss of coolant rate. The flow regime is largely determined by liquid and gas superficial velocities and contains slug flow that causes high-pressure pulsations to the infrastructure of the main cooling circuit. Experimental and numerical investigations on adiabatic slug flow of a water-air system were carried out in a horizontal pipe of about 10 m length and 54 mm diameter at atmospheric pressure and room temperature. Stereoscopic high-speed Particle Image Velocimetry in combination with Laser Induced Fluorescence was successfully applied on round pipe geometry to determine instantaneous three-dimensional water velocity fields of slug flows. After grid independence studies, numerical simulations were run with the open-source CFD program OpenFOAM. The solver uses the VOF method (Volume of Fluid) with phase-fraction interface capturing approach based on interface compression. It provides mesh refinement at the interfacial area to improve resolution of the interface between the two phases. Furthermore, standard k-ε turbulence model was applied in an unsteady Reynolds averaged Navier Stokes (URANS) model to resolve self-induced slug formation. The aim of this work is to present the feasibility of both relatively novel possibilities of determining two-phase slug flows in pipes. Experimental and numerical results allow the comparison of the slug initiation and expansion process with respect to their axial velocities and cross-sectional void fractions.

Vapor Bubble Condensation Characteristics of Subcooled Flow Boiling in Vertical Rectangular Channel

In this study, the behavior of condensing single vapor bubble in subcooled boiling flow within two different vertical rectangular channels has been numerically investigated by using the VOF (Volume of Fluid) multiphase flow model. The mass and energy transfer model of bubble condensing process induced from the interfacial heat transfer was proposed to describe the interfacial transportation between the two phases. The results of VOF simulations showed good agreements with previous experimental data in the bubble size variation and lifetime. The bubble lifetime was almost proportional to bubble initial size and prolonged by system pressure. With the increase of subcooling, the bubble lifetime reduces significantly, and the effect of mass flux could be negligible. When bubble size increased, the bubble shape tends to be changed in a large-scale channel. The VOF simulation results of deformation have good agreement with those of Kamei’ experiment and the results of MPS (Moving Particle Semi-implicit) simulation in the large-scale channel. Furthermore, the initial bubble size, subcooling of liquid and system pressure play an important role to influence the bubble deformation behaviors significantly. The bubble could deform sharper with the increase of subcooling and initial diameter, or could breakup when the subcooling and the initial diameter reached a certain value at the last bubble stage. Whereas the trend of bubble deformation would be weaken with the increase of system pressure.

Uncertainty Evaluation of the Rod Withdrawal at Power Accident Analysis Including 3D Neutron Kinetics

A continuous uncontrolled Rod Cluster Control Assembly (RCCA) bank withdrawal at power belongs to group of Reactivity Initiated Accidents (RIA). It will cause an increase in core heat flux and a reactor coolant temperature rise. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise could eventually result in departure from nucleate boiling (DNB) and/or fuel centreline melt. The accident can be DNBR or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. The Rod Withdrawal At Power (RWAP) accident was analyzed for NPP Krško to evaluate possible Resistance Temperature Detectors (RTD) bypass removal and introduction of thermowell for the average temperature measurement. The influence of different coolant temperature measurement delays to related protection system response and limiting system variables was studied first using point kinetics model as implemented in RELAP5 code. The selected scenario (maximum insertion rate with rods in manual mode) has been re-calculated using RELAP5/PARCS coupled code. Core wide departure from nucleate boiling ratio (DNBR) calculation has been performed at the end of the coupled code calculation using COBRA based model to determine minimum DNBR for hot channel. In order to assess available safety margins following such accident CIAU methodology has been applied to evaluate the uncertainty of RELAP5 analysis and modified CIAU/TN methodology to evaluate uncertainty of the three-dimensional neutronics/thermal-hydraulics calculations. Differences between system and coupled code results and uncertainties is discussed.

Performance Evaluation of a High Thermal Conductivity Fuel With Graphene Additives During the LBLOCA

The fuel rod performance of enhanced thermal conductivity UO2/graphene composites is investigated through a LBLOCA analysis. The benefits increased monotonically with increasing thermal conductivity in terms of reduced fuel center temperature and PCT. The performance of the UO2/graphene composite fuel is assessed in OPR-1000 (Optimized Power Reactor-1000) during a LBLOCA. Graphene can be a promising material for developing advanced nuclear fuel because of its property about the high thermal conductivity and low absorption cross section. The results confirm a LBLOCA performance related to PCT of the UO2/graphene composite fuel and its potential while maintaining large safety margins.