AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

Feasibility Assessment of Air-Cooling System As an Ultimate Heat Sink of the ATOM System

Nowadays Small Modular Reactors (SMRs) have been receiving considerable attentions worldwide for potential advantages of an excellent flexibility for siting, low capital investment, and advanced safety. In Korea, a new research project has launched for the development of a conceptual design of a further advanced SMR which aims for a naturally-safe and autonomous operation, so called Autonomous Transportable On-demand reactor Module (ATOM). Major design objectives of the ATOM system are focused on the soluble boron-free (SBF) primary coolant system which enables the SMR to operate automatically in a load following mode. For the secondary system, the SCO2 power conversion cycle with air-cooling system as a final heat sink is being considered. The air-cooling system is expected to show flexible response even to extreme environmental conditions, such as a desert where utilization of cooling water is limited. The objective of this study is a feasibility assessment for applying the air-cooling system as a final heat sink of the ATOM by means of experimental work. As a 1st phase of the ATOM development, we first conducted the experiments using a typically considered primary coolant, water-steam, to verify that air flow has enough cooling capability to remove developed heat which the coolant carries. An Integrated Condensation Loop with Air-cooling System (ICLASS) experimental facility with three pressure boundaries (Steam, coolant, and air) was established. The cooling capability of the air-cooling system was evaluated by varying steam mass flow rate, coolant flow rate, and air environment temperature as experiment variables. Overall heat transfer rate by condensation was compared with numerical simulations of a 1D thermal-hydraulics analysis code, using the MARS model of the ICLASS facility.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

Development and Application of Multi-Physics Safety Analysis Code for Advanced Liquid Metal Cooled Reactor

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

Effects of Parameters on the Two-Phase Flow Instability in a Microchannel

Series of experiments are conducted in a single microchannel, where subcooled water flows upward inside a transparent and vertical microchannel. The cross section of the channel is rectangle with the hydraulic diameter of 2.8mm and the aspect ratio of 20. The working fluid is 3–15K subcooled and surface heat flux on the channel is between 0–3.64 kW/m2, among which two-phase instability at low vapor quantity may occur. By using a novel transparent heating technique and a high-speed camera, visualization results are obtained. The parameters are acquired with a National Instruments Data Acquisition card. In the experiments, long-period oscillation and short-period oscillation are observed as the primary types of instability in a microchannel. Instability characteristics represented from signals correspond well with the flow pattern. Moreover, effects of several parameters are investigated. The results indicate that the oscillating period generally increases with the heat flux density and decreases with inlet subcooling, while the effects of inlet resistance are more complex.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

Numerical Study on Effect of Pressure on Behavior of Bubble Coalescence by Using CMFD Simulation

The mechanism of critical heat flux (CHF) for higher system pressure remains to be clarified, even though it is important to evaluate the CHF for the light water reactor (LWR) which is operated under the high pressure condition. In this study, the process of bubble coalescence was simulated by using a computational multi-fluid dynamics (CMFD) simulation code TPFIT under various system pressure in order to investigate the behavior of bubbles as a basic study. The growth of bubbles was simulated by blowing of vapor from a tiny orifice simulating bubble bottom. One or four orifices were located on the bottom surface in this simulation study. The numerical simulations were conducted by varying the pressure and temperature.

Sensitivity of CTF Solution to Subchannel Window Size

In this work, a so-called mini-core consisting of a 3 × 3 array of 17 × 17 pressurized water reactor (PWR) fuel assemblies (FA) is considered with the aim of identifying the most conservative window size for hot channel analysis of bowed fuel assemblies. Overall, five different mini-core configurations are analyzed: one is the reference case, i.e. without FA displacement and four different cases with diagonal and parallel FA displacements. Rod power maps for these mini-cores were exported from neutronic calculations with CASMO-SIMULATE codes. Subchannel modelling with COBRA-TF code of all five mini-cores allows one to identify the rod position with a minimum departure from the nucleate boiling ratio (DNBR) and to construct input decks with different rod window sizes around the previously identified rod position. Overall, eight different window sizes are considered: 3 × 3, 5 × 5, 7 × 7, 9 × 9, 11 × 11, 13 × 13, 15 × 15 and 17 × 17. Results of subchannel analysis for a mini-core and different subchannel window configurations are compared with the help of DNBR parameter, which is the ratio between the critical heat flux (CHF) and the actual local heat flux on a rod. An assessment of three different CHF models is applied in this work: Groeneveld CHF look-up table (LUT), W3 CHF correlation, and Doroschuk CHF LUT. The general conclusion of this work is that for deformed core configurations, an appropriate rod window size needs to be determined to adequately capture the local flow redistribution. For large displacements (the largest displacement considered in this work is 10 mm), the DNBR ratio can drop to one. DNBRs obtained with the W3 CHF correlation give the most conservative results.

Suppression Methods of Acoustic Noise Generated in Main Steam Stop Valve

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.

Study of Fission Product Behavior in Containment Vessel Using Modified ART Mod 2: Update of Cesium and Iodine Compound Models

From the Fukushima accident, Thailand has studied fission product behavior in containment vessel using ART Mod 2 code. Cesium iodide (CsI), cesium hydroxide (CsOH) and iodine (I2) behaviors are studied using modified ART Mod 2 code. However, there are other compounds which are not included in the codes especially cesium and iodine compounds such as from Phébus FPT3 experiment including cesium molybdate (Cs2MoO4), cesium telluride (Cs2Te), methyl iodide (CH3I) and iodine pentoxide (I2O5). The paper objective is to add the four compounds in the codes in order to enlarge the coverage of the code in evaluation fission product behavior in the containment vessel. Physical parameters and models of the four compounds are updated in the codes. It is found that deposition phenomena of Cs2MoO4, Cs2Te CH3I and I2O5 are close to the experiment in case of no chemical reaction.