Assessment of Different Turbulence Models in Helically Coiled Pipes Through Comparison With Experimental Data

This paper deals with a comprehensive study of fully developed single-phase turbulent flow and pressure drops in helically coiled channels. To the aim, experimental pressure drops were measured in an experimental campaign conducted at SIET labs, in Piacenza, Italy, in a test facility simulating the Steam Generator (SG) of a Generation III+ integral reactor. Very good agreement is found between data and some of the most common correlations available in literature. Also more data available in literature are considered for comparison. Experimental results are used to assess the results of Computational Fluid Dynamics (CFD) simulations. By means of the commercial CFD package FLUENT, different turbulence models are tested, in particular the Standard, RNG and realizable k-ε models, Shear Stress Transport (SST) k-ω model and second order Reynolds Stress Model (RSM). Moreover, particular attention is placed on the different types of wall functions utilized through the simulations, since they seem to have a great influence on the calculated results. The results aim to be a contribution to the assessment of the capability of turbulence models to simulate fully developed turbulent flow and pressure drops in helical geometry.

Comparison of Existing Supercritical Carbon Dioxide Heat Transfer Correlations for Horizontal and Vertical Bare Tubes

This paper presents a thorough analysis of ability of various heat transfer correlations to predict wall temperatures and Heat Transfer Coefficients (HTCs) against experiments on internal forced-convective heat transfer to supercritical carbon dioxide conducted by Koppel [1], He [2], Kim [3] and Bae [4]. It should be noted the Koppel dataset was taken from a paper which used the Koppel data but was not written by Koppel. All experiments were completed in bare tubes with diameters from 0.948 mm to 9 mm for horizontal and vertical configurations. The datasets contain a total of 1573 wall temperature points with pressures ranging from 7.58 to 9.59 MPa, mass fluxes of 400 to 1641 kg/m2s and heat fluxes from 20 to 225 kW/m2. The main objective of the study was to compare several correlations and select the best of them in predicting HTC and wall temperature values for supercritical carbon dioxide. This study will be beneficial for analyzing heat exchangers involving supercritical carbon dioxide, and for verifying scaling parameters between CO2 and other fluids. In addition, supercritical carbon dioxide’s use as a modeling fluid is necessary as the costs of experiments are lower than supercritical water. The datasets were compiled and calculations were performed to find HTCs and wall and bulk-fluid temperatures using existing correlations. Calculated results were compared with the experimental ones. The correlations used were Mokry et al. [5], Swenson et al. [6] and a set of new correlations presented in Gutpa et al. [7]. Statistical error calculations were performed are presented in the paper.

Suitability of Eutectic Field’s Metal for Use in an Electromagnetohydrodynamically-Enhanced Experimental Two-Phase Flow Loop

Several Generation IV nuclear reactors, such as sodium fast reactors and lead-bismuth fast reactors, use liquid metal as a coolant. In order to better understand and improve the thermal hydraulics of liquid metal cooled GEN IV nuclear reactors liquid metal flow needs to be studied in experimental circulation loops. Experimental circulation loops are often located in a laboratory setting. However, studying liquid metal two phase flow in laboratory settings can be difficult due to the high temperatures and safety hazards involved with traditional liquid metals such as sodium and lead-bismuth. One solution is to use a low melt metal alloy that is as benign as reasonably achievable. Field’s metal is a eutectic alloy of 51% Indium, 32.5% Bismuth and 16.5% Tin by weight and has a melting point of 335K making it ideal for use in a laboratory setting. A study is undertaken to determine its suitability to use in a two-phase experimental flow loop enhanced by magnetohydrodynamic forces. The study investigated its reactivity with air and water, its ability to be influenced by magnetic fields, its ability to flow, and its ease of manufacture. The experiments melted reference samples of Field’s metal and observed its behaviour in a glass beaker, submerged in water and an inclined stainless steel pipe. Then Field’s metal was manufactured in the laboratory and compared to the sample using the same set of experiments and standards. To determine Field’s metal degree of magnetism permanent neodymium magnets were used. Their strength was determined using a Gaussmeter. All experiments were recorded using a COHU digital camera. Image analysis was then performed on the video to determine any movements initiated by the magnetic field forces. In conclusion, Field’s metal is more than suitable for use in experimental settings as it is non-reactive, non-toxic, simple to manufacture, easy to use, and responds to a magnetic force.

Numerical Simulation of the Self-Leveling Phenomenon by Modified SIMMER-III

During a hypothetical core-disruptive accident in a sodium-cooled FBR, degraded core material can form debris beds on the core-support structure and/or in the lower inlet plenum of the reactor vessel, due to the rapid quenching and fragmentation of the core material melt. Heat convection and vaporization of the sodium will lead ultimately to leveling the debris bed that is of crucial importance to the relocation of the molten core, the recriticality evaluation and the heat removal capability of the debris bed. There is, therefore, a great need for more studies focusing on this topic, especially the much needed numerical simulation. The widely-used fast reactor safety analysis code, SIMMER-III, has difficulties in this simulation because of the lack of modeling for mechanistic interactions among particles in the current version. However, the extensive experimental analysis and the previously-proposed analytical model provide SIMMER-III the possibility of taking consideration of the extra influence of solid particles in this phenomenon. Thus, the debris fluidization model and the boiling regulation model are proposed and introduced into SIMMER-III. Calculations, by the modified SIMMER-III, against several representative experiments with typical self-leveling behavior have been performed and compared with the evaluated items recorded in experiments. The good agreements on these items suggest the modified SIMMER-III can simulate the self-leveling behavior with reasonable precision, especially on the onset of self-leveling, although further model improvement is necessary to represent the transient behavior of bed leveling more reasonably.

Sensitivity Study of the South Texas Project Power Plant Steady-State Simulations Using RELAP5-3D Coupled With Dakota

A RELAP5-3D input deck of the South Texas Project (STP) power plant was created in order to study the thermal-hydraulic behavior of the plant during normal operation (steady-state) and during a Loss of Coolant Accident (LOCA). It is important to study the sensitivity of selected output parameters such as the total coolant mass flow rate, the peak clad temperature, the secondary pressure, as a function of specific input parameters (reactor nominal power, vessel inlet temperature, steam generators primary side heat transfer coefficient, primary pressure etc.) in order to identify the variables that play a role in the uncertainty of the thermal-hydraulic calculations. RELAP5-3D, one of the most used best estimate thermal-hydraulic system codes, was coupled with DAKOTA, developed by Sandia National Laboratory for Uncertainty Quantification and Sensitivity Analysis in order to simplify the simulation process and the analysis of the results. In the present paper, the results of the sensitivity study for selected output parameters of the steady-state simulations are presented. The coupled software was validated by repeating one set of simulations using the RELAP5-3D standalone version and by analyzing the simulation results with respect of the physical expectations and behavior of the power plant. The thermal-hydraulic parameters of interest for future uncertainty quantification calculations were identified.

Integration of Direct/Indirect Influences of Severe Accidents for Improvements of Nuclear Safety

Risk evaluation is an important assessment tool of nuclear safety, and a common index of direct/indirect influences of severe accidents as a compound of risk is necessary then. In this research, various influences of severe accidents are converted to monetary value and integrated. The integrated influence is calculated in a unit of “cost per severe accident” and “cost per kWh”. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. To calculate the “cost per severe accident” and the “cost per kWh”, typical sequences of severe accidents are picked-up first. Containment failure frequency (CFF) and source terms of each sequence are taken from the results of level 2 probabilistic risk assessment (PRA). The source terms of each sequence is input into the level 3 PRA code OSCAAR which was developed by Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the results presented in this study are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated. It consists of various costs and other influences converted into monetary values. This methodology is applied to a virtual 1,100 MWe BWR-5 plant. Seismic events are considered as the initiating events. The data obtained from the open documents on the Fukushima Accident are utilized as much as possible. Sensitivity analyses are carried out to identify the dominant influences, sensitive assumptions/parameters to the cost per accident or per kWh. Based on these findings, optimization of radiation protection countermeasures is recommended. Also, the effects of sever accident management are investigated.

Severe Accident Analysis for BWR With Containment Outer Pool

Newly designed plants, e.g., next-generation light water reactor or ESBWR, employ a passive containment cooling system and have an enhanced safety with RHRs (Residual Heat Removal system) including active components. Passive containment cooling systems have the advantage of a simple mechanism, while materials used for the systems are too large to employ these systems to existing plants. Combination of passive system and active system is considered to decrease amount of material for existing plants. In this study, alternatives of applying containment outer pool as a passive system have been developed for existing BWRs, and effects of outer pool on BDBA (Beyond Design Basis Accident) have been evaluated. For the evaluation of containment outer pool, it is assumed that there would be no on-site power at the loss of off-site power event, so called “SBO (Station BlackOut)”. Then, the core of this plant would be uncovered, heated up, and damaged. Finally, the reactor pressure vessel would be breached. Containment gas temperature reached the containment failure temperature criteria without water injection. With water injection, containment pressure reached the failure pressure criteria. With this situation, using outer pool is one of the candidates to mitigate the accident. Several case studies for the outer pool have been carried out considering several parts of containment surface area, which are PCV (Pressure Containment vessel) head, W/W (Wet Well), and PCV shell. As a result of these studies, the characteristics of each containment outer pool strategies have become clear. Cooling PCV head can protect it from over-temperature, although its effect is limited and W/W venting can not be delayed. Cooling suppression pool has an effect of pressure suppressing effect when RPV is intact. Cooling PCV shell has both effect of decreasing gas temperature and suppressing pressure.

Supercritical-Water Experimental Setup for In-Pile Operation

The main goal of the Generation-IV nuclear-energy systems is to address the fundamental research and development issues necessary for establishing the viability of next-generation reactor concepts to meet future needs for clean and reliable energy production. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a SuperCritical Water-cooled Reactor (SCWR), which continues the utilization of well-known light-water-reactor technologies. Research Centre Rez Ltd. has taken part in a large European joint-research project dedicated to Generation-IV light-water reactors with objectives to contribute to the fundamental research and development of the SCWRs by designing and building a test facility called “SuperCritical Water Loop (SCWL)”. The main objective of this loop is to serve as an experimental facility for in-core and out-of-core corrosion studies of structural materials, testing and optimization of suitable water chemistry for future SCWRs, studies of water radiolysis at supercritical conditions and nuclear fuels. This paper summarizes the concept of the SCWL, its design, utilization and first results obtained from non-active tests already performed within the supercritical-water conditions.

Application of Fine Depletion Calculation for LLFPs In-Core Transmutation Evaluation in PWRs

An in-core transmutation analysis and evaluation code, named CATE, considering in-core fine flux calculation and fine depletion process, is verified and validated in the present paper. Verification and Validation of implementations for the OECD/NEA PWR cell benchmark for actinides transmutation, IAEA PWR benchmark and infinite homogenized plate problem to confirm reliability and numerical accuracy for the code have been performed in presented paper. The numerical performance of the code system is demonstrated in the analyses of the in-core fuel management calculation. It is found that the present code system gives stability in prediction of critical concentration of boric solution and radial power distribution. Based on the verifications and validations, a preliminary LLFP transmutation pattern is calculated. Numerical results indicate that CATE can be used not only for the fuel management calculation, but also for in-core transmutation evaluation of PWR.

Code Simulation of Quenching of a High Temperature Debris Bed: Model Improvement and Validation With Experimental Results

The loss of coolant accidents with core degradation e.g. TMI-2 and Fukushima demonstrated that the nuclear safety analysis has to cover accident sequences involving a late reflood activation in order to develop appropriate and reliable mitigation strategies for both, existing and advanced reactors. The reflood (injection of water) is possible if one or several water sources become available during the accident. In a late phase of accident, no well-defined coolant paths would exist and a large part of the core would resemble to a debris bed e.g. particles with characteristic length-scale: 1 to 5 mm, as observed in TMI-2. The French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation at a late stage of an accident. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core for ICARE-CATHARE code. The comparison of the calculations with PRELUDE experimental results is presented. It is shown that the quench front exhibits either a 1D behavior or a 2D one, depending on injection rate or bed characteristics. The PRELUDE data cover a rather large range of variation of parameters for which the developed model appears to be quite predictive.