In severe accidents of nuclear power plants, large amounts of gas containing radioactive particles are generated. In the process of gas release into the atmosphere, it is needed to suppress the leakage of radioactive material. The gas is decontaminated by moving radioactive particles from the gas phase to the liquid phase. This effect of capturing particles is called pool scrubbing, and it has been verified great decontamination effect. Therefore, it is extremely important to analyze the effect in evaluating the influence to the environment. But study on its principle is not carried out sufficiently. And also we don’t have enough experimental date to analyze the effect. The purpose of this study is to clarify the gas-liquid two-phase flow behavior which is important in elucidating the mechanism of pool scrubbing. Particularly, this study focused on the behavior of bubble generation and breakup after being injected from the nozzle and the flow structure of rising bubbles in the still water. Furthermore, we evaluate the validity of the model used in the existing severe accidents analysis code such as the MELCOR by comparing the model with experimental data. We measure the gas phase jet injected from the upward nozzle inserted to a test water tank. Nozzle diameter, gas phase flow rate, liquid phase temperature, and water depth were used as parameters. Bubble behavior was observed via a high-speed camera. The bubble rising speed, bubble distribution and void fraction were measured by a wire mesh sensor. In previous studies, experiments using non-condensable gas in normal temperature water have been mainly conducted. In order to conduct the experiment under conditions that simulate actual equipment, steam which is a condensable gas was used in this study. Moreover, it is assumed that thermal stratification is formed in the pressure suppression pool during severe accidents. To reproduce this situation, thermal stratification is formed in the test water tank. For bubble behavior and flow phenomenon, the result of using non-condensable gas was compared with that using steam. We consider the influence of formation of a thermal stratification. As described above, the flow phenomenon in the pool scrubbing was visualized and measured. Finally, we discuss the validity of the analysis code by comparing the calculation formula and model in the analysis code with the experiment data.

Thermal hydraulic behavior in the upper plenum of pool-type sodium-cooled fast reactors (SFRs) is a major concern, as many design challenges are concentrated in this region. As SFR designs aim for licensing and commercialization, it is important to accurately analyze and predict the thermal-hydraulic behavior in this region during accident scenarios, specifically thermal stratification. Thermal stratification models are currently a major source of uncertainty in most system codes for all types of power plants. Most system codes, including SAS4A/SASSYS-1, a system level code developed by Argonne National Laboratory (Argonne), use very coarse meshes that cannot capture the complexities of the stratification phenomena. While the commonly employed lumped-volume based models for thermal stratification are able to run in a matter of seconds, they result in approximate results and can only handle simple cases. Other 2-D and 3-D methods, such as computational fluid dynamics (CFD) models, can analyze simple configurations with higher fidelity, but come with a relatively large computational expense. Finding a modeling solution that is both accurate and computationally efficient has proven difficult. This paper provides details of a review and gap analysis of the various modeling approaches proposed to date and explores a path forward for future thermal stratification modeling efforts, with a focus on developing new models for the SAS4A/SASSYS-1 system code.

Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k–ε model (SKE) and the realizable k–ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.

In nuclear plant piping, when high-temperature water penetrates from main pipe, a thermal stratification may be formed in a branch pipe with a closed end. The penetrating flow of high-temperature water is called as a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur due to periodic temperature fluctuations. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In the present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable range is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe. In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping. This paper presents experimental results and consideration of thermal hydraulics phenomena about thermal stratification. The horizontal piping test and the vertical-horizontal branch piping test are conducted. Pipes are made of acrylic resin for visualization of water flow. The water temperature is less than 60°C. The horizontal piping test reproduces natural circulation flow by installing a heater which imitated heating from cavity flow, and a chiller which imitated heat dissipation from a blockage valve. Temperature profiles are obtained by using thermocouples and optical fibers. Moreover, water flow in piping is visualized by injected ink. From the visualization and the temperature measuring result in the horizontal piping test, flow characteristic of the natural circulation region of large diameter piping is clarified The vertical-horizontal branch piping test reproduces cavity flow penetration from main pipe. Temperature profiles are obtained at cross-section direction and axial direction. Moreover, the amount of mass transfer at the thermal stratification which is formed between a cavity flow and natural circulation is evaluated experimentally by using a lithium tracer method. The second paper[Ref.2] explains modeling of evaluation method for cavity flow penetration depth in large diameter piping.

In nuclear plant piping, when the high-temperature water penetrates from main pipe, a thermal stratification interface may be formed in branch pipe with a closed end. The penetrating flow of high-temperature water is called a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur by a periodic temperature fluctuation. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable piping is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe. In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping. This paper presents about modeling of evaluation method for cavity flow penetration depth in large diameter piping. First, the mass transport model at the interface of layer in natural circulation region is arranged with Reynolds number and Richardson number based on visualization and the temperature measuring result. Second, the mass transport model at a thermal stratification interface between a cavity flow domain and a natural circulation domain is built based on time variation of lithium tracer concentration. The evaluation model of cavity flow penetration depth applicable to large diameter piping is proposed by combining new mass transportation models and existing evaluation guideline. In conclusion, predictive accuracy of cavity flow penetration depth in 200 mm diameter piping became the same order as the evaluation result of 50 mm diameter piping by using present evaluation guideline.

In this study, RELAP5’s capability to simulate thermal stratification under different conditions is assessed. In nuclear power plants (NPPs), thermal stratification can occur in the following locations: pressurizer, piping systems such as hot legs, cold legs, surge lines, and cooling tanks if available. In general, thermal stratification in a horizontal pipe could not be simulated by RELAP5 due to the inherent one-dimensional setting. Moreover, RELAP5 failed to simulate turbulent penetration which was often a pre-requisite prior to thermal stratification in a pipe. This type of situation could arise in connection between hot leg and surge line, spray lines, feed water lines, etc. It is recommended that for this type of problem CFD be used. In the literature, it was found that RELAP5 was capable of simulating thermal stratification in a pool or a tank-like component if multiple channels and crossflow junctions were used. However, due to uncertainties associated with the input model, the current RELAP5 model failed to reproduce experimental data and therefore further investigation would be required to identify the sources of error.

Thermal stratification of pressurizer surge line in nuclear power plant induced by the stratification of fluid results in additional stresses (system-dependent and system-independent stresses). These additional stresses are related to the thermal stratification distribution, such as temperature difference, stratification location, stratification length and boundary layer slope and thickness. This thermal stratification distribution is effected by several parameters such as turbulent penetration length and surge line slope. These influence factors are studied with numerical method. For the additional stresses, this study theoretically establishes the solution of the system-dependent stresses according to the thermal stratification distribution assumption. With the system-dependent and system-independent stresses, a complex three-dimension problem is simplified into a one-dimension. Using this effective one-dimension method, fatigue evaluation can be fast processed and accurate improved in engineering analyses.

The In-containment Refueling Water Storage Tank (IRWST) provides low-pressure safety injection flow for passive CAP1400 Nuclear Power Plant (NPP) during Loss-Of-Coolant-Accident (LOCA) and subsequent Long Term Core Cooling (LTCC). The Passive Residual Heat Removal Heat Exchanger (PRHR HX) and the spargers of Automatic Depressurization System (ADS) stage 1∼3 are submerged in the IRWST. During small break LOCA, heat and mass are delivered through PRHR HX and ADS spargers to IRWST, and IRWST is heated up before its safety injection. However, numerical and experimental investigation has shown that IRWST is not perfect mixing, and thermal stratification exists. During ADS-4/IRWST initiation phase, the temperature of IRWST injection flow is of great importance, and is affected greatly by IRWST simulation method when modeling with system code like RELAP5. In this paper, two different IRWST simulation methods where one use multi channels in horizontal direction while the other use only one, are analyzed for CAP1400 SBLOCA with RE-LAP5, and their effects are compared. Finally, the better method which uses only one channel in horizontal direction is recommended.

The accident progression and melt behavior at the Fukushima Daiichi Units 1 and 2 were investigated using MELCOR 2.1. In the modeling the lower head failure mechanism by penetration tube rupture and ejection was modeled. In the modeling of Unit 2, according to the latest findings by TEPCO investigation, the possibilities of torus room flooding, RCIC piping leakage and thermal stratification in suppression pool were taken into account. The analysis results indicate that for Unit 1 when considering penetration tube failure, a part of debris still remained in the lower head after debris discharge; otherwise all the debris discharged out. The present MELCOR modeling of Unit 2 well reproduced the RPV and PCV pressure. A part of the core was damaged and the debris that slumped into the lower head was sufficiently cooled down. The pressure vessel kept intact.

Pressurizer surge lines are essential pipeline structure in NPPs, and the thermal stratification in surge line is recognized as one of the possible cause of thermal fatigue. In this paper, a Computational Fluid Dynamic (CFD) method has been adopted to simulate temperature fluctuations on the process of temperature rising in a pressurizer surge line under rolling motion of single degree of freedom. This work focuses on a fundamental description of differences of thermal stratification between the surge line rolling around the coordinate X-axis condition and that in a static state. The Large-eddy simulation (LES) model is employed to capture the details of temperature change in surge line. Temperature distributions near the inner wall of a surge line pipe with or without swinging were monitored and compared. The temperature differences between the top and bottom of the pipe sections are employed to represent the maximum temperature differences at all the monitored sections. As the surge line swinging, the pattern of temperature distribution and the length of thermal stratification development are different from that in a static. Fluid temperature fluctuation in surge line occur periodically during the fluid temperature rising when the surge line is rotated with the X-axis, and the temperature difference between top and bottom of the surge line is reduced in the same motion mode compared with the static state.

For AP1000 reactor, passive containment cooling system is a vital way to release heat to the environment, so an accurate prediction of distribution of temperature or density in large layered space plays an important role on the reactor optimization design and safety analysis. This paper investigates in comparing the results of different kinds of models in FLUENT with the experiment results in order to find out a more effective model and more suitable mesh number to simulate the mixing and stratification phenomenon. When LOCA or MSLB occurs in containment, the radius, position, and angle of the break can affect the containment mixing and thermal stratification. So this paper also studies the influence of height of the break and angle of the break on stratification with Fluent, and makes comparative analysis with the experiment results.

Pipelines are widely used in many fields including power industry, petroleum system etc. Pipelines such as the surge line and main pipe are easily subjected to thermal stratification as a result of the non-uniform temperature distribution in the nuclear power plants. Furthermore, pipelines can suffer from thermal fatigue in virtue of long-term uneven stress distribution. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen because of the existence of pipeline crack. The thermal pipeline crack leakage mainly appears in the region with stress concentration. As the pipe system is always covered with thermal insulation layer in the actual nuclear power plants, it is hard for workers to observe pipeline leak, which can have a bad effect on the normal operation. Since the temperature and humidity close to the pipe crack due to leakage can change compared to the normal operation, we can infer from the temperature and humidity changes that the pipe leakage occurs. Based on this idea, the temperature and humidity near the crack of the pipe need to be measured to detect the leakage fields. As the fluids with high pressure and high temperature flow in the pipe system in an actual nuclear power plant, the pipe leakage experiment was performed in the high pressure and high temperature condition. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The temperature and humidity variations were monitored in the experiment with temperature and humidity probes which have the advantage of responding to the change of temperature and humidity sensitively. The data collection program was mainly written based on the LABVIEW platform. The collecting time step was set 1s. As the measuring position and leakage flux are two key factors for the pipe leakage, the experiment was carried out with different measuring positions and leakage fluxes conditions. The experimental results showed that the leak flux had an important influence on the temperature and humidity near the pipe crack. The temperature and humidity started to change in a very short time with large leak flux. At the same time, the velocity of the temperature and humidity change was high with large leak flux. When the pipe leakage occurred in the location near the temperature and humidity probe, the temperature and humidity responded quickly and the velocity of temperature and humidity change was large. The experiment data can be used for the prediction of the pipe leakage in the nuclear power plants.

This study was based on the passive containment thermal hydraulic facility, which locating in Beijing Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing. The platform was built in the year of 2010, Pro. Niue has carried out a series of studies about the thermal stratification and concentration stratification for air and water vapour and achieved remarkable results. In this paper, the experimental test is divided into two consecutive stages: phase A and phase B. Phase A is the injection phase of steam, however, phase B is divided into three steps, include hydrogen injection, no injection, opening the external spray, to study the stratification and transport of hydrogen (replaced with helium) in a cylindrical steel container which has ellipsoid dome, and also discussing the change of hydrogen concentration near the wall over time. Except for this, the distribution of hydrogen during the third step is also emphasized. For the transport of gas, it has not given the direct result of measuring in this paper, but it gave the analysis of transport which is based on the hydrogen concentration at different stages, different location of the containment over the time. It can be found that: 1) during the stages of hydrogen injection, no injection, the stratification is obvious, but, the helium concentration in the dome space is uniform, and the transport of the gas is mainly affected by the thermal buoyancy and jet, 2) near the wall-area, the trend of droplet entrainment is obvious, 3) during the stage of opening the external spray, the mixture of gas is intensified and the stratification is weakened.

Thermal stratification can occur when hot and cold fluids mix inside the pipeline of nuclear power plant, which can lead to temperature fluctuations with large amplitude and high frequency with time-varying wall heat stress and even induce thermal fatigue phenomenon. Therefore, how to accurately get temperature fluctuations of the inner wall without damaging the whole structure of the pipe system becomes the prime problem to be solved in the study of thermal fatigue. Base on energy conservation law, the non-iterative mathematic model of inverse heat conduction using the volume-control method has been developed in this paper for pipe in nuclear power plant. Numerical and experiment results show that the volume-control method can accurately catch the temperature fluctuation of different positions on the inner wall.

Scaled down thermal-hydraulic test facility is widely used in the nuclear reactor safety analysis. The wall stored energy is one of the common problem in a scaled down test facility, which will cause distortions in the simulation of the transients or local phenomena. For the natural-circulation mode in the Core Makeup Tank (CMT), the cold wall will work as heat sink and absorb heat from the heated water, and then has an influence in energy balance on the simulation of the prototype phenomena with scaled down test facility. In order to study the influence of the wall stored energy on the natural-circulation mode in the CMT test facility, this paper established the CMT test facility model with RELAP5/MOD 3.4, and studied the influence caused by the wall stored energy distortion. The simulation results show that the natural-circulation process in the CMT test facility can be divided into two stages: 1. Before the thermal stratification reaches the outlet, the natural-circulation flow rate decreases gradually. 2. Once the thermal stratification reaches the outlet, the natural-circulation flow rate has an apparent turning point and decreases faster. That is because the heated water reaches the outlet and fills the discharge line quickly, which reduces the density difference between the hot leg and cold leg, thus causing the natural-circulation flow rate decrease faster. Besides, before the turning point, the wall stored energy basically has no influence on the natural-circulation process, and the influence can be neglected in the design of test facility. However, after the turning point, the wall stored energy distortion results in a slower descent speed of natural-circulation flow rate, which is not conservative. Therefore, we can conclude that the scaled down CMT test facility can only simulate the prototype properly and conduct the natural-circulation simulation experiment before the thermal stratification reaches the exit, while afterwards, the simulation of the test facility won’t be accurate and conservative.

In AP1000 plant, the automatic depressurization system (ADS) works to discharge the high-temperature and high-pressure steam from the Reactor Coolant System (RCS) primary side to the In-containment Refueling Water Storage Tank (IRWST) in the LOCA conditions. In particular, for the AP1000 plant, both the IRWST and ADS spargers are specially designed, and the ADS spargers are located in one corner of the IRWST. All the factors lead to the special and complicated thermal and flow behavior in the IRWST, which in turn have great influences on the condensation effects of the ejected steam. In the present work, an overall scaled-down IRWST and ADS sparger models are built to study the condensation and mixing phenomena in the accidental depressurization events in AP1000. Thermocouples matrix with more than 200 T-type sheathed thermocouples are utilized to measure the three dimensional temperature in the large tank. The Particle Image Velocimetry (PIV) is employed for the measurement of the natural convection flow velocity. Based on the experimental data, the local spraying flow patterns, flow behavior, and thermal stratification phenomena in IRWST etc. are analyzed. The results indicate that the spraying steam condensation flow patterns are closely related to the subcooling degree in the IRWST. In addition, the stratification number is developed to evaluate the thermal stratification extent in the IRWST, which indicates that only part of the fluid are used efficiently for condensing the spraying steam directly.

Evaluation of accidental sodium leak, combustion, and its thermal consequence is one of the important issues to be assessed in the field of sodium-cooled fast reactor (SFR) since the liquid sodium is chemically active and might give thermal load to plant building structure due to its exothermic reaction with oxygen in air atmosphere. Therefore, many experimental investigations and numerical simulation tools development have been and still now are being carried out to understand the details of sodium fire behaviors and to contribute to the investigation and preparation of appropriate mitigation measures in the plant design. From various kinds of sodium fire situations, the present paper treats the sodium pool fire and subsequent heat transfer behavior in air atmosphere two-cell geometry both experimentally and analytically because such two-cell configuration is considered as the typical one to possess important characteristic of multi-compartment system seen in an actual plant. Main description of this paper consists of a sodium pool fire experiment that was performed in a rectangular-shaped two-cell system with an opening between the cells, and the discussion of the experimental results. Inner volume of the experimental cells is about 70 m 3 . The amount of used sodium and the prepared pool surface area in the experiment are about 55 kg and 2.25 m 2 , respectively. The experiment has provided the temperature data measured in more than 100 positions for atmospheric gas and structures other than the data of oxygen concentration and suspended sodium aerosols concentration in the cells. The analyses of the measured data clarify the basic characteristics of sodium pool combustion and consequential heat and mass transfer in the cells, for instance, suggesting several features of multidimensional thermal-hydraulic behaviors such as thermal stratification near the opening between the two cells. In the discussion, numerical analysis using a lumped-parameter based zonal model safety analysis code ‘SPHINCS’ and the comparison of its results with the experimental data are also carried out to investigate the validity and applicability of the code to this type of sodium fire situation.

MYRRHA (Multi-purpose hybrid research reactor for high-tech applications) is a lead-bismuth eutectic (LBE) cooled research reactor currently under development at SCK•CEN, the Belgian Nuclear Research Centre. The compact design of the pool-type primary system implies the presence of pronounced 3D thermal fluid-dynamic phenomena, which can affect the evolution of certain accidental transients such as loss of flow (LOF). System thermal-hydraulics (STH) codes, conceived to carry out global NPP safety analyses, present severe limitations in taking into account local 3D phenomena including flow mixing, thermal stratification, etc. To overcome this limitation, a promising solution is coupling STH codes with CFD codes, which can calculate complex flow fields but result, on the other hand, in too expensive computational resources for whole-plant simulations. A domain decomposition method that couples the STH code RELAP5-3D and the CFD code Ansys FLUENT has been developed and implemented. Proof-of-principle tests on simple configurations have been carried out to demonstrate its validity and to identify modeling and numerical issues. The experimental campaign carried out at the test facility TALL-3D, operated by the KTH Royal Institute of Technology in Sweden, has been selected for preliminary verification and validation (V&V) of this method. This paper presents the results of the coupled 1D-3D simulation of a forced-to-natural circulation transient event, whose evolution results to be strongly affected by flow mixing and stratification phenomena. The experimental validation, based on a high-quality set of experimental data, is currently on-going. Further development and validation activities will be carried out in the experimental facility ESCAPE, under commissioning at SCK•CEN, within the recently launched EU project MYRTE (Horizon 2020 programme).

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved in the HLM technology development. In particular, several experimental campaign employing HLM loop and pool facilities (CIRCE [1], NACIE [2], HELENA [3], HERO [4]) are carried out in order to support HLM technologies development. In this frame, suitable experiments were carried out on the CIRCE pool facility refurbished with the Integral Circulation Experiment (ICE) test section in order to investigate the thermal hydraulics and the heat transfer in grid spaced Fuel Pin Bundle cooled by liquid metal providing, among the others aim, experimental data in support of codes validation for the European fast reactor development. The study of thermal stratification in large pool reactor is relevant in the design of HLM nuclear reactor especially for safety issue. Thermal stratification should induce thermomechanical stresses on the structures and in accidental scenario conditions, could opposes to the establishment of natural circulation which is a fundamental aspect for the achievements of safety goals required in the GEN-IV roadmap. In the present work, a Protected Loss of Heat Sink with Loss Of Flow (PLOHS+LOF) scenario is experimentally simulated and the mixed convection with thermal stratification phenomena is investigated during the simulated transient, as foreseen in the frame of Horizon 2020 SESAME project [5]. A full characterization of thermal stratification inside the pool is presented, and the main results gained during the run are reported. The two tests named A (20 h) and B (6 h) here reported, essentially differs for the power supplied to the fuel bundle during the full power run (800 kW and 600 kW respectively). After the transition to natural circulation conditions, the power supplied to the bundle is decreased to about 30 kW simulating the decay heat. Finally the Nusselt number for the central subchannel of the fuel bundle simulator (FPS) is evaluated and compared with values obtained from Ushakov and Mikityuk correlations [6–7].

Large volumes or enclosures are utilized in nuclear systems such as the sodium fast reactor (SFR) upper plenum, spent fuel pools, and containment structures seen in common nuclear reactor designs. These volumes may be quiescent or turbulent the injection of a hotter or cooler jet of fluid. This can lead to turbulent mixing of the entire volume or thermal stratification of a hotter layer of fluid. Due to the large thermal gradients experienced by the structures due to a stratified volume, the structures can experience degradation which can lead to severe consequences such as a loss of coolant accident. In order to prevent or mitigate these consequences, computational fluid dynamics (CFD) codes can be utilized to predict the occurrence of turbulent mixing and thermal stratification. This data can be used in conjunction with structural analysis codes to determine the severity of phenomena occurrence on structures. Unfortunately, the validity of CFD codes to predict this type of behavior is limited due to the lack of experimental data to validate the predicted behavior. In response, the twin jet water facility (TJWF) created by the University of Tennessee was repurposed to create data sets of temperature using thermocouples and particle image velocimetry (PIV) for this effort. The temperature data collected using the thermocouples is similar to previous experiments conducted within an older version of the facility with a different geometric configuration. The most recently collected data was created by conducting several runs of each each set of experiment conditions and ensemble averaged. This was done to confirm the observed behavior is due to the physical processes and not due to noise or random happenstance. The spectral frequency responses of the temperature data were determined to observe frequencies or spectral behavior corresponding to turbulent mixing and stratification. The temperature and frequencies are reported to compare the experiments to simulations being conducted in conjunction with this study.