Plenum Model Research Articles

Low-frequency fluctuation propagation of rotating stall in the centrifugal compressor and pipe system

The characteristic signal judgment of rotating stall in a centrifugal compressor is necessary to avoid the compressor becoming instability in operation. This study investigates rotating stall in a centrifugal compressor and pipe system using experimental and numerical simulation methods. In the experiment, a low-frequency (approximately 10% of the impeller's rotational frequency) pressure fluctuation is observed in the inlet and outlet pipes as the pressure ratio curve declined at small flow rate. The frequency spectrum results of different measuring locations suggest that the low-frequency disturbance is in the flow direction within the pipe system during rotating stall. To further analyze this pressure fluctuation, a transient numerical simulation of the centrifugal compressor with plenum model is conducted. Rotating stall can be captured by the numerical model, and the low-frequency pressure fluctuation is also observed in the transient simulation, aligning with the experimental results. The periodic evolution of the tip leakage vortex influences the flow in the impeller passage, causing a fluctuation in the flow direction that propagates upstream and downstream as revealed by flow field analysis. The low-frequency pressure fluctuation in the inlet and outlet pipe system is a characteristic signal, which can be a new stall judgment of rotating stall in the operation of centrifugal compressor.

Benchmark analysis of the FFTF LOSWOS test #13 with OpenMC and THACS

In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi’an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie’ outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR.

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Abstract Feedback reactivity caused by core deformation is one of the inherent safety features in a sodium-cooled fast reactor (SFR). To validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code, named Super-COPD, we have been conducting the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of balance of plant (BOP)-302R and BOP-301 as one of the representative issues in Experimental Breeder Reactor-II (EBR-II), a pool-type experimental SFR in the United States. During the transient of the ULOHS tests, the reactor power decreased to the decay heat level due to the negative reactivity caused by the radial expansion of the core. Developing the evaluation models of the reactivity feedback is essential for predicting the plant behavior in unprotected events, such as ULOHS events, and the numerical analyses modeling the feedback reactivity were conducted. By comparing the numerical results and the experimental data, the increment temperature profiles of the core inlet and the decreasing reactor power profile calculated by Super-COPD were comparable with experimental data. The applicability of the evaluation models for the reactivity feedback was indicated during the ULOHS event. In addition, through the sensitivity analyses on the cold pool model, it was found that the modification of the plenum model of the cold pool to take into account the thermal stratification was required to improve the core inlet temperature profile.

Numerical simulation method of surge experiments on gas turbine engines

In order to obtain the surge margin of an aero-engine during its operation, an engine surge experiment is required. A multi-dimensional simulation method for an aero-engine is established in this paper. The simulation of a surge experiment using high-pressure air-injection is then carried out on a turbo-shaft engine to obtain the surge boundary using this method. More specifically, firstly, a body-force model is employed to calculate the compressor performance owing to its capability of capturing the main three-dimensional features of compressor surge and avoiding excessive simulation time required by the traditional fully-three-dimensional Reynolds Averaged Navier-Stokes (RANS) method. Then, a one-dimensional model combining a lumped-parameter plenum model is used for the combustor to account for the propagation of pressure waves and the heat-release process, and a zero-dimensional throttle model is used to mimic the choking effect at the turbine nozzle. Finally, the air-injection system is modeled by imposing an injection boundary condition, which can be used conveniently in changing injection parameters. Based on the established method, the influences of different test parameters, such as the air-injection location, the pressure, the orifice size, the number of injection orifices, and the injection time duration on the surge characteristics and boundary are further studied, which offer effective guidance to optimize an actual experimental design.

On the prediction of the reverse flow and rotating stall characteristics of high-speed axial compressors using a three-dimensional through-flow code

This paper presents the development of a low-order three-dimensional through-flow code created at Cranfield University in the UK, named ACROSS (Axial Compressor Rotating Stall and Surge simulator), and its application to create the reverse flow and rotating stall characteristics of a modern high-speed compressor. The compressor modelled is a six-stage axial-flow machine, representative of a modern aero-engine high-pressure compressor. The tool has been previously validated using experimental data from two low-speed compressor rigs. This article describes how the tool's robustness and computational speed have been improved by introducing higher-order schemes to model the circumferential fluxes and the rigid movement of the flow in the rotating blade rows. Further improvements include variable axial discretization, an algorithm to introduce random flow perturbations in the flow field and an improved plenum model. The compressor is first modelled in reverse flow conditions to create its reverse-flow characteristics and these are then compared against results from high-fidelity 3D CFD simulations. Results obtained suggest that despite the presence of three-dimensional flow features, 2D axi-symmetric simulations are adequate to generate the full range of reverse flow characteristics of the compressor. The rotating stall characteristics at 77% and 100% corrected rotational speed are created by modelling several steady rotating stall cases in 3D. Using the code ACROSS, the complete map of the compressor modelled, comprising of forward flow, reverse flow and rotating stall characteristics, was created in only 5 days using 3 desktop workstations. For comparison, state-of-the-art high-fidelity 3D CFD requires several days to simulate a single rotating stall case on a high performance computing facility.

Effect of ribbed and smooth coolant cross-flow channel on film cooling

The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M=0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M=1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M=0.5 than at M=1.0. The film cooling with the cross-flow model is better at a blowing ratio of M=0.5 in the near hole region, while further downstream, it is better at M=1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

The use of ducts to improve the control of supply air temperature rise in UFAD systems: CFD and lab study

Cool supply air flowing through the underfloor plenum is exposed to heat gain from both the concrete slab (conducted from the warm return air on the adjacent floor below the slab) and the raised floor panels (conducted from the warmer room above). The magnitude of this heat gain can be quite high, resulting in undesirable loss of control of the supply air temperature from the plenum into the occupied space. These warmer supply air temperatures can make it more difficult to maintain comfort in the occupied space (without increasing airflow rates), particularly in perimeter zones where cooling loads reach their highest levels. How to predict plenum thermal performance is one of the key design issues facing practicing engineers – evidence from completed projects indicates that excessive temperature rise in the plenum can be a problem.One of the recommended strategies for addressing temperature rise in UFAD systems is the use of ductwork (flexible or rigid) within the underfloor plenum to deliver cool air preferentially to perimeter zones or other critical areas of high cooling demand.Several experiments were carried out in a full-scale underfloor plenum test facility, in order to characterize all the phenomena that take place in an underfloor plenum equipped with a fabric or metal duct. Experimental data were collected for validation of a computational fluid dynamics (CFD) model of the plenum. This paper describes the first part of a more comprehensive work, whose aim is to use the validated CFD plenum model to conduct simulations of a broader range of plenum design and operational parameters. This work proves that using ductwork within the underfloor plenum reduce the temperature rise in the plenum.

Experimental and Numerical Study of Film Cooling with Internal Coolant Cross-Flow Effects

This study experimentally and numerically analyzes the effect of the coolant cross-flow on film-cooling effectiveness. The three kinds of film holes considered are cylindrical, fan-shaped, and trenched holes. The results show that the internal coolant flow has a large effect on film cooling. The film-cooling effectiveness of the cross-flow coolant model is larger than that of the plenum model for the cylindrical and trenched holes, but it is less for the fan-shaped hole. A comparison of the experimental and numerical results shows that the Reynolds averaged Navier–Stokes model exaggerated the coolant diffusion downstream but underestimates the lateral distribution.

Numerical and modeling issues in application of CFD to flow in a simplified plenum relevant to a prismatic VHTR

Abstract Computational Fluid Dynamics (CFD) calculations have been performed for turbulent flow inside a plenum model that resembles a section of the lower plenum of a typical helium-cooled prismatic Very High Temperature Reactor (VHTR). Different Reynolds Averaged Navier–Stokes (RANS) based turbulence models are employed to investigate the capability in capturing unsteady large scale coherent structures due to vortex shedding at two different Reynolds numbers. A grid convergence study is conducted with those models which were able to capture unsteady vortex shedding. The non-linear interaction of mesh quality, turbulence model and numerical scheme lead to flow regime changes with significantly different unsteady behaviors. This makes it difficult to assess numerical and modeling uncertainty using the procedures available in the literature. Some remedies to overcome this difficulty are recommended. The numerical uncertainty in the local values of velocity components at selected locations inside the plenum, as well as the uncertainty associated with derived quantities such as wall shear stress at critical locations are calculated and reported. Since there are no experiments corresponding to the present cases simulated, the current analysis can be considered as a blind application of the proposed uncertainty estimation procedures.

Measurement of Flow Phenomena in a Lower Plenum Model of a Prismatic Gas-Cooled Reactor

Mean-velocity-field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor similar to a General Atomics gas-turbine-modular helium reactor design. The data were obtained in the Matched-Index-of-Refraction (MIR) Facility at Idaho National Laboratory (INL) and are offered for assessing computational fluid dynamics software. This experiment has been selected as the first standard problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined cross flow—with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate geometry scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages in and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolutions compared with similar facilities at smaller scales. A three-dimensional particle image velocimetry system was used to collect the data. Inlet-jet Reynolds numbers (based on the jet diameter and the time-mean bulk velocity) are approximately 4300 and 12,400. Uncertainty analyses and a discussion of the standard problem are included. The measurements reveal developing, nonuniform, turbulent flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices), and presentations that describe the component flows at specific regions in the model. Information on inlet conditions is also presented.

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.

ICONE15-10808 Scaled Experimental Modeling of VHTR Plenum Flows

The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. Various scaled heated and unheated gas and water flow facilities were investigated for modeling VHTR upper and lower plenum flows during the decay heat portion of a pressurized conduction-cooldown scenario and for modeling thermal mixing and stratification ("thermal striping") in the lower plenum during normal operation. It was concluded, based on phenomena scaling and instrumentation and other practical considerations, that a heated water flow scale model facility is preferable to a heated gas flow facility and to unheated facilities which use fluids with ranges of density to simulate the buoyancy effect of heating. For a heated water flow lower plenum model, both the Richardson numbers and Reynolds numbers may be approximately matched for conduction-cooldown natural circulation conditions. Thermal mixing during normal operation may be simulated but at lower, but still fully turbulent, Reynolds numbers than in the prototype. Natural circulation flows in the upper plenum may also be simulated in a separate heated water flow facility that uses the same plumbing as the lower plenum model. However, scaling distortions will occur due primarily to the necessity of using a reduced number of channels connected to the upper plenum than in the prototype in an otherwise geometrically scaled model. Experiments conducted in either or both facilities will meet the objectives of providing benchmark data for the validation of codes proposed for NGNP designs and safety studies, as well as providing a better understanding of the complex flow phenomena in the plenums.

Hydromechanical investigation on 3 PWR upper plenum core structures

In this paper, experiments with three upper plenum core structures—the upper plenum core structure of active 300 MW PWR in the People's Republic of China and its two improved core structures, respectively—are carried out in a 1/4 scale transparent model of the active 300 MW PWR upper plenum with water. The horizontal across velocity in the upper plenum model is measured with an N-J type dynamic resistance strain foil velocity meter (N-J DRSFV). The axial velocity is obtained by using a laser Doppler velocity measuring system (LDV). The experimental results are compared and analyzed carefully. The complex flow velocity distribution is obtained in the experiment with each core structure. Finally, it is concluded that the improved upper plenum core structure can reduce the hydraulic load on the drop of control rods.

A methodology for the design of perforated tiles in raised floor data centers using computational flow analysis

An important aspect of the design of data centers involves sizing of the perforated floor tiles for return of cold air, the size of the space under the raised floor, and placement of the DP equipment and modular chillers. The flow through individual perforated tiles needs to fulfil the cooling requirements of the computer equipment placed adjacent to them. The novelty of this paper lies in the treatment of the volume under the raised floor as a uniformly pressurized plenum. The accuracy of the Pressurized Plenum model is demonstrated with reference to a computational fluid dynamics (CFD) analysis of the recirculating flow under the raised floor and the limits of its validity are also identified. The simple model of the volume under the raised floor enables use of the technique of flow network modeling (FNM) for the prediction of the distribution of flow rates exiting from the various tiles. An inverse design method is proposed for one-step design of the perforated tiles and flow balancing plates for individual chillers. Subsequent use of the FNM technique enables assessment of the performance of the actual system. Further, required design changes to an existing system can also be evaluated using the FNM analysis in a simple, quick, and accurate manner. The resulting design approach is very simple and efficient, and is well suited for the design of modern data centers.

Effects of Reynolds number and Richardson number on thermal stratification in hot plenum

Thermal transient water tests were performed using a simplified hot plenum model of a pool-type LMFBR to examine the fundamental characteristics of thermal stratification during reactor trips. These tests were divided into two series to clarify the effects of (1) Reynolds number and (2) Richardson number on the behavior of thermal stratification. The results of the first series indicated that the characteristics of thermal stratification were independent of Reynolds number in the range of Re > 10 4. It was concluded from the results of the second series that the rising speed of the thermal interface and the temperature gradient at the interface varied proportionally with Ri −1 2 and Ri 1 2 , respectively. In addition, large scale internal waves, called internal standing waves, were observed under the condition of Re > 10 4 and 1 < Ri < 5.

ＡＣＶ上下揺れにおけるクッション安定性の非線形解析について 続報

Anonlinear analysis of the cushion stability of slowly oscillating ACVs has been made. The 'inertance' of the air slug in the ducting has been considered to investigate the unsteady effect. It is found that the unsteady effect becomes noticeable in a frequency range higher than 1 Hz. The result of the analysis is compared with the result of the forced heaving test of a plenum model. The agreement between the two is excellent. It is thus coucluded that the inertance of the air in the ducting is the dominant unsteady effect in a frequency range (up to 1 Hz) of the present study. The effect of compressibility of air in the ducting and the cushion has also been considered quasistatically but it is found to be negligible for the conventional marine ACVs of low cushion pressure and low oscillation frequency.

Thermo-hydraulic model test of the first nuclear ship reactor in Japan

A series of thermo-hydraulic model tests were carried out to confirm that the first nuclear ship reactor in Japan was properly designed. The major test models were a half-size core structural model, a full-size fuel assembly model and a half-size steam generator plenum model. The measured results were compared with the design calculation of the reactor. Most of the calculated results were found to be in good agreement with experiment. However, the pressure drop in the reactor vessel was found to be overestimated, and a necessary correction was made on the system design. Also flow velocity distribution was found to be much more distorted than expected in the fuel assembly, and an appropriate design modification was made in the lower end plate. Flow coastdown in the primary loop and coolant mixing in the fuel assembly are also discussed.