Effect Of Spacer Grids Research Articles

The effect of spacer grids on the stress applied to a post-LOCA cladding tube under horizontal vibrations

ABSTRACT There is a recognized need to discuss the preservation of the coolable geometry of fuel rods during long-term cooling after a loss-of-coolant accident (LOCA). Mechanical load induced by an earthquake may be a primary cause of loss of the coolable geometry after a LOCA. In this research, aiming to confirm the preservation of coolable geometry of fuel rods during an earthquake after a LOCA, we investigated the stress applied to the cladding tube with a rupture opening under a vibration condition via finite element analyses for a fuel assembly that considered the mechanical interaction between a fuel rod and spacer grids. The analyses demonstrated that the stress on the cladding tube increased with the axial holding force of the spacer grid on the fuel rod. The stress concentrating at the periphery of the rupture opening was compared with the strength of the post-LOCA cladding tubes determined via bending tests. This research concludes that the fuel rods are likely to be prevented from fracture due to bending arising from an earthquake during post-LOCA cooling unless the oxidation of the cladding tubes exceeds 15% equivalent clad reacted.

Measurement of Local Void Fraction of Air-Water Flow in an 8 × 8 Rod Bundle Using High-Resolution Gamma-Ray Tomography

High-resolution two-phase flow data in the rod bundle are important in the development and validation of high-fidelity models for computational fluid dynamics and subchannel codes, in particular, those pertaining to light water reactor cooling systems. The Michigan Adiabatic Rod Bundle Flow Experiment (MARBLE) has been constructed as a modular assembly of an 8 × 8 lattice rod bundle to simulate scaled pressurized water reactor and boiling water reactor subchannel assemblies. To establish a high-spatial resolution database of the void fraction in the reactor fuel assembly geometries, tomographic measurements were performed with the High-Resolution Gamma-ray Tomography System, which was designed and built in house; the detector system has a spatial resolution of less than 1.0 mm using 240 LYSO (Lu1.8Y0.2SiO5) scintillators with a fan-beam array. In the present study, the local void fraction was measured with the MARBLE facility under various air-water flow conditions (jg = 0.04 to 0.85 m/s and jl = 0.12 to 0.77 m/s) covering from bubbly to cap-turbulent flows. The local void fraction was also successfully measured under nonuniform and asymmetric air bubble distribution conditions with an investigation of the effect of spacer grids and mixing vanes on void drift across subchannels.

Impact of circumferential variation in power, neutron flux and spacer grids on structural behavior of SiC-SiC cladding

Silicon carbide (SiC) cladding is being studied for its potential application in light water reactors for improved accident tolerance. SiC-SiC composite material undergoes irradiation-induced, temperature-dependent swelling. The circumferentially non-uniform neutron flux and power generation in the fuel rod, will produce circumferentially non-uniform swelling and thermal expansion, ultimately leading to bending of the cladding. This bending will cause interaction between the cladding and the spacer grids and can be detrimental to the structural integrity of both the cladding and spacer grids. In this study we have analyzed the bending behavior of the cladding, the stress development in the cladding and the effects of spacer grids. The sensitivity of the cladding stress to the spacer grid material, extent of circumferential variation in power have been also evaluated. The results indicate that the spacer grids do not significantly affect the cladding stresses. The stress generated in the spacer grids due to its interaction with the cladding is found to be well below the strength of the spacer grid material. The sensitivity studies showed that changing the spacer grid material have negligible impact on the stresses in both cladding and spacer grid, while the extent of circumferential variation in power can lead to significant increase in the stresses in spacer grid.

Detailed Analysis of the Effects of Spacer Grid and Mixing Vanes on Turbulence in a PWR Subchannel Under DFFB Conditions Based on DNS Data

Spacer grids and mixing vanes exhibit a significant role in the thermal hydraulics of pressurized water reactors (PWRs), especially in the post loss-of-coolant accident regimes. A detailed analysis of the contrasting upstream and downstream turbulent flow features is of great importance to both system codes and computational fluid dynamics (CFD)–Reynolds-averaged Navier–Stokes (RANS) modeling. Further, with the advent of supercomputing resources and machine learning research, a data-driven approach to turbulence modeling is gaining popularity. However, owing to the complexities associated with large-scale, high-fidelity data collection capabilities, the application of machine learning–based turbulence models has been limited to simple geometries. In this work, using a highly scalable CFD code PHASTA, we have collected data from direct numerical simulations of a PWR subchannel with high spatial and temporal resolution. From the collected data we extract key turbulent flow features, including mean velocities and Reynolds stresses that highlight the effects of spacer grids and mixing vanes on downstream turbulence in a typical PWR subchannel. An invariant analysis of the anisotropic stress tensor is also presented, which further elucidates their effect on the nature of turbulence in the immediate upstream and downstream vicinity. The high-resolution data from the simulations are archived and intended for the development of data-driven RANS closure models that are capable of capturing the evolution of anisotropy in PWR subchannels.

Visualization of spacer grid effect on bubble behavior and CHF in a single-rod channel

The effect of spacer grids on bubble behavior and critical heat flux (CHF) has been investigated through visualization of a heated single-rod channel cooled with refrigerant R-134a over a wide range of conditions. The test section consisted of a heated rod inside a square channel, which was equipped with a viewing window. A high-speed video camera was used to record bubble-growth phenomena and evolution process in flow boiling. Several spacer grids were installed along the heated rod to maintain its location and minimize vibration. Two different configurations of spacer grid (the full span and half span structures) were used. The half-span arrangement of spacer grids intensified the “tearing effect” on bubbles and reduced the size of bubbles. The strip, rigid convex and spring have been shown to be the main pushers of the “tearing effect” of bubbles in addition to maintaining the position of the spacer grid. The mixing vane promoted the spiral rise of bubbles and thus enhanced the cross-flow mixing. “Tearing effect” and cross-flow mixing facilitated the heat transfer enhancement leading to improvement in CHF. A comparative study was performed for the experiment with and without mixing vane by quantitative analysis of bubble characteristics. The results could support the design of a spacer grid with mixing vanes.

CFD study on crossflow and heat transfer characteristics of single phase flow in rod bundle with spacer grid under rolling conditions

In order to study the thermal and hydraulic characteristics of nuclear reactors under ocean conditions and meet the urgent needs of the design of floating nuclear power plants, the crossflow and heat transfer characteristics of rod bundle with the spacer grid under rolling conditions are studied by using computational fluid dynamics (CFD) code. Firstly, revise and verify CFD codes suitable for rolling conditions. According to CFD results, the influence of heat flux on heat transfer coefficient is greater than that of pressure drops under this condition. And with the spacer grid, the total pressure drop between inlet and outlet is higher. Furthermore, the heat transfer coefficient is also higher due to the mixing effect of spacer grid. Under periodic inlet velocity, the fluctuation amplitude of total pressure, the temperature difference between inlet and outlet and heat transfer coefficient is increased obviously with the spacer grid. Under the condition of nonuniform heat flux, the variation law of crossflow with rolling motion will become more obvious and the spacer grid can enhance the crossflow mixing and improve the heat transfer capacity of the fluid under rolling conditions.

Experimental and numerical investigation on heat transfer of supercritical water flowing upward in 2 × 2 rod bundles

This paper presents the investigation of heat transfer of supercritical water flowing upward in 2 × 2 rod bundles, both experimentally and numerically. The experiment conditions are as follows: pressures from 22.9 MPa to 26.1 MPa, mass fluxes ranging from 450 kg/m2/s to 1500 kg/m2/s and heat fluxes ranging from 0.4 MW/m2 to 1.5 MW/m2. Four heated-rods of 8 mm outer diameter formed 2 set of experiment sections with pitch to diameter ratio (P/D ratio) 1.18 and 1.3. Heat transfer coefficients between the heated rod and fluid were conducted by 3 different methods and the effect of spacer grids was clearly observed. The experiment results show that heat transfer to supercritical water in rod bundles is strongly affected by flow channels and spacer grids. Moreover, a large database of heat transfer in rod bundles was carried out and HTC were compared with tube and annuli with the equivalent hydraulic diameter. The phenomenon of heat transfer difference in rod bundle circumferential direction and enhancement heat transfer in downstream of spacer grids were observed and investigated. In addition, Computational Fluid Dynamics (CFD) method was applied to give a deep insight of velocity profile, spacer effect and heat transfer enhancement in this research.

Numerical investigation of flow and heat transfer characteristics of subcooled boiling in a single rod channel with/without spacer grid

A multiphase flow numerical model based on Eulerian is built to simulate the characteristics of subcooled boiling. The thermodynamic performance in a single rod channel with/without spacer grid are obtained by numerical simulation, and the influence of spacer grid on flow and heat transfer characteristics of subcooled boiling is studied. The model is verified by comparing with the experimental results. After passing through the spacer grid, the high-temperature zone departures from the surface of fuel rod moves towards the cold wall. This impact extends to a certain downstream distance, and then it gradually declines. The influence of different components in spacer grid on the pressure drop is detailed. The pressure drop with spacer grid is 5% higher than that without spacer grid, within the allowable design limitation. The maximum value of section average secondary flow can reach 13.5% of the inlet velocity, indicating that the mixing vanes have stronger ability to produce transverse flow. The influence length of spacer grid on heat flux partitioning is about 27Dh. The effect of spacer grid on the wall heat flux partitioning mainly lies in the decrease of the evaporation heat flux, benefited to reduce the risk of nuclear reactor approaching DNB.

Development and assessment of a new rod-bundle CHF correlation for China fuel assemblies

One of the main objectives of subchannel analysis for a pressurized water reactor (PWR) core or fuel assembly is to predict the critical heat flux (CHF) which usually employs the correlation based on the local parameters. The effects of spacer grid, cold wall, and non-uniform heat flux are essential for the sake of accurate calculation of local parameters. The subchannel code CORTH is applied to determine the local thermal-hydraulic parameters at the occurrence of departure from nucleate boiling (DNB) for each set of bundle CHF data of China fuel (CF) assemblies. In order to develop a new CHF correlation for CF assembly, the minimum DNB ratio (MDNBR) point and the burn-out (BO) point methods are both employed. The three-step form was adopted for the correlation which considers the effects of spacer grids, cold walls, and non-uniform heat flux. The coefficients of the correlation were determined based on the nonlinear regression method. The analysis and assessment results against the experimental data indicate that the correlation can predict the CHF with high prediction accuracy and correct parametric trends. In comparison with MDNBR point method, the BO point method is recommended to predict the risk of DNB because it is more reasonable and conservative and has a better predication rate. When implemented with the subchannel code CORTH, the developed correlation can predict the thermal-hydraulic performances of CF assemblies.

The Effect of Spacer Grids on the Formation of Added Mass under the Bending Vibrations of a Fuel-Element Bundle in a Covered Fuel-Element Assembly of a VVER-440 Reactor

A significant effect of spacer gratings exerted on the dynamic characteristics of the covered fuel-element assemblies is shown. The calculated and experimental values of the added mass in the course of the bending vibrations of a bundle of fuel elements (fuel elements) are in good agreement between each other only when taking into account the flow of liquid in a narrow gap between the hexahedral surface of the spacer gratings and the hexahedral border of the cover of fuel-element assemblies. The small dimensions inherent in the spacer gratings in the axial direction of fuel-element assemblies lead to uneven distribution of the added mass along the length of the fuel-element assemblies, which is a significant factor in the study of bending oscillations of a bundle of fuel elements in liquid.

Rod Bundle Heat Transfer Thermal-Hydraulic Program

The U.S. Nuclear Regulatory Commission has been conducting thermal-hydraulic research using the Rod Bundle Heat Transfer (RBHT) facility at the Pennsylvania State University since 2001. The facility has been used for five individual test programs: forced reflood, steam cooling, mixture level swell, dispersed droplet injection, and oscillatory reflood test series. While rod bundle thermal hydraulics has been extensively studied in the past, the RBHT data have provided new insights into rod bundle phenomena especially on the effects of spacer grids. This paper provides a summary of the RBHT test program and discusses some of the major findings from this research with the emphasis on reflood thermal hydraulics and the effect of spacer grids. Of particular interest are data that enable model and correlation development. Recent efforts have focused on the evaluation of RBHT data and development of improved models and correlations suitable for systems thermal-hydraulic codes such as TRACE and RELAP. Because of detailed instrumentation on and about spacer grids, RBHT data have enabled improved models for convective heat transfer enhancement and droplet breakup. New correlations for the inverted annular and the inverted slug film boiling regimes have also been developed as an initial step toward an improved model for dispersed droplet film boiling.

Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR

Abstract Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. The most important aspect of turbulent motion is that the velocity and pressure at a fixed point do not remain constant with time even in steady state but go through very irregular high frequency fluctuations. These fluctuations influence the diffusion of scalar and vector quantities. The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type, heavy water moderated and boiling light water cooled natural circulation based reactor. The fuel bundle of AHWR contains 54 fuel rods set in three concentric rings of 12, 18 and 24 fuel rods. This fuel bundle is divided into number of imaginary interacting flow channel called subchannels. Alteration from single phase to two phase flow situation occurs in reactor rod bundle with raise in power. The two phase flow regimes like bubbly, slug-churn, and annular flow are generally encountered in reactor rod bundle. Prediction of thermal margin of the reactor has necessitated the investigation of turbulent mixing rate of coolant between these subchannels under these flow regimes. Thus, it is fundamental to estimate the effect of spacer grids on turbulent mixing between subchannels of AHWR rod bundle.

Modeling of spacer grid mixing effects through mixing vane crossflow model in subchannel analysis

Spacer grids, especially those with mixing vanes or other mixing devices in the rod bundle, have great influence on local flow fields in the reactor core. This mixing effect has been demonstrated by both experimental and numerical simulations, however, the spacer grids’ effect on flow field is not well reflected in most subchannel codes that are currently used as main instruments in calculating local conditions and safety analysis for the nuclear industry. This paper focuses on improvement of the mixing vane crossflow model in subchannel analysis for simulating the mixing effects of spacer grids. The distributed resistance method (DRM) is applied, and the source terms associated with the spacer grids are added to momentum equations to account for the mixing vane grids’ influence on local flow fields in rod bundles (U.S. patent, application number: 61841961, confirmation number: 8224; U.S. patent, application number: 62375023, confirmation number: 2306; Chinese patent, application number: 201611179395.1). The improved DRM model is used to analyze the thermal hydraulic performance in the 5×5 rod bundle. The calculation results are compared with experimental data and CFD simulation results respectively. The subchannel calculation results are also compared with those from the conventional code without the DRM model. The flow characteristics, such as the crossflow direction, the crossflow values at grid zone, and the decay trend of the crossflow downstream of the mixing vane grid are analyzed. The improved mixing vane crossflow model is assessed by comparison analysis, and some conclusions are obtained.

Bubbly, Slug, and Annular Two-Phase Flow in Tight-Lattice Subchannels

Abstract An overview is given on the work of the Laboratory of Nuclear Energy Systems at ETH, Zurich (ETHZ) and of the Laboratory of Thermal Hydraulics at Paul Scherrer Institute (PSI), Switzerland on tight-lattice bundles. Two-phase flow in subchannels of a tight triangular lattice was studied experimentally and by computational fluid dynamics simulations. Two adiabatic facilities were used: (1) a vertical channel modeling a pair of neighboring subchannels; and (2) an arrangement of four subchannels with one subchannel in the center. The first geometry was equipped with two electrical film sensors placed on opposing rod surfaces forming the subchannel gap. They recorded 2D liquid film thickness distributions on a domain of 16 × 64 measuring points each, with a time resolution of 10 kHz. In the bubbly and slug flow regime, information on the bubble size, shape, and velocity and the residual liquid film thickness underneath the bubbles were obtained. The second channel was investigated using cold neutron tomography, which allowed the measurement of average liquid film profiles showing the effect of spacer grids with vanes. The results were reproduced by large eddy simulation + volume of fluid. In the outlook, a novel nonadiabatic subchannel experiment is introduced that can be driven to steady-state dryout. A refrigerant is heated by a heavy water circuit, which allows the application of cold neutron tomography.

Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

Abstract A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 × 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

CFD analysis on mixing effects of spacer grids with different dimples and sizes for advanced fuel assemblies

Abstract The thermal hydraulic characteristics of a mixing vane grid are largely dependent on the structure of key components, such as strip, spring, dimple, weld nugget, as well as the mixing vane configuration. In this paper, several types of spacer grids with different dimple shapes are modeled under subcooled boiling conditions. Prior to the application of CFD on the dimple shape analysis, the mixing effects of spacer grids were studied. After the dimple shape analysis, the side channel effect is discussed by comparing the simulation results of a 3 × 3 and a 5 × 5 spacer grid. The two phase flow CFD models in this study are validated through simple geometry showing that the calculated void fraction is in good agreement with the experimental data. The dimple comparison result shows that varying dimple structures can result in different temperatures, lateral velocities and void fraction distributions downstream of the spacer grids. Comparison of two sizes of spacer grids demonstrate that the side channel generates different flow distribution pattern in the center channel.

A Study of Neutronics Effects of the Spacer Grids in a Typical PWR via Monte Carlo Calculation

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the “volume-preserving” streamlined heterogeneous spacer grids, but the “banded” dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.

Experiments on the Effects of a Spacer Grid in Air-Water Two-Phase Flow

Understanding the effects of spacer grids on the coolant flow through a nuclear reactor core is required for best-estimate design and analysis of the plant. The impact of a spacer grid on two-phase flows is of particular importance because the geometric effects of the grid can alter the two-phase flow structure and, consequently, the mass, momentum, and energy transfer characteristics. Therefore, a scaled separate-effects test facility is constructed to investigate the effects of a spacer grid on the hydrodynamics of air-water two-phase flow through a rod bundle. The test facility is scaled to maintain hydrodynamic and geometric similarity to single- and two-phase flows in a conventional pressurized water reactor and to facilitate detailed local measurements of two-phase flow parameters around the simulant fuel rods with a four-sensor conductivity probe. This paper presents measurements of local time-averaged two-phase flow parameters acquired upstream and downstream of the spacer grid with the conductivity probe in a representative subchannel of a 1×3 rod bundle for eight flow conditions. Characteristic features of the development of the two-phase flow parameters along the test section are discussed.

Innovative model of annular fuel design for lead-cooled fast reactors

The results of an investigation of the analytical solutions of the default – solid fuel and innovative – annular fuel approach are summarized in this paper. Innovative annular fuel concept is described in details and shows that it is a promising technology in terms of safety aspects. Liquid metal cooled reactors are considered to burn used nuclear fuel from conventional nuclear power plants or processed weapon grade plutonium and therefore, higher safety standards must be assured.During the design process seven fuel assemblies with annular fuel elements were created. The investigation was conducted in case of both commonly used fuel lattices i.e. square and hexagonal, moreover effect of spacer grids was taken into considerations. Results show that the annular fuel is superior in case of maximum fuel temperature, which is up to 757 °C lower than in default base design – with use of spacer grids. The most promising designs are hexagonal lattice with 91 fuel elements and square lattice 18×18, where the maximum temperatures are 822 °C and 732 °C, while pressure drop of 185 kPa and 128 kPa, respectively.The square lattice proved better performance according to our evaluation, it is possible to obtain very similar or smaller pressure drop than default – solid fuel – design. Hence, use of the same or even smaller coolant pumps is possible in case of annular fuel elements. Moreover, the innovative fuel also allows to overpower reactor up to 110% of nominal power, while securing all safety limits. Therefore, it is promising concept and further investigation would be interesting.

Single-phase convective heat transfer enhancement by spacer grids in a rod bundle

The spacer grids within a fuel assembly of a nuclear reactor core disrupt and re-establish the momentum and thermal boundary layers so that they enhance the local heat transfer within and downstream of the spacer grids. An experimental study in a 6×6 rod bundle has been performed to investigate the effects of spacer grids on the single-phase convective heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. The conventional correlations showed poor predictions of the heat transfer enhancement by spacer grids at low Reynolds numbers; in particular, the maximum heat transfer rate at the top end of the spacer grids was significantly overestimated. Furthermore, the conventional correlations did not properly account for the effects of the Reynolds numbers on the heat transfer enhancement. Therefore, more systematic experiments should be performed using various spacer grids with large blockage ratios at low Reynolds numbers, considering an early phase of the reflood conditions.