Rod Surface Temperature Research Articles

A heat transport benchmark problem for predicting the impact of measurements on experimental facility design

This work presents the application of the “Predictive Modeling of Coupled Multi-Physics Systems” (PM_CMPS) methodology conceived by Cacuci (2014) to a “test-section benchmark” problem in order to quantify the impact of measurements for reducing the uncertainties in the conceptual design of a proposed experimental facility aimed at investigating the thermal-hydraulics characteristics expected in the conceptual design of the G4M reactor (GEN4ENERGY, 2012). This “test-section benchmark” simulates the conditions experienced by the hottest rod within the conceptual design of the facility's test section, modeling the steady-state conduction in a rod heated internally by a cosinus-like heat source, as typically encountered in nuclear reactors, and cooled by forced convection to a surrounding coolant flowing along the rod. The PM_CMPS methodology constructs a prior distribution using all of the available computational and experimental information, by relying on the maximum entropy principle to maximize the impact of all available information and minimize the impact of ignorance. The PM_CMPS methodology then constructs the posterior distribution using Bayes’ theorem, and subsequently evaluates it via saddle-point methods to obtain explicit formulas for the predicted optimal temperature distributions and predicted optimal values for the thermal-hydraulics model parameters that characterized the test-section benchmark. In addition, the PM_CMPS methodology also yields reduced uncertainties for both the model parameters and responses.As a general rule, it is important to measure a quantity consistently with, and more accurately than, the information extant prior to the measurement. For the test-section benchmark, it is shown that the maximum coolant and rod surface temperatures responses are the most important measurable quantities. Using all of the available information, the PM_CMPS formulas yield optimally predicted best-estimate nominal model parameter values and reduced predicted standard deviations for the predicted parameters, such that: (i) the model parameters displaying the largest relative sensitivities and largest relative standard deviations are affected the most; and (ii) the model parameters with zero sensitivities remain unaffected by applying the PM_CMPS methodology. The PM_CMPS formulas also yield optimally predicted best-estimate nominal values for the model responses, along with correspondingly reduced predicted standard deviations.Assimilating within the PM_CMPS methodology an accurate measurement of the benchmark's maximum coolant temperature has the following impacts on the quantity that has been measured (i.e., on the maximum coolant temperature): (i) the nominal value of the predicted maximum coolant temperature is closer to the more “accurate” experimental value; (ii) the originally computed standard deviation is decreased significantly. In addition, the measurement of the maximum coolant temperature impacts not only the quantity being measured, but also impacts the other responses of interest, e.g., the maximum temperature in the rod and the maximum rod surface temperature. In all cases, the predicted standard deviations were decreased substantially, indicating that the predicted values are more accurate than the originally computed ones.The PM_CMPS formulas were also used to predict the impact of the simultaneous assimilation of the maximum coolant and the maximum rod's surface temperature measurements. As expected in view of the fact that the PM_CMPS methodology uses Bayes’ theorem to combine information, the assimilation of two accurate experiments reduces the predicted standard deviations of the maximum rod, rod surface and coolant temperatures even further. The test-section benchmark and methodology presented in this work can serve as a paradigm for quantifying the impact of experimental measurements in the analysis, design and/or operation of other thermal-hydraulics experimental facilities.

가열탱크 내부 스크류 회전속도에 따른 전기보일러의 열성능에 관한 연구

Abstract This study was aimed at the heating tank with a screw-rotation device for improving the thermalefficiency of electric boiler. In the proposed system, analysis items were the heater rod surfacetemperature variation, reaching time for set temperature and thermal efficiency. The followingconclusions are obtained from this experimental study. (1) When screw speed increases, the timereaching for set temperature tended to be shorter. (2) When the rotation speed becomes 300 rpm, thesurface temperature difference between the right and left heater rod decreases by 49%, from 19.7 o C to9.7 o C in average. (3) When the rotation speed is over 250 rpm, proposed heating tank structure appearedto be effective in terms of thermal efficiency. Thermal efficiency with the rotation speed 300 rpm isimproved by 3.8% compared to the case of rotation speed 0 rpm.Keywords: Electric boiler(전기보일러 ), Heater rod(히터봉 ), Heating tank(가열탱크 ), Screw(스크류 ),rpm(회전수 ), Thermal Efficiency(열효율 ) 기호설명 G : 유량 [kg/s]C : 유체의 비열 [kJ/kg

Experimental study on the convective heat transfer enhancement in single-phase steam flow by a support grid

Single-phase flow occurs in the fuel rod bundle of a pressurized water reactor, during the normal operation period or at the early stage of the reflood phase in a loss-of-coolant accident scenario. In the former period, the flow is single-phase water flow, but in the latter case, the flow is single-phase steam flow. Support grids are required to maintain a proper geometry configuration of fuel rods within nuclear fuel assemblies. This study was conducted to elucidate the effects of support grids on the convective heat transfer in single-phase steam flow. Experiments were made in a square array 2×2 rod bundle. The four electrically-heating rods were maintained by support grids with mixing vanes creating a swirl flow. Two types of support grids were considered in this study. The two types are geometrically similar except the blockage ratio by different mixing vane angles. For all test runs, 2kW power was supplied to each rod. The working fluid was superheated steam with Re = 2,301–39,594. The axial profile of the rod surface temperatures was measured, and the convective heat transfer enhancement by the presence of the support grids was examined. The peak heat transfer enhancement was a function of not only the blockage ratio but also the Reynolds number. Given the same blockage ratio, the heat transfer enhancement was sensitive to the Reynolds number in laminar flow, whereas it was nearly independent of the Reynolds number in turbulent flow.

Experimental investigations of single and two-phase flow in a heated rod bundle

Abstract An experimental facility for the study of boiling flows in a 3 × 3 rod bundle geometry was setup. The bundle resembles in essential geometrical parts the geometry in a pressurized water reactor fuel element. The facility is operated with a refrigerant fluid. Beside standard instrumentation for temperature, pressure and flow rate we employed particle image velocimetry for single phase flow studies, gamma ray densitometry for integral gas fraction measurement sand ultrafast X-ray tomography for the study of the void dynamics in the cross-section. Moreover extensive thermo-instrumentation allows axial rod surface temperature measurements for the central heated rod. First results will be discussed in this article.

An experimental assessment of cooling of a 54-rod bundle by in-bundle injection

The performance of an in-bundle coolant injection system for the quenching of dry heated rods has been experimentally investigated. The rod bundle contains 54 fuel rods of 11.2mm diameter, 3700mm long, arranged in three concentric rings with a central coolant supply tube. The coolant tube supplies the coolant in the form of jets through a series of circumferential holes at different axial levels inside the rod bundle. Visualization during cold state injection tests ensures that the liquid spray can reach different levels of all the rods above a certain flow rate of water through the coolant tube. Extensive cooling experiments were done to assess the suitability of the proposed scheme of in-bundle coolant injection. Time–temperature curves have been derived from rods at different locations, from different heights of the rods, over a range of coolant flow rate as well as for different rod temperatures. The effect of the presence of the spacers on local cooling has also been investigated. The cooling curves follow a general trend of a rapid temperature drop up to almost 100°C of the rod surface temperature irrespective of the operating parameters and the location of the rod. Thereafter, the temperature falls slowly reaching the coolant temperature almost asymptotically. Moreover, the second phase of cooling is often marked by temperature fluctuations of random nature. It was also observed that a large volume of steam generates during cooling and comes out through the top of the test section expelling a significant amount of the coolant. This counter current flow of steam–water mixture has an adverse effect on cooling. Further, comparison of the cooling curves obtained from the present experiment with those obtained from quenching of single rod by top flooding brings out the similarities and differences of the two processes.

Experimental Study of Quenching Process During Bottom Reflooding Using “Queen” Test Section

Phenomenon of quenching of hot fuels in core during bottom reflooding following loca event is investigated in order to understand the performance cooling process. the study is conducted experimentally using queen test section which allow study of rod surface temperature histories based on which the heat fluxes are estimated. the visual observation is also done to study the boiling regimes. the test variables are initial rod temperature, i.e. 400 o c, 500 o c and 600 o c, and coolant flow rate, i.e. 0,01kg/s, 0.02 kg/s and 0.04 kg/s with constant water inlet temperature of 30 o c. the results shows different heat transfer regimes such as film boiling, transition boiling, nucleate boiling and convective single phase heat transfer regimes. for specified initial rod temperature, the higher flow rate provides high rewet velocity and higher maximum heat flux, then quenching process is more effective.

Numerical Analysis on Thermal-Hydraulics of Supercritical Water Flowing in a Tight-Lattice Fuel Bundle

To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of a supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D which has been enhanced by Japan Atomic Energy Agency. In ACE-3D, a two-phase flow turbulent model based on the k- model was adopted. In this calculation, a one-twelfth model is adopted as the computational domain and takes advantage of symmetry. As boundary conditions, mass velocity, inlet enthalpy and power per rod are to be the same as the steady state condition of the Super Fast Reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. The effect of grid spacers is taken into account in the analysis. Calculated rod surface temperatures take peak values near the top of the rods. The maximum cladding surface temperature (MCST) is observed at the position facing the narrowest gap on the center rod near the outlet and the value is 901 K (628 °C). It was confirmed that the predicted MCST satisfies the thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650 °C.

ICONE19-43300 Fuel Rod Surface Temperature Distributions in Liquid Metal Sub-channel Flows

A numerical study on fuel rod surface temperature distributions in liquid metal cooled sub-channel flow is presented in this paper. Special interests are focused on the effects of misaligned central fuel rod in typical fuel rod cluster geometries. Numerical models based on FLUENT, a Computational Fluid Dynamics (CFD) commercial software, are developed to model steady state fuel thermal conductions, fully-developed turbulent convective flows, and local flow mixing. Numerical results show that fuel rod surface temperatures are higher for surfaces facing the narrow gap of the sub-channel. The temperature differences are mainly dependent on the fuel pitch-to-diameter ratio, flow patterns, and other flow physical properties. Situations are worse for cases where misaligned fuel rods exist in tightly packed fuel rod clusters. Even for a small amount of rod center deviation, such temperature differences can be as high as 54 K around a fuel rod whose diameter is only 4.7 mm. The situations could be very serious and might generate unnecessary thermal stresses and cause fuel failure.

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes. In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.

Modifications to COBRA-TF to model dispersed flow film boiling with two flow, four field Eulerian–Eulerian approach – Part 1

In the case of a postulated loss of coolant accident (LOCA) in a nuclear reactor, an accurate prediction of clad temperature is needed to determine the safety margins. During the reflood phase of the LOCA, when the local void fraction is greater than 80% with the wall temperature above minimum film boiling temperature (Tmin), the heat transfer process is dispersed flow film boiling (DFFB). This study has been performed to model DFFB in the reflood phase of a LOCA in a pressurized water reactor (PWR) rod bundle. The COBRA-TF computer code is utilized, since it has a detailed reflood package which takes into account the effect of spacer grids on the local heat transfer. The COBRA-TF code has also been improved to include a four field Eulerian–Eulerian modeling for the two-phase dispersed flow film boiling heat transfer regime. The modifications include adding a small droplet field to COBRA-TF as the fourth field. In addition, the spacer grid models of COBRA-TF have been revised and modified. In the first part of the paper, the results of the code predictions are presented by comparing the experimental data from rod bundle heat transfer (RBHT) experiments with the results of code simulations performed with original and modified code. Measurements and calculations for the heater rod, vapor temperatures and quench front progression have been compared and the results are described in detail. The results of the analysis performed with the modified code indicate the improvement in code predictions for the rod surface temperature, vapor temperature and quench front behavior. The results also indicate the need for improvement in the entrainment and interfacial drag models for the drop fields. The effects of spacer grids on the heat transfer, the models improved and developed for spacer grids and the results of the code calculations with these models are described in the part 2 of the paper.

Two forms of attachment of an atmospheric-pressure direct-current arc in argon to a thermionic cathode

Comparative investigation of two forms of attachment of a dc (20 < I < 200 A) atmospheric-pressure arc in argon to a thermionic cathode made of pure tungsten is carried out. The current-voltage characteristics of the arc, axial distribution of the cathode rod surface temperature (except for the site of arc attachment), and plasma temperature axial distribution in the cathode region are measured, and the current density on the cathode surface is estimated. The measurements of current-voltage curves shows that the voltages of the arc with different forms of cathode attachment differ distinctly (but not too much) one from another, the curves for two modes iontersect. This confirms the results of theoretical analysis carried out earlier by M.S. Benilov who has showed that an existence of different forms of attachment is associated with the presence of branching points in a solution to the cathode heat balance problem. The point of intersection should be viewed as one such point. Optical measurements disclose that the temperature and its distribution over the cathode rod surface differ greatly for the two forms of attachment considered. The plasma temperature in the cathode region of the contracted attachment far exceeds that in the diffuse attachment, exceeding 3 eV in the immediate vicinity of the cathode surface. The maximal temperature of the plasma in the contracted attachment does not depend on the current. Analysis of erosion prints shows that the current density on the cathode does not depend on the current for both forms of cathodic attachment. For the contracted attachment, the current density is roughly four times higher (∼104 A/cm2) than for the diffuse form. The experimental data are in good agreement with present-day calculations of the cathode plasma parameters and temperature conditions.

Thermal-hydraulic performance of primary system of RBMK in case of accidents

Within the framework of an European Commission sponsored activity, an overall assessment of the deterministic safety technology of the ‘post-Chernobyl modernized’ RBMK has been completed. The accident analysis, limited to the area of design basis accident, constituted the key subject for the study; therefore, the notorious Chernobyl unit 4 event was outside the scope of the investigation. The present paper deals with the evaluation of the accident performance of the primary system of the RBMK. Ignalina-2 and Smolensk-3 nuclear power plants (NPP) are considered. The documented activity includes four main parts: (a) description of key features of the primary system for the two NPP; (b) identification and characterization of relevant accident scenarios part of the (DBA) area; (c) key features and qualification level of adopted computation tools primarily including codes and input decks; (d) results from the analysis of selected accident scenarios. In the first part of the paper the layout complexity of the RBMK systems, including the emergency core cooling and related logics is pointed out together with main differences between Smolensk-3 and Ignalina-2 units. Accident scenarios suitable for the analysis, second part of the paper, are derived from recently issued International Atomic Energy Agency (IAEA) reports, from the available Safety Analysis Reports (SAR) of existing RBMK NPP and from the list of bounding accidents derived from the already mentioned EC Project. The features of Korsar and Relap5 thermal-hydraulic system codes developed in Russia and US, respectively, are outlined in the third part of the paper, giving emphasis to the qualification level in the range of parameters of interest to RBMK systems. The adopted nodalisations resembling Ignalina-2 and Smolensk-3 NPP are also described including the demonstration of the qualification level. The results, fourth part of the paper, basically show the overall RBMK system robustness within the DBA area, when the nominal operating conditions constitute the starting conditions for the accident scenarios. However, parallel channel instabilities may trigger dry-out phenomena and consequent rod surface temperature excursion in case of special accident types.

Analysis of the FIST integral tests 4DBA1, 6SB2C and T1QUV with TRAC-BFl/v2001.2

As part of the assessment of the frozen version of the PSU TRAC-BFl/v2001.2 and its qualification as a LB-LOCA and SB-LOCA code, in this work, we shall outline the comparisons between measurements and code predictions for three FIST tests: The LB-LOCA test 4DBA1, the SB-LOCA test 6SB2C and the failure to maintain water level test T1QUV. We shall study the effect of the number of axial levels in the active core as well as (in the case of the SB-LOCA test 6SB2C) the effect of the timing of the activation of the reflooding options/heat transfer package on the code predictions. Furthermore, we shall show that by using the upwinding option of some terms of the three-dimensional momentum equations, severe mass–error problems appearing in the analysis of the test T1QUV can be resolved. Generally, we shall show that although there are some differences between measurements and predictions, TRAC-BF1 captures quite well the overall behaviour of the LB-LOCA transient (depending on the number of axial nodes in the core) but underpredicts the rod surface temperatures of the SB-LOCA test 6SB2C.

Study of the thermal plasma etching at atmospheric pressure on silica rods

Etching of SiO2 rods has been obtained with a dc torch with argon as the process gas in an air environment at atmospheric pressure; the high temperature of the plasma jet causes vaporization of the exposed area. The apparatus and torch operative parameters have been set up to obtain a depth etch rate of up to 0.6 mm min−1 corresponding to 0.826 g min−1.An enthalpy probe has been employed to monitor the plasma conditions before the thermal plasma etching process and from the experimental etch rate a surface rod temperature of Tsur = 2057 K has been derived. Etching has been obtained with uniformity over the entire exposed area with peak to peak differences below 1%.The plasma to rod heat transfer has been simulated using a commercial CFD code Fluent©. The model consists of a non-steady two-dimensional simulation for a compressible turbulent fluid, with an adapted grid calculation. Boundary conditions have been set out using the enthalpy probe plasma jet map. In particular for the flow, the continuity equation for chemical species, the species transport equations (Fick's Law), the momentum conservation equations, the energy equation and the turbulent kinetic energy have been solved. In the solid regions the convective energy transfer due to the rotational motion of the rod, the conduction and radiation energies have been considered. The heat transfer between the fluid and the silica walls (continuity temperature between the fluid and the solid) accounts for the energy loss due to surface radiation (qr) and vaporization (qv) obtained from experimental data.A rod surface temperature of 2057 K obtained from experimental etch rate data agrees well with the calculated temperature obtained using Fluent© (Tsim = 2010 K).

Boiling Heat Transfer Characteristics with Highly Wettable Heated Surface under Forced Convection Conditions

Under forced convection and atmospheric pressure conditions, heat transfer characteristics were studied using an annulus channel of a heater rod with highly-wettable surface. Improvement of phase change heat transfer under boiling and condensation process requires that either liquid heated or vapor cooled can contact a heat transfer surface effectively, but either a vapor layer or a liquid film can be formed repeatedly on the surface during boiling and condensation processes. As oxide semiconductor materials are known to be highly wettable, TiO 2 was coated on a heater rod surface. The Leidenfrost temperature for the TiO 2 coated surface was higher than that for the uncoated surface. Under mass flux conditions from 600 to 1 700 kg/m 2 s, the heater rod surface temperature at variable power conditions was measured. A higher heat flux at the point where the forced convection vaporization started and a larger critical heat flux were observed for the TiO 2 coated surface than that for the uncoated surface.

Development, assessment and application of TRAC-BF1/v2001.2 for beyond design basis BWR LOCA transients

In preparation for the possible transition to risk based licensing, it is increasingly important to demonstrate the applicability of Best Estimate codes to more extreme conditions corresponding for example to limited equipment availability. With this idea in mind, we have reviewed the application of TRAC-BF1 to Large and Small Break LOCAs. In this context this work describes the assessment and applications of the Penn State University (PSU) version 2001.2 of TRAC-BF1 with all the PSI updates, to the analysis of hypothetical Large Break (LB) LOCAs in a BWR/4 with postulated limited ECC availability. Since in contrast to a LB-LOCA in a BWR with full ECC availability, the rod surface temperatures reach relatively high values, additional assessment of the code under such conditions is required. Hence, after analyzing bottom flooding separate-effect experiments in two different heater rod bundles and a TLTA LB-LOCA test, we shall present and discuss the results of the analysis of a LB-LOCA with highly restricted Emergency Core Cooling flow in a BWR/4. In this context, we shall also assess different descretization of some terms of the 3D momentum equations, which was found to be important in the analysis of BWR Small Break LOCAs.

Effects of dispersed flow interfacial area/shear modeling and node size in the core on TRAC-BFl rod surface temperature (rst) histories of separate-effect and system calculations

As part of the task of assessing and qualify the frozen version of PennState University TRAC-BF1/v2001.2 as a BWR LOCA code, in this work, we shall study the extent to which changes in physical models in the code can affect the predicted rod surface temperature histories during separate-effect tests and system calculations. In particular, we shall concentrate our attention to the reflooding phase and after first demonstrating by two examples the necessity of using the code with the PSI updates, we shall implement in the code a dispersed flow interfacial shear model which is consistent with the corresponding interfacial heat transfer model, i.e. it utilizes the interfacial area per unit volume for both closure relations and not only for the interfacial heat transfer. For this, we shall implement in the code the dispersed flow interfacial shear model of RELAP5/MOD3; furthermore, we shall assume that the average droplet diameter is a function of the distance from the bottom quench front. Having done this, we shall analyze one FLECHT bottom flooding test with both code versions and show that the differences between measured and predicted PCTs are for both cases approximately the same. Finally, we shall study the effect of the number of nodes (8, 24 and 48) in the core by analyzing a hypothetical LB-LOCA in a BWR with limited ECC availability both with the frozen and the modified versions of the code and we shall show that when the frozen version of the code is used, the predicted peak clad temperatures (PCTs) are 120 K lower if 48 rather than when 8 nodes are used in the active core.

Development of a Transient Boiling Transition Analysis Method Based on a Film Flow Model

A new single-channel, transient boiling transition (BT) prediction method based on a film flow model has been developed for a core thermal-hydraulic code. This method could predict onset and location of dryout and rewetting under transient conditions mechanically based on the dryout criterion and with consideration of the spacer effect. The developed method was applied to analysis of steady-state and transient BT experiments using BWR fuel bundle mockups for verification. Comparisons between calculated results and experimental data showed that the developed method tended to predict occurrence of rewetting earlier, however, onset time of BT and maximum rod surface temperature were well predicted within 0.6 s and 20°C, respectively. Moreover, it was confirmed that consideration of the spacer effect on liquid film flow rate on the rod surface was required to predict dryout phenomena accurately under transient conditions.

2446 非定常サブチャンネル解析コード NASCA の開発・改良 : (4) クロスフローモデルの検証

A cross flow model was developed for the three-fluid subchannel analysis code NASCA. The adequacy of the model was confirmed by comparing with calculated and measured subchannel mass flux and quality distributions obtained by 3×3 rod bundle experiment. To evaluate the applicability of the code to the post BT condition, the heater rod surface temperature in 8×8 mock-up bundle at the post BT condition was analyzed.

Metal flow in the deformation gap at primary swaging

Taking into consideration technological conditions typical for the real swaging process an analysis of metal flow in the deformation gap has been performed using finite elements method, and assuming flatness of sections. The stress and strain distributions, as well as deformation rates and temperature distributions have been calculated for swaging according to sizing schemes: circle → circle, circle → square and circle → octagon. The obtained results of local strain distribution calculations have been verified experimentally. The local strains were measured using the method of combined rods being swaged in practical conditions. Moreover, the experimental verification of the calculated temperature distributions in the swaged material cross-section has been carried out by comparison with the results of pyrometric measurements of the rod surface temperature. The calculated strain distributions can be used to design efficiently the swaging process because the knowledge of the strain distribution indicates some potential crack nucleation areas and the respective level of acceptable strain limits.