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This paper presents the results of a numerical study of the effects of swirling flow in coolant jets on film cooling performance. Some combined-hole designs with swirling coolant flow entering the delivery hole are proposed and analyzed. Adiabatic film cooling effectiveness values for cases with various blowing ratios are compared. Detailed flow structures and underlying mechanisms are discussed. The results show that film cooling effectiveness is improved with jet swirl at high blowing ratios, and that swirl strength has significant influence on film cooling performance. Combined-hole designs can further improve film cooling performance using swirling jets due to mixing of coolant flows and interaction of vortices. The largest improvements of area-averaged film cooling effectiveness for a single-hole swirl case and a combined-hole swirl case over corresponding non-swirling case results are 157% and 173%, respectively.

Objective of the vehicle thermal management system is to achieve precise coolant temperature control that ensures operation of the combustion engine in the region with the maximum thermal efficiency. In this study it is achieved by a combined effect of coolant temperature control algorithm that is operating an advanced electronic valve with one input and three outputs that balances distribution of the coolant flow between the heat source and sink components. The objective is to present an application of a nonlinear model predictive control (NMPC) approach for coolant temperature control. It is a simulation study to quantify benefits of precise coolant temperature control with the possible inclusion of preview information.

The method of computer analysis of system parameters at formation of protective coating on the shell of fuel element of fast reactors is presented, which can serve as a prerequisite for development of cognitive architecture of control of nuclear power engineering processes by control of composition of gas phase in reactor core, containment, other tanks and making decisions on results of process parameters deviation from stationary function. The work was carried out to justify the prospects of using a liquid metal sublayer (LMS) instead of a gas sublayer in fuel elements, which allows, due to the high thermal conductivity of the liquid metal, to increase the width of the gap between the fuel and the shell and, therefore, to increase the fuel element life limited by the contact of the fuel with the shell. At the same time, safety in case of UTOP (uncontrolled increase of power), ULOF (termination of coolant flow through the reactor) accidents is increased due to significant decrease of temperature in the fuel center. However, the possibility of using LMS is limited by the substantial mass transfer - corrosion dissolution in the lead of the components of the fuel element shell in the hot zone and precipitation in the cold zone. In this article, the basis for computer analysis was studies on the absorption of gas (nitrogen) when forming a protective ceramic coating of zirconium nitride on the surface of the shell of a pipe-in-pipe type fuel element simulator to suppress the corrosion process from eutectic alloy lead-magnesium-zirconium. Abnormal intense decrease of nitrogen pressure (p, kPa) in expansion tank with Pb-Mg-Zr alloy at increased temperature (about 950 K) according to linear kinetic law (t, c), connected with its interaction with system components, was revealed: p, kPa=−0.0002⋅t−77.867 R2=0.942. Method of diagnostics of system parameters at formation of anticorrosive coating by nitrogen absorption is developed, empirical dependencies of process flow in wide range of temperature are built.

Carbon fiber reinforced plastic (CFRP) composites are the first choice for high-end equipment in fields of aviation and aerospace pursuing weight reduction. However, producing high quality holes in drilling of CFRP remains a long-standing challenge mainly due to its weak interlaminate bonding and sensitivity to temperature. This article proposed a novel cooling process for drilling of CFRP, deploying reverse air suctioning, to reduce drill-exit damages. The concept of the proposed cooling process is to change the coolant flow direction, with the benefits of reducing the force of the coolant acting on the drill exit and increasing the constraint of the drill-exit material. Comparative experiments are conducted under dry and reverse air suctioning conditions. Results show that the proposed cooling process can significantly reduce the drilling temperature, below Tg. Despite the cooling process has led to increased thrust force, the drill damages have been reduced as compared with the dry drilling, and the potential damage suppression mechanisms have been discussed.

The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a subchannel void distribution correction model for the two-stage core analysis method using an assembly-based thermal-hydraulics calculation in the core analysis stage. The model assumes two kinds of subchannel void distribution gradients along with the two diagonal lines in the horizontal cross section of a BWR fuel assembly. The model appends and tabulates the difference of the subchannel perturbation condition from the base condition in the lattice physics, and evaluates the tilts within the 2D lattice physics scheme, and couples those results with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. The developed model is evaluated using a heterogeneous and a small core problem. The model gives a better power distribution compared with that of the authors’ previous model. As a result, the model can incorporate the subchannel effect into the current two-stage core calculation method.

This paper investigates the thermal efficiency of converging microchannels with superhydrophobic walls under various coolant flow rates. To enhance the thermal performance of microchannels, geometric optimization is performed for different values of pumping power, Ω, ranging from 0.01 up to 10 Watts. For this purpose, Navier-Stokes and energy equations with slip boundary conditions are solved for various design points by the Finite Volume Method. Response functions for pumping power and thermal resistance are then constructed. Finally, desirability function approach is employed to minimize the total thermal resistance for a fixed pumping power to give the optimum design set. It is found that for each value of Ω, by the use of superhydrophobic walls, optimal number of channels (N), width-tapered ratio (Λz) and height-tapered ratio (Λy) increase, whereas the optimal width ratio (β) decreases. It is also found that with an increase in pumping power, the optimal N increases, whereas the optimal values of β, Λz, and Λy decrease. Finally, it is shown that the thermal performance enhancement by the use of superhydrophobic walls is only achieved for a low pumping power (Ω 5 Watts), thermal performance is deteriorated by their use.

In order to analyze the influence of four parameters on the cooling effect of a liquid-cooled battery thermal management model, a L16(44) orthogonal array is selected to design sixteen models to perform and quantify the parametric to identify the main and secondary factors, then the optimal combination model is found. The results show that the number of the pipe has most obvious effect on the average temperature of the cooling plate in the four parameters, and velocity of coolant flow is second, pipe height has the minimal effect. In terms of temperature uniformity, amount of pipe and velocity of coolant have similar effects which are both main factors, pipe width and pipe height have similar effects which are both secondary factors. The optimal case by this orthogonal analysis can meet various cooling conditions.

This paper investigates effects of geometrical and coolant flow parameters for a three-row compound angle hole design over a flat plate. These included β = +45° and −45°, in-line and staggered arrangement, hole spacing (4d, 6d and 8d), five blowing ratios (0.5–1.5) and three density ratios (1.0, 1.5, 2.0). Spanwise averaged effectiveness plots and contours were developed using Pressure-Sensitive Paint (PSP) technique and cross comparisons were carried out. The parametric results obtained from experiment generally agreed with widely accepted trends: the cascade effect of coolant jets for multi-hole design increases film effectiveness especially at large x/d, increasing density ratio increases effectiveness particularly at higher blowing ratio and increasing hole spacing has a detrimental effect on film cooling. However, some interesting observations are discovered including: staggered arrangement is not always the most optimum design (β = +45°, −45°, −45° with staggered 2nd and 3rd row reported the lowest effectiveness for all blowing ratios and density ratios); in-line arrangement of holes with opposing orientation angles (β = +45°, −45°, −45°) yields better effectiveness; blowing ratio effect is strongly dependent on geometric design. Moreover, hole spacing effect is closely related to neighboring coolant jet coalescence and interaction with mainstream flow. Currently, little data is available on three-row design in open literature; this study provides systematic, baseline information for future studies.

Three different kinds of coolant chamber configuration for film cooling are proposed to develop the swirling coolant flow at blowing ratios ranging from 0.5 to 2.0. The results show that the difference of film cooling effectiveness for three kinds of coolant chamber configuration is little at low blowing ratio, but the advantage of swirling film cooling becomes obviously with the increase of blowing ratio. When the blowing ratio is 2.0, the jet momentum of original coolant chamber configuration is large and uniform, which leads to the lowest cooling effectiveness due to the formation of a strong kidney vortex. The first coolant chamber configuration has a low jet momentum region at upstream of the film hole, the coolant in this region interacts with high temperature mainstream and bypasses the large jet momentum coolant to attach cooling surface at downstream. The second coolant chamber configuration is sprayed with the structure of unidirectional vortex, which forms a vortex pressing on other vortex, making the coolant in pressed vortex attach surface better, producing the best coverage and the higher film cooling effectiveness.

Experimental tests are performed in the current study to explore the overall cooling performance of a combined slot injection/effusion array cooling structure. The pure slot injection and pure effusion array cooling schemes are taken into consideration for the comparison purpose. In addition, numerical simulations on the adiabatic and overall cooling effectiveness are also performed to illustrate the conjugated effects, such as internal convective heat transfer inside the coolant flow channel, thermal conduction inside the perforated plate, and internal cooling in the multiple holes, on the film cooling performance. The combined cooling scheme overcomes respective drawback of slot injection cooling in its decayed stage and effusion array cooling in its developing stage, thus producing a more ideal cooling performance over the entire protected surface under the same coolant usage. Due to the conjugated effects, the influence of adding slot injection on improving overall cooling effectiveness is significantly weaker than that on the adiabatic cooling effectiveness. Related to the pure slot injection or effusion array cooling scheme, the benefit of combined cooling scheme on the film cooling improvement is dependent on the blowing ratio. Under a combination of Msl = 0.5 and Meff = 0.4, it produces more obvious improvement on overall cooling effectiveness. While under a combination of Msl = 0.5 and Meff = 1.2, the overall cooling effectiveness in the front zone of effusion array is a little less than the pure injection scheme.

A series of numerical computations are conducted to investigate the film cooling from a row of round-to-slot holes on a flat surface. Particular attention is paid to explore effect of slot width (0.2 ≤ s/d ≤ 0.5) on the film cooling performances of round-to-slot holes with a fixed slot length of l/d = 2.0 under typical blowing ratios ranged from 1.0 to 2.0. Besides, several experimental tests are also performed to obtain the film cooling effectiveness and discharge coefficient of the round-to-slot holes with specific geometries. The results show that the coolant flow turning towards both lateral sides due to the transitional flow passage of film-hole is dominated inside the round-to-slot hole. Owing to the specific interaction mechanism between the coolant jet issued from round-to-slot holes and the primary flow, the hot primary flow is mainly entrained into the coolant layer from the film-hole centerline. In general, the diffusing round-to-slot hole shows its favorable feature on the film cooling enhancement near the film hole. However, beyond a certain distance, the converging round-to-slot hole achieves a little higher laterally-averaged film cooling effectiveness than the diffusing round-to-slot hole. The converging round-to-slot hole with s/d = 0.2 produces higher laterally-averaged heat transfer coefficient than the cylindrical hole. While the diffusing round-to-slot hole produces less laterally-averaged convective heat transfer coefficient than the cylindrical hole beyond x / d = 10. The discharge coefficient decreases significantly as the slot width reduces.

The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL 1 ) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal operating conditions (NOC). Due to fissioning, Xe and Kr gases are produced, which contribute to the internal gas pressure within an element. The increase in gas pressure may jeopardize the fuel element integrity (sheath rupture). On the other hand, fission gas is produced as a function of neutron flux/power and diffuses to the grain boundaries and fuel-pellet interface as a function of fuel temperature. Therefore, it is important to know the fuel temperatures and distributions within the fuel.

This paper presents the analysis and controlled modification of the coolant flow in tapping processes by means of Computational Fluid Dynamics (CFD). First, a conventional straight flute tapping tool was analyzed and the results of the CFD simulation show, that the cutting edges are not sufficiently supplied with coolant. Therefore, the design of the internal coolant channels was modified based on these simulation results. To validate the CFD simulation, experimental tests were performed, using an optimized tool. The applied modifications lead to a reduction of the tool wear and an increase of the tool's performance of about 36% was achieved.

The laminated cooling configuration with integrated impingement, rib-roughed and film cooling can result in high overall cooling effectiveness and is believed to be a promising cooling configuration for the next generation of advanced gas turbines. Typical laminated cooling configuration usually employs discrete film holes which have the similar hole topology along the streamwise and the spanwise direction. Because the functioning distance of film cooling along the streamwise direction is much longer than that along the spanwise direction. So the discrete film cooling cannot perform well in the spanwise direction. A novel laminated cooling layout in which the film outlet spacing in the spanwise direction is much shorter than that in the typical layout was presented in this paper. The angled film holes are also employed to enhance the film cooling effectiveness and to reduce the coolant flow resistance. Conjugate heat transfer computations were performed to assess the overall cooling effectiveness and the coolants resistance. The computational results show that by employing angled film hole on laminated configurations with novel layout, the area-averaged overall cooling effectiveness was considerably improved.

In this study, the use of the CANFLEX fuel bundle in the CANDU6 reactor has been investigated with high-fidelity Monte Carlo simulations in view of major safety parameters. The CANFLEX fuel bundle is designed such that it has a smaller linear power and flatter bundle power distributions than those of the standard design. This results in a significantly lower average fuel temperature. In addition, with slightly larger coolant flow area than the standard design, the CANFLEX bundle is expected to improve fuel temperature coefficient (FTC) and coolant temperature coefficient (CTC) during normal operation. This study presents 2-D neutronic lattice evaluations of these safety parameters of the CANDU6 reactor loaded with CANFLEX fuel bundle against with the standard fuel design. To adequately model the equilibrium CANDU6 core, a number of different 2-D fuel bundle lattices were simulated on a huge number of neutron histories. For a reasonably accurate evaluation of the FTC, thermal motion of 238 U was considered with the Doppler Broadened Rejection Correction (DBRC) scheme. Based on these numerical evaluations, it can reasonably be concluded that the CANFLEX fuel design can possibly enhance safety of the CANDU6 reactor as its PCR is clearly negative for a wide range of reactor power. All depletion calculations were completed using the Monte Carlo Serpent2 code with ENDF/B-VII.0 library.

To analyze the typical characteristics of accidents occurring during the startup of a supercritical water reactor (SCWR), a comprehensive SCWR system model was established with the SCTRAN analysis code, based on models of the high-performance light-water reactor (HPLWR) steam cycle, SCWR circulation startup loop, passive safety systems, and CSR1000 core parameters. According to the sequence of the loss of coolant flow accident (LOFA), without a safety system and trigger signal of the safety system under the rated condition, a new trigger signal during the startup was designed and then used to assess “LOFA,” “uncontrolled CR withdrawal accident,” and “loss of coolant accident (LOCA)” during the startup process. The results show that the new trigger signal can ensure the effective and timely response of the safety system, as well as reactor safety during the startup process. Moreover, the maximum cladding surface temperature (MCST) of the reactor peaks (850 °C, well below the safety criterion of 1260 °C) at the end of the fourth stage of the startup process in the case of LOCA.

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17 × 17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12 × 12 and 13 × 13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13 × 13 and 12 × 12 arrays) are several times larger than that of the referenced solid fuel (17 × 17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.

To develop tritium breeding technology for a Fusion Reactor, Korea has participated in the Test Blanket Module (TBM) program in ITER. The He Cooled Ceramic Reflector (HCCR) TBM consists of functional components such as First Wall (FW), Breeding Zone (BZ), Side Wall (SW), and Back Manifold (BM) and it was designed based on the separate analyses for each component in 2012. Based on the each component analysis model, the integrated model is prepared and thermal-hydraulic analysis for the HCCR TBM is performed in the present study. The coolant flow distribution from BM and SW to FW and BZ, and resulted structure temperatures are obtained with the integrated model. It is found that the non-uniform flow rate occurs at FW and BZ and it causes excess of the design limit (550 °C) at some region. Based on this integrated model, we will perform the design optimization for obtaining uniform flow distribution for satisfying the design requirements.

CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the result of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analyses, it was found that the MCST does not exceed 1260 °C, meets the design safety requirements. The 2nd MCST are higher than 1st MCST. The isolation of main steam line has the least safety margin.

Designing cooling channels for the thermoplastic injection process is a very important step in mold design. A conformal cooling channel can significantly improve the efficiency and the quality of production in plastic injection molding. This paper introduces an approach to generate spiral channels for conformal cooling. The cooling channels designed by our algorithms has very simple connectivity and can achieve effective conformal cooling for the models with complex shapes. The axial curves of cooling channels are constructed on a free-form surface conformal to the mold surface. With the help of boundary-distance maps, algorithms are investigated to generate evenly distributed spiral curves on the surface. The cooling channels derived from these spiral curves are conformal to the plastic part and introduce nearly no reduction at the rate of coolant flow. Therefore, the channels are able to achieve uniform mold cooling. Moreover, by having simple connectivity, these spiral channels can be fabricated by copper duct bending instead of expensive selective laser sintering.

The present work has been carried out in the framework of the European FP7 project SEARCH, in which the MYRRHA demonstrator reactor is designed to be able to operate both in ADS mode and in critical mode using lead–bismuth eutectic (LBE) as primary coolant. According to the project task definition, the pin failure and fuel dispersion scenarios in severe accidents had to be extensively studied for reactor safety analysis. In this paper, the unprotected severe transients analyses for the MYRRHA–FASTEF critical core were performed using the SIMMER-III code. The aim of the current work is to obtain a deeper understanding of core material redistribution processes after pin damage. Since the fuel has almost the same density as the coolant, its pellets, chunks and particles will essentially be carried by the coolant flow, thus moving upwards out of the core and redistributing into the upper pool region and peripheral structures. Starting the simulations from the steady state configuration, relevant parameters reflect good agreement with the design operational conditions. For the transients, the most severe accident scenario proposed, that may possibly lead to pin failure and furthermore core damage, is the unprotected blockage accident (UBA). The calculation results show that after pin failure, the mobile fuel starts to re-distribute. In the meantime, the reactor stabilizes to shut-down status because of the fuel loss. Our results show that the blockage propagation is impossible thanks to the gap between fuel subassemblies.

Limited results of studies on coolant channel blockage accidents in a prismatic-gas cooled reactor core are available, although coolant channel blockage of the fuel block in a prismatic gas-cooled reactor has been regarded as a highly important phenomenon. In this study, therefore, intensive and extensive thermal analyses were carried out to enhance the understanding and knowledge about the thermal performance of multi-hole type prismatic fuel blocks under coolant channel blockage accidents. A series of thermo-fluid calculations were performed using the CORONA code. In contrast to existing works, a full core model was used to consider an accurate coolant flow redistribution after the blockage in this work. The results of the CORONA calculations show that the single channel blockage accidents of PMR200 are not a safety concern regardless of the location. However, the maximum fuel temperature exceeds the safety limit in the case of more than ∼10 channel blockage at the same region of the fuel column. It was also found that the selection of the computational domain is crucial. A full core analysis is required for accurate results in the case of severe blockage accidents.

It is important to understand the packing characteristics and behaviour of the randomly packed pebble bed to further analyse the reactor physical and thermal-hydraulic behaviour and to design a safe and economically feasible pebble bed reactor. The objective of this work was to establish methods to model and analyse the pebble packing in detail to provide useful tools and data for further analyses. Discrete element method (DEM) is a well acknowledged method for analysing granular materials, such as the fuel pebbles in a pebble bed reactor. In this work, a DEM computer code was written specifically for pebble bed analyses. Analysis methods were established to extract data at various spatial scales from the pebble beds resulting from the DEM simulations. A comparison with available experimental data was performed to validate the DEM implementation. To test the code implementation in full-scale reactor calculations, DEM packing simulations were done in annular geometry with 450,000 pebbles. Effects of the initial packing configuration, friction and restitution coefficients and pebble size distribution to the resulting pebble bed were investigated. The packing simulations revealed that from the investigated parameters the restitution coefficient had the largest effect on the resulting average packing density while other parameters had smaller effects. Detailed local packing density analysis of pebble beds with different average densities revealed local variations especially strong in the regions near the walls. The implemented DEM code can be used for further analyses of pebble beds and can be customised for particular needs. The results of the packing simulations are useful when determining suitable values to obtain a pebble bed with the desired packing density for subsequent analyses. It is recommended to consider the local packing density variations especially in further coolant flow and heat transfer analyses.

A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R&D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

Total loss of flow accident characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with all upward flow core cooling in two paths is investigated at supercritical pressure. Upward flow cooling has advantage that flow stagnation does not occur at total loss of coolant flow accident due to the buoyancy of the coolant. It is found that maximum cladding surface temperature (MCST) is sensitive to the volume of the gap between two paths the. By open the automatic depressurization system (ADS), depressurization will induce flow in the RPV, and it can improve the MCST in this accident.

A model to describe deformation behaviour of a CANDU 37-element bundle has been developed under the COMSOL Multiphysics finite-element platform. Beam elements were applied to the fuel elements (composed of fuel sheaths and pellets) and endplates in order to calculate the bowing behaviour of the fuel elements. This model is important to help assess bundle-deformation phenomena, which may lead to more restrictive coolant flow through the sub-channels of the horizontally oriented bundle. The bundle model was compared to the BOW code for the occurrence of a dry-out patch, and benchmarked against an out-reactor experiment with a variable load on an outer fuel element.

Axial coolant flow inside a tightly packed rod bundle presents a complex behavior; experimental analysis had clearly shown that when reducing the pitch-to-diameter ratio ( P / D ) the turbulence field in rod bundles deviates significantly from that in a circular tube. Moreover for extremely tight configurations the existence of large-scale periodic “flow oscillations” has been shown, which is responsible for the high inter-sub-channel heat and momentum exchange. A complete understanding of these oscillations has still to be achieved; the evidence shown to this point suggests that the oscillations are connected to interactions between coherent structures in adjacent sub-channels. Moreover, the coherent structures show a truly three-dimensional pattern (i.e., structures in different gaps tend to interact) that has not been fully investigated up to this point. A fully transient simulation of turbulence has been performed for an infinite tight triangular lattice. In this case it has been performed with Large Eddy Simulation (LES) at Re = 6400 and P / D = 1.05. A database of snapshots of the flow field has then been collected and proper orthogonal decomposition (POD), a powerful statistical technique, has been applied in order to obtain the most energetic modes of turbulence. The results obtained highlight the presence of several travelling waves propagating in the streamwise direction. The spatial modulation of the travelling waves offers a phenomenological explanation for observed de-coherence effects between the velocity signal in different gaps. It also provides additional insight into the three dimensional structure of the flow oscillations.

In the design of advanced light water reactors (ALWRs) and in the safety assessment of currently operating nuclear power plants, it is necessary to evaluate the possibility of experiencing a degraded core accident and to develop innovative safety technologies in order to assure long-term debris cooling. The objective of this experimental study is to investigate the enhancement factors of dryout heat flux in debris beds by coolant injection from below. The experimental facility consists mainly of an induction heater, a double-wall quartz-tube test section containing a steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of the particle bed was achieved in the radial direction and the axial variation was within 20%. This paper reports the experimental data for 3.2 mm and 4.8 mm particle beds with a 300 mm bed height. The dryout heat density data were obtained for both the top-flooding and the forced coolant injection from below with an injection mass flux of up to 1.5 kg/m²s. The dryout heat density increased as the rate of coolant injection increased. At a coolant injection mass flux of 1.0 kg/m²s, the dryout heat density was ~6.5 MW/m3 for the 4.8 mm particle bed and ~5.6 MW/m3 for the 3.2 mm particle bed. The enhancement factors of the dryout heat density were 1.6-1.8.

The fuel assemblies used in the OPR1000s in Korea employ four coil-shaped hold-down springs to exert compressive load at the top of fuel assembly so that the assemblies may not be damaged by preventing its hydraulic-induced lifting-off from its lower seating surface. However, the coolant flow generates the flow-induced vibration at the coil-shaped hold-down springs which may cause wear on the spring surfaces. A hold-own spring may be fractured if torsional stress acting on its worn area exceeds a stress limit, resulting in the loss of hold-down spring force of the fuel assembly. In this paper, flow-induced vibration tests were performed for standard and improved coil type hold-down springs to investigate the effects of these two hold-down spring designs on flow-induced vibration wear. In parallel, a wide spectrum of mechanical tests was performed to obtain vibration-related characteristics of these two hold-down springs, which can be used as input data for the fuel assembly static and dynamic analysis. It is found that the improved hold-down spring design is better against flow-induced vibration wear than the standard one. With the use of the three-dimensional Solidwork model, the stress-related design lifetime of the improved hold-down spring was estimated by extrapolating its wear data measured from the flow-induced vibration tests, which indicates that the improved HD spring design will maintain integrity during the fuel design lifetime in OPR1000s in Korea.

In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235 U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235 U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900–950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

A numerical investigation into the effect of a coastdown flow on the early stage cooling of the reactor pool in Korea Advanced Liquid Metal Reactor (KALIMER)-600 during a loss of normal heat sink accident has been carried out. Based on the design values of KALIMER-600, thermal-hydraulic calculations for steady and transient states have been done using the COMMIX-1AR/P code. Coastdown flow effect was evaluated based on a transient analysis of reactors employing various flywheels, which had coastdown flow time (CDT) values ranging from 0 (without a flywheel) to 300 s. The transient analysis has been done from a reactor trip to the onset of an overflow into the DHX support barrel. It was found that the coastdown flow range could be divided into three zones, based on its effect. Among them an excessive core coolant peak temperature and a reversed flow at the core region were observed for a medium coastdown flow range. The medium ranged coastdown flow induces the development of a high density layer near the core exit. This layer contributes to the development of an adverse effect in the core coolant flow, and finally results in increasing the core peak temperature. It was also found that the initiation of heat removal by DHX could be accelerated by the increase of the CDT, although it needs a large flywheel. From this analysis the best CDT is determined to be 25 s.

The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.

The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5–11 MW) and different core coolant flow rates (500–921 m 3 /h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW th with the total power peaking factor maintained at less than or equal to 3.0.

Nuclear vendors and utilities perform numerous simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes, most of which were developed based on one-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and three-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. The use of advanced commercial CFD codes is considered beneficial in the safety analysis and design of NPPs. The present work analyzes the flow distribution in the downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of a PWR is used. The results give a clear figure about flow fields in the downcomer and lower plenum of a PWR, which is one of major safety concerns.

One postulated accident scenario for the Advanced CANDU Reactor (ACR-700™) is the complete coolant flow blockage of a single pressure-tube (PT). The flow is not monitored within each individual PT, thus during the early stages of this accident the reactor remains at full power and full pressure, resulting in rapid coolant boil-off and fuel overheating. Melting of the Zircaloy (Zry) components of the fuel bundle can occur, with relocation of some molten material to the bottom of the PT, which may cause failure of the PT and/or the calandria-tube (CT). We analyzed several key phenomena occurring after the blockage, including coolant boil-off, cladding heat-up and melting, dripping of molten Zircaloy (Zry) from the fuel pin, thermal interaction between the molten Zry and the PT, localized bulging of the PT, and interaction of the bulged PT with the CT. The main findings of the study are as follows: (1) Most coolant boils off within 3 s of accident initiation. (2) The Zry cladding starts to melt between 7 and 10 s after accident initiation. (3) The very high heat-up rate typical of the flow blockage accident sequence ensures that the molten Zry would drip to the bottom of the PT. (4) After contacting the molten Zry, the PT and CT bulge out radially under the effect of the reactor pressure. (5) PT/CT failure occurs only if the postulated mass of molten Zry in contact with the PT is sufficiently large, i.e., >100 g. The characteristic time scales for this 100-g case are as follows: - PT bulging starts within 3 s of Zry/PT contact; - PT makes contact with the CT in another 3 s; - CT bulging starts in approximately 1 s; - CT failure occurs within another 6 s. Thus, the duration of the PT/CT deformation transient is 13 s, which gives a total duration of the accident (from PT blockage to PT/CT failure) of 20–23 s. The relatively simple models developed in this study and the estimates generated with these models provide a solid physical framework for the key phenomena in the single-channel flow blockage event in ACR-700. As such, they can also assist in the interpretation and verification of future analyses of this event conducted with more sophisticated codes and tools.

A simulation model is developed to simulate the coupled kinetics and thermal-hydraulics of Material Testing Reactors (MTR) with simple means. The developed model is used to analyze the series of benchmark transients specified for IAEA 10 MW research reactor. The computed results for ramp positive reactivity insertion and loss of coolant flow transients with scram are presented and discussed. The model predictions are compared with the calculations conducted in various institutions using different codes. It is shown that the model can provide accurate predictions as long as the clad temperature remains below the onset of the nucleate boiling temperature. However, the results are very encouraging and the model is useful for the analysis of most research reactors encountered in practice.

This work experimentally investigates the heat transfer characteristics of Taylor–Couette–Poiseuille flow in an annular channel by mounting longitudinal ribs on the rotating inner cylinder. The ranges of the axial Reynolds number (Re) and the rotational Reynolds number (ReΩ) are Re = 30–1200 and ReΩ = 0–2922, respectively. Three modes of the inner cylinder without/with longitudinal ribs are considered. A special entry and exit design for the axial coolant flow reveals some interesting findings. The value of Nusselt number (Nu) is almost minimal at the inlet of the annular channel, and then sharply rises in the axial direction. The average Nusselt number ( Nu ¯ ) increases with Re. Nu increases rapidly with ReΩ at low Re (such as at Re = 30 and 60) but that the effect of ReΩ decreases as the value increases (such as at Re = 300–1200). The ratio Nu ¯ / Nu ¯ 0 increases with ReΩ and exceed two at all Re and in the test modes. The heat transfer is typically promoted by mounting longitudinal ribs on the rotating inner cylinder, especially at Re = 300 and 600. When Re = 300 or 600 and ReΩ > 2000, the Nu ¯ of the system with ribs reaches around 1.4 times that of Nu ¯ A ( Nu ¯ in mode A). Under a given pumping power constraint (PRe3), the Nu ¯ of the system with ribs (modes B and C) generally exceeds that without ribs (mode A), while the difference between the values of Nu ¯ in modes B and A slowly falls as PRe3 increases. Additionally, mode B (with ribs) is preferred for high heat transfer when PRe3 4.5 × 1013.

This study investigates experimentally and analytically the thermal hydraulic phenomena during the transition from design basis accident (DBA) to beyond-DBA, particularly, the depletion of core coolant inventory. To investigate the overall thermal hydraulic behavior after safety injection (SI) failure during a large-break loss-of-coolant accident (LBLOCA) in a cold leg, an integral loop test was performed at the Seoul National University Integral Test Facility (SNUF), which was scaled down to 1/6.4 in length and 1/178 in area from the advanced power reactor 1400 MWe (APR1400) according to the three-level scaling method. The plant condition at 200.0 s as the base case and those at 625.0 and 1950.0 s as test cases after the initiation of LBLOCA were applied as initial conditions in the experiments, respectively. The experimental results showed that the sweepout increased the coolant flow discharged to the break depending on the steam flow rate in intact cold legs and the initial downcomer coolant level and expedited the depletion of the core coolant inventory. In the meantime, since RELAP5/MOD3.3 uses the average properties of donor volume as those of its connected junction, this scheme causes the mass and the energy flux in a junction to be calculated incorrectly if significant phase separation occurs in the donor volume such as in the downcomer during the LBLOCA. The sweepout model was developed and implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory during the LBLOCA. To assess the applicability of the modified RELAP5/MOD3.3 to the actual system, the experiments in SNUF were simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the discharge flow rate at the break larger than that of the experiment. On the other hand, the modified one calculated the discharge flow rate more similar to that of the experiment than the original one did. As a result, the modified RELAP5/MOD3.3 reduced the coolant flow discharged to the break to delay the initiation time of heater heat-up after SI failure during LBLOCA in a cold leg. This improved RELAP5/MOD3.3 will support a more realistic thermal hydraulic analysis in an integrated analysis system.

The design of nuclear fuel, which is mostly subjected to loads due to high temperature, pressure and flow under radiation environment, is of immense importance in nuclear engineering. Generally, the nuclear fuel consists of number of fuel pins, which carry the fuel pellets inside. A number of such pins are grouped together to a specific size to form a bundle. Flow-induced vibration in the bundle due to high velocity coolant flow cause interaction between the fuel pins. Commonly, spacers are provided between the fuel pins to maintain the gap between the neighboring pins. Large number of such spacers in the reactor core is considered to be a load on thermal-hydraulics and on physics of reactor control. This paper addresses the work done with an objective to help optimization of the number of spacers required on a typical fuel bundle by experimental and analytical study.

Safety demonstration tests using the high temperature engineering test reactor (HTTR) are conducted for demonstrating inherent safety features of high temperature gas-cooled reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005 and the second phase tests will be carried out from FY 2006.

Heat transfer phenomena in two types of eutectic crystallizers have been analyzed. This equipment is equipped with rotating scraper blades and the effect of the blade action on heat transfer phenomena in both crystallizing and non-crystallizing conditions is studied. It was found that the coolant flow in the equipment can be described by fitted Dittus–Boelter relations. In both crystallizers the transport of heat from process liquid to cooled wall surfaces at various rotational speeds of the scrapers can generally be accurately described with penetration theory. Exceptions are approximated with fitted Re – Pr -relations. Both increasing and decreasing heat transfer rates have been found in crystallizing conditions at increasing scraping rates. Differences are attributed to geometrical crystallizer characteristics and solid content.

Background Surface active agents added to handpiece coolants enhance bur cutting rates, or CRs, through chemomechanical effects, or CMEs. The authors evaluate the effect of CMEs on long-term cutting by diamond burs. Methods The authors tested medium-grit diamonds in a high-speed handpiece under a 147.5-gram load at 350,000 revolutions per minute, with 22 milliliters per minute coolant flow consisting of distilled water or a 1:10 mouthwash/water mixture. Repeated 6-millimeter–long edge cuts were made through machinable glass ceramic bars until the CR (determined as the time to transect the bars) had decreased by more than 75 percent. The authors used six burs for each coolant and analyzed the data via one-way analysis of variance with post hoc Scheffe tests. Results CRs with water irrigation continuously decreased with the number of cuts and declined by 87 percent over eight cuts. The authors found faster CRs with CMEs; after 12 cuts, the CR was still close to 40 percent of the initial CR, compared with only 13 percent of the initial CR for water irrigation after eight cuts. These differences between CME-enhanced CRs and those found with water irrigation were statistically significant. Conclusions Diluted alcohol- and glycerol-based mouthwash/water mixtures significantly enhanced the CRs of diamond burs and prolonged their service life by more than 50 percent compared with water irrigation alone. Clinical Implications Dentists can increase bur CRs and extend bur cutting life by as much as 200 percent through the addition of diluted alcohol and glycerol mouthwash to the handpiece coolant.

The paper is devoted to the description of the regular neutron noise diagnostics measurements performed by KFKI AEKI at the four units of the Hungarian Paks NPP. Theoretical background, practical details and some results of the fuel assembly coolant flow surveillance measurements are outlined, emphasising their importance in practical plant operation. Control rod vibration monitoring methods and results are treated shortly, as well.

A curve matching technique is proposed to process experimentally obtained images to permit analysis of coolant flow patterns, which will be used to determine suitable parameters for an in-process optical measurement method to be used in the machining processes. In the proposed method, an array of ideal images is created based on the theoretical results and the disparities with the actual images are examined through a recursive process. To quantify the assessment of the disparities, an error index is used to determine the most suitable match. The working principle of the proposed technique is presented and the recursive algorithm to examine the image disparities described. To improve the computational efficiency, techniques to offer the necessary enhancement are described. The results of the experimental test demonstrated the effectiveness of the proposed method.

There are a few transient and loss-of-coolant accident conditions in RBMK-1500 reactors that lead to a local flow decrease in fuel channels. Because the coolant flow decreases in fuel channels (FC) leads to overheating of fuel claddings and pressure tube walls, mitigation measures are necessary. The accident analysis enabled the suggestion of the new early reactor scram actuation and emergency core cooling system (ECCS) initiation signal, which ensures the safe shutdown of the reactor and compensates the stagnation flow. Analysis of such conditions is presented in this paper. Thermal-hydraulic analysis was conducted using the state-of-the-art RELAP5 code. Results of the analysis demonstrated that, after implementation of the developed management strategy for destruction of local flow stagnation, the Ignalina nuclear power plant (NPP) would be adequately protected following accidents, leading to local coolant flow decrease in the primary circuit.

The objective is to control the internal temperature of a continuous stirred tank reactor (CSTR) with jacket and analyze the potentiality that offers manipulating the thermodynamic conditions of the coolant flow. In this paper four predictive functional control (PFC) strategies are analyzed, two of them are conventionally structured by manipulating flow in one case and temperature of the jacket coolant in the other. The other two manipulate the enthalpic conditions of the coolant in a cascade form by setting flow and temperature optimally. The servo and regulator problem are analyzed including an statistical economic studio for defining the best control structure.

A parameter that is often used in the assessment of cooling efficiency in grinding is the amount of fluid that flows through the grinding zone. This paper shows that when grinding with wheels possessing no bulk porosity, only a portion of the flow through the grinding zone is effective. Even at the highest wheel speeds employed, it is observed that ample fluid travels on the wheel surface and is re-directed into the grinding zone — a phenomenon that may adversely affect cooling and wheel cleaning. Therefore fluid rejection off the wheel is important. A theoretical model for fluid rejection is proposed, that describes the drop formation process, presents the operating instabilities, identifies characteristic time and length scales, and predicts the amount of rejected fluid. Although the surface roughness of the wheel is important in quantifying coolant flow rates, a smooth wheel is employed in this work to study first order effects. Preliminary experimental results obtained for a rough grinding wheel are in conformance with those for a smooth wheel.

Although most commercial chillers are designed and installed to operate at constant coolant flow rates, one can alter the power consumption of their compressors by modifying the design to incorporate variable coolant flow rates. The issue addressed here is how to model that explicit flow-rate dependence within an analytic chiller model. We expand our earlier thermodynamic model to explicitly account for the influence of condenser coolant flow rate, and validate model predictions against an extensive set of experimental measurements from a large commercial centrifugal chiller.