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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 37, Issue 6 - Dec 2005
Volume 37, Issue 5 - Oct 2005
Volume 37, Issue 4 - Aug 2005
Volume 37, Issue 3 - Jun 2005
Volume 37, Issue 2 - Apr 2005
Volume 37, Issue 1 - Feb 2005
Selecting the target year
STATUS AND PERSPECTIVE OF TWO-PHASE FLOW MODELLING IN THE NEPTUNE MULTISCALE THERMAL-HYDRAULIC PLATFORM FOR NUCLEAR REACTOR SIMULATION
BESTION DOMINIQUE ; GUELFI ANTOINE ; DEN/EER/SSTH CEA-GRENOBLE, ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 511~524
Thermalhydraulic reactor simulation of tomorrow will require a new generation of codes combining at least three scales, the CFD scale in open medium, the component scale and the system scale. DNS will be used as a support for modelling more macroscopic models. NEPTUNE is such a new generation multi-scale platform developed jointly by CEA-DEN and EDF-R&D and also supported by IRSN and FRAMATOME-ANP. The major steps towards the next generation lie in new physical models and improved numerical methods. This paper presents the advances obtained so far in physical modelling for each scale. Macroscopic models of system and component scales include multi-field modelling, transport of interfacial area, and turbulence modelling. Two-phase CFD or CMFD was first applied to boiling bubbly flow for departure from nucleate boiling investigations and to stratified flow for pressurised thermal shock investigations. The main challenges of the project are presented, some selected results are shown for each scale, and the perspectives for future are also drawn. Direct Numerical Simulation tools with Interface Tracking Techniques are also developed for even smaller scale investigations leading to a better understanding of basic physical processes and allowing the development of closure relations for macroscopic and CFD models.
DEVELOPMENT OF INTERFACIAL AREA TRANSPORT EQUATION
ISHII MAMORU ; KIM SEUNGJIN ; KELLY JOSEPH ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 525~536
The interfacial area transport equation dynamically models the changes in interfacial structures along the flow field by mechanistically modeling the creation and destruction of dispersed phase. Hence, when employed in the numerical thermal-hydraulic system analysis codes, it eliminates artificial bifurcations stemming from the use of the static flow regime transition criteria. Accounting for the substantial differences in the transport mechanism for various sizes of bubbles, the transport equation is formulated for two characteristic groups of bubbles. The group 1 equation describes the transport of small-dispersed bubbles, whereas the group 2 equation describes the transport of large cap, slug or chum-turbulent bubbles. To evaluate the feasibility and reliability of interfacial area transport equation available at present, it is benchmarked by an extensive database established in various two-phase flow configurations spanning from bubbly to chum-turbulent flow regimes. The geometrical effect in interfacial area transport is examined by the data acquired in vertical fir-water two-phase flow through round pipes of various sizes and a confined flow duct, and by those acquired In vertical co-current downward air-water two-phase flow through round pipes of two different sizes.
SAFETY STUDIES ON HYDROGEN PRODUCTION SYSTEM WITH A HIGH TEMPERATURE GAS-COOLED REACTOR
TAKEDA TETSUAKI ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 537~556
A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.
AN EXPERIMENTAL STUDY ON AIR-WATER COUNTERCURRENT FLOW LIMITATION IN THE UPPER PLENUM WITH A MULTI-HOLE PLATE
NO HEE CHEON ; LEE KYUNG-WON ; SONG CHUL-HWA ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 557~564
Air-water countercurrent flow limitation at perforated plates with four holes was investigated in a vertical tank to see the effects of the plate thickness, the number of hole, and the diameter of the hole on the onset of CCFL. The thickness of plates was 1 cm and 4 cm, with a relatively large hole diameter of 5 cm. The collapsed water level formed on the perforated plate and its distribution in the upper plenum were measured. The gas flow rate in the multi-hole plate is relatively higher than one in the single tube because some of holes in the multi-hole plate provide a flow path fur liquid with less air-liquid resistance than in the single tube. The onset of CCFL occurred at nearly the same air flow rate regardless of the plate thickness. The negligible effect of the plate thickness on CCFL means that the flooding is initiated at the top of the plate rather than at its bottom. It turns out that
better fit the data than
when hole diameter is greater than 2.86 cm. In our experimental ranges, the collapsed water levels at the onset of CCFL ranged from 7.5 cm to 10.5 cm. There was no three dimensional distribution of water level before and after the onset of CCFL.
COMPARISON OF DRYOUT POWER DATA BETWEEN CANFLEX MK-V AND CANFLEX MK-IV BUNDLE STRINGS IN UNCREPT AND CREPT CHANNELS
JUN JI SU ; LEUNG L.K.H. ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 565~574
The CANFLEX Mk-V bundle is designed to improve upon the critical heat flux (CHF) characteristics of the CANFLEX Mk-IV bundle. The main difference between these two bundles is an increase in bearing pad height of about 0.3 mm in the CANFLEX Mk-IV bundle. This change in bearing pad height leads to an increase in gap flow at the bottom of the bundle, primarily eliminating the localized narrow-gap effect that limits the CHF of the CANFLEX Mk-IV bundle. The objective of this paper is to examine the effects of bearing pad height and pressure tube creep on the sheath-temperature distribution, dryout power, and dryout location, as observed ken full-scale bundle tests, between CANFLEX Mk-IV and Mk-V bundles In uncrept and crept channels. A comparison of surface-temperature differences between the top and bottom elements of the bundles showed that increasing the bearing pad height has led to a more homogeneous enthalpy distribution in subchannels of the bundle. Initial dryout locations of the CANFLEX Mk-V bundle were mainly observed at the mid-spacer plane of either the
) bundle in the 12-bundle string, as compared to the mid-spacer and downstream-button planes for the CANFLEX Mk-IV bundle. Dryout power and boiling-length-average (BLA) CHF values exhibit consistent trends and little scatter with varying flow conditions for both types of CANFLEX bundles in uncrept and crept channels. An increase in pressure tube creep has led to a reduction in dryout power (about
crept channel and
crept channel as compared to dryout powers for the uncrept channel). Increasing the bearing pad height of the CANFLEX bundle has led to an increase in the dryout power. Overall, the dryout power of the CANFLEX Mk-V bundle is 7 to
higher than that of the CANFLEX Mk-IV bundle at the inlet temperature range of interest (i.e., between 243 and
ANALYSES OF FLUID FLOW AND HEAT TRANSFER INSIDE CALANDRIA VESSEL OF CANDU-6 REACTOR USING CFD
YU SEON-OH ; KIM MANWOONG ; KIM HHO-JUNG ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 575~586
In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.
DEVELOPMENT OF MARS-GCR/V1 FOR THERMAL-HYDRAULIC SAFETY ANALYSIS OF GAS-COOLED REACTOR SYSTEMS
LEE WON-JAE ; JEONG JAR-JUN ; LEE SEUNG-WOOK ; CHANG JONGHWA ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 587~594
In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide (
) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.
MEASUREMENT OF THE SINGLE AND TWO PHASE FLOW USING A NEWLY DEVELOPED AVERAGE BIDIRECTIONAL FLOW TUBE
Yun, Byong-Jo ; Euh, Dong-Jin ; Kang, Kyunc-Ho ; Song, Chul-Hwa ; Baek, Won-Pil ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 595~604
A new instrument, an average BDFT (Birectional Flow Tube), was proposed to measure the flow rate in single and two phase flows. Its working principle is similar to that of the Pilot tube, wherein the dynamic pressure is measured. In an average BDFT, the pressure measured at the front of the flow tube is equal to the total pressure, while that measured at the rear tube is slightly less than the static pressure of the flow field due to the suction effect downstream. The proposed instrument was tested in air/water vertical and horizontal test sections with an inner diameter of 0.08m. The tests were performed primarily in single phase water and air flow conditions to obtain the amplification factor(k) of the flow tube in the vertical and horizontal test sections. Tests were also performed in air/water vertical two phase flow conditions in which the flow regimes were bubbly, slug, and churn turbulent flows. In order to calculate the phasic mass flow rates from the measured differential pressure, the Chexal drift-flux correlation and a momentum exchange factor between the two phases were introduced. The test results show that the proposed instrument with a combination of the measured void fraction, Chexal drift-flux correlation, and Bosio & Malnes' momentum exchange model could predict the phasic mass flow rates within a
error. A new momentum exchange model was also proposed from the present data and its implementation provides a
improvement to the measured mass flow rate when compared to that with the Bosio & Malnes' model.
RHODIUM SELF-POWERED NEUTRON DETECTOR'S LIFETIME FOR KOREAN STANDARD NUCLEAR POWER PLANTS
YOO CHOON SUNG ; KIM BYOUNG CHUL ; PARK JONG-HO ; FERO ARNOLD H. ; ANDERSON S. L. ;
Nuclear Engineering and Technology, volume 37, issue 6, 2005, Pages 605~610
A method to estimate the relative sensitivity of a self-powered rhodium detector for an upcoming cycle is developed by combining the rhodium depletion data from a nuclear design with the site measurement data. This method can be used both by nuclear power plant designers and by site staffs of Korean standard nuclear power plants for determining which rhodium detectors should be replaced during overhauls.