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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 38, Issue 8 - Dec 2006
Volume 38, Issue 7 - Oct 2006
Volume 38, Issue 6 - Aug 2006
Volume 38, Issue 5 - Jul 2006
Volume 38, Issue 4 - Jun 2006
Volume 38, Issue 3 - Apr 2006
Volume 38, Issue 2 - Feb 2006
Volume 38, Issue 1 - Feb 2006
Selecting the target year
MULTIPHASE FLOW IN EX-VESSEL COOLABILITY: DEVELOPMENT OF AN INNOVATIVE CONCEPT
CORRADINI MICHAEL L. ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 1~10
The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability.
STATE OF THE ART IN USING BEST ESTIMATE CALCULATION TOOLS IN NUCLEAR TECHNOLOGY
D'AURIA FRANCESCO ; ANIS BOUSBIA-SALAH ; PETRUZZI ALESSANDRO ; NEVO ALESSANDRO DEL ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 11~32
System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.
A COMPARATIVE OVERVIEW OF THERMAL HYDRAULIC CHARACTERISTICS OF INTEGRATED PRIMARY SYSTEM NUCLEAR REACTORS
NINOKATA HISASHI ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 33~44
This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.
ASSESSMENT OF A NEW DESIGN FOR A REACTOR CAVITY COOLING SYSTEM IN A VERY HIGH TEMPERATURE GAS-COOLED REACTOR
PARK GOON-CHERL ; CHO YUN-JE ; CHO HYOUNGKYU ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 45~60
Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.
EXPERIMENTAL STUDY ON CHF CHARACTERISTICS OF WATER-TI02 NANO-FLUIDS
Kim, Hyung-Dae ; Kim, Jeong-Bae ; Kim, Moo-Hwan ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 61~68
CHF characteristics of nano- fluids were investigated with different volumetric concentrations of
nanoparticles. Pool boiling experiments indicated that the application of nano-fluids, instead of pure water, as a cooling liquid significantly increased the CHF. SEM (scanning electron microscope) observations subsequent to the pool boiling experiments revealed that nanoparticles were coated on the heating surface during pool boiling of nano-fluids. In order to investigate the roles of nanoparticles in CHF enhancement ofb nano-fluids, pool boiling experiments were performed using (a) a nanoparticle-coated heater, prepared by pool boiling of nano-fluids, immersed in pure water and (b) a nanoparticle-coated heater immersed in nano-fluids. The results demonstrated two different roles of nanoparticles in CHF enhancement using nano-fluids: the effect of nanoparticles coated on the heater surface and the effect of nanoparticles suspended in nano- fluids.
FLUID-ELASTIC INSTABILITY OF ROTATED SQUARE TUBE ARRAY IN AN AIR-WATER TWO-PHASE CROSSFLOW
CHUNG HEUNG JUNE ; CHU IN-CHEOL ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 69~80
Fluid-elastic instability in an air-water two-phase cross-flow has been experimentally investigated using two different arrays of straight tube bundles: normal square (NS) array and rotated square (RS) array tube bundles with the same pitch-to-diameter ratio of 1.633. Experiments have been performed over wide ranges of mass flux and void fraction. The quantitative tube vibration displacement was measured using a pair of strain gages and the detailed orbit of the tube motion was analyzed from high-speed video recordings. The present study provides the flow pattern, detailed tube vibration response, damping ratio, hydrodynamic mass, and the fluid-elastic instability for each tube bundle. Tube vibration characteristics of the RS array tube bundle in the two-phase flow condition were quite different from those of the NS array tube bundle with respect to the vortex shedding induced vibration and the shape of the oval orbit of the tube motion at the fluid-elastic instability as well as the fluid-elastic instability constant.
DESIGN OF A PWR POWER CONTROLLER USING MODEL PREDICTIVE CONTROL OPTIMIZED BY A GENETIC ALGORITHM
Na, Man-Gyun ; Hwang, In-Joon ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 81~92
In this study, the core dynamics of a PWR reactor is identified online by a recursive least-squares method. Based on the identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to designing an automatic controller for the thermal power control of PWR reactors. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, this procedure for solving the optimization problem is repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired temperature, as well as minimizing the variation of the control rod positions. In addition, the objectives are subject to the maximum and minimum control rod positions as well as the maximum control rod speed. Therefore, a genetic algorithm that is appropriate for the accomplishment of multiple objectives is utilized in order to optimize the model predictive controller. A three-dimensional nuclear reactor analysis code, MASTER that was developed by the Korea Atomic Energy Research Institute (KAERI) , is used to verify the proposed controller for a nuclear reactor. From the results of a numerical simulation that was carried out in order to verify the performance of the proposed controller with a
ramp increase or decrease of a desired load and a
step increase or decrease (which were design requirements), it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.
RADIAL UNIFORMITY OF NEUTRON IRRADIATION IN SILICON INGOTS FOR NEUTRON TRANSMUTATION DOPING AT HANARO
KIM MYONG-SEOP ; LEE CHOONG-SUNG ; OH SOO-YOUL ; HWANG SUNG-YUL ; JUN BYUNG-JIN ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 93~98
The radial uniformity of neutron irradiation in silicon ingots for neutron transmutation doping (NTD) at HANARO is examined by both calculations and measurements. HANARO has two NTD holes named NTD1 and NTD2. We have been using the NTD2 hole for 5 in. NTD commercial service, and we intend to use two holes for 6 in. NTD. The objective of this study is to predict the radial uniformity of 6 in. NTD at the two holes. The radial neutron flux distributions inside single crystal and noncrystal silicon loaded at the NTD2 hole are calculated by the VENTURE code. For NTD1, the radial distributions of the reaction rate for a 6 in. NTD with a neutron screen are calculated by MCNP, and measured by gold wire activation. The results of the measurements are compared with those of the calculations. From the VENTURE calculation, it is confirmed that the neutron flux distribution in the single crystal silicon is much flatter than that in the non-crystal silicon. The non-uniformities of the measurements for radial neutron irradiation are slightly larger than those of the calculations. However, excluding local dips in the measurements, the overall trends of the distributions are similar. The radial resistivity gradient (RRG) for a 5 in. silicon ingot is estimated to be about
. For a 6 in. ingot, the RRG of a silicon ingot irradiated at HANARO is predicted to be about
. Also, from the experimental results, we expect that the RRG would not be larger than
EXPERIMENTAL STUDY ON MEASUREMENT OF EMISSIVITY FOR ANALYSIS OF SNU-RCCS
CHO YUN-JE ; KIM MOON OH ; PARK GOON-CHERL ;
Nuclear Engineering and Technology, volume 38, issue 1, 2006, Pages 99~108
SNU-RCCS is a water pool type RCCS (Reactor Cavity Cooling System) developed for VHTR (Very High Temperature Reactor) application by SNU (Seoul National University). Since radiation heat transfer is the major process of passive heat removal in a RCCS, it is important to determine the precise emissivity of the reactor vessel. Review studies have used a constant emissivity in the passive heat removal analysis, even though the emissivity depends on many factors such as temperature, surface roughness, oxidation level, wavelength, direction, atmosphere conditions, etc. Therefore, information on the emissivity of a given material in a real RCCS is essential in order to properly analyze the radiation heat transfer in a VHTR. The objectives of this study are to develop a method for compensation of the factors affecting the emissivity measurement using an infrared thermometer and to estimate the true emissivity from the measured emissivity via the developed method, especially in the SNU-RCCS environment. From this viewpoint, we investigated factors such as the attenuation effect of the window, filling gas, and the effect of background radiation on the emissivity measurements. The emissivity of the vessel surface of the SNU-RCCS facility was then measured using a sight tube. The background radiation was subsequently removed from the measured emissivity by solving a simultaneous equation. Finally, the calculated emissivity was compared with the measured emissivity in a separate emissivity measurement device, yielding good agreement with the emissivity increase with vessel temperature in a range of 0.82 to 0.88.