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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 41, Issue 10 - Dec 2009
Volume 41, Issue 9 - Nov 2009
Volume 41, Issue 8 - Oct 2009
Volume 41, Issue 7 - Sep 2009
Volume 41, Issue 6 - Aug 2009
Volume 41, Issue 5 - Jun 2009
Volume 41, Issue 4 - May 2009
Volume 41, Issue 3 - Apr 2009
Volume 41, Issue 2 - Mar 2009
Volume 41, Issue 1 - Feb 2009
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ANALYSIS OF THE NODALISATION INFLUENCE ON SIMULATING ATMOSPHERIC STRATIFICATIONS IN THE EXPERIMENT THAI TH13 WITH THE CONTAINMENT CODE SYSTEM COCOSYS
Burkhardt, Joerg ; Schwarz, Siegfried ; Koch, Marco K. ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1135~1142
DOI : 10.5516/NET.2009.41.9.1135
The activities related to this paper are to investigate the influence of nodalisation on simulating atmospheric stratification in the THAI experiment TH13 (ISP-47) with the German containment code COCOSYS. This article focuses on different nodalisations of the vessel dome, where an atmospheric stratification occurred due to a high helium content. The volume of the dome was divided into several levels that were varied horizontally into different geometries. These geometries differ in the number of zones as well as in the existence of zones that enable the direct rise of an ascending steam plume into the vessel dome. Additionally, the vertical subdivision of the vessel dome was increased to simulate density gradients in a more detailed way. It was pointed out that the proper simulation of atmospheric stratifications and their dissolution depends on both a suitable horizontal as well as vertical nodalisation scheme. Besides, the treatment of fog droplets has an influence if their settlement is not simulated correctly. This report gives an overview of the gained experience and provides nodalisation requirements to simulate atmospheric stratifications and their proper dissolution.
DEVELOPMENT OF A WALL-TO-FLUID HEAT TRANSFER PACKAGE FOR THE SPACE CODE
Choi, Ki-Yong ; Yun, Byong-Jo ; Park, Hyun-Sik ; Kim, Hee-Dong ; Kim, Yeon-Sik ; Lee, Kwon-Yeong ; Kim, Kyung-Doo ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1143~1156
DOI : 10.5516/NET.2009.41.9.1143
The SPACE code that is based on a multi-dimensional two-fluid, three-field model is under development for licensing purposes of pressurized water reactors in Korea. Among the participating research and industrial organizations, KAERI is in charge of developing the physical models and correlation packages for the constitutive equations. This paper introduces a developed wall-to-fluid heat transfer package for the SPACE code. The wall-to-fluid heat transfer package consists of twelve heat transfer subregions. For each sub-region, the models in the existing safety analysis codes and the leading models in literature have been peer reviewed in order to determine the best models which can easily be applicable to the SPACE code. Hence a wall-to-fluid heat transfer region selection map has been developed according to the non-condensable gas quality, void fraction, degree of subcooling, and wall temperature. Furthermore, a partitioning methodology which can take into account the split heat flux to the continuous liquid, entrained droplet, and vapor fields is proposed to comply fully with the three-field formulation of the SPACE code. The developed wall-to-fluid heat transfer package has been pre-tested by varying the independent parameters within the application range of the selected correlations. The smoothness between two adjacent heat transfer regimes has also been investigated. More detailed verification work on the developed wall-to-fluid heat transfer package will be carried out when the coupling of a hydraulic solver with the constitutive equations is brought to completion.
AN AXIOMATIC DESIGN APPROACH OF NANOFLUID-ENGINEERED NUCLEAR SAFETY FEATURES FOR GENERATION III+ REACTORS
Bang, In-Cheol ; Heo, Gyun-Young ; Jeong, Yong-Hoon ; Heo, Sun ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1157~1170
DOI : 10.5516/NET.2009.41.9.1157
A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.
LARGE SCALE FINITE ELEMENT THERMAL ANALYSIS OF THE BOLTS OF A FRENCH PWR CORE INTERNAL BAFFLE STRUCTURE
Rupp, Isabelle ; Peniguel, Christophe ; Tommy-Martin, Michel ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1171~1180
DOI : 10.5516/NET.2009.41.9.1171
The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The
De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code_Saturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer.
ESTIMATION OF THE POWER PEAKING FACTOR IN A NUCLEAR REACTOR USING SUPPORT VECTOR MACHINES AND UNCERTAINTY ANALYSIS
Bae, In-Ho ; Na, Man-Gyun ; Lee, Yoon-Joon ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1181~1190
DOI : 10.5516/NET.2009.41.9.1181
Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models' uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.
TOWARD AN ACCURATE APPROACH FOR THE PREDICTION OF THE FLOW IN A T-JUNCTION: URANS
Merzari, E. ; Khakim, A. ; Ninokata, H. ; Baglietto, E. ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1191~1204
DOI : 10.5516/NET.2009.41.9.1191
In this study, a CFD methodology is employed to address the problem of the prediction of the flow in a T-junction. An Unsteady Reynolds Averaged Navier-Stokes (URANS) approach has been selected for its low computational cost. Moreover, Unsteady Reynolds Navier-Stokes methodologies do not need complex boundary formulations for the inlet and the outlet such as those required when using Large Eddy Simulation (LES) or Direct Numerical Simulation (DNS). The results are compared with experimental data and an LES calculation. In the past, URANS has been tried on T-junctions with mixed results. The biggest limit observed was the underestimation of the oscillatory behavior of the temperature. In the present work, we propose a comprehensive approach able to correctly reproduce the root mean square (RMS) of the temperature directly downstream of the T-junction for cases where buoyancy is not present.
TRANSIENT CHF PHENOMENA DUE TO EXPONENTIALLY INCREASING HEAT INPUTS
Park, Jong-Doc ; Fukuda, Katsuya ; Liu, Qiusheng ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1205~1214
DOI : 10.5516/NET.2009.41.9.1205
The critical heat flux (CHF) levels that occurred due to exponential heat inputs for varying periods to a 1.0-mm diameter horizontal cylinder immersed in various liquids were measured to develop an extended database on the effect of high subcoolings for quasi-steady-state and transient maximum heat fluxes. Two main mechanisms of CHF were found. One mechanism is due to the time lag of the hydrodynamic instability (HI) which starts at steady-state CHF upon fully developed nucleate boiling, and the other mechanism is due to the explosive process of heterogeneous spontaneous nucleation (HSN) which occurs at a certain HSN superheat in originally flooded cavities on the cylinder surface. Steady-state CHFs were divided into three regions for lower, intermediate and higher subcooling at pressures resulting from HI, transition and HSN, respectively. HSN consistently occurred in the transient boiling CHF conditions that correspond to a short period. It was also found that the transient boiling CHFs gradually increased, then rapidly decreased and finally increased again as the period became shorter.
TRIGGERING AND ENERGETICS OF A SINGLE DROP VAPOR EXPLOSION: THE ROLE OF ENTRAPPED NON-CONDENSABLE GASES
Hansson, Roberta Concilio ;
Nuclear Engineering and Technology, volume 41, issue 9, 2009, Pages 1215~1222
DOI : 10.5516/NET.2009.41.9.1215
The present work pertains to a research program to study Molten Fuel-Coolant Interactions (MFCI), which may occur in a nuclear power plant during a hypothetical severe accident. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography. The current study is concerned with the MISTEE-NCG test campaign, in which a considerable amount of non-condensable gases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzed and special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt drop preconditioning. Energetics of the vapor explosion (conversion ratio) were also evaluated. The MISTEE-NCG tests showed two main aspects when compared to the MISTEE test series (without entrapped air). First, analysis showed that the melt preconditioning still strongly depends on the coolant subcooling. Second, in respect to the energetics, the tests consistently showed a reduced conversion ratio compared to that of the MISTEE test series.