Go to the main menu
Skip to content
Go to bottom
REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
Editor in Chief :
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
Selecting the target year
HORIZON EXPANSION OF THERMAL-HYDRAULIC ACTIVITIES INTO HTGR SAFETY ANALYSIS INCLUDING GAS-TURBINE CYCLE AND HYDROGEN PLANT
No, Hee-Cheon ; Yoon, Ho-Joon ; Kim, Seung-Jun ; Lee, Byeng-Jin ; Kim, Ji-Hwang ; Kim, Hyeun-Min ; Lim, Hong-Sik ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 875~884
DOI : 10.5516/NET.2009.41.7.875
We present three nuclear/hydrogen-related R&D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data.
ACCURACY AND EFFICIENCY OF A COUPLED NEUTRONICS AND THERMAL HYDRAULICS MODEL
Pope, Michael A. ; Mousseau, Vincent A. ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 885~892
DOI : 10.5516/NET.2009.41.7.885
This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid heat conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional "operator split" approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to
order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant limit required of operator-split methods. In this work, a pilot code was used which employs this tightly coupled, fully implicit method to simulate a reactor core. Results are presented from a simulated control rod movement which show
order accuracy in time. Also described in this paper is a simulated rod ejection demonstrating how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.
DEVELOPMENT OF AN LES METHODOLOGY FOR COMPLEX GEOMETRIES
Merzari, Elia ; Ninokata, Hisashi ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 893~906
DOI : 10.5516/NET.2009.41.7.893
The present work presents the development of a Large Eddy Simulation (LES) methodology viable for complex geometries and suitable for the simulation of rod-bundles. The use of LES and Direct Numerical Simulation (DNS) allows for a deeper analysis of the flow field and the use of stochastical tools in order to obtain additional insight into rod-bundle hydrodynamics. Moreover, traditional steady-state CFD simulations fail to accurately predict distributions of velocity and temperature in rod-bundles when the pitch (P) to diameter (D) ratio P/D is smaller than 1.1 for triangular lattices of cylindrical pins. This deficiency is considered to be due to the failure to predict large-scale coherent structures in the region of the gap. The main features of the code include multi-block capability and the use of the fractional step algorithm. As a Sub-Grid-Scale (SGS) model, a Dynamic Smagorinsky model has been used. The code has been tested on plane channel flow and the flow in annular ducts. The results are in excellent agreement with experiments and previous calculations.
PREDICTION OF A HEAT TRANSFER TO CO
FLOWING IN AN UPWARD PATH AT A SUPERCRITICAL PRESSURE
Cho, Bong-Hyun ; Kim, Young-In ; Bae, Yoon-Yeong ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 907~920
DOI : 10.5516/NET.2009.41.7.907
This study was performed to evaluate the prediction capability of a commercial CFD code and to investigate the effects of different geometries such as a 4.4 mm tube and an 8/10 mm annular channel on the detailed flow structures. A numerical simulation was performed for the conditions, at which the experimental data was produced by the test facility SPHINX. A 2-dimensional axisymmetric steady flow was assumed for computational simplicity. The RNG
turbulence model (RNG) with an enhanced wall treatment option, SST
(SST) and low Reynolds Abid turbulence model (ABD) were employed and the numerical predictions were compared with the experimental data generated from the experiment. The effects of the geometry on heat transfer were investigated. The flow and temperature fields were also examined in order to investigate the mechanism of heat transfer near the wall. The local heat transfer coefficient predicted by the RNG model is very close to the measurement result for the tube. In contrast, the local heat transfer coefficient predicted by the SST and ABD models is closer to the measurement for the annular channel.
DETAILED EVALUATION OF THE IN-VESSEL SEVERE ACCIDENT MANAGEMENT STRATEGY FOR SBLOCA USING SCDAP/RELAP5
Park, Rae-Joon ; Hong, Seong-Wan ; Kim, Sang-Baik ; Kim, hee-Dong ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 921~928
DOI : 10.5516/NET.2009.41.7.921
As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.
SIMULATION OF CORE MELT POOL FORMATION IN A REACTOR PRESSURE VESSEL LOWER HEAD USING AN EFFECTIVE CONVECTIVITY MODEL
Tran, Chi-Thanh ; Dinh, Truc-Nam ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 929~944
DOI : 10.5516/NET.2009.41.7.929
The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.
ANALYSIS OF TMI-2 BENCHMARK PROBLEM USING MAAP4.03 CODE
Yoo, Jae-Sik ; Suh, Kune-Yull ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 945~952
DOI : 10.5516/NET.2009.41.7.945
The Three Mile Island Unit 2 (TMI-2) accident provides unique full scale data, thus providing opportunities to check the capability of codes to model overall plant behavior and to perform a spectrum of sensitivity and uncertainty calculations. As part of the TMI-2 analysis benchmark exercise sponsored by the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD NEA), several member countries are continuing to improve their system analysis codes using the TMI-2 data. The Republic of Korea joined this benchmark exercise in November 2005. Seoul National University has analyzed the TMI-2 accident as well as the currently proposed alternative scenario along with a sensitivity study using the Modular Accident Analysis Program Version 4.03 (MAAP4.03) code in collaboration with the Korea Hydro and Nuclear Power Company. Two input files are required to simulate the TMI-2 accident with MAAP4: the parameter file and an input deck. The user inputs various parameters, such as volumes or masses, for each component. The parameter file contains the information on TMI-2 relevant to the plant geometry, system performance, controls, and initial conditions used to perform these benchmark calculations. The input deck defines the operator actions and boundary conditions during the course of the accident. The TMI-2 accident analysis provided good estimates of the accident output data compared with the OECD TMI-2 standard reference. The alternative scenario has proposed the initial event as a loss of main feed water and a small break on the hot leg. Analysis is in progress along with a sensitivity study concerning the break size and elevation.
ONE-DIMENSIONAL ANALYSIS OF THERMAL STRATIFICATION IN THE AHTR COOLANT POOL
Zhao, Haihua ; Peterson, Per F. ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 953~968
DOI : 10.5516/NET.2009.41.7.953
It is important to accurately predict the temperature and density distributions in large stratified enclosures both for design optimization and accident analysis. Current reactor system analysis codes only provide lumped-volume based models that can give very approximate results. Previous scaling analysis has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by jets modeled using integral techniques. This allows very large reductions in computational effort compared to three-dimensional CFD simulation. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was developed to implement such ideas. This paper summarizes major models for the BMIX++ code, presents the two-plume mixing experiment simulation as one validation example, and describes the codes' application to the liquid salt buffer pool system in the AHTR (Advanced High Temperature Reactor) design. Three design options have been simulated and they exhibit significantly different stratification patterns. One of design options shows the mildest thermal stratification and is identified as the best design option. This application shows that the BMIX++ code has capability to provide the reactor designers with insights to understand complex mixing behavior with mechanistic methods. Similar analysis is possible for liquid-metal cooled reactors.
PREDICTIONS OF CRITICAL HEAT FLUX USING THE ASSERT-PV SUBCHANNEL CODE FOR A CANFLEX VARIANT BUNDLE
Onder, Ebru Nihan ; Leung, Laurence Kim-Hung ; Rao, Yanfei ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 969~978
DOI : 10.5516/NET.2009.41.7.969
The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced
reactor fuel bundle. Based primarily on the
fuel bundle, several geometry changes (such as element sizes and pitch-circle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures.
NEUTRON-INDUCED CAVITATION TENSION METASTABLE PRESSURE THRESHOLDS OF LIQUID MIXTURES
Xu, Y. ; Webster, J.A. ; Lapinskas, J. ; Taleyarkhan, R.P. ;
Nuclear Engineering and Technology, volume 41, issue 7, 2009, Pages 979~988
DOI : 10.5516/NET.2009.41.7.979
Tensioned metastable fluids provide a powerful means for low-cost, efficient detection of a wide range of nuclear particles with spectroscopic capabilities. Past work in this field has relied on one-component liquids. Pure liquids may provide very good detection capability in some aspects, such as low thresholds or large radiation interaction cross sections, but it is rare to find a liquid that is a perfect candidate on both counts. It was hypothesized that liquid mixtures could offer optimal benefits and present more options for advancement. However, not much is known about radiation-induced thermal-hydraulics involving destabilization of mixtures of tensioned metastable fluids. This paper presents results of experiments that assess key thermophysical properties of liquid mixtures governing fast neutron radiation-induced cavitation in liquid mixtures. Experiments were conducted by placing liquid mixtures of various proportions in tension metastable states using Purdue's centrifugally-tensioned metastable fluid detector (CTMFD) apparatus. Liquids chosen for this study covered a good representation of both thermal and fast neutron interaction cross sections, a range of cavitation onset thresholds and a range of thermophysical properties. Experiments were devised to measure the effective liquid mixture viscosity and surface tension. Neutron-induced tension metastability thresholds were found to vary non-linearly with mixture concentration; these thresholds varied linearly with surface tension and inversely with mixture vapor pressure (on a semi-log scale), and no visible trend with mixture viscosity nor with latent heat of vaporization.