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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 33, Issue 6 - Dec 2001
Volume 33, Issue 5 - Oct 2001
Volume 33, Issue 4 - Aug 2001
Volume 33, Issue 3 - Jun 2001
Volume 33, Issue 2 - Apr 2001
Volume 33, Issue 1 - Feb 2001
Selecting the target year
Fault-tolerance Performance Evaluation of Fieldbus for NPCS Network of KNGR
Jung, Hyun-Gi ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 1~11
In contrast with conventional fieldbus researches which are focused merely on real-time performance, this study aims to evaluate the real-time performance of the communication system including fault-tolerant mechanisms Maintaining performance in presence of recoverable faults is very important in case that the communication network is applied to a highly reliable system such as next generation Nuclear. Power. Plant (NPP). If the tie characteristics meet the requirements of the system, the faults will be recovered by fieldbus recovery mechanisms and the system will be safe. If the time characteristics can not meet the requirements, the faults in the fieldbus can propagate to the system failure. In this study, for the purpose of investigating the time characteristics of fieldbus, the recoverable faults are classified and then the formulas that represent delays including recovery mechanisms are developed. In order to validate the proposed approach, we have developed a simulation model that represents the Korea Next Generation Reactor (KNGR) NSSS Process Control System (NPCS). The results of the simulation show us the reasonable delay characteristics of the fault cases with recovery mechanisms. Using the simulation results and the system requirements, we also can calculate the failure propagation probability from fieldbus to outer system.
Evaluation of Post-LOCA Long Term Cooling Performance in Korean Standard Nuclear Power Plants
Bang, Young-Seok ; Jung, Jae-Won ; Seul, Kwang-Won ; Kim, Hho-Jung ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 12~24
The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.
Pin Power Reconstruction of HANARO Fuel Assembly via Gamma Scanning and Tomography Method
Seo, Chul-Gyo ; Park, Chang-Je ; Cho, Nam-Zin ; Kim, Hark-Rho ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 25~33
To determine the pin power distribution without disassembling, HANARO fuel assemblies are gamma-scanned and then the distribution is reconstructed tv using the tomography method. The iterative least squares method (ILSM and the wavelet singular value decomposition method (WSVD) are chosen to solve the problem. An optimal convergence criterion is used to stop the iteration algorithm to overcome the potential divergence in ILSM. WSVD gives better results than ILSM , and the average values from the two methods give the best results. The RMSE (root mean square errors) to the reference data are 5.1, 6.6, 5.0, 6.5, and 6.4% and the maximum relative errors are 10.2, 13.7, 12.2, 13.6, and 14.3%, respectively. It is found that the effect of random positions of the pins is important. Although the effect can be accommodated by the iterative calculations simulating the random positions, the use of experimental equipment with a slit covering the whole range of the assembly horizontally is recommended to obtain more accurate results. We made a new apparatus using the results of this study and are conducting an experiment in order to obtain more accurate results.
Development of Performance Analysis System (NOPAS) for Turbine Cycle of Nuclear Power Plant
Kim, Seong-Kun ; Park, Kwang-Hee ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 34~45
We have needs to develop a performance analysis system that can be used in domestic nuclear power plants to determine performance status of turbine cycle. We developed new NOPAS system to aid performance analysis of turbine cycle . Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). Robustness in the performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful aids for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis reports.
Evaluation of Neutron Cross Sections of Dy Isotopes in the Resonance Region
Oh, Soo-Youl ; Gil, Choong-Sup ; Jonghwa Chang ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 46~61
The neutron cross sections of
Dy have been evaluated in the resonance region of which upper energy is set to several tens of keV. The cross sections are formulated with resonance parameters in the energy region under consideration. In the resolved resonance region, the positive-energy resonance parameters were adopted from the BNL compilation published in 1984 with slight, if any, modifications. A bound level resonance for each isotope except
Dy was invoked to reproduce the reference 2200 m/s cross sections and the bound coherent scattering length. Subsequently, the statistical behavior of the resolved resonance parameters was analyzed, and thus obtained s-wave average parameters were adopted in the unresolved resonance region. In addition, recent measurements of the capture cross sections in the unresolved region were taken into account in adjusting the average resonance parameters for high orbital angular momentum resonances. The present evaluation resulted in large improvements in the cross sections over the ENDF/B-Vl release 6.6.
Numerical Simulations of the Moisture Movement in Unsaturated Bentonite Under a Thermal Gradient
Park, J.W. ; K. Chang ; Kim, C.L. ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 62~72
The one-dimensional finite element program was developed to analyze the coupled behavior of heat, moisture, and air transfer in unsaturated porous media. By using this program, the simulation results were compared with those from the laboratory infiltration tests under isothermal condition and temperature gradient condition, respectively. The discrepancy of water uptake was found in the upper region of a bentonite sample under isothermal condition between numerical simulation and laboratory experiment. This indicated that air pressure was built up in the bentonite sample which could retard the infiltration velocity of liquid. In order to consider the swelling phenomena of compacted bentonite which cause the discrepancy of the distribution of water content and temperature, swelling and shrinkage factors were incorporated into the finite element formulation. It was found that these factors could be effective to represent the moisture diffusivity and unsaturated hydraulic conductivity due to volume change of bentonite sample.
Verification Test and Model Updating for a Nuclear Fuel Rod with Its Supporting Structure
H. S. Kang ; K. N. Song ; Kim, H. K. ; K. H. Yoon ; Y. H. Jung ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 73~82
Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.
Evaluation of 0ff-gas Characteristics in Vitrification Process of ion-Exchange Resin
Park, S. C. ; Kim, H. S. ; K. H. Yang ; C. H. Yun ; T. W. Hwang ; S. W. Shin ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 83~92
The properties of off-gas generated from vitrification process of ion-exchange resin were characterized. Theoretical composition and flow rate of the off-gas were calculated based on chemical composition of resin and it＇s burning condition inside CCM. The calculated off-gas flow rate was 67.9Nm
/h at the burning rate of 40kg/h. And the composition of off-gas was avaluated as
(13.3%), NO(3.6%), and SO
(1.6%) in order. Then, actual flow rate and composition of off-gas were measured during pilot-scale demonstration tests and the results were compared with theoretical values. The actual flow rate of off-gas was about 1.6 times higher than theoretical one. The difference between theoretical and actual flow rates was caused by the in-leakage of air to the system, and the in-leakage rate was evaluated as 36.3Nm
/h. Because of continuous change in the combustion parameters inside CCM, during demonstration tests, the concentration of toxic gases showed wide fluctuation. However, the concentration of CO, a barometer of incompleteness of combustion inside CCM, was stabilized soon. The result showed quasi-equilibrium state was achieved two hours after feeding of resin.
Nonlinear Dynamic Buckling Behavior of a Partial Spacer Grid Assembly
Yoon, Kyung-Ho ; Kang, Heung-Seok ; Kim, Hyung-Kyu ; Song, Kee-Nam ; Jung, Yeon-Ho ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 93~101
The spacer grid is one of the main structural components in the fuel assembly, which supports the fuel rods, guides cooling water, and protects the system from an external impact load, such as earthquakes. Therefore, the mechanical and structural properties of the spacer grids must be extensively examined while designing them. In this paper, a numerical method for predicting the buckling strength of spacer grids is presented. Numerical analyses on the buckling behavior of the spacer grids are performed for a various array of sizes of the grids considering that the spacer grid is an assembled structure with thin-walled plates and imposing proper boundary conditions by nonlinear dynamic finite element method using ABAQUS/Explicit. Buckling tests on several numbers of specimens of the spacer grid were also carried out in order to compare the results between the test and the simulation result. The drop test is accomplished by dropping a carriage on the specimen at a pre-determined position. From this test, the specimens are buckled only at the uppermost and the lowermost layer among the multi-cells, which is similar to the local buckling at the weakest point of the grid structure. The simulated results also similarly predicted the local buckling phenomena and were found to give good correspondence with the experimental values for the thin-walled grid structures.
A Model Predictive Controller for The Water Level of Nuclear Steam Generators
Na, Man-Gyun ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 102~110
In this work, the model predictive control method was applied to a linear model and a nonlinear model of steam generators. The parameters of a linear model for steam generators are very different according to the power levels. The model predictive controller was designed for the linear steam generator model at a fixed power level. The proposed controller at the fixed power level showed good performance for any other power levels by designed changing only the input-weighting factor. As the input-weighting factor usually increases, its relative stability does so. The steam generator has some nonlinear characteristics. Therefore, the proposed algorithm has been implemented for a nonlinear model of the nuclear steam generator to verify its real performance and also, showed good performance.
Dynamic Characteristics of the Integral Reactor SMART
Kim, Tae-Wan ; Park, Keun-Bae ; Jeong, Kyeong-Hoon ; Lee, Gyu-Mahn ; Park, Suhn ;
Nuclear Engineering and Technology, volume 33, issue 1, 2001, Pages 111~120
In this study, a dynamic analysis of the integral reactor SMART (System-integrated Modular Advanced ReacTor) under postulated seismic events is performed to review the response characteristics of the major components. To enhance the feasibility of an analysis model, a detailed finite element model is synchronized with the products of concurrent design activities. The artificial time history, which has been applied to the seismic analysis for the Korean Standard Nuclear Power Plant (KSNP), is chosen to envelop broad site specifics in Korea. Responses in the horizontal direction are found slightly amplified, while those in the vertical direction are suppressed. Since amplified response is monitored at the control element drive mechanism (CEDM), minor design provision is considered to enhance the integrity of the subsystem.