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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 30, Issue 6 - Dec 1998
Volume 30, Issue 5 - Oct 1998
Volume 30, Issue 4 - Aug 1998
Volume 30, Issue 3 - Jun 1998
Volume 30, Issue 2 - Apr 1998
Volume 30, Issue 1 - Feb 1998
Selecting the target year
Improvement of Liquid Droplet Entrainment Model in the COBRA-TF Code
Ha, Kwi-Seok ; Jeong, Jae-Jun ; Sim, Suk-Ku ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 181~193
The COBRA-TF liquid droplet entrainment models have been assessed and improved through various experiments. The COBRA-TF code uses the Wurtz entrainment model in the film mist flow regime and the mechanistic model based on the critical Weber number and critical vapor velocity in the hot wall flow regimes, respectively. The Wurtz model has been replaced with the modified Sugawara model. The assessment against the experiments by Hewitt, Keeys, Yanai, and Whalley showed the modified Sugawara model better predicts the steam-water as well as the air-water experiments for the film mist flow regime. For hot wall flow regime, the COBRA-TF entrainment model was modified using two methods, one with an increased critical Weber number and the other with the Yonomoto's critical vapor velocity model. The modified models were assessed using the FLECHT-SEASET bottom reflood tests. The results showed that the Yonomoto model best predicts the quenching time, whereas the local maximum rod temperature was not affected much.
Application of Dynamic Reliability Analysis Method to the CANDU Pressurizer System
Lee, Sook-Hyung ; Oh, Se-Ki ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 194~201
DYLAM (Dynamic Logical Analytical Methodology) and its related methodologies are reviewed and found to have many favorable characteristics. Previous studies have shown that the DYLAM methodology represents an appropriate tool to study dynamic analysis. A hybrid model which is a synthesis of the DYLAM model, a system thermodynamic simulation model and a neural network predicative model, is implemented and used to analyze dynamically the CANDU pressurizer system. This study demonstrates that the hybrid model for system reliability analyses is effective.
A Study on the Improvement of Stress Field Analysis in a Domain Composed of Dissimilar Materials
Song, Kee-Nam ; Lee, Jin-Seok ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 202~211
Interfacial stresses at two-material interfaces and initial displacement field over the entire domain are obtained by modifying the potential energy functional with a penalty function, which enforces continuity of the stresses at the interface of two materials. Based on the initial displacement field and interfacial stresses, a new methodology to generate a continuous stress field over the entire domain has been proposed by combining the modified projection method of stress-smoothing and Loubignac's iterative method of improving the displacement field. Stress analysis is carried out on two examples made of dissimilar materials : one is a two-material cantilever composed of highly dissimilar materials and the other is a zirconium-lined cladding tube made of slightly dissimilar materials. Results of the analysis show that the proposed method provides an improved continuous stress field over the entire domain, and accurately predicts the nodal stresses at the interface, while the conventional displacement-based finite element method produces significant stress discontinuities at the interface. In addition, the total strain energy evaluated from the improved continuous stress field converges to the exact value in a few iterations.
Structural Integrity of PWR Fuel Assembly for Earthquake
Jhung, M.J. ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 212~221
In the present study, a method for the dynamic analysis of a reactor core is developed. Peak responses for the motions induced from earthquake are obtained for a core model. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are investigated. Prediction of fuel assembly stress during an earthquake requires development of a fuel assembly stress analysis model capable of interfacing with the models and results discussed in the dynamic analysis of a reactor core. This analysis uses beam characteristics which describe the overall fuel assembly response. The stress analysis method and its application for the case of an increased seismic level are also presented.
On-line Estimation of DNB Protection Limit via a Fuzzy Neural Network
Na, Man-Gyun ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 222~234
The Westinghouse OT
T DNB protection logic heavily restricts the operation region by applying the same logic for a full range of operating pressure in order to maintain its simplicity. In this work, a fuzzy neural network method is used to estimate the DNB protection limit using the measured average temperature and pressure of a reactor core. Fuzzy system parameters are optimized by a hybrid learning method. This algorithm uses a gradient descent algorithm to optimize the antecedent parameters and a least-squares algorithm to solve the consequent parameters. The proposed method is applied to Yonggwang 3&4 nuclear power plants and the proposed method has 5.99 percent larger thermal margin than the conventional OT
T trip logic. This simple algorithm provides a good information for the nuclear power plant operation and diagnosis by estimating the DNB protection limit each time step.
Use of MAAP in Generating Accident Source Term Parameters
Kim, Jong-Wok ; Yun, Joeng-Ik ; Kang, Chang-Sun ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 235~244
The parametric model method determines the accident source term which is Presented by a set of source term parameters. In this method, the cumulative distribution of each source term parameter should be derived for its uncertainty analysis. This paper introduces a method of generating the parameters in the form of cumulative distribution using MAAP version 4.0. In MAAP, there are model parameters which could incorporate uncertain physical and/or chemical phenomena. In general, the model parameters do not have a point value but a range. In this paper, considering that, the input values of model parameters influencing each parameter are sampled using LHS. Then, the computation results are shown in cumulative distribution form. For a case study, the CDFs of FCOR and WES of Kori Unit 1 are derived. The target scenarios for the computation are the ones whose initial events are large LOCA, small LOCA and transient, respectively. It is found that the computed CDF's in this study are consistent to those of NUREG-1150 and the use of MAAP is proven to be adequate in assessing the parameters of the severe accident source term.
A Study on Validation of Variable Aperture Channel Model: Migration Experiments of Conservative Tracer in Parallel and Wedge-Shaped Fracture
Keum, D.K. ; Hahn, P.S. ; Vandergraaf, T.T. ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 245~261
In order to validate the variable aperture channel model that can deal with the non-uniform How rate in flow domain, migration experiments of conservative tracer were performed in two artificial fractures, a parallel and a wedge-shaped fracture. These different fracture shapes were designed to give different flow pattern. The fractures were made from a transparent acrylic plastic plate and a granite slab with dimensions of 10
61 cm. Uranine (Fluorescein sodium salt) was used as a conservative tracer. The volumetric flow rates of uranine feed solution were 30 mL/ hr, giving a mean residence time in the fracture of approximately 24 hours for the parallel fracture and 34 hours for the wedge-shaped fracture. The migration plumes of uranine were photographed to obtain profiles in space and time for movement of a tracer in fractures. The photographed migration plume was greatly affected by the geometric shape of fractures. The variable aperture channel model could have predicted the experimental results for the parallel fracture with a large accuracy. It is expected that the variable aperture channel model would be effective to predict the transport of the contaminant, especially, with the flow rate variation in a fracture.
A Review of Pressure Tube Failure Accident in the CANDU Reactor and Methods for Improving Reactor Performance
Yoo, Ho-Sik ; Chung, Jin-Gon ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 262~272
The experiences and causes of pressure tube cracking accidents in the CANDU reactors and the development of the fuel channel at AECL(Atomic Energy Canada Limited) have been described. Most of the accidents were caused by Delayed Hydride Cracking(DHC). In the cases of the Pickering units 3&4 and the Bruce unit 2, excessive residual stresses induced by an improper rolled joint process played a role in DHC. In the Pickering unit 2, cracks formed by contact between the pressure and calandria tubes due to the movement of the garter spring were the direct cause of the failure. To extend the life of a fuel channel, several R&D programs examining each component of the fuel channel have been carried out in Canada. For a pressure tube, the main concern is focused on changing the fabrication processes, e.g., increasing cold working rate, conducting intermediate annealing and adding a third element like Fe, V, and Cr to the tube material. In addition to them, chromium plating on the end fitting and increasing wall thickness at both ends of the calandria tube are considered. There has also been much interest in the improvement of fuel channel performance in our country and several development programs are currently under way.
Analysis of High Burnup Fuel Behavior Under Rod Ejection Accident in the Westinghouse-Designed 950 MWe PWR
Chan Bock Lee ; Byung Oh Cho ;
Nuclear Engineering and Technology, volume 30, issue 3, 1998, Pages 273~286
As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident(RIA) may occur at the energy lower than the expected, fuel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod turnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the conventional zero dimensional analysis methodology and the fraction of fuel failure in the core is less than 4 %. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied.