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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
Editor in Chief :
Volume & Issues
Volume 27, Issue 6 - Dec 1995
Volume 27, Issue 5 - Aug 1995
Volume 27, Issue 4 - Aug 1995
Volume 27, Issue 3 - Jun 1995
Volume 27, Issue 2 - Apr 1995
Volume 27, Issue 1 - Feb 1995
Selecting the target year
An Analysis of the Loss of Residual Heat Removal System Event for Pressurized Water Reactor at Reduced Inventory Operation
Han, Kee-Soo ; Song, Jin-Ho ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 645~660
The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.
Automatic Stair-Climbing Algorithm of the Planetary Wheel Type Mobile Robot in Nuclear Facilities
Kim, Byung-Soo ; Kim, Seung-Ho ; Lee, Jongmin ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 661~669
A mobile robot, named KAEROT, has been developed for inspection and maintenance operations in nuclear facilities. The main feature of locomotion system is the planetary wheel assembly with small wheels. This mechanism has been designed to be able to go over the stairs and obstacles with stability. This paper presents the inverse kinematic solution that is to be operated by remote control. The automatic stair climbing algorithm is also proposed. The. proposed algorithms generates the moving pathes of small wheels and calculates the angular velocity of 3 actuation wheels. The results of simulations and experiments are given for KAEROT peformed on the irregular stairs in laboratory. It is shown that the proposed algorithm provides the lower inclination angle of the robot body and increases its stability during navigation.
Real-Time Diagnosis of Incipient Multiple Faults with Application for Kori Nuclear Power Plant
Chung, Hak-Yeong ; Zeungnam Bien ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 670~686
This paper provides an improvement on our previous study ［1］ for multi-fault diagnosis in real time in large-scale systems. In the method, fault propagation probability(FPP) and fault propagation time(FPT) in a fuzzy sense are additively used to describe the fault propagation model(FPM) in more practical manner. A modified fault diagnosis procedure is also given. This method is applied for diagnosis of the primary system in the Kori nuclear power plant unit 2 under a transient condition in case of unit value of FPP on each branch of the FPM.
Transient Analysis of the CANDU-9 480/SEU Reactor
J. C. Shin ; Park, J. H. ; K. N. Han ; H. C. Suk ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 687~700
The thermal-hydraulic transient analysis of the proposed CANDU-9 plant was peformed. Several major transients ore analyzed if they meet the heat transport system design requirements. The proposed heat transport system configuration and the preliminary sizes of system equipment are justified by analysis in terms of the fuel integrity and the high system pressure limit during transients. The compliance with AECB R-77 requirements for CANDU-9 reactor was estimated. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements and the fuel overheating is prevented. One pump start-up during the reactor start-up operation was analyzed to investigate the How reversal through the fuel channel, which is specific in the proposed CANDU-9 plant
Quantification of Realistic Discharge Coefficients for the Critical Flow Model of RELAP5/MOD3/KAERl
T. S. Kwon ; B. D. Chung ; Lee, W. J. ; Lee, N. H. ; J. Y. Huh ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 701~709
The realistic discharge coefficient for the critical How model of RELAP5/AOD3/KAERI are determined for the subcooled and too-phase critical flow by assessments of nine MARVIKEN Critical flew Test(CFT). The selected test runs include a high initial subcooling and large nozzle aspect rat-io(L/D). The code assessment results show that RELAP5/MOD3/KAERI over-predicts the subcooled critical flow and under-predicts the two-phase critical flow. Using these result, the realistic discharge coefficients of critical flow models are quantified by an iterative method. The realistic discharge coefficients are determined to be 0.89 for the subcooled critical How and 1.07 for the two-phase critical flow, and the associated standard deviations are 0.0349 and 0.1189, respectively. The results obtained from this study can be applied to calculate the realistic system response of Large Break Loss of Coolant Accident and to evaluate the realistic Emergency Core Cooling System performance.
A Theoretical Study on the Fluid-Structure Interaction Due to the Pump in the Pressurized Water Reactor
Lee, Kye-Bock ; Jong Ryul park ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 710~720
The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.
Optimization of the Korean Nuclear Fuel Cycle Using Linear Programming
Kim, J. I. ; K. N. Chae ; Lee, B. W. ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 721~729
The Korean optimal nuclear fuel cycle strategy from the year 2000 to 2030 is derived using linear programming. The fuel cycle cost, the cost uncertainty, and the natural uranium consumption are used as the criteria for the optimization. These objectives are compromised by fuzzy decision-making technique which maximizes the minimum degree of satisfaction among the three objectives. The options for the back-end fuel cycle are direct disposal, reprocessing, and DUPIC. The optimal fuel cycle strategy of Korea is to start reprocessing in around 2010 and increase its capacity with the maximum of 800 tHM in around 2025, and to star DUPIC processing in 2025. The cot uncertainty and the natural uranium consumption of the optimal fuel cycle strategy are reduced by 7.1％ and 6.1％, respectively, at the cost penalty of 5.4％ compared with the cost-only optimal solution.
The LQG/LTR Dynamic Digital Control System Design for the Nuclear Steam Generator Water Level
Lee, Yoon-Joon ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 730~742
The steam generator feedwater and level control system is designed by two steps of the feedwater control design and the feedback loop controller design. The feedwater sen system is designed by the optimal LQR/LQG approach and then is modified by the LTR method to recover the robustness. The plant characteristics are subject to change with the power variation and these dynamic properties are considered in the design of the feedback controller. All the designs are made in the continuous domain and are digitalized by applying the proper sampling period. The system is simulated for the two cases of power increase and decrease. From the results of simulation, it is found that the controller constants would rather be invariable during the power increase, while for the case of power decrease they should be changed with the power variation to keep the system stability.
Fuel Assembly Modelling for Dynamic Analysis of Reactor Internals and Core
Jhung, Myung-Jo ; Hwang, Jong-Keun ; Kim, Yeon-Seung ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 743~752
This paper investigates the effects of fuel groupings in the coupled internals and core model on the internals and fuel responses due to pipe breaks. The 177 fuel assemblies for Korean standard nuclear power plant are grouped into several stick models and the responses of internals components are calculated. The analysis results show that the fuel model groupings in the coupled internals and core model have no significant effects on the internals and fuel responses for pipe break excitation. Also, in order to determine the feasibility of constructing a single equivalent stick representation of In or more adjacent fuel bundles, the reduced models, each of which employs a different stiffness lumping rule, are constructed. It is shown that the equivalent stiffness calculated to get the first natural frequency of the original model while preserving net gap between grouping centers gives the minimum modelling error.
A Quantitative Evaluation of Chemical and Volume Control System Design Simplification
Son, Han-Seong ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 753~759
One of the important features of the advanced nuclear power plants is the system simplification. In this work, a model has been introduced to quantitatively evaluate the system simplification. A few models have been developed for quantitative evaluation of design simplification and the design enhancements of CVCS of the advanced reactors have been evaluated with models based on the entropy concept and the system availability. In addition, operational interface of CVCS with peripheral systems has been considered to develop a new evaluation model in this work. The quantification results for the design of the System 80+ and KSNPP indicate that the simplicity of the CVCS is primarily dependent on the type and number of charging pumps.
Elastic Stiffness Analysis of Leaf Type Holddown Spring Assemblies
Lim, Hyun-Tae ; Kim, Jae-Won ; Song, Kee-Nam ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 760~766
A general method is proposed for elastic stiffness analysis of the leaf type holddown springs using only the geometric data and Young's modulus of the springs. In this method, an engineering beam theory and Castigliano's theory are applied to elastic stiff analysis of the leaf type holddown springs. To show reliability and effectiveness of this method, the elastic stiffness from the proposed method is compared with test result and from the comparison, the unposed method has been proven to be effective for estimating the elastic stiffness of the leaf springs.
Trial Burns of Low-Level Radioactive Wastes the Demonstration-Scale Incineration Plant at KAERI
Yang, Hee-Chul ; Kim, In-Tae ; Kim, Jeong-Guk ; Kim, Joon-Hyung ; Seo, Yong-Chil ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 767~774
Behavior of radionuclides such
Cs in the incineration Process was Studied by trial burns of simulated wastes with radio-isotope tracers. Behavior of nonvolatiles,
Mn, was similar to that of particulate matters in the process. Decontamination factors(DFs) for
Mn were 4.7
, respectively. Behavior of semivolatile radio-isotope,
Cs, was temperature dependent. DFs for
Cs at In different incineration temperature of 85
, respectively. Trial bums of dry active waste(DAW) transported from nuclear power station(NPS) Kori 3,4 were also performed. DF for gross
radioactivity in DAW was 1.1
. This was a little higher than the estimated value, which was calculated from the tracer test results and nuclides distribution in the DAW. Average emission concentration was 0.019 Bq/N
, which could meet the maximal permissible concentration(MPC) in stack emission.n.
Safety Margin Improvement Against Failure of Zr-2.5Nb Pressure Tube
Jeong, Yong-Hwan ; Kim, Young-Suk ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 775~783
This study is to assess the effects of increasing wall thickness on the safety margin of pressure tube in operating and of lowering initial hydrogen concentration on the DHC growth in respect to the improvement of the reliability of pressure tube in CANDU reactors. The pressure tube with thicker wall of 5.2 mm shows much higher safety margin for flaw tolerance by 25％ than the current 4.2mmm tube. The thicker pressure tubes have a great benefit in LBB assessment including the initial crack depth at which DHC occurs, the crack length at onset of leaking and the available time for action. The resistance for the pressure tube ballooning at LOCA accident is also increased with the thicker tube. The calculations for Heq concentration after 20 years of operation as a function of wall thickness and initial hydrogen concentration show that the 5.2 mm nil thickness tube with 5 ppm initial hydrogen concentration is the most resistant to DHC. with the lower initial hydrogen concentration, TSS temperature for the precipitation or hydride decreases and the crack growth during cooldown reduces.
A Study on Electronic Circuit for Liwe-Time Correction in Multi-Channel Analyzer : Survey and Analysis
I. K. Hwang ; K. H. Kwon ; S. J. Song ;
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 784~791
This paper describes the counting-loss problem for radiation measurement Multi-channel analyzers and spectrometers adopt various techniques for compensation for counting-losses in process-ing the radiation pulses from a detector. Researchers hate tried to seek the best solution for the problem. However, any absolute solution has not been reached and vendors of radiation instruments use their own algorithms individually. This survey explains the various compensation algorithms with electronic implementation approach. Shortcomings and merits of each algorithm are also reviewed and a direction is suggested of the recommendable development strategy for counting-loss compensation.
발전용원자로 안전규제기술요건 개발
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 792~802
핵 연료집합체 Holddown Spring 특성해석
Nuclear Engineering and Technology, volume 27, issue 5, 1995, Pages 803~810