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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 36, Issue 6 - Dec 2004
Volume 36, Issue 5 - Oct 2004
Volume 36, Issue 4 - Aug 2004
Volume 36, Issue 3 - Jun 2004
Volume 36, Issue 2 - Apr 2004
Volume 36, Issue 1 - Feb 2004
Selecting the target year
Determination of Performance Indicator Thresholds Based on Typical PSA Results
Kang, Dae-Il ; Kim, Kil-Yoo ; Hwang, Mee-Jung ; Sung, Key-Yong ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 485~496
Typical probabilistic safety assessment (PSA) results were used to estimate the performance indicator (PI) thresholds of unplanned reactor scram (URS) and safety system unavailability (SSU) for Korean nuclear power plants (NPPs). The changes in core damage frequency (
were adopted as the risk criteria in setting up the PI thresholds. The PI thresholds for the URS were estimated using information pertaining to the initiating event frequencies, the CDF, and the CDF contribution of each initiating event. The PI thresholds of the SSU were estimated using information on the unavailability, the Fussell-Vesely importance, and the CDF.
The Effect of an Aggressive Cool-Down Following A Refueling Outage Accident in which a Pressurizer Safety valve is Stuck Open
Lim, Ho- Gon ; Park, Jin-Hee ; Jang, Seung-Cheol ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 497~511
A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.
Nonlinear Finite Element Analysis of Containment Vessel by Considering the Tension stiffening Effect
Lee, Hong-Pyo ; Choun, Young-Sun ; Seo, Jeong-Moon ; Shin, Jae-Chul ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 512~527
This paper describes the finite element (FE) analysis results of a 1/4 scale model of a prestressed concrete containment vessel (PCCV) by considering the tension stiffening effect, which is a result of the bond effect between the concrete and the steel. The tension stiffening model is assumed to be an exponential form based on the relationship between the average stress and the average strain of the concrete. The objective of the present FE analysis is to evaluate the ultimate internal pressure capacity of the PCCV, as well as its failure mechanism, when the PCCV model is subjected to a monotonous internal pressure beyond is design pressure capacity. With the commercial code ABAQUS, the FE analysis used two concrete failure criteria: a 2-dimensional axi-symmetric model with modified Drucker-Prager failure criteria and a 3-dimensional model with a damaged plasticity mod디. The results of our FE analysis on the ultimate pressure capacity and failure modes of PCCV have a good agreement with the experimental data.
On-line Generation of Three-Dimensional Core Power Distribution Using Incore Detector Signals to Monitor Safety Limits
Jang, Jin-Wook ; Lee, Ki-Bog ; Na, Man-Gyun ; Lee, Yoon-Joon ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 528~539
It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.
Development of a Diagnostic Algorithm with Acoustic Emission Sensors and Neural networks for Check Valves
Seong, Seung-Hwan ; Kim, Jung-Soo ; Hur, Seop ; Kim, Jung-Tak ; Park, Won-Man ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 540~548
Check valve failure is one of the worst problems in nuclear power plants. Recently, many researches have been based on new technology using accelerometers and ultrasonic and magnetic flux detection have been carried out. Here, we have suggested a method that uses acoustic emission sensors for detecting the failures of check valves through measuring and analyzing backward leakage flow, a system that works without disassembling the check valve. For validating the suggested acoustic emission sensor methodology, we designed a hydraulic test loop with a check valve. We have assumed in this study that check valve failure is caused by disk wear or by the insertion of a foreign object. In addition, we have developed diagnostic algorithms by using a neural network model to identify the type and size of the failure in the check valve. Our results show that the proposed diagnostic algorithm with acoustic emission sensors is a good solution for identifying check valve failure without necessitating any disassembly work.
Application of Electromagnetic Fields to Improve the Removal Rate of Radioactive Corrosion Products
Kong, Tae-Young ; Lee, Kun-Jai ; Song, Min-Chul ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 549~558
TTo comply with increasingly strict regulations for protection against radiation exposure, many nuclear power plants have been working ceaselessly to reduce and control both the radiation sources within power plants and the radiation exposure experienced by operational and maintenance personnel. Many research studies have shown that deposits of irradiated corrosion products on the surfaces of coolant systems are the main cause of occupational radiation exposure in nuclear power plant. These corrosion product deposits on the fuel-clad surface are also known to be main factors in the onset of axial offset anomaly (AOA). Hence, there is a great deal of ongoing research on water chermistry and corrosion processes. In this study, a magnetic filter with permanent magnets was devised to remove the corrosion products in the coolant stream by taking advantage of the magnetic properties of the corrosion products demonstrated a removal efficiency of over 90% for particles above 5
. This finding led to the construction of an electromagnetic device that causes the metallic particulates to flocculate into larger aggregates of about 5
in diameter by using a novel application of electromagnetic flocculation on radioactive corrosion products.
3-D CFD Analysis of the CANDU-6 Moderator Circulation Under Nnormal Operating Conditions
Yoon, Churl ; Rhee, Bo-Wook ; Min, Byung-Joo ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 559~570
A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard
turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is
at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is
A Coherent Methodology for the Evaluation of a Steam Explosion Load Using TEXAS-V
Song, Jin-Ho ; Park, Ik-Kyu ; Kim, Jong-Hwan ;
Nuclear Engineering and Technology, volume 36, issue 6, 2004, Pages 571~581
A methodology is proposed for the evaluation of a steam explosion load on a reactor scale by evaluating the steam explosion model against the experimental data. Being part of the OECD/SERENA program,, appropriate data was selected by international experts and the analytical model of TEXAS-V was chosen. The procedure consists of two steps. the pre-mixing model was verified against the FARO L-14 and FARO L-28 data. The explosion model was verified against the experimental data of KROTOS-44, FARO L-33, TROI-13, and TROI-34. The capabilities and deficiencies of the fundamental models of the TEXAS-V are reviewed in terms of their adequacy in a simulation of steam explosion on a reactor scale.