• Journal of KSNVE
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• v.6 no.6
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• pp.827-834
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• 1996

The effect of pipe breaks in the secondary side is investigated as a part of the fuel assembly qualification program. Using the detailed dynamic analysis of a reactor core, peak responses for the motions induced from pipe breaks are obtained for a detailed core model. The secondary side pipe breaks such as main steam line and economizer feedwater line braksare considered because leak-before-break methodology has provided a technical basis for the elimination of double ended guillotine breaks of all high energy piping systems with a diameter of 10 inches or over in the primary side from the design basis. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are carefully investigated. Also, the stress analysis is performed and the effect of the secondary side pipe breaks on the fuel assembly structural integrity under the faulted condition is addressed.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.622-627
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• 1996

Detailed analysis of Feedwater Line Break (FLB) event for the fuel failure point of view are lack because the event was characterized as the increase in reactor coolant system (RCS) pressure. Up to now, the potential of the rapid system heatup case has been emphasized and comprehensively studied. The cooldown effects of FLB event is considered to be bounded by the Steam Line Break (SLB) event since the cooldown effect of SLB event is larger than that of the FLB event. This analysis provides a new possible path which can cause the fuel failure. The new path means that the fuel failure can occur under the heatup scenario because the Pressurizer Safety Valves (PSVs) open before the reactor trips. The 1000 MWe typical C-E plant FLB event assuming Loss of Offsite Power (LOOP) at the turbine trip has been analyzed as an example and the results show less than 1% of the fuel failure. The result is well within the acceptance criteria. In addition to that, a study was accomplished to prevent the fuel failure for the heatup scenario case as an example. It is found that giving the proper pressure gap between High Pressurizer Pressure Trip (HPPT) analysis setpoint and the minimum PSV opening pressure could prevent the fuel failure.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.93-100
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• 2003

An estimation methodology of the mass and energy (M/E) release due to the main steam line break (MSLB) has been developed with the RETRAN-3D code. In the case of equipment qualification (EQ), the over-estimated temperature would exceed the design limits of some cables or valves. In order to have a more flexible EQ profiles from the MSLB M/E release, the methodology with the best-estimated code was used. The major conditions affecting the MSLB M/E were found to be the initial SG level, heat transfer between primary and secondary sides, power level, operable protection system, main or auxiliary feedwater availability, and break conditions. The RETRAN-3D models were developed for the Kori unit 1 (KRN-1) which is typical two loop Westinghouse (WH) designed plant. Particularly, a detailed model of the steam generators was developed to estimate a more realistic two-phase heat transfer effect of the steam flow. After the modeling, the methodology has been developed through the sensitivity analyses. The M/E release data generated from the analyses have been used as the input to the inside containment pressure and temperature (P/T) analysis. According to the results at the point of view containment P/T, the Kori unit 1 can have more margin of 5∼15 ㎪ in pressure and 8∼15$^{\circ}C$ in temperature.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.85-90
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• 1996

In this paper, the self-organizing fuzzy logic controller is developed for water level control of steam generator. In comparison with conventional fuzzy logic controllers, this controller performs control task with no control rules at initial and creates control rules as control behavior goes on, and also modifies its control structure when uncertain disturbance is suspected. Selected parameters in the fuzzy logic controller are updated on-line by the gradient descent loaming algorithm based on the performance cost function. This control algorithm is applied to water level control of steam generator model developed by Lee, et al. The computer simulation results confirm good performance of this control algorithm in all power ranges. This control algorithm can be expected to be used for automatic control of feedwater control system in the nuclear power plant with digital instrumentation and control systems.

• Proceedings of KOSOMES biannual meeting
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• 2009.06a
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• pp.177-178
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• 2009

Hydraulic Transients would be occurred since pressure is increased or decreased when water speed inside of pipeline is rapidly changed A study on water hammer has become more important because the pumping stations were big and the systems conveying the fluid through the large and long transmission pipelines were complex. In this study, the method of characteristic line was adopted to evaluate the valve-induced water hammer phenomena in a pumps feedwater system.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.617-621
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• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.235-241
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• 1995

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor Power Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1173-1183
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• 2018

Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

• Progress in Superconductivity and Cryogenics
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• v.23 no.3
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• pp.14-19
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• 2021

The reduction of carbon dioxide emissions becomes a global issue, the main source of carbon dioxide emissions in the Asian region is the energy conversion sector, especially coal-fired power plants. We are working to develop technologies that will at least limit the increase in carbon dioxide emissions from the thermal power plants as one way to reduce carbon dioxide emissions. Our research aims to reduce carbon dioxide emissions by removing iron oxide scale from the feedwater system of thermal power plants using a superconducting high-gradient magnetic separation (HGMS) system, thereby reducing the loss of power generation efficiency. In this paper, the background of thermal power plants in Asia is outlined, followed by a case study of the introduction of a chemical cleaning line at an actual thermal power plant in Japan, and the possibility of introducing it into the thermal power plants in China based on the results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.11-21
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• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.