• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2003년도 유체기계 연구개발 발표회 논문집
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• pp.161-166
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• 2003

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of coolant pump power has been introduced in this study. Up to now, we did not found out a precedent which the coolant pump power is used for the real-time flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the real-time flow-rate calculation method by the measurement of coolant pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs, So, it has been concluded that it is possible to calculate the real-time flow-rate by the measurement of pump motor inputs. In addition, the compensation for a above new method can be made by HBM being now used in the commercial nuclear power plants.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2757-2772
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• 2023

The fluid retention effect of the Volume Control Tank (VCT) leads to a long time delay in Reactor Coolant System (RCS) concentration during the boration process. A coupled model combining a lumped-parameter sub-model and a computational fluid dynamics sub-model is currently used to investigate the concentration dynamic characteristic of RCS during the boration process. This model is validated by comparison with experimental data, and the predicted results show excellent agreement with experimental data. We provide detailed fields in VCT and concentration variations of RCS to study the interaction between mixing in VCT and the transient responses of RCS. Moreover, the impacts of the inlet flow rate, inlet nozzle diameter, original concentration, and replenishing temperature of VCT on the RCS concentration characteristic are studied. The inlet flow rate and nozzle diameter of VCT remarkably affect the RCS concentration characteristic. Too-large or too-small inlet flow rates and nozzle diameters will lead to unacceptable long delays. In this work, the optimal inlet flow rate and nozzle diameter of VCT are 5 m³/h and 58.8 mm, respectively. Besides, the impacts of the original concentration and replenishing temperature of VCT are negligible under normal operating conditions.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.181-184
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• 2014

Loose Parts Monitoring System(LPMS) monitors loosened or detached parts and foreign parts inside the pressure boundary of a reactor coolant system (RCS). It is difficult to discriminate valid signal from LPMS alarms at full power since the signal pattern by thermal shocks and structure friction are similar to those by loose metal impacts. In addition, It is more difficult to discriminate the impact signals induced by the rod driving, sensor hard-line movement and loosened component since they have similar frequency characteristics with valid signals. This paper classifies the signal patterns by analyzing actual LPMS signal captured during nuclear power plant operation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.110-114
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• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• 설비공학논문집
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• 제18권3호
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• pp.202-210
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• 2006

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thormal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, when the turbulence penetration occurs in the branch pipe, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine.

• 방사성폐기물학회지
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• 제19권2호
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• pp.243-253
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• 2021

In the dismantling process of a reactor coolant system (RCS) piping, a radiation protection plan should be established to minimize the radiation exposure doses of dismantling workers. Hence, it is necessary to estimate the individual effective dose in the RCS piping dismantling process when decommissioning a nuclear power plant. In this study, the radiation exposure doses of the dismantling workers at different positions was estimated using the MicroShield dose assessment program based on the NUREG/CR-1595 report. The individual effective dose, which is the sum of the effective dose to each tissue considering the working time, was used to estimate the radiation exposure dose. The estimations of the simulation results for all RCS piping dismantling tasks satisfied the dose limits prescribed by the ICRP-60 report. In dismantling the RCS piping of the Kori-1 or Wolsong-1 units in South Korea, the estimation and reduction method for the radiation exposure dose, and the simulated results of this study can be used to implement the radiation safety for optimal dismantling by providing information on the radiation exposure doses of the dismantling workers.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2748-2754
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• 2022

Because existing reactant coolant system (RCS) leakage detection mechanisms are insensitive to small leaks, a real-time, direct detection system with a detection threshold below 0.5 gpm·hr^-1 was studied. A beta-ray detection system using a silicon detector with good energy resolution for beta rays and a low gamma-ray response was proposed. The detection performance in the leakage condition was evaluated through experiments and simulations. The concentration of ¹⁶N in the coolant corresponding to a coolant leakage of 0.5 gpm was calculated using the analytic method and ORIGEN-ARP. Based on the concentration of ¹⁶N and the measurement of the silicon detector with ⁹⁰Sr/⁹⁰Y, the beta-ray count rate was estimated using MCNPX. To evaluate the effect of gamma rays inside the containment building, the signal-to-noise ratio (SNR) was calculated. To evaluate the count rate ratio, the radiation field inside the containment building was simulated using MCNPX, and response evaluation experiments were performed using beta and gamma rays on the silicon detector. The expected beta-ray count rate at 0.5 gpm leakage was 7.26 × 10⁵ counts/sec, and the signal-to-background count rate ratio exceeded 88 for a transport time of 10 s, demonstrating its suitability for operation inside a reactor containment building.