• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1153-1160
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• 2022

The application of video processing techniques are useful for the safety of nuclear power plants by tracking the people online on video to estimate the dose received by staff during work in nuclear plants. Nuclear reactors remotely visually controlled to evaluate the plant's condition using video processing techniques. Internal reactor components should be frequently inspected but in current scenario however involves human technicians, who review inspection videos and identify the costly, time-consuming and subjective cracks on metallic surfaces of underwater components. In case, if any frame of the inspection video degraded/corrupted/missed due to noise or any other factor, then it may cause serious safety issue. The problem of missing/degraded/corrupted video frame estimation is a challenging problem till date. In this paper a systematic literature review on video processing techniques is carried out, to perform their suitability analysis for NPP applications. The limitation of existing approaches are also identified along with a roadmap to overcome these limitations.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.575-576
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• 2005

• Transactions of the Korean Society of Mechanical Engineers B
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• v.39 no.11
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• pp.905-911
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• 2015

In this study, to examine the effect of a hole size change(smaller hole diameter) in the outer region of the lower-support-structure-bottom plate(LSSBP) on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.15. The predicted results were compared with those of the original LSSBP. Through these comparisons, it was concluded that a more uniform distribution of the mass flow rate at the core-inlet plane could be obtained by reducing the hole size in the outer region of the LSSBP. Therefore, from the nuclear regulatory perspective, design change of the hole pattern in the outer region of the LSSBP may be desirable in terms of improving both the mechanical integrity of the fuel assembly and the core thermal margin.

• Journal of Mechanical Science and Technology
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• v.19 no.2
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• pp.634-648
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• 2005

Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel. A round robin consisting of one prerequisite deterministic study and five cases for probabilistic approaches is proposed, and all organizations interested are invited. The problems are solved by the participants and their results are compared to issue some recommendation of best practices and to assure an understanding of the key parameters in this type of approach, like transient description and frequency, material properties, defect type and distribution, fracture mechanics methodology etc., which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria. Six participants from 3 organizations responded to the problem and their results are compiled and analyzed in this study.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

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.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1738-1748
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• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• Journal of the Korean Institute of Gas
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• v.17 no.4
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• pp.1-8
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• 2013

The aim of this study is to assess thermal hazards of polystyrene polymerization process by bulk polymerization with accelerating rate calorimeter(ARC) and Multimax reactor system(MM). From this study, we found out that the polymerization process should be operated at reaction temperature of $120^{\circ}C{\sim}130^{\circ}C$. At reaction temperature over $130^{\circ}C$, there was a runaway reaction hazard due to the temperature control failure following a viscosity increase of reaction products. With a cooling failure of a reactor in the early stage of process operation at the reaction temperature ($120^{\circ}C{\sim}130^{\circ}C$), there was a high thermal hazard of burst of a reactor's rupture disk or explosion of a reactor caused by the rapid rise of temperature and pressure to $340^{\circ}C$, 5.3 bar respectively within 30 - 50 minutes.

• Journal of Hydrogen and New Energy
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• v.23 no.1
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• pp.73-82
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• 2012

A 2-dimensional heterogeneous reactor model was developed and simulated for a tube reactor of packed bed where the steam-$CO_2$ combined reforming reaction of natural gas proceeded to produce synthesis gas. Under the reactor feeding rate, 45 $Nm^3$/h, of the reactant gas stream, the 2-dimensional heterogeneous reactor model showed the similar results to those from the ASPEN simulator although there were some discrepancies between the two in the temperature and the $H_2$/CO ratio of the reformed gas at the reactor exit. The calculated enthalpy difference between the reformed gas at the reactor exit and the reactant gas fed to the reactor was closely correspondent to the total amount of heat transferred to the reactor interior from the furnace. This supports that the 2-dimensional heterogeneous reactor model was reasonably established and the numerical solution was properly obtained.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.855-864
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• 2017

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.1-6
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• 2011

Flame extinction and the chemistry of stoichiometric methane/air mixture were investigated numerically in the PSR(perfectly stirred reactor). For the study, PSR code was modified to be possible to unsteady calculation, and the sinusoidal fluctuation was subjected to the residence time. In the region of residence time far from the extinction limit, combustion mode was strongly dependent on the frequency. The low frequency excitation provided the quasi-steady behavior on the temperature and the concentrations of related species, but small variation of temperature was observed under high frequency. In the region of residence time near the extinction limit, the mixture subjected above 1 KHz was still reacting even though extinction had to be occurred under quasi-steady concept. The attenuation of extinction limit resulted from that chemical time was comparable to the flow time. The mean mole fractions of both NO and CO were almost same regardless of imposed frequency. However, the average mole fraction of $C_2H_2$ was decreased as increasing frequency, which implies that soot yield might be reduced at the higher frequency of flow excitation. The result provides the basic concept for flame stabilization, and it will be used to design a mild combustor.