• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1796-1809
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• 2021

The subchannel of a research reactor used to generate high power density is designed to be narrow and rectangular and comprises plate-type fuels operating under downward flow conditions. Critical heat flux (CHF) is a crucial parameter for estimating the safety of a nuclear fuel; hence, this parameter should be accurately predicted. Here, machine learning is applied for the prediction of CHF in a narrow rectangular channel. Although machine learning can effectively analyze large amounts of complex data, its application to CHF, particularly for narrow rectangular channels, remains challenging because of the limited flow conditions available in existing experimental databases. To resolve this problem, we used four CHF correlations to generate pseudo-data for training an artificial neural network. We also propose a network architecture that includes pre-training and prediction stages to predict and analyze the CHF. The trained neural network predicted the CHF with an average error of 3.65% and a root-mean-square error of 17.17% for the test pseudo-data; the respective errors of 0.9% and 26.4% for the experimental data were not considered during training. Finally, machine learning was applied to quantitatively investigate the parametric effect on the CHF in narrow rectangular channels under downward flow conditions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1273-1279
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• 2010

Pressure tubes at the CANada Deuterium Uranium (CANDU) nuclear power plants are periodically inspected in accordance with the CSA N285.4 code. If flaws that do not satisfy the criteria given in CSA N285.4 are detected, the code permits a fitness-for-service assessment to determine the acceptability of the flawed pressure tubes. In this paper, the Wolsong In-service Evaluation System (WIES) is introduced; this system has been developed for the assessment of the flawed pressure tubes and is based on CSA N285.8. Since the system evaluates the integrity of flawed pressure tubes exactly and promptly during an in-service inspection, it will help in operating the Wolsong nuclear power plants without prolonging the outage period.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.440-452
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• 1994

Nuclear fuel bundles are designed such that the heat flux at a-fuel pin surface should not exceed the critical heat flux (CHF) during normal operation and anticipated transient. Therefore, evaluation of the CHF for fuel bundle is demanded in an exact and reliable manner. One of the major concerns with the current application of CHF correlations is that the CHF based on circular tubes is applied to the fuel bundle subchannel analysis, mainly in terms of the hydraulic diameter with correction factors which may result in a source of possibly large uncertainties in CHF prediction. The hydraulic diameter does not recognize the local properties of fluid nor such effect as the surface curvature; the turbulence action on the convex surface is much more pronounced than that on the concave surface. Even for the tube having concave curvature, the effect of tube diameter on CHF becomes important with decreasing diameter. These facts imply that the convex curvature effect is significant and crucial to the reliable CHF prediction. This paper reviews and discusses analytical and experimental aspects of effect of transverse convex curvature in incompressible turbulent flow and heat transfer, and on CHF. Flow models to quantify this effect are briefly mentioned and future works are recommended.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.526-531
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• 2001

The flow of remote field eddy current energy is studied at nonmagnetic coaxial tubes by using both experiments and finite element calculations based on commercial software package. The results showed that remote field eddy current energy at coaxial tubes flow along over the outer surface of external tube, not through the gap between internal and external tubes. This means that the through wall transmission characteristic of remote field eddy current testing (RFECT) is still valid at tube in tube configurations and the RFECT could be potential nondestructive technique for crack detection, spacer location and gap sizing at the coaxial CANDU fuel channel tubes.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.66-80
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• 1991

The LOFT intermediate break experiment L5-1, which simulates 12 inch diameter ECC line break in a typical PWR, has been analyzed using the reactor thermal/hydraulic analysis code RELAP5/MOD2, Cycle 36.04. The base calculation, which modeled the core with single flow channel and two heat structures without using the options of reflood and gap conductance model, has been successfully completed and compared with experimental data. Sensitivity studies were carried out to investigate the effects of nodalization at reactor vessel and core modeling on major thermal hydraulic parameters, especially on peak cladding temperature(PCT). These sensitivity items are : single flow channel and single heat structure (Case A), two flow channel and two heat structures (Case B), reflood option added (Case C) and both reflood and gap conductance options added (Case D). The code, RELAP5/MOD2 Cycle 36.04 with the base modeling, predicted the key parameters of LOFT IBLOCA Test L5-1 better than Cases A,B,C and D. Thus, it is concluded that the single flow channel modeling for core is better than the two flow channel modeling and two heat structure is also better than single heat structure modeling to predict PCT at the central fuel rods. It is, therefore, recommended to use the reflood option and not to use gap conductance option for this L5-1 type IBLOCA.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.449-454
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• 2006

Neutron imaging technique was used to investigate the water distribution and movement in Polymer Electrolyte Membrane Fuel Cell (PEMFC) at HANARO, KAERI. The Feasibility tests were performed in the first and second exposure rooms at the neutron radiography facility (NRF) at HANARO in order to check the ability of each exposure room, respectively. The feasibility test apparatus was composed of water and pressurized air before making up the actual test apparatus. Due to the low neutron intensity in the second exposure room, the exposure time was too long to investigate the transient phenomena of PEMFC. Although the exposure time was improved to 0.1 sec in the first exposure room, it was difficult to discriminate detail water movement at the channel due to the high noise level. Therefore, the experimental setup must be optimized according to the test conditions. Water discharge characteristics were investigated under different flow field geometries by using feasibility test apparatus and the neutron imaging technique. The water discharge characteristics of a 3-parallel serpentine are superior to those of a 1-parallel serpentine, but water at Membrane Electrode Assembly (MEA) was not removed, regardless of the flow field type.

• Nuclear Engineering and Technology
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• v.2 no.1
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• pp.19-25
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• 1970

A study was conducted on the coolant mixing between water flowing in two adjacent subchannels. Measurements were made of the quantity of mass transferred between a larger rectangular channel and a smaller triangular channel in a 19-rod fuel bundle under the conditions of single phase flow and air-water two-phase flow. The results of the experiments showed that the low mixing rate appears in single phase flow, and high mixing rate was measured in air-water two-phase flow Mixing rate decreases with the increasing of air void fraction during the air-water flow. It seems that the high mixing rate in the air-water flow was caused due to adequate agitation of the chaotic air void.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.57-61
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• 2008

During the period of retubing work for the licensing renewal, the fuel channels, calandria tubes and feeders of CANDU Reactors will be replaced, and calandria visual examination will be performed. This period is a unique opportunity to inspect the inside of the calandria. The visual inspection for the calandria vessel and its internals of Wolsong NPP is scheduled for confirming the calandria integrity. The first visual inspection for the calandria is planned in Pt. Lepreau led by AECL. The visual inspection for Wolsong NPP, led by NETEC(Nuclear Engineering & Technology Institute) of KHNP, will employ 3D laser scanner and 3D CAD Mock-up for the first time in the world, in addition to a conventional video camera. The inspection system is composed of a robot with the 3D laser scanner, a video camera and a hardness meter.