• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.202-210
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• 2016

A thermal barrier coating (TBC) has been widely applied to machine components working under high temperature as a thermal insulator owing to its critical financial and safety benefits to the industry. However, the nondestructive evaluation of TBC damage is not easy since sensing of the microscopic change that occurs on the TBC is required during an evaluation. We designed an eddy current probe for evaluating damage on a TBC based on the finite element method (FEM) and validated its performance through an experiment. An FEM analysis predicted the sensitivity of the probe, showing that impedance change increases as the TBC thermally degrades. In addition, the effect of the magnetic shield concentrating magnetic flux density was also observed. Finally, experimental validation showed good agreement with the simulation result.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.493-501
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• 1994

An experimental study has been peformed on the pool-boiling critical heat flux (CHF) phenomenon on downward -facing plates. The CHF for inclinations of -90$^{\circ}$(horizontally downward position), -88$^{\circ}$, -86$^{\circ}$, -84$^{\circ}$, -60$^{\circ}$ and -40$^{\circ}$ were measured using plate-type test sections of 20mm 200mm and 25mm 200mm in a pool of saturated water under atmospheric pressure. The measured CHF was lower for the wider test section and decreased as its orientation approached to the horizontally downward position. The lower CHF can be attributable to the increased difficulty for the bubbles in escaping from the heater surface. When compared with the previous works, the overall trends were similar; however, a transition angle, at which the decrease rate in the CHF was changed, was observed in the vicinity of -80$^{\circ}$.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.56-65
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• 1991

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.439-445
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• 2021

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

• Membrane Journal
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• v.32 no.4
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• pp.209-217
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• 2022

Ion exchange membrane (IEM) is an important class of membrane applied in batteries, fuel cells, chloride-alkali processes, etc to separate various mono and multivalent ions. The membrane process is based on the electrically driven force, green separation method, which is an emerging area in desalination of seawater and water treatment. Electrodialysis (ED) is a technique in which cations and anions move selectively along the IEM. Anion exchange membrane (AEM) is one of the important components of the ED process which is critical to enhancing the process efficiency. The introduction of cross-linking in the IEM improves the ion-selective separation performance due to the reduction of free volume. During the desalination of seawater by reverse osmosis (RO) process, there is a lot of dissolved salt present in the concentrate of RO. So, the ED process consisting of a monovalent cation-selective membrane reduces fouling and improves membrane flux. This review is divided into three sections such as electrodialysis (ED), anion exchange membrane (AEM), and cation exchange membrane (CEM).

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.518-531
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• 1995

A study has been Peformed to investigate the thermal margin increase by replacing the single-channel analysis model with a multichannel analysis model. h new critical heat flux(CHF) correlation, which is applicable to a 17$\times$17 Korean Fuel Assembly(KOFA)-loaded core, was developed on the basis of the local conditions predicted by the subchannel analysis code, TORC. The hot sub-channel analysis was carried out by using one-stage analysis methodology with a prescribed nodal layout of the core. The result of the analysis shooed that more than 5％ of the thermal margin can be recovered by introducing the TORC/KRB-1 system(multichannel analysis model) instead of the PUMA/ERB-2 system(single-channel anal)sis model). The thermal margin increase was attributed not only to the effect of the local thermal hydraulic conditions in the hot subchannel predicted by the code, but also to the effect of the characteristics of the CHF correlation.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.939-946
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• 2023

Comprehensive Research Facility for Fusion Technology (CRAFT) is an integration of different demonstrating or testing facilities, which aim to develop the critical technology or composition system towards the fusion reactor. Due to the importance and challenge of the negative ion based neutral beam injection (NNBI), a NNBI test facility is included in the framework of CRAFT. The initial object of CRAFT NNBI test facility is to obtain a H⁰ beam power of 2 MW at the energy of 200-400 keV for the pulse duration of 100 s. Inside the negative ion accelerator of NNBI system, the interactions of the negative ions with the background gas and electrodes can generate abundant stray electrons. The stray electrons can be further accelerated and dumped on the electrodes or eject from the accelerator. The stray electrons, including the ejecting electrons, cause the unwanted particle and heat flux onto the electrodes and the inner components of beamline (especially the temperature sensitive cryopump). The suppression of the stray electrons from the CRAFT accelerator is carried out via a series of design and simulation works. The paper focuses the influence of different magnetic field configurations on the stray electrons and the character of the ejecting electrons.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Journal of IKEEE
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• v.27 no.4
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• pp.518-523
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• 2023

The diagnosis of Inter Turn Short Circuits (ITSC) in induction motors is critical due to the escalating severity of faults resulting from even minor disruptions in the stator windings. However, diagnosing ITSC presents significant challenges due to similarities in noise and losses shared with 3-phase induction motors. Although artificial intelligence techniques have been explored for efficient diagnosis, practical applications heavily rely on model-based methods, necessitating further research to enhance diagnostic performance. This study proposed a diagnostic method applied the Clarke Transformation approach, focusing solely on current components while disregarding changes in rotating flux. Experimental results conducted over a 30-minute period, encompassing both normal and ITSC conditions, demonstrate the effectiveness of the proposed approach, with FAR(False Accept Rates) of 0.2% for normal-to-ITSC FRR(False Rejection Rates) and 0.26% for ITSC-to-normal FRR. These findings underscore the efficacy of the proposed approach.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.