• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.425-431
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• 1996

KT-2 poloidal-field (PF) system is designed to cope the up-down symmetric double-null (DN) and asymmetric single-null (SN) discharges with typical plasma parameters, in which three sets of "design-basis" scenarios - the ohmic heating (OH), the 5MW and the high bootstrap (HIBS) baseline modes - are applied. The power and energy demand for each cases are also deduced. The peak power and the maximum energy requirements for the KT-2 magnet system, incorporating the PF and the toroidal-field (TF) coils, are proven to be 123MW and 1601MJ, respectively when it is driven in DN configuration. The KT-2 PF system is capable of achieving the machine mission of creating a 500kA heated plasma with a current flattop of $\geq$20 seconds.

• Analytical Science and Technology
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• v.8 no.4
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• pp.443-448
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• 1995

Three anode configurations of six-jet, cone-jet and cylindrical-jet are tested for their analytical performance under power mode operation. The effect of pressure, power and gas flow rate on atomic absorption signals have been studied. The increase of atomic absorption signal of sample element is observed at a fixed pressure in all configurations as the gas flow rate increase up to 300-600 seem, and as the power dissipated in the glow discharge cell increase. The lower the pressure is in the glow discharge cell at a fixed discharge power and argon flow rate, the greater the absorbance of sample element is. The optimum conditions are taken from these data and a calibration curve of Cu in low-alloy steel sample is obtained. In this calibration curve, six-jet configuration shows the best analytical results varies as the sample element.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.1008-1013
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• 1995

The CANFLEX fuel bundle has been developed by KAERI/AECL jointly to facilitate the use of various fuel cycles in CANDU-6 reactor. As one of the design evaluations, the structural analysis of the fuel bundles by hydraulic drag force is performed to evaluate the fuel integrity in the period of the refuelling in CANDU-6. The structural integrity is evaluated by FEM modelling for the complicated bundles configuration in channel. It is noted that the present analysis method is newly developed for the structural integrity evaluation. The analysis results show that the fuel bundle is shown to keep its structural integrity during the refuelling.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.29-33
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• 1997

Neutron room scattering corrections that should be made when neutron detectors are calibrated with a $D_2O$ moderated $^{252}Cf$ neutron source in the center of a calibration room are considered. Such room scattering corrections are dependent on specific neutron source type, detector type, calibration distance, and calibration room configuration. Room scattering corrections for the responses of a thermoluminescence dosimeter and two different types of spherical detectors to neutron source in the Radiation Calibration Laboratory(RCL) neutron calibration facility at the Korea Atomic Energy Research Institute(KAERI) were experimentally determined and are presented. The measured room scattering results are then compared with theoretical results calculated by predicting room scattering effects in terms of parameters related to the specific configuration. Agreement between measured and calculated scattering correction is generally about 10% for three kinds of detectors in the calibration facility.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.1-5
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• 2017

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.1033-1039
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• 1995

The magnetic system design of the KT-2 tokamak has been performed at KAERI. Design goal has been set to facilitate the so-called "advanced tokamak" studies, which is essential to secure the economy of the tokamak fusion reactors. Design features include a large-aspect-ratio machine configuration, long-pulse operation capability with heavy plasma shaping, hybrid magnetic field control and machine/in-vacuum structures for MHD stability.

• Journal of the Korean Society of Safety
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• v.20 no.1 s.69
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• pp.62-67
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• 2005

An eccentrically-loaded single edge crack tension specimen (ESET) is similar to a compact tension(CT) specimen loaded in tension-tension. The standard ESET specimen exhibits advantages over other types of cracked specimen, such as n, single-edge crack, and middle-crack tension specimen. The details of ESET specimen configuration, test procedure, and calculations are described in ASTM E647 standard. However, a difficulty in attaching COD gage to the knife-edge on the front foe of the specimen can be found when the size of ESET specimen is small for rapid cooling and heating in thermal fatigue testing. The finite element analysis is performed for the ESET specimen with projected knife-edge on the front foe and a crack-length-compliance equation is suggested for the new specimen configuration. Calibration test are conducted with 347 stainless steel to compare the measured crack length with the calculated crack length from the suggested compliance expression. The test results showed good agreements with those of analysis.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.148-157
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• 1997

A new capsule that has a unique structure in which the test environments including neutron flux and fluence, and irradiation temperature can be controlled precisely during irradiation, was conceptually designed. The capsule structure and instrumentation were successfully designed according to the JMTR's standard procedures of capsule design. Based on the target irradiation, the details of the irradiation such as neutron fluence and irradiation temperature ore calculated and the related capsule safety was evaluated. In addition, the effects of design parameters including the changes in inner-capsule configuration, heater capacity, and Helium gas pressure on the specimen temperature were analyzed with a computer program. Through these thermal and strength evaluations, this capsule was proved to be safe during the irradiation in the JMTR.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.83-100
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• 1999

As disposal packaging concepts of spent fuels generated from the domestic NPP, two types, one is to package PWR and CANDU spent fuels in different containers and the other is to package them together, were proposed. The configuration of the containers and the layout of underground repository, such as the container spacing and the deposition tunnel spacing, were developed. The layout of underground repository satisfies the thermal constraint of the bentonite buffer surrounding disposal container, which should be lower than $100^{\circ}C$ in order to keep the physical and chemical properties of bentonite From the spent fuel packaging concepts and container emplacement methods, seven options were developed. With a typical pair-wise comparison methods, AHP, the most promising disposal concept was selected based on the technology Point of view.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.328-341
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• 1998

This paper presents in detail the hydraulic characteristic measurements using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids and has a cylindrical configuration. Axial velocity and turbulent intensity were measured. The effects of the spacer grids on the turbulent flow were investigated using the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of the fuel bundle and the loss coefficients for the spacer grids were estimated from the measured pressure drops. The turbulent thermal mixing phenomena were discussed.