• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.5
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• pp.520-525
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• 2003

Fast Breeder Test Reactor (FBTR) is a 40MW (thermal) / 13.2MW (electrical), Plutonium - Uranium mixed carbide fuelled, sodium cooled, loop type nuclear reactor operating at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. Its main aim is to generate experience in operation of fast reactors and sodium systems and to serve as an irradiation facility for development of fuels and structural materials fur fast reactors. Nuclear reactors pose difficulties to the NDT techniques used to monitor the conditions of the internal components. Sodium cooled fast breeder reactors have their own typical difficulties in using the NDT techniques. These are due to the need for operation in aggressive environment of nuclear radiation and sodium (molten/vapour), as well as the need to maintain leak tightness of a very high order during all states of reactor operation and shutdown for fuel handling, maintenance and remote inspection. This paper discusses the following NDT techniques, which have been successfully used for the past 15 years in FBTR: (i) Periscope and Projector, (ii) Core Co-ordinate Measuring Device and, (iii) Optical fiberscope. The inspection using these techniques have given confidence for further reactor operation at high power by giving useful data on the conditions of the components inside the reactor vessel.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.921-928
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• 2009

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.221-230
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• 2007

Wavelet theory was applied to detect a singularity in a reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, using wavelet transform after de-noising, singular points can easily be found. To test this theory, reactor power signals were generated using the HANARO(a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations contaminated with Gaussian noise. Wavelet transform decomposition and de-noising procedures were applied to these signals. It was possible to detect singular events such as a sudden reactivity change and abrupt intrinsic property changes. Thus, this method could be profitably utilized in a real-time system for automatic event recognition(e.g., reactor condition monitoring).

• Journal of Radiation Protection and Research
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• v.30 no.1
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• pp.39-44
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• 2005

A new developing reactor isn't fixed the structure and the materials of reactor components. To perform the shielding analysis for a reactor vessel by $R-\theta$ geometry, it takes much effort and time to modeling of source term according to the change of reactor components every time. Therefore, we developed the shielding analysis system for the reactor vessel by $R-{\theta}$ geometry, which wasn't affected by the reactor core geometry. By using the developed shielding analysis system, we performed the shielding analysis for the reactor vessel of an integral reactor which has the hexagonal geometry of nuclear fuel assemblies in reactor core. We compared the results obtained from the developed system with those obtained from MCNP analysis. Because the results of developed shielding analysis system were more conservative than those of MCNP calculation, it is useful for shielding analysis. As we had developed the new shielding analysis system for a reactor vessel by $R-{\theta}$ geometry, we reduced error of model for reactor core which was formerly designed by hand and saved the time and the effort to design source term model of reactor core.

• Korean Membrane Journal
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• v.6 no.1
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• pp.10-15
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• 2004

We developed a water-selective ceramic composite membrane for use as a dehydration membrane reactor for dimethylether (DME) synthesis from methanol. The membranes were modified on the porous stainless steel support by the sol-gel method accompanied by a suction process. The improved membrane modification process was effective in increasing the vapour permselectivity by removal of defects and pinholes. The optimized alumina/silica composite membrane exhibited a water permeance of 1.14${\times}$10$^{-7}$ mol/$m^2$.sec.Pa and a water/methanol selectivity of 8.4 at permeation temperature of 25$0^{\circ}C$. The catalytic reaction for DME synthesis from methanol using the membrane was performed at 23$0^{\circ}C$, and the reaction conversion was compared with that of the conventional fixed-bed reactor. The reaction conversion of the membrane reactor was much higher than that of the conventional fixed-bed reactor. The reaction conversion of the membrane reactor and the conventional fixed-bed reactor was 82.5 and 68.0%, respectively. This improvement of reaction efficiency can last if the water vapour produced in the reaction zone is removed continuously.

• Journal of the Institute of Electronics and Information Engineers
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• v.51 no.9
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• pp.215-220
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• 2014

Study and application result of power controller to research reactor is presented. Considering safety-oriented design concept and other control environment, we developed a simple closed-loop controller that provides limiting function of power-change-rate as well as low-overshoot and fine tracking performance. The design result has been well-proven via simulation and actual application to a research reactor.

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.135-143
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• 2017

The domestic unique research reactor, HANARO (Hi-flux Advanced Neutron Application ReactOr), has been constructed with the open-pool, the core is submerged in, for the multi-purpose neutron application. The reactor has a primary cooling system to remove the fission heat from the core and its connected fluidic systems. Since the works are required at the reactor pool top as a characteristic of the research reactor, the radiation shall be minimized with the operation of the hot water layer system to avoid unnecessary radiation exposure on the workers during work at the pool top. Moreover, the pool water management system is connected to the reactor pool to maintain the pool temperature below $50^{\circ}C$ to minimize the uprising radioactive gas or impurity from the colder pool bottom. For the efficient flow rate of the PWMS, the thermal capacity of heat exchanger is selected with 260 kW in the normal operation condition. In this paper, the modeling is formulated to figure out whether or not each pool temperature maintains below the temperature limit and the calculation results show that the designed PWMS heat exchanger has enough capacity with the design margin regardless of the reactor operation mode.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.963-967
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• 2019

Accurate measurement of the reactivity worth of safety rods is very important for the safe reactor operation, in normal and emergency conditions. In this paper, the reactivity worth of safety rods in Heavy Water Zero Power Reactor (HWZPR) in the new lattice pitch is measured by advanced source jerk method. The average of the results related to two different detectors is equal to 29.88 mk. In order to verify the result, this parameter was compared to the previously measured value by subcritical to critical approach. Different experiment results are finally compared with corresponding calculated result. Difference between the average experimental and calculated results is equal to 2.2%.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2958-2968
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• 2020

Understanding public perception on nuclear research reactor is necessary for the policy maker to adopt such technology in Thailand, especially the locals who live in the proposed location. The study compared perceptions between the locals living near the proposed nuclear research reactor location (within 5 km) and those living in the outer region (5-15 km). Structural equation modeling technique was adopted by assuming casual relationships between latent variables including social status, information perception, trust, benefit perception and risk perception on the local acceptance of research reactor. The results showed that the strongest relationships for both the inner and the outer perimeters were from information perception toward technology acceptance via trust and benefit perception. While both zones showed similar results, the outer perimeter seemed to show slightly stronger effects than those in the inner perimeter.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.473-478
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• 1997

A severe accident management strategy, in-vessel retention corium through external reactor vessel cooling(ERVC) is being studied worldwide as a means to prevent reactor vessel failure following a core melt accident. An evaluation of feasibility of this ERVC for a large Advanced Light Water Reactor (ALWR) is presented. To account for the coolability of corium and metal in the reactor vessel, a thermal analysis is performed using an existing method. Results show that the peak heat flux along the inner surface of the reactor vessel lower head has a relatively smaller margin than a small capacity reactor such as AP600 in regards with the critical heat flux attainable at the outer surface of the reactor vessel lower head.