• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.776-783
• /
• 2020

As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures.

• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.87-92
• /
• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Journal of Radiation Protection and Research
• /
• v.41 no.4
• /
• pp.354-358
• /
• 2016

Background: To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition Materials and Methods: MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. Results and Discussion: It was found that the leak rate is $0.1%{\cdot}day^{-1}$ of air, $0.004%{\cdot}day^{-1}$ of noble gas and $3.7{\times}10^{-5}%{\cdot}day^{-1}$ of aerosol during typhoon passing. The air leak rate of $0.1%{\cdot}day^{-1}$ can be converted into $1.36m^3{\cdot}hr^{-1}$, but the design leak rate in HANARO safety analysis report was considered as $600m^3{\cdot}hr^{-1}$ under the condition of $20m{\cdot}sec^{-1}$ wind speed outside of the building by typhoon. Conclusion: Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

• Nuclear Engineering and Technology
• /
• v.41 no.7
• /
• pp.921-928
• /
• 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
• /
• v.39 no.3
• /
• pp.221-230
• /
• 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).

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

• Nuclear Engineering and Technology
• /
• v.34 no.4
• /
• pp.344-357
• /
• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• International Journal of Control, Automation, and Systems
• /
• v.4 no.1
• /
• pp.105-112
• /
• 2006

The main function of a reactor protection system is to maintain the reactor core integrity and the reactor coolant system pressure boundary. Generally, the reactor protection system adopts the 2-out-of-m redundant architecture to assure a reliable operation. This paper describes the safety assessment of a digital reactor protection system using the fault tree analysis technique. The fault tree technique can be expressed in terms of combinations of the basic event failures such as the random hardware failures, common cause failures, operator errors, and the fault tolerance mechanisms implemented in the reactor protection system. In this paper, a prediction method of the hardware failure rate is suggested for a digital reactor protection system, and applied to the reactor protection system being developed in Korea to identify design weak points from a safety point of view.

• Journal of Microbiology and Biotechnology
• /
• v.6 no.1
• /
• pp.30-35
• /
• 1996

Distillery wastewater was used in a thermophilic laboratory-scale two stage anaerobic digester to test the effects of the redox potential of the first acidogenic reactor on the performance of the system. The digester consisted of first a acidogenic reactor and the an upflow anaerobic sludge blanket (UASB) reactor. The digestor was operated at a hydraulic retention time (HRT) of 48 h. Under these conditions, about 90% of the chemical oxygen demand as measured by the chromate method ($COD_{cr}$) was removed with a gas production yield of 0.4 l/g-COD removed. The redox potential of the acidogenic reactor was increased when the reactor was purged with nitrogen gas or agitation speed was increased. The increase in reduction potential was accompanied by an increase in acetate production and a decrease in butyrate formation. A similar trend was observed when a small amount of air was introduced into the acidogenic reactor. It is believed that the hydrogen partial pressure in the acidogenic reactor was decreased by the above mentioned treatments. The possible failure of anaerobic digestion processes due to over-loading could be avoided by the above mentioned treatments.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1071-1082
• /
• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.