• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.37-44
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• 1989

An improved water chemistry technique was studied to reduce radioactivity build-up in reactor coolant system. The technique is convering the current coordinated lithium-boron chemistry regime to the elevated lithium chemistry regime in order to maintain high pH. Correlations between reactor coolant pH and radioactivity build-up were analized by using pH data from domestic PWRs. Consequently, it was founded that high pH chemistry was moer effective for radioactivity build-up reduction than current chemistry regime. This fact had revealed that much portion of reactor coolant corrosion products were nickel ferrite rather than magnetite, and that pH value ranging 7.0-7.4 was appropriate for high-pH chemistry operation.

• Proceedings of the Korea Institute of Fire Science and Engineering Conference
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• 2003.04a
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• pp.259-264
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• 2003

The fire protection system of Nuclear Power Plants(NPPs) is an integrated system that is applied multi-field technology. So, it needs synthetic design and analysis, that is, the plan of fire protection, fire compartment, fire detection, fire suppression, and success of safety shut down, etc. In case of a fire in NPPs, secure the safety of reactor and minimize the radioactivity contamination. For this purpose, perform the fire risk analysis and make up the deducted problem through the improvement of design or the change of operation process.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2135-2146
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• 2022

A nuclear energy facility is one of the most critical facilities to be safely protected during and after operation because the physical destruction of its barriers by an external attack could release radioactivity into the environment and can cause harmful effects. The barrier walls of nuclear energy facilities should be sufficiently robust to protect essential facilities from external attack or sabotage. Physical protection system (PPS) vulnerability assessment of a typical small nuclear research reactor was carried out by simulating an external attack with a tri-nitro toluene (TNT) shaped charge and results are presented. The reinforced concrete (RC) barrier wall of the research reactor located at a distance of 50 m from a TNT-shaped charge was the target of external attack. For the purpose of the impact assessment of the RC barrier wall, a finite element method (FEM) is utilized to simulate the destruction condition. The study results showed that a hole-size of diameter 342 mm at the front side and 364 mm at the back side was created on the RC barrier wall as a result of a 143.35 kg TNT-shaped charge. This aperture would be large enough to let at least one person can pass through at a time. For the purpose of the PPS vulnerability assessment, an Estimate of Adversary Sequence Interruption (EASI) model was used, which enabled the determination of most vulnerable path to the target with a probability of interruption equal to 0.43. The study showed that the RC barrier wall is vulnerable to a TNT-shaped charge impact, which could in turn reduce the effectiveness of the PPS.

• Journal of Life Science
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• v.12 no.5
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• pp.584-588
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• 2002

Asymmetric enantioselective resolution system using epoxide hydrolase activity of Aspergillus niger LK was developed and operated for the production of optically pure styrene oxide. Two-phase hollow-fiber reactor system was employed for the enhanced solubility of racemic styrene oxide in organic phase and protection of epoxide hydrolase activity in aqueous phase. For the removal of phenyl-1,2-ethandiol, the inhibitor of epoxide hydrolase, cascade hollow-fiber reactor system was also developed. Chiral (S)-styrene oxide (39 mM in dodecane) could be asymmetrically resolved with high enantiopurity (> 99% ee) using these reactor system.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1442-1450
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• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.40.2-40
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• 2001

The protection system of the system communication, nuclear reactor and chemical reactor are representative of reliable system. This reliable system must be designed based on reliability as well as concept of safety, which is a failed system go a way of safe. Reliable system is composed of part of data acquisition, calculator, communication with redundancy, and a voter is important factor of reliability. Because it is serially connected. This paper presents a Design and Analysis of a Voting Mechanism considering Safety for reliable system Using EPLD. In the case of digital implementation a coincidence logic (voter) of reliable system, it needs CPU and memory, so increase a number of units. Therefore the failure rate and cost are increased on contrary when it is designed EPLD or FPGA.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.813-820
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• 2009

A rapid pneumatic transfer system (PTS) for an instrumental neutron activation analysis (INAA) is developed as an automatic irradiation facility involving the measurement of a short half-life nuclide and a delayed neutron counting system. Three new PTS designs with improved functions were constructed at the HANARO research reactor in 2006. The new system is composed of a manual system and an automatic system for both an INAA and a delayed neutron activation analysis (DNAA). The design and basic conception of a modified PTS are described, and the functions of system operation and control, radiation protection and emissions of radioactive gas are improved. In addition, a form of capsule transportation of these systems is tested. The experimental results pertaining to the irradiation characteristics with variation of the neutron flux and the temperature of the irradiation position with the irradiation time are presented, as is an analysis of the reference material for analytical quality control and uncertainty assessments.

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.165-175
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• 1995

The overpressure protection system for PWR should be designed with sufficient capacity to limit the pressure to less than 110％ of the reactor coolant system design pressure during the most severe abnormal operational transient. In this study, the feasibility of adopting the SEBIM POSRV instead of the current spring loaded pop-opening safety valves to the ABB-CE designed 2825 MWt PWR is investigated for its overpressure protection capability. The required SEBIM POSRV size as well as its opening/closing setpoints are determined through a series of computer analyses using the LTC code which has been used for the overpressure protection analysis for Yonggwang units 3&4. The analysis results show that the overpressure protection system with monobloc SEBIM POS-RV can maintain the RCS pressure below 110％ of the design pressure demonstrating its overpressure protection capability for the ABB-CE designed 2825 MWt PWRs.