• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.23 no.1
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.249-256
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• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

• Journal of Energy Engineering
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• v.17 no.2
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• pp.67-76
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• 2008

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). Recently several integral effect tests for the reflood period of a LBLOCA (Large Break LOss of Coolant Accident) of the APR1400 have been performed with the ATLAS. In the APR1400 a high flow condition is changed to a low flow condition due to an fluidic device during an operation of the SIT. As the self-controlled fluidic device was not installed in the ATLAS, a set of characterization tests was performed to simulate its injection capability from the SIT for the APR1400 simulation. In the ATLAS the required SIT flow rate in the high flow condition was acquired by installing orifices with an optimized flow area to throttle the SIT discharge line and the low flow condition was achieved by changing the opening of the flow control valve in the SIT injection line. The test results showed that the safety injection systems of the ATLAS could simulate the required high and low flow rates of the SIT for the APR1400 simulation efficiently.

• Journal of Microbiology and Biotechnology
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• v.27 no.1
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• pp.9-18
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• 2017

Nine bacilli with fibrinolytic activities were isolated from doenjang, a traditional Korean fermented soy food. Among them, RSB34 showed the strongest activity and was identified as Bacillus amyloliquefaciens by 16S rRNA and recA gene sequencing. During growth on LB up to 96 h, RSB34 showed the highest fibrinolytic activity ($83.23mU/{\mu}l$) at 48 h. Three bands of 23, 27, and 42 kDa in size were observed when the culture supernatant was analyzed by SDS-PAGE and 27 and 42 kDa bands by fibrin zymography. The gene encoding the 27 kDa fibrinolytic enzyme AprE34 was cloned by PCR. BLAST analyses confirmed that the gene was a homolog to genes encoding AprE-type proteases. aprE34 was overexpressed in Escherichia coli BL21(DE3) using pET26b(+). Recombinant AprE34 was purified and examined for its properties. The $K_m$ and $V_{max}$ values of recombinant AprE34 were $0.131{\pm}0.026mM$ and $16.551{\pm}0.316{\mu}M/l/min$, respectively, when measured using an artificial substrate, N-succinyl-ala-ala-pro-phe-p-nitroanilide. aprE34 was overexpressed in B. subtilis WB600 using pHY300PLK. B. subtilis transformants harboring pHYRSB34 (pHY300PLK with aprE34) showed higher fibrinolytic activity than B. amyloliquefaciens RSB34.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.04a
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• pp.589-595
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being verified on the integrity of RVI for the design life of the plant by performing the non-prototype category-2 type on the US Nuclear Regulatory Commission Guide(NRC RG) 1.20, for which consists of a vibration and stress analysis program, a limited vibration measurement program, an inspection program, and the correlation of these programs. The aim of this paper is to describe the plan for the vibration measurement, test and acceptance criteria portion, and documentation and results of the APR1400 RVI CVAP. We will conduct the limited vibration measurement program of the APR1400 RVI CVAP according to the measurement plan and the vibration measurement testing in this paper.

• Journal of Microbiology and Biotechnology
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• v.23 no.7
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• pp.974-983
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• 2013

Bacillus amyloliquefaciens CB1 was isolated from cheonggukjang, a Korean fermented soy food. B. amyloliquefaciens CB1 secretes proteases with fibrinolytic activities. A gene homologous to aprE of Bacillus subtilis, aprECB1, was cloned from B. amyloliquefaciens CB1, and DNA sequencing showed that aprECB1 can encode a prepro-type serine protease consisting of 382 amino acids. When aprECB1 was introduced into B. subtilis WB600 using an E. coli-Bacillus shuttle vector, pHY300PLK, transformants showed fibrinolytic activity and produced a 28 kDa protein, the size expected for the mature enzyme. The 28 kDa fibrinolytic enzyme was purified from the culture supernatant of B. subtilis WB600 transformant. AprECB1 was completely inhibited by phenylmethylsulfonyl fluoride and almost completely inhibited by EDTA and EGTA, indicating that it is a serine metalloprotease. AprECB1 exhibited the highest specificity for N-succinyl-Ala-Ala-Pro-Phe-p-nitroanilide, a known substrate for ${\alpha}$-chymotrypsin. $A{\alpha}$ and $B{\beta}$ chains of fibrinogen were quickly degraded by AprECB1, but the ${\gamma}$-chain was resistant.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.24-35
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• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.56.2-56.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.171-171
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• 2002