• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2812-2820
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• 2020

House load operation (HLO) occurs when the generator supplies power to the house load without triggering reactor trips during grid disturbances. In Korea, the HLO capability of optimized power reactor 1000 (OPR1000) plants has prevented several reactor trips. Operational experiences demonstrate the difference in the reactor trip incidence due to grid disturbances between OPR1000 plants and Westinghouse plants in Korea, attributable to the availability of the HLO capability. However, probabilistic safety assessments (PSAs) for OPR1000 plants have not considered their specific design features in the initiating event analyses. In an at-power PSA, the HLO capability can affect the initiating event frequencies of general transients (GTRN) and loss of offsite power (LOOP), resulting from transients within the grid system. The initiating event frequencies of GTRN and LOOP for an OPR1000 plant are reduced by 17.7% and 78.7%, respectively, compared to the Korean industry-average initiating event frequencies, and its core damage frequency from internal events is reduced by 15.2%. The explicit consideration of the HLO capability in initiating event analyses makes significant changes in the risk contributions of the initiating events. Consequently, for more realistic at-power PSAs in Korea, we recommend incorporating plant-specific HLO-related design features when estimating initiating event frequencies.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.403-410
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• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1993년도 한국자동제어학술회의논문집(국내학술편); Seoul National University, Seoul; 20-22 Oct. 1993
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• pp.749-753
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• 1993

It is very difficult to control batch reactor with conventional linear controller due to its severe nonlinearity. To control the nonlinearity of batch reactor, we applied with relay feedback method and SOAS. The SOAS can be designed to work quite well, but it requires engineering effect and some knowledge about the process in order to get a satisfactory performance of the closed loop system For the applications to more reliable, further studies on robustness in various situations and process noises and would be required.

• 한국환경과학회지
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• 제14권2호
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• pp.157-165
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• 2005

This study was conducted to determine optimum design parameters in nitrification and denitrfication of chemical fertilizer wastewater using pilot plant, Jet Loop Reactor. The chemical fertilizer wastewater which contains low amounts of organic carbon and has a high nitrogen concentration requires a post-denitrfication system. Organic nitrogen is hydrolyzed above $86\%$, and the concentration of organic nitrogen was influent wastewater 126mg/L and of effluent wastewater 16.4mg/L, respectively. The nitrification above $90\%$ was acquired to TKN volumetric loading below $0.5\;kgTKN/m^3{\cdot}d$, TKN sludge loading below $0.1\;kgTKN/kgVSS{\cdot}d$ and SRT over 8days. The nitrification efficiency was $90\%$ or more and the maximum specific nitrification rate was $184.8\;mgTKN/L{\cdot}hr$. The denitrification rate was above $95\%$ and the concentration of $NO_3-N$ was below 20mg/L. This case was required to $3\;kgCH_3OH/kgNO_3-N$, and the effluent concentration of $NO_3^--N$ was below 20mg/L at $NO_3^--N$ volumetric loading below $0.7\;kgNO_3^--N/m^3{\cdot}d$ and v sludge loading below $0.12\;kgNO_3^-N/kgVSS{\cdot}d$. At this case, the maximum sludge production was $0.83\;kgTS/kgT-N_{re}$ and the specific denitrfication rate was $5.5\;mgNO_3-N/gVSS{\cdot}h$.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.358-373
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• 1995

중수로용 개량핵 연료집합체인 CANFLEX 핵연료다발의 CANDU-6 원자로 장전시 열수송계통에 대한 유동안정성이 분석되었다. CANFLEX 핵연료다발은 기존의 37개봉 핵연료다발과 원자로출력 및 압력강하 측면에서 거의 일치되며, 이로인해 수력적 거동이 양립하는 반면, CANFLEX핵연료다발은 기존의 37개봉 핵연료다발 보다 임계채널 출력이 증가하며, 반경방향 출력분포의 평탄화로 인해 균일한 엔탈피 분포를 확보할 수 있게 된다. CANFLEX 핵연료다발 및 출구모관들의 상호연결관에 대한 SOPHT 모델을 개발하였으며, 이 모델을 이용하여 CANFLEX 핵연료다발이 장전된 월성 1호기의 유동 안정성 거동이 해석되었다. 해석결과, 열수송계통의 출구모관들의 상호연결관이 없을 경우에는 기존의 37개봉 핵연료다발과 같이 유동이 불안정함을 보였으며, 출구모관들의 상호연결관이 있을 경우에는 정격출력의 $\pm$1% 내에서 안정함을 보였다. 따라서 CANFLEX 핵연료다발의 월성 1호기 장전시 열수송계통의 유동안정성 측면에서는 건전할 것으로 판단되었다.

• KSBB Journal
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• 제8권3호
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• pp.193-199
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• 1993

공기부양 생물반응기에서 천연 청색 색소원인 쪽 (Poly,앵wm tinctorium) 세포를 대량 배양하기 위한 배양조건을 찾기 위해 영양성분 빛 색소 전 구체를 포함한 배지 성분의 생육에 미치는 영향 과 여러 배양조건에서 세포의 생육 특성에 대하 여 설험하였다. 최적 생육조건과 생육 특성을 규 명하기 위 해 external 1oop와 internal loop 두 가지의 arr-lift bioreactor 종류를 사용하여 14일간 배양하였다. 초기 접종량을 달리한 결과 각 reactor에서 공히 접종량이 0.5-2.5% packed cell volume 보다 많을수록 생육에 좋은 결과를 나타내였다. 초기 당농도의 영향을 살펴보기 위하여 2-4%로 그 조건을 달리하여 14일간 배양한 결과 4% 초기 당농도가 생육에 가장 좋은 조건임을 알 수 있었다. Sucrose는 대수기 초기단계에서 모두 소비되었다. Tryptophan(lmM)을 첨가한 경우 14일 배양 후에 internal loop reactor에서 3.8g/l, external loop reactorr에서 3.5g/l 의 cell mass를 얻었다. Indole을 첨가했을 때에는 세포 생육 저해효과 를 나타내였는데 external과 internal Ioop reactor에서 각각 2.5g/l와 2.2g/l를 얻었다. 무기질소원 으로는 potassium nitrate가 선발되었고, contrl이에 비해 110%d의 생육도 증가가 있였다. 최적 조건 하에셔의 14일간 internal loop bioreactor 배양에서 세포 건조 중량 16.34 g/l를 얻었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.158.2-158
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• 1999

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.353-363
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• 2023

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2001년도 ICCAS
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• pp.97.2-97
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• 2001

This work is concerned with the improvement of the productivity and the product quality in the polymerization reactors by using model-on-demand predictive control(MoDPC). This technique is applied to a continuous styrene polymerization reactor and a semibatch methyl methacrylate (MMA)/vinyl acetate(VAc) copolymerization reactor. The regress is constructed with the most influential variables the conversion and the jacket inlet temperature for the styrene polymerization reactor, and the free volume and the reactor temperature for the MMA/VAc copolymerization reactor through open loop operations. From the simulation results for setpoint tracking and disturbance rejection problems, it is demonstrated that the MoDPC shows ...

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2005년도 연구개발 발표회 논문집
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• pp.509-514
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• 2005

Main coolant pump (MCP) is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel rods and steam generators in an integral type reactor. The reactor is designed to operate under condition of 310 oC and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition in order to verify its performance and safety. In present work, a test loop to simulate real operating situation of the reactor has been designed and constructed to test MCP. And then, as a part of qualification test, canned motor functional test and pump hydraulic performance test have been carried out upon a prototype MCP. Canned motor efficiency and pump hydraulic characteristics including homologous curves and NPSH curves were obtained from the qualification test.