• 전기저널
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• 통권463호
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• pp.48-53
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• 2015

• 원자력산업
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• 제26권9호통권283호
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• pp.38-46
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• 2006

• 원자력산업
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• 제26권9호통권283호
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• pp.47-52
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• 2006

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• 한국지구과학회:학술대회논문집
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• 한국지구과학회 2005년도 추계학술발표회 논문집
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• pp.136-141
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• 2005

The K Ar ages, whole rock geochemistry and Sr Nd Pb isotopes have been determined for the submarine basalts dredged from the P2 and P3 segments of the Antarctic-Phoenix Ridge (APR), Drake Passage, Antarctica, for better understanding on temporal variation of magma chemistry in association with extinction of seafloor spreading. The fossilized APR is distant from the known hot spots, and consists of older N-MORB prior to extinction of spreading and younger E-MORB after extinction. The older N-MORB (3.5-6.4 Ma) occur in the southeast flank of the P3 segment (PR3) and the younger E-MORB (1.4-3.1 Ma) comprise a huge seamount at the P3 segment (SPR) and a big volcanic edifice at the P2 segment (PR2). The N-type PR3 basalts have higher Mg#, K/Ba, and CaO/Al2O3 and lower Zr/Y, Sr, and Na8.0 with slight enrichment in incompatible elements and almost flat REE patterns. The E-type SPR and PR2 basalts are highly enriched in incompatible elements and LREE. The extinction of spreading occurring at 3.3 Ma seems to have led to a temporal magma oversupply with E-MORB signatures. Geochemical signatures such as Ba/TiO2, Ba/La, and Sm/La suggest heterogeneity of upper mantle and formation of E-MORB by higher contribution of enriched materials to mantle melting, compared to N-MORB environment. E-MORB magmas beneath the APR seem to have been produced by low melting degree (up to 1% or more) at deeper low-temperature regime, where metasomatized veins consisting of pyroxenites have preferentially participated in the melting. The occurrence of E-MORB at the APR is a good example to better understand what kinds of magmatism would occur in association with extinction of spreading.

• 시스템엔지니어링학술지
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• 제12권1호
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• pp.73-87
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• 2016

Combining power generation and water production by desalination is economically advantageous. Most desalination projects use fossil fuels as an energy source, and thus contribute to increased levels of greenhouse gases. Environmental concerns have spurred researchers to find new sources of energy for desalination plants. The coupling of nuclear power production with desalination is one of the best options to achieve growth with lower environmental impact. In this paper, we will per-form a sensitivity study of coupling nuclear power to various combinations of desalination technology: {1} thermal (MSF [Multi-Stage Flashing], MED [Multi-Effect Distillation], and MED-TVC [Multi-Effect Distillation with Thermal Vapour Compression]); {2} membrane RO [Reverse Osmosis]; and {3} hybrid (MSF-RO [Multi-Stage Flashing & Reverse Osmosis] and MED-RO [Multi-Effect Distillation & Reverse Osmosis]). The Korean designed reactor plant, the APR1400 will be modeled as the energy production facility. The economical evaluation will then be executed using the computer program DEEP (Desalination Economic Evaluation Program) as developed by the IAEA. The program has capabilities to model several types of nuclear and fossil power plants, nuclear and fossil heat sources, and thermal distillation and membrane desalination technologies. The output of DEEP includes levelized water and power costs, breakdowns of cost components, energy consumption, and net saleable power for any selected option. In this study, we will examine the APR1400 coupled with a desalination power plant in the Kingdom of Saudi Arabia (KSA) as a prototypical example. The KSA currently has approximately 20% of the installed worldwide capacity for seawater desalination. Utilities such as power and water are constructed and run by the government. Per state practice, economic evaluation for these utilities do not consider or apply interest or carrying cost. Therefore, in this paper the evaluation results will be based on two scenarios. The first one assumes the water utility is under direct government control and in this case the interest and discount rate will be set to zero. The second scenario will assume that the water utility is controlled by a private enterprise and in this case we will consider different values of interest and discount rates (4%, 8%, & 12%).

• 대한기계학회논문집B
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• 제36권7호
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• pp.767-774
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• 2012

신형경수로인 APR1400의 안전주입탱크에는 유량조절장치가 설치되어있다. 이러한 유량조절장치는 안전주입탱크 내부의 수위에 따라 유량조절장치 내부에 위치한 와류실의 유동양식이 변하도록 설계되어 있어, 피동적으로 유량이 조절되는 특성을 가지고 있다. 그러나 고유량에서 저유량으로 전환되는 과정에서 유동의 관성에 의해 상부 기체가 방출되는 현상이 존재하여, FD 성능에 영향을 줄 수 있다. 본 논문에서는 전산유체역학 코드인 CFX 코드를 활용하여 안전주입탱크의 유량전환현상시 기체의 방출현상에 대해 이상유체 자유수면모델 적용하여 계산하였다. 이를 통해 안전주입수의 수위 및 상부 충전기체의 거동을 평가하여 FD 성능특성에 미치는 영향을 분석하였다. 그 결과, 유량전환구간에서 자유수면의 높이가 순간적으로 낮아지게 되어, 상부의 기체가 일부 방출되는 것으로 평가되었으나, FD 성능특성에는 큰 영향을 미치지 않는 것으로 평가되었다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• Nuclear Engineering and Technology
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• 제44권4호
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• pp.369-378
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• 2012

A load-following operation in APR+ nuclear plants is necessary to reduce the need to adjust the boric acid concentration and to efficiently control the control rods for flexible operation. In particular, a disproportion in the axial flux distribution, which is normally caused by a load-following operation in a reactor core, causes xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. A model predictive control (MPC) method was used to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for APR+ nuclear plants. Some tracking controllers employ the current tracking command only. On the other hand, the MPC can achieve better tracking performance because it considers future commands in addition to the current tracking command. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at the current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The support vector regression (SVR) model that is used widely for function approximation problems is used to predict the future outputs based on previous inputs and outputs. In addition, a genetic algorithm is employed to minimize the objective function of a MPC control algorithm with multiple constraints. The power level and ASI are controlled by regulating the control banks and part-strength control banks together with an automatic adjustment of the boric acid concentration. The 3-dimensional MASTER code, which models APR+ nuclear plants, is interfaced to the proposed controller to confirm the performance of the controlling reactor power level and ASI. Numerical simulations showed that the proposed controller exhibits very fast tracking responses.