• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1514-1524
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• 2019

Hydrogen-steam gas mixture may be injected into containment with flow regime varying both spatially and transiently due to wall effect and pressure difference between primary loop and containment in severe accidents induced by loss of coolant accident. Preliminary CFD analysis is conducted to gain information about the helium flow regime transition process from jet to buoyancy plume for forthcoming experimental study. Physical models of impinging jet and wall condensation are validated using separated effect experimental data, firstly. Then helium transportation is analyzed with the effect of jet momentum, buoyancy and wall cooling discussed. Result shows that helium distribution is totally dominated by impinging jet in the beginning, high concentration appears near gas source and wall where jet momentum is strong. With the jet weakening, stable light gas layer without recirculating eddy is established by buoyancy. Transient reversed helium distribution appears due to natural convection resulted from wall cooling, which delays the stratification. It is necessary to concern about hydrogen accumulation in lower space under the containment external cooling strategy. From the perspective of experiment design, measurement point should be set at the height of connecting pipe and near the wall for stratification stability criterion and impinging jet modelling validation.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Clean Technology
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• v.17 no.1
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• pp.69-77
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• 2011

The intrinsic $CO_2$ separation and hydrogen production system is a novel concept using oxidation and reduction reactions of oxygen carrier for both $CO_2$ capture and high purity hydrogen production. The process consists of a fuel reactor (FR), a steam reactor (SR) and an air reactor (AR). The natural gas ($CH_4$) is oxidized to $CO_2$ and steam by the oxygen carrier in FR, whereas the steam is reduced to hydrogen by oxidation of the reduced oxygen carrier in SR. The oxygen carrier is fully oxidized by air in AR. In the present study, the chemical looping moving bed reactor having 200 L/h hydrogen production capacity is designed and the hydrodynamic properties were determined. Compared with other reactors, two moving bed reactors (FR, SR) were used to obtain high conversion and selectivity of the oxygen carrier. The desirable solid circulation rates are calculated to be in the range of $20{\sim}100kg/m^2s$ from the conceptual design. The solid circulation rate can be controlled by aeration in a loop-seal. To maintain the gas velocity in the moving beds (FR, SR) at the minimum fluidization velocity is found to be suitable for the stable operation. The solid holdup in moving beds decrease with increasing gas velocity and solid circulation rate.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.383-387
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• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.1
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• pp.152-159
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• 2010

This paper describes the actual application of a feedback control loop as a means for minimizing turbine impulse chamber pressure variation during the turbine steam valve tests at a 1,000 MW nuclear power plant. The chamber pressure control loop was implemented in the new digital control system which was installed as a replacement for the old analog type control system. There has been about 40MW of the generator output change during the steam valve tests, especially the high pressure governing valve tests, because the old control system had not the impulse chamber pressure control so the operators had to compensate steam flow drop manually. The process of each valve test consists of a closing process and an reopening process and the operators can make sure that the valves are in their sound conditions by checking the valves movement. The control algorithm described in this paper contributed to keep the change in megawatt only to 6MW during the steam valve tests. Thereby, the disturbance to reactor control was reduced, and the overall plant control system's stability was greatly improved as well.

• Journal of the Korean Society of Safety
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• v.33 no.3
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• pp.92-98
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• 2018

The flow of cooling water in a passive containment cooling system (PCCS), used to remove heat released in design basis accidents from a concrete containment of light water nuclear power plant, was conducted in order to investigate the thermo-fluid equilibrium among many parallel tubes of PCCS. Numerical simulations of the subcooled boiling flow within a coolant loop of a PCCS, which will be installed in innovative pressurized-water reactor (PWR), were conducted using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. As the first step, the simplified geometry of PCCS with 36 tubes was modeled in order to reduce computational resource. Even and uneven thermal loading conditions were applied at the outer walls of parallel tubes for the simulation of the coolant flow in the PCCS at the initial phase of accident. It was observed that the natural circulation maintained in single-phase for all even and uneven thermal loading cases. For uneven thermal loading cases, coolant velocity in each tube were increased according to the applied heat flux. However, the flows were mixed well in the header and natural circulation of the whole cooling loop was not affected by uneven thermal loading significantly.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.311-318
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• 1992

During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.137-143
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• 2011

Fuel Test Loop(FTL) is a facility which could conduct a fuel irradiation test at HANARO(High-flux Advanced Neutron Application Reactor). FTL simulates commercial NPP's operating conditions such as the pressure, temperature and neutron flux levels to conduct the irradiation and thermo-hydraulic tests. The In-Pile Test Section(IPS) installed in HANARO FTL is designed as a pressure vessel design conditions of $350^{\circ}C$, 17.5MPa. The instrumentation MI-cables for thermocouples, SPND and LVDT are passed through the sealing plug, which is in the pressure boundary region and is a part of instrumentation feedthrough of MI-cable. In this study, the brazing method and performance test results are introduced to the sealing plug with BNi-2 filler metal, which is selected with consideration of the compatibility for the coolant. The performance was verified through the insulation resistance test, hydrostatic test, and helium leak test.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1429-1435
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• 2021

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation. The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.