• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1291-1302
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• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• Nuclear Engineering and Technology
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• 제28권5호
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• pp.488-499
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• 1996

We have extended the source projection analytic nodal discrete ordinates method (SPANDOM) for more flexible applicability in analysis of hexagonal assembly cores. The method (SPANDOM-FH) does not invoke transverse integration but instead solves the discrete ordinates equation analytically after the source term is projected and represented in hybrid form of high-order polynomials and exponential functions. SPANDOM-FH which treats a hexagonal node as one node is applied to two fast reactor benchmark problems and compared with TWOHEX. The results of comparison indicate that the present method SPANDOM-FH predicts accurately $k_eff$ and flux distributions in hexagonal assembly cores. In addition, SPANDOM-FH gives the continuous two dimensional intranodal scalar flux distributions in a hexagonal node. The reentering models between TWOHEX and SPANDOM were also compared and it was confirmed that SPANDOM's model is more realistic. Through the results of benchmark problems, we conclude that SPANDOM-FH has the sufficient accuracy for the nuclear design of fast breeder reactor (FBR) cores with hexagonal assemblies.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2004년도 하계학술대회 논문집 A
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• pp.633-635
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• 2004

The current limiting characteristics of the flux-lock type superconducting fault current limiters(SFCLs) were investigated. The flux-lock type SFCL consists of a flux-lock reactor and high-T_c superconducting (HTSC) element. In this SFCL the initial limiting current level can be controlled by adjusting the inductances of two coils. In this paper, the operational characteristics of the flux-lock type SPCL were analyzed and the current Limiting characteristics of it were investigated through the experiments.

• 멤브레인
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• 제17권2호
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• pp.112-117
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• 2007

The application of membrane bioreactors for the depuration of wastewater coming from the washing of mineral oil storage tanks is described. Microfiltration hollow-fibre membranes were used in the submerged configuration. Filtration tests were carried out with a biomass concentration of about 15 g/L in order to assess the critical flux of the hollow fibre membrane used. Then particular care was taken in carrying out the performance runs in the sub-critical flux region. The reactor performance was very high, with removal efficiencies ranging between 93% and 97% also when the concentration of hydrocarbon was very high. Some kinetic parameters for the COD and the hydrocarbon removal were estimated.

• Progress in Superconductivity
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• 제3권2호
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• pp.229-234
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• 2002

In this paper, we analyzed the fault current performance in a $high-T_{c}$ superconductor(HTS) which was installed on flux-lock reactor with an external magnetic field coil covering the HTS. In this HTS fault current limiter using flux-lock concepts, the initial limiting current level can be controlled by adjusting the inductance of the coils. Furthermore, the current limiting characteristics of $high-T_{c}$ superconducting FCL can be improved by applying the external magnetic field into the $high-T_{c}$ superconductor. This paper discusses current limiting performance according to the inductance of the coil 1 in two cases with ac magnetic field coil or not and suggests the methods to improve the current limiting factor $P_{limit}$, which is defined as the ratio of the limited current $I_{FCL}$ at the current limiting phase to the prospective short -circuit current $I_{PSC}$.TEX> PSC/.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.348-359
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• 1997

This paper reviews cooling aspects of the diverter system in Tokamak fusion devices with primary emphasis on the critical heat flux (CHF) issues for oater-cooled designs. General characteristics of four (4) coolant options for diverter cooling gases, oater, liquid metal, and organic liquid - are discussed first, focusing on the comparison of advantages and disadvantages of those options. Then results of recent studies on the high-heat-flux CHF of water at subcooled high-velocity conditions are reviewed to provide a general idea on the feasibility of the water-cooled diverter concept for future Tokamak fusion reactors. Water is assessed to be the most viable and practical coolant option for diverters of future experimental Tokamaks.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1100-1108
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• 2021

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirable high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal margins for safe reactor operation. In the present work, the heat flux corresponding to onset of HTD for vertically upward flow of supercritical water in a pipe is obtained over a wide range of system parameters, namely pressure, mass flux, and pipe diameter. This is done by performing large number of simulations using an in-house CFD code, which is especially developed and validated for this purpose. The identification of HTD is based on observance of one or more peak/s in the computed wall temperature profile. The existing correlations for predicting the onset of HTD are compared against the results obtained by present simulations as well as available sets of experimental data. It is found that the prediction accuracy of the correlation proposed by Dongliang et al. is best among the existing correlations.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.479-490
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• 2005

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.111-112
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• 2005