• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.119-128
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• 2006

We start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing the advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the twophase flow simulation code TPFIT with the advanced interface tracking method. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results of numerical simulation, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values obtained by the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.426-427
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• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.248-254
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• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.847-858
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• 2016

An overview is given on the work of the Laboratory of Nuclear Energy Systems at ETH, Zurich (ETHZ) and of the Laboratory of Thermal Hydraulics at Paul Scherrer Institute (PSI), Switzerland on tight-lattice bundles. Two-phase flow in subchannels of a tight triangular lattice was studied experimentally and by computational fluid dynamics simulations. Two adiabatic facilities were used: (1) a vertical channel modeling a pair of neighboring sub-channels; and (2) an arrangement of four subchannels with one subchannel in the center. The first geometry was equipped with two electrical film sensors placed on opposing rod surfaces forming the subchannel gap. They recorded 2D liquid film thickness distributions on a domain of $16{\times}64$ measuring points each, with a time resolution of 10 kHz. In the bubbly and slug flow regime, information on the bubble size, shape, and velocity and the residual liquid film thickness underneath the bubbles were obtained. The second channel was investigated using cold neutron tomography, which allowed the measurement of average liquid film profiles showing the effect of spacer grids with vanes. The results were reproduced by large eddy simulation + volume of fluid. In the outlook, a novel nonadiabatic subchannel experiment is introduced that can be driven to steady-state dryout. A refrigerant is heated by a heavy water circuit, which allows the application of cold neutron tomography.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.56-71
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• 1996

The local hydrulic characteristics in subchannels of 5$\times$5 nuclear fuel bundles with spacer grids were measured at upstream and downstream of the spacer grid for the investigation of the spacer grid effects on turbulent flow structure by using an LDV(Laser Doppler Velocimeter). The measured parameters are axial velocity and turbulent intensity, skewness factor, and flatness factor. Pressure drops were also measured to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. From these data, it was found that the turbulent mixing and forced mixing occur up to $x/D^h=10$ and 20 from the spacer grid, respectively. The turbulence decay behind spacer grid behaves in the similar decay rate as turbulent flow through mesh grids or screens. Mixing factors useful in subchannel analysis code were correlated from the data and show the highest value near spacer grid and then have a stable values.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.301-307
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• 2005

The morphology of the high burnup $UO_2$ spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent $UO_2$ fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at $1400^{\circ}C$ for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to $500^{\circ}C$ and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the $UO_2$ matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to $10\;{\mu}m$.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.579-588
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• 2022

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module Invessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for the formulation of severe accident mitigation measures, the article select three experiments to evaluate the updated severe accident models of MIDAC. Among them, QUENCH-06 is the international standard No.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod melting experiment recently carried out by Xi'an Jiaotong University. The heating and melting model with lumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarev correlations combination in MIDAC could better meet the needs of severe accident analysis. Although the influence of nitrogen still need to be further improved, the air oxidation model with NUREG still has the ability to provide guiding significance for engineering practice.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.184-189
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• 1997

A prompt fission gas release model incorporating the resistance to gas flow in the gap was developed and the effects of gap resistance and failure location on prompt fission gas release from the cladding breach were assessed. The process of prompt fission gas release from the plenum and gap into the coolant was modeled in accordance with three major phenomena: (1) transient gas flow in the gap, (2) the growth of the fission gas bubble while it is still attached to the breach, and (3) the detachment of the fission gas bubble from the breach and mixing with the coolant. The cumulative mass release fraction by the present model was calculated for the case of Young-Gwang 3 & 4 nuclear fuel rod as a typical example. The results showed that the release behavior of prompt fission gas with time was different from the frictionless model which has frequently been used in a simplified approach, and that the location of cladding failure was another key factor for the prompt fission gas release process due to the resistance in the gap.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.445-450
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• 1997

Self-Powered Neutron Detectors(SPNDs) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. In this paper, Monte Carlo simulation is accomplished to calculate the escape probability of beta particle as a function of their birth position fur the typical geometry of rhodium-based SPNDs. Also, a simple numerical method calculates the initial generation rate of beta particles and the change of generation rate due to rhodium burn-up. Using the simulation and the numerical method, the burn-up profile of rhodium density and the neutron sensitivity are calculated as a function of burn-up time in the reactor. The sensitivity of the SPNDs decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. In addition, for improvement of some properties of rhodium-based SPNDs which are currently used, this paper presents a new material. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long term usage.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.178-182
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• 1992

Spent PWR fuel rods have been scanned axially and sectionally to measure the relative gamma-ray intensity of Cs-137 and then bumups of the scanned rods determined by measuring Nd-148 which has been chemically separated. From these experimental results, a linear relation(LR) between the gamma-ray intensity of Cs-137 and the bumup in the range of 10∼35 GWD/MTU was obtained. In order to validate the LR, the Cs-137 gamma-ray intensity of unknown sample was nondestructively measured and the bumup obtained by the LR was compared with that of the Nd-148 method. It is revealed that the results from both methods are in good agreement, and thus it seems to be possible to estimate the bumup of spent PWR fuel rod by measuring nondestructively gamma-ray of fission product Cs-137.