• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.39-48
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• 2024

Research reactors are operated under ambient temperature and atmospheric pressure, which is much less severe conditions compared to those in typical nuclear power plants. Due to the high temperature, heat resistant materials such as austenite stainless steel should be used for the reactors in typical nuclear power plants. Whereas, as the effect of temperature is low for research reactors, materials with high resistance to neutron irradiation, such as zircaloy and beryllium, are used. Therefore, these conditions should be considered when performing integrity assessment for research reactors. In this study, a computational technique through finite element (FE) analysis was developed considering the operating conditions and materials of research reactor when conducting integrity assessment. Neutron irradiation analysis techniques using thermal expansion analysis were proposed to consider neutron irradiation growth and swelling in zirconium alloys and beryllium. A user subroutine program that can calculate the strain rate induced by neutron irradiation creep was developed for use in the commercial analysis program Abaqus. To validate the proposed technique and the user subroutine, FE analysis results were compared with hand-calculation results, and showed good agreement. Consequently, developed technique and user subroutine are suitable for evaluating structural integrity of research reactors.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.523-533
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• 2015

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-averaged Navier-Stokes flow simulation using a shear stress transport turbulence model. The main purpose of the current study is to understand the three-dimensional complex flow phenomena in a wire-wrapped fuel assembly to support the license issue for the core design. Computational fluid dynamics results show good agreement with friction factor correlation models. The secondary flow in the corner and edge subchannels is much stronger than that in an interior subchannel. The axial velocity averaged in the corner and edge subchannels is higher than that averaged in the interior subchannels. Three-dimensional multiscale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Behavior of the large-scale vortex structures in the corner and edge subchannels is closely related to the relative position between the hexagonal duct wall and the helically wrapped wire spacer. The small-scale vortex is axially developed in the interior subchannels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and the wire spacer.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.140-148
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• 1994

A qualification test was performed for the iron removal chemical cleaning of the secondary side of nuclear steam generators at the selected temperature, 1$25^{\circ}C$, higher than the standard application temperature, 93$^{\circ}C$. The field cleaning condition for a nuclear unit was tested in a bench scale test loop including a SUS 316 stainless steel autoclave with one gallon capacity as a test vessel. The kinetics of sludge dissolution, corrosion of the secondary side materials and change of solvent chemistry were monitored. Test results indicated that more thorough cleaning was accomplished in less than half of the cleaning time required at 93$^{\circ}C$. And the total corrosions of the secondary side materials were found to be less than the values at 93$^{\circ}C$. While the solvent is recirculated and heated by an external chemical cleaning equipment for the conventional 93$^{\circ}C$ process, the secondary side is heated by the lateral heat of the primary coolant without the recirculation of the cleaning solution, and the solvent is mixed by vigorous boiling induced by periodic ventilation for the high temperature process. The requirement that the reactor coolant pumps should be running during the cleaning operation is the major disadvantage of the high temperature process which also should be considered when chemical cleaning is planned for steam generators under operation.

• Journal of Energy Engineering
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• v.17 no.4
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• pp.218-226
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• 2008

A component-scale two-phase analysis code has been developed for a realistic simulation of two-phase flow transients in a light water nuclear reactor component. In the code, a two-fluid three-field model is adopted and the governing equations are solved on an unstructured mesh. For the numerical solution scheme, the semi-implicit method used in the RELAP5 code was selected, which has been proved to be very stable and accurate for most of practical applications. However, some modifications were needed for its application to an unstructured non-staggered grid. This paper presents the modified semi-implicit numerical method for unstructured grid and the preliminary results of the calculations.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.645-660
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• 1995

The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2001.09a
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• pp.67-74
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• 2001

This paper presents the Safe Shutdown Earthquake(SSE) input motion for the seismic design of the Advanced Power Reactor 1400(APR1400). The Design Ground Response Spectra(DGRS) far the SSE is based on the design spectrum specified in regulatory Guide(RG) 1.60 of U.S. Nuclear Regulatory Commission(US NRC), anchored to a Peak Ground Acceleration(PGA) of 0.3g and enriched in the high frequency range. This SSE seismic input motion is to be applied to the seismic analysis as the free-field seismic motion at the ground surface of both the rock and generic soil sites fur APRI1400. The enrichment for APR1400 seismic input motion is performed considering the current US NRC regulations, the seismic hazard studies performed by the Lawrence Livermore National Laboratory (LINL) and Electric Power Research Institute(EPRI) for the Central and Eastern United States nuclear power plant sites, and the seismic input motions used in the design certifications of the three existing U.S. advanced standard plants. It is represented by a set of DGRS and the accompanying Target Power Spectral Density(PSD) Function in both the horizontal and vertical directions.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.252-262
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• 1992

The intent of this study is to develop an efficient calculation method which can be used to analyze the heterogeneous time-dependent reactor problems. By using the nodal theory one can not only reduce the calculational efforts, but accurately determine the group dependent flux densities averaged over the entire homogeneous nodes. This method uses correction factors(called“discontinuity factors”) in a rigorous manner to obtain the relationship between the node-averaged flux and the surface-averaged fluxes and currents. The discontinuity factors are calculated from the node-averaged fluxes, diffusion coefficients, and the discontinuity factors of the previous time step. The test results for two benchmark problems demonstrate the accuracy and efficiency of the method developed for the transient application in which assembly-size nodes can be used.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.30-41
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• 2002

The neutron beam is fully characterized for the prompt gamma activation analysis facility at Hanaro in the Korea Atomic Energy Research Institute(KAERI). The facility uses thermal neutrons which are diffracted vertically from a horizontal beam port by a set of pyrolytic graphite(PG) crystals positioned at the Bragg angle of 45" Neutron spectra, neutron flux and Cd-ratio are determined for the three extraction modes of diffracted beam by means of the theoretical and experimental efforts. To obtain theoretical result, the reflectivity of pyrolytic graphite is calculated in the diffraction model for mosaic crystal and the angular divergence after diffraction by mosaic crystal is estimated from Monte Carlo simulation. The time-of-flight spectrometer and gold activation wire are used for measuring the neutron spectra. Both the calculated and measured spectra have proven that the unique feature of polychromatic beam obtained by PG crystals are useful for PGAA. The thermal neutron flux of 7.9$\times$107 n/cm$^2$s and the Cd-ratio of 266 for gold have been achieved at the sample position while the reactor operates at 24 MW The uniformity of beam flux is 12% in the central 1$\times$1 cm$^2$ area. Finally, the beam is briefly characterized by the effective velocity and temperature which are determined by measuring the prompt Y-ray spectra for thin and thick boron samples.ples.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.106-113
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• 2024

Understanding the tritium release and retention behavior of candidate tritium breeder materials is crucial for breeder blanket design. Recently, a melt spraying process was developed to prepare Li₄SiO₄ pebbles, which were subsequently subjected to the in-pile tritium production and extraction platform in China Mianyang Research Reactor (CMRR) to investigate their in-situ tritium release behavior and irradiation performance. The results demonstrate that HT is the main tritium release form, and adding hydrogen to the purge gas reduces tritium retention while increasing the HT percent in the purge gas. Post-irradiation experiments reveal that the irradiated pebbles darken in color and their grains swell, but the mechanical properties remain largely unchanged. It is concluded that the tritium residence time of Li₄SiO₄ made by melt spraying method at 467 ℃ is approximately 23.34 h. High-density Li₄SiO₄ pebbles exhibit tritium release at relatively low temperatures (<600 ℃) that is mainly controlled by bulk diffusion. The diffusion coefficient at 525 ℃ and 550 ℃ is 1.19 × 10^-11 cm²/s and 5.34 × 10^-11 cm²/s, respectively, with corresponding tritium residence times of 21.3 hours and 4.7 hours.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.239-245
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• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.