• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2670-2689
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• 2022

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

• Korean Journal of Computational Design and Engineering
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• v.9 no.4
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• pp.416-424
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• 2004

In this paper, the heat transfer analysis and thermal stress analysis of the PCB(Printed Circuit Board) equipped in electronic Packages are carried out for various may types of chips on the PCB. And two structural PCB models are used in the analyses. The electronic chips on the PCB usually emit heat and this heat generates the thermal stress around the chip. The thermal load due to the heat generation of chips on the PCB may cause the malfunction of the electronic packages such as a monitor. a computer etc. Hence, the PCB should be designed to withstand these thermal loads. In this paper, the heat transfer analysis and thermal stress analysis are executed for the PCB model with pins and the analysis results are compared with the results for the PCB model without pins. The analysis results show that the PCB model without pins is not good for the thermal stress analysis of PCB, even though these two models have similar heat transfer characteristics. The analysis results also show that the highest thermal stress occurs in the pin especially attached to the highest temperature chip, and the PCB constrained to the electronic package on the long side is structurally more stable than other cases. The analyses of the PCB are executed using the finite element analysis code, NISA.

• International Journal of Concrete Structures and Materials
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• v.10 no.3
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• pp.393-406
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• 2016

Axially loaded reinforced concrete columns are hardly exist in practice due to the development of some bending moments. These moments could be produced by gravity loads or the lateral loads. First, the current paper presents a detailed analysis on the overall structural behavior of 15 eccentrically loaded columns as well as one concentrically loaded control one. Columns bent in either single curvature or double curvature modes are tested experimentally up to failure under the effect of different end eccentricities combinations. Three end eccentricities ratio were studied, namely, 0.1b, 0.3b and 0.5b, where b is the column width. Second, an expression correlated the decay in the normalized axial capacity of the column and the acting end eccentricities was developed based on the experimental results and then verified against the available formula. Third, based on the equivalent column concept, the equivalent pin-ended columns were obtained for columns bent in either single or double curvature modes. And then, the effect of end eccentricity ratio was correlated to the equivalent column length. Finally, a simplified design procedure was proposed for eccentrically loaded braced column by transferring it to an equivalent axially loaded pin-ended slender column. The results of the proposed design procedure showed comparable results against the results of the ACI 318-14 code.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.751-757
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• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Journal of the Korean Institute of Telematics and Electronics D
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• v.36D no.1
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• pp.47-56
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• 1999

As the density of logic gates of VLSI chips has rapidly increased, more number of I/O pins has been required. This results in bigger package size and higher packager cost. The package cost is higher than the cost of bare chips for high I/O count VLSI chips. As the density of logic gates increases, the reduction method of the number of I/O pins for a given complexity of logic gates is required. In this paper, we propose the novel I/O interface circuit using CTR (Constant-Transition-Rate) code to reduce 50% of the number of I/O pins. The rising and falling edges of the symbol pulse of CTR codes contain 2-bit digital data, respectively. Since each symbol of the proposed CTR codes contains 4-bit digital data, the symbol rate can be reduced by the factor of 2 compared with the conventional I/O interface circuit. Also, the simultaneous switching noise(SSN) can be reduced because the transition rate is constant and the transition point of the symbols is widely distributed. The channel encoder is implemented only logic circuits and the circuit of the channel decoder is designed using the over-sampling method. The proper operation of the designed I/O interface circuit was verified using. HSPICE simulation with 0.6 m CMOS SPICE parameters. The simulation results indicate that the data transmission rate of the proposed circuit using 0.6 m CMOS technology is more than 200 Mbps/pin. We implemented the proposed circuit using Altera's FPGA and confimed the operation with the data transfer rate of 22.5 Mbps/pin.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.44-60
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• 2021

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• Korean Journal of Optics and Photonics
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• v.26 no.5
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• pp.269-274
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• 2015

In this paper we design the optical system for an integrated two-channel TO-type optical transmitter to apply the HDMI interface using the code V simulator. The proposed integrated two-channel optical transmitter has two VCSELs attached in parallel on an 8-pin TO-CAN package, on top of which is a lens filter block ($1mm{\times}2mm{\times}4mm$) composed of hemispherical lenses and WDM filters. Considering two-channel transmitters manufactured with wavelength combinations of 1060nm/1270nm and 1330nm/1550nm, we obtain the optimum value of the diameter of the hemispherical lens as 0.6 mm for both combinations, and the distances L between the lens filter block and ball lens as 1.7 mm and 2.0 mm for the 1060nm/1270nm and 1330nm/1550nm wavelength combinations, respectively. At this time, the focal length f0 of the lens filter blocks for wavelengths of 1060, 1270, 1330, and 1550 nm are 0.351, 0.354, 0.355, and 0.359 mm, respectively, and the focal lengths F of light passing through the lens filter block and ball lens are 0.62 mm for 1060nm/1270nm and 0.60-0.66 mm for 1330nm/1550nm wavelength combinations.