• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.167-181
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• 2019

Currently, the development of evaluation technology for vibration and shock load characteristics and spent nuclear fuel structural integrity under normal conditions of transport is being conducted in the Republic of Korea. This is the first such research conducted in the Republic of Korea and, thus, previous international studies need to be investigated and will be referred to in the ongoing project. Before 2000, several studies related to measurement of vibration and shock loads on spent nuclear fuel were conducted in the US. US national research institutes conducted uniaxial fuel assembly shaker tests, concrete block tests, and multi-axis fuel assembly tests between 2009 and 2016. In 2017, multi-modal transportation tests including road, sea, and rail transport were also performed by research institutes from the US, Spain and the Republic of Korea. Therefore, test preparation procedures, acceleration and strain measurement results, and finite-element and multi-body dynamics analysis were investigated. Based on the measured strain data, the preliminary conclusion was obtained that the measured strain was too small to cause damage to spent nuclear fuel rods. However, this conclusion is a preliminary conclusion that only reviews part of the results; a detailed review is being conducted in the US. The investigation of international studies on spent nuclear fuel structural integrity tests considering vibration and shock loads under normal conditions of transport in the US will be useful data for the project being conducted in the Republic of Korea.

• Corrosion Science and Technology
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• v.19 no.5
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• pp.231-238
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• 2020

Radioactive corrosion product specimens were analyzed using an electron probe microanalyzer (EPMA) and X-ray image mapping. It is difficult to analyze the composition of radioactive corrosion products using an EPMA due to the size and rough shape of the surfaces. It is particularly challenging to analyze the composition of radioactive corrosion products in the form of piled up, small grains. However, useful results can be derived by applying a semi-quantitative analysis method using an EPMA with X-ray images. A standard-less, semi-quantitative method for wavelength dispersive spectrometry. EPMA analysis was developed with the objective of simplifying the analytical procedure required. In this study, we verified the reasonable theory of semi-quantitative analysis and observed the semi-quantitative results using a sample with a good surface condition. Based on the validated results, we analyzed highly rough-surface radioactive corrosion products and assessed their composition. Finally, the usefulness of the semi-quantitative analysis was reviewed by verifying the results of the analysis of radioactive corrosion products collected from spent nuclear fuel rods.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.15-20
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• 1997

We present the results HELIOS verification against VENUS PWR critical experiments loaded with high plutonium content mixed oxides fuels. The effective multiplication factors are calculated to be slightly supercritical within an acceptable error bound. In the prediction of power shape, HELIOS results are in close agreement with the measured values. The RMS errors of re-normalized calculated fission rate distribution are less than 1.4 % with either explicit or implicit models or micro tubes/rods in each fuel assembly for both ALL-MOX and GD-MOX mock-up cores.

• 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.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.509-514
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• 2005

Main coolant pump (MCP) is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel rods and steam generators in an integral type reactor. The reactor is designed to operate under condition of 310 oC and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition in order to verify its performance and safety. In present work, a test loop to simulate real operating situation of the reactor has been designed and constructed to test MCP. And then, as a part of qualification test, canned motor functional test and pump hydraulic performance test have been carried out upon a prototype MCP. Canned motor efficiency and pump hydraulic characteristics including homologous curves and NPSH curves were obtained from the qualification test.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.22-22
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• 1999

The HELlOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC "Kurchatov Institute" with water-moderated hexagonally pitched lattices of highly enriched Uranium fuel rods (8Ow/o). We also used the same input by using the MCNP code that was described in the evaluation report, and compared our results with those of the evaluation report. HELlOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves neutronics model with the AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that the HELIOSMASTER code system is fast and accurate enough to be used as nuclear core analysis tool for hexagonal geometry.ometry.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.225-230
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• 1996

A three-dimensional reactor core simulation code, MASTER has been developed as a part of ADONIS which is the Korean core design package in KAERI. CASMO-3 is used as a precedent lattice code for two-group microscopic cross section and heterogeneous formfunctions. The pin power reconstruction capability of CASMO-3/MASTER was evaluated for a validation and verification Five B&W critical experiments were selected as benchmark problems. These problems included two experiments for CE-type and three for WH-type fuel assemblies. Two of them contained gadolinia rods as burnable absorber. Comparison of the calculated pin power distributions with the measured ones demonstrate that CASMO-3/MASTER can predict the pin power distribution as well as CASMO-3/SIMULATE-3.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.5
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• pp.437-447
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• 2016

In a nuclear power plant, the fuel assembly, which is composed of fuel rods, burns, and the high temperature can generate power. The fuel rod consists of pellets and a cladding that covers the pellets. It is important to understand the pellet-cladding mechanical interaction with regard to nuclear safety. This paper proposes simulation of the PCMI. The gap between the pellets and the cladding, and the contact pressure are very important for conducting thermal analysis. Since the gap conductance is not known, it has to be determined by a suitable method. This paper suggests a solution. In this study, finite element (FE) contact analysis is conducted considering thermal expansion of the pellets. As the contact causes plastic deformation, this aspect is considered in the analysis. A 3D FE module is developed to analyze the PCMI using FORTRAN 90. The plastic deformation due to the contact between the pellets and the cladding is the major physical phenomenon. The simple analytical solution of a cylinder is proposed and compared with the fuel rod performance code results.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.444-455
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• 1991

An accurate and fast running NEDAR model for calculating radial power profile throughout fuel life in both solid and annular pellets for existing and advanced CANDU-PHWR-fuel was developed in this work. This model contains resultant flux depression equations and neutron depression data tables which have been developed for CANDU-PHWR fuel of pellet with the diameter 8.0 to 19.5 mm and enrichment 0.71-6.0 wt % U-235, over a bumup range of 0 to 840 MWh /kgU (35000 MWD/T). In order to obtain the neutron flux distribution in the fuel pellet, the CE-HAMMER physics code was run for a neutron flux spectrum appropriate to a CANDU-PHWR to give predictions of radial power profile for several ranges of fuel design parameters. The results, which were calculated by the CE-HAMMER physics code, were fitted to an equation which is solved in terms of Bessel and exponential functions in order to obtain the parameters, $textsc{k}$, $\beta$ and λ in the resultant equation. The present NEDAR model produce a radial profile which, when normalized to unity at the pellet surface, is slightly higher than the profile of the original ELESIM data table. The predictions of the fission gas release by KAFEPA-NEDAR are in slightly better agreement with the experiments than those of ELESIM. The NEDAR model described in this study has been shown to provide an effective, reliable, and accurate method for determining radial power profiles in CANDU-PHWR fuel rods without incurring a significant increase in computing time.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.344-351
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• 2020

AgInCd alloy is widely used as neutron absorber in nuclear reactors. However, the AgInCd control rods may fail during service due to the irradiation swelling. In the present study, a calculational method is proposed to calculate the composition change of the AgInCd absorber. Calculated results show that neutron fluence has significant impact on the chemical compositions. Ag and In contents gradually decrease while Cd and Sn conversely increases from the center to the rim of AgInCd absorber due to the depression of neutron flux. The composition change at the surface is higher almost two times than that at the center. Based on the calculated compositions, six simulated AgInCdSn alloys were prepared and examined. With the increase of Cd and Sn, the simulated AgInCdSn alloys transform from a single fcc phase into the mixed fcc and hcp phases, and finally into the single hcp phase. The atomic volume of the hcp phase is obviously larger than the fcc phase. The fcc-hcp transformation results in considerable volume swelling of the AgInCd absorber. Moreover, the lattice parameters of the fcc and hcp phases gradually increase with Cd and Sn contents, which also can induce small volume swelling.