• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.10a
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• pp.859-862
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• 2002

The Core Support Barrel (CSB) is a major component of Reactor Internals, and is designed to support and protect the Reactor Core. In this study, Reactor Core, Core Shroud and CSB were simplified to coaxial cylinders and then the offset of Reactor Core & Core Shroud to the dynamic characteristic of CSB was analyzed. For the beam modes, natural frequencies of the cantilevered cylinder are compared with those of the cantilevered beam. And it was found out that shear modulus must be used correctly to convert the shell model to the equivalent beam model. From the dynamic characteristics of the beam model, it was found out that natural frequencies are proportional to the length of Reactor Core & Core Shroud and inversely proportional to the mass. From the comparison with the dynamic characteristics of a beam model and a lumped-mass model it was found out that the size of lumped-mass must be determined considering both the length and the mass of Reactor Core & Core Shroud.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2716-2727
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• 2021

The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using the ANSYS Design Parametric Language. Using the El Centro earthquake, seismic analyses are performed for the simplified and detailed core shroud models. Modal characteristics are obtained and their results are used for a time history analysis. Response spectrum analyses are also performed to access the degree of seismic excitation. The results of these analyses are compared to investigate the response characteristics between the simplified and detailed core shroud models from the time history and response spectrum analyses.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.80-99
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• 2023

This paper presents the results of investigating the change in welding residual stresses of the core shroud, which is one of subcomponents in reactor vessel internals, performing finite element analysis. First, the welding residual stresses of the core shroud were calculated by applying the heat conduction based lumped pass technique and finite element elastic-plastic stress analysis. Second, the temperature distribution of the core shroud during the normal operation was calculated by performing finite element temperature analysis considering gamma heating. Third, through the finite element viscoelastic-plastic stress analysis using the calculated temperature distribution and setting the calculated residual stresses as the initial stress state, the variation of the welding residual stresses was derived according to repeating the normal operation. In the viscoelastic-plastic stress analysis, the effects of neutron irradiation on mechanical properties during the cyclic normal operations were considered by using the previously developed user subroutines for the irradiation agings such as irradiation hardening/embrittlement, irradiation-induced creep, and void swelling. Finally, the effect of neutron irradiation on the welding residual stresses was analysed for each irradiation aging. As a result, it is found that as the normal operation is repeated, the welding residual stresses decrease and show insignificant magnitudes after the 10^th refueling cycle. In addition, the irradiation-induced creep/void swelling has significant mitigation effect on the residual stresses whereas the irradiation hardening/embrittlement has no effect on those.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.1236-1240
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• 2004

Thermal vacuum test for satellites should be performed before launch to verify the feasibility of satellites' operation in a harsh space environment which is represented as an extremely cold temperature and vacuum condition. A large space simulator(${\Phi}8m{\times}L10m$) has been demanded to accomplish the thermal vacuum test for the huge satellites designed in compliance with the national space program of Korea. In this paper, the design and calculation of thermal shroud which is the core part of large space simulator were discussed. The characteristics of the large space simulator being constructed at Korea Aerospace Research Institute(KARI) were depicted.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.355.1-355
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• 1999

• Structural Engineering and Mechanics
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• v.2 no.2
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• pp.157-171
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• 1994

A general approach to the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced form earthquake. The dynamic responses such as fuel assembly deflected shapes and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.637-642
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• 1996

This paper presents the analysis results for the core degradation processes and the fission product release of the PHEBUS FPT0 experiment using MELCOR1.8.3. The objective of this study is to assess models associated with the core damage and fission product behavior in MELCOR. The calculation results were much improved through sensitivity studies. Thermal/hydraulic behavior in the core and the circuit was well predicted under the intact core geometry. In non-eutectic model case. the UO$_2$ dissolution model in the MELCOR always showed such a tendency that the resulting dissolved UO$_2$ mass was small at the highly oxidized condition due to the model logic. Total H$_2$ generation mass was underpredicted because the stiffner was not modeled and the liner in the shroud was not allowed to be oxidized in MELCOR. Some difficulties were found in modeling the activation product were solved by manipulating the RN input associated with the initial fission product inventory. These problem were occurred because there are no control rod model in MELCOR. Generally the fission product release ratio showed a similar trend compared with the measured data except the activation product. which have no model to simulate in MELCOR.

• Nuclear Engineering and Technology
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• v.20 no.4
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• pp.233-240
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• 1988

The hafnium shroud is used to control the excess reactivity and power distribution in KMRR. The core analysis is performed by the diffusion code VENTURE using the 5 group macroscopic cross sections homogenized for an assembly. Investigated are the applicability of the diffusion calculation by homogenized cross sections to the analysis of control assembly which features unusual geometry such that hafnium shroud surrounds a multiplying medium inside. Comparative calculation is performed for the excess reactivity and power levels by the transport code TWOTRAN. The results show the acceptability of the diffusion calculation by the homogenized cross sections without significant error.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.111-120
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• 1992

This paper investigates the dynamic characteristics of spacer grid impact loads and the effects of variations in the amplitude and frequency of the core plate motions on the resultant impact loads. A model of the longest row (15 fuel assemblies) across the core is analyzed using the input motions generated from safe shutdown earthquake. Input excitations consist of time history motions applied to the core support plate, fuel alignment plate and core shroud. The responses are determined for a set of four parameter runs with respect to the amplitude and frequency changes. Spacer grid impact loads and normalized input values for all cases are presented. The results show that changing the natural frequency has negligible effect but changing the amplitude of the input motions has a significant effect on the grid impact loads Therefore, time history analysis is not necessary for a shifted case to get the core responses under the seismic excitation.