• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2002년도 추계학술대회 논문집
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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.

• 한국압력기기공학회 논문집
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• 제15권2호
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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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• 제53권8호
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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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• 제55권1호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.355.1-355
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• 1999

• Structural Engineering and Mechanics
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• 제2권2호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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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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• 제20권4호
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• pp.233-240
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• 1988

KMRR은 잉여반응도 및 출력분포조절을 위하여 Hafnium관을 사용한다. 현재 노심해석은 핵연료집합체 단위로 균질화된 5군 군정수를 이용하여 중성자 확산코드인 VENTURE을 이용한다. Hafnium관내에 핵연료가 들어있는 특수한 조정 집합체에 대해서도 이러한 균질화된 군정수를 사용한 중성자 확산계산이 적용될 수 있는가를 조사하였다. 비교계산은 중성자 수송코드인 TWOTRAN을 사용하여 잉여반응도 및 출력 준위에 대해 수행하였다. 계산결과 현재의 균질화된 군정수를 사용하는 중성자 확산계산이 큰 오차없이 적용할 수 있는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제24권2호
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• pp.111-120
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• 1992

본 논문은 안전정지지진에 대한 spacer grid충격하중의 동적특성을 고찰한 것이다. 안전정지지 진 (SSE： Safe Shutodown Earthquake)을 가진력으로 사용하였고 모델로는 가장 긴 15 row의 핵연료집합체 모델을 사용하였다. 외부에서 제공되는 입력데이타의 불확실성을 보상하기 위하여 흔히 사용되는 10%증가된 입력값과, NRC 권고사항인 입력응답스펙트럼의 이동효과의 영향이 평가되었다. 평가대상으로는 spacer grid의 충격력이 사용되었다. 해석결과로서 입력응답 스펙트럼의 이동효과의 영향은 거의 무시할 수 있으나 증가된 입력값의 영향은 spacer grid의 충격력에 큰 영향을 줌을 알 수 있었다. 따라서 원자로노심의 동적응답을 위해서는 입력응답 스펙트럼을 이동시키면서 해석할 필요가 없음을 확인하였다.