• Proceedings of the Materials Research Society of Korea Conference
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• 2002.11a
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• pp.193-193
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• 2002

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3485-3490
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• 2007

The TEXAS-V code tuned for TROI-13 was used for analyzing the parametric findings in TROI experiments. The calculations on the melt composition are relatively similar to the TROI experimental results. The water depth effect in TEXAS-V code seems to be consistent with TROI experiments in some degree. The water area effect of TEXAS-V calculations seems not to be harmonious to that in TROI experiments. This seems to indicate that TEXAS-V as 1-dimensional code or as the numerical steam explosion has a limitation on estimating area effect. Thus, TEXAS-V tuned for TROI-13 seems to have an ability to estimate the parametric effect of TROI experiments. The evaluated TEXAS-V was used for estimating the ex-vessel steam explosion load. The calculated explosion pressure and load were about 40 MPa and 75 kPa.sec, which are not much threatening level for containment integrity, but are arguable value for the integrity.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.10
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• pp.1043-1047
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• 2012

The European nuclear design requirements, which should be satisfied by nuclear reactors in Europe, usually recommend a so-called core catcher, which is a molten core ex-vessel cooling facility, to manage a severe accident at a nuclear reactor. Two different types of core catcher concepts are compared to determine their abilities to manage severe accidents and cool core melts. The study reveals that direct cooling is better for cooling capacity and is convenient to construct, while indirect cooing is better for the management of a severe accident.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1854-1858
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• 2007

The reliability assurance with respect to the test procedure and results of the out-pile mechanical performance test for the nuclear fuel assembly is an essential task to assure the test quality and to get a permission for fuel loading into the commercial reactor core. For the case of vibration test, which is carried out to obtain basic dynamic characteristics of the fuel assembly, proper management and appropriate calibration of instruments and devices used in the test, various efforts to minimize the possible error during the test and signal acquisition process are needed. Additionally, the deep understanding both of the theoretical assumption and simplification cation for the signal processing/modal analysis and of the functions of the devices used in the test were highly required. Finally, to verify the test result to represent the accurate natural characteristics of the structure, the proper correlation analysis between the theoretical and experimental method has to be carried out. In this study, the overall procedure and result of lateral vibration test for the fuel assembly's mechanical characterization were briefly introduced. A series of measures to assure and improve the reliability of the vibration test were discussed.

• Transactions of the KSME C: Technology and Education
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• v.3 no.3
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• pp.209-215
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• 2015

The effects of the mounting location of ICI cables on severe accident mitigation systems, specially IVR-ERVC (In-Vessel Retention by External Reactor Vessel Cooling) and core catcher (Ex-vessel corium retention and cooling system), are investigated. The effects of bottom-mounted ICI strategy on severe accident mitigation are summarized and advantages of top-mounted ICI to improve severe accident mitigation are also highlighted.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.8
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• pp.622-628
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• 2009

The TEXAS-V code tuned for TROI-13 was used for analyzing the parametric findings in TROI experiments. The calculations on the melt composition are relatively similar to the TROI experimental results. The water depth effect in TEXAS-V code seems to be consistent with TROI experiments in some degree. The water area effect of TEXAS-V calculations seems not to be harmonious to that in TROI experiments. This seems to indicate that TEXAS-V as 1-dimensional code or as the numerical steam explosion has a limitation on estimating area effect. Thus, TEXAS-V tuned for TROI-13 seems to have an ability to estimate the parametric effect of TROI experiments. The evaluated TEXAS-V was used for estimating the ex-vessel steam explosion load. The calculated explosion pressure and load were about 40 MPa and 75 kPa.sec, which are not much threatening level for containment integrity.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.102.1-102.1
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.703-710
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• 1996

본 연구에서는 영광 3,4호기 원자로의 Core Support Barrel 거동을 중성자 잡음해석을 이용하여 분석하였다. 분석 방법은 원자로 노외계측기에서 취득한 교류 성분의 중성자 잡음 신호를 주파수분석하므로서 얻어진 PSD, Phase, Coherence 등을 이용하였다. 영광 3,4호기의 1 주기 동안의 신호를 분석결과, CSB의 Beam Mode 주파수는 영광 3호기의 경우 BOL, MOL, EOL에서 각각 7.75∼8.5Hz, 7.75Hz, 7.25∼7.75Hz로 나타났고, 영광 4호기 BOL에서 8.5∼8.75Hz 임이 도출되었다. 본연구 결과는 한국형 원전의 원자로 내부구조물의 진동 특성을 파악하고 운전중 CSB건전성 진단을 위한 기초 자료로 활용할 수 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.663-668
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• 1996

하나로를 이용한 재료조사시험용 계장캡슐 개발에 앞서 캡슐 mock-up (96M-01K)을 제작하였으며, 이 캡슐 mock-up의 실제 하나로 조사시험공 장입시를 가정하여 강도 및 열적 건전성 평가를 수행하였다. 평가 결과 하나로 정상출력시 (30MW) 캡슐 mock-up 내 조사시료의 온도는 진공 및 heating system을 사용하여 279~473$^{\circ}C$ 범위로 조절될 수 있었으며, 목표 조사기간 동안 캡슐 mock-up은 강도적으로 허용기준을 충분히 만족함으로써 안전한 것으로 판명되었다. 향후 본 캡슐 mock-up을 이용한 노외 simulation 실험 등을 통하여 기존 캡슐 mock-up의 건전성을 확인한 후 이를 기준으로 하여 표준형 하나로 캡슐을 설계 .제작하고자 한다.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.158-167
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• 2007

It is required that the neutron dosimetry be present to monitor the reactor vessel throughout its plant life. The Ex-vessel Neutron Dosimetry Systems which consist of sensor sets, radiometric monitors, gradient chains, and support hardware have been installed for 3-Loop plants after a complete withdrawal of all six in-vessel surveillance capsules. The systems have been installed in the reactor cavity annulus in order to characterize the neutron energy spectrum over the beltline region of the reactor vessel. The installed dosimetry were withdrawn and evaluated after a irradiation during one cycle and then compared to the cycle specific neutron transport calculations. The reaction rates from the measurement and calculation were compared and the results show good agreements each other.