• Nuclear Engineering and Technology
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• 제53권1호
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.264-267
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• 2008

Dual-cooled fuel with inner and outer flow channel was proposed for high burup, next generation nuclear fuel design. The annular cylinder of dual cooled fuel has higher structural strength compared to the conventional one, but also have concerns about flow induced vibration due to an additional flow of inner channel and the difference of flow velocity in between inner and outer channel. In this study, the dynamic stability of flexible, annular cylinder was evaluated according to the flow variation and compared to the that of the conventional PWR fuel rod. Centrifugal and Coriolis force by the additional flow in the inner channel were added in the dynamic equation of flexible beam in uniform, external, and axial flow. Complex eigenfrequency was calculated by the finite element method. Stability margin of annular cylinder compared to the solid cylinder and change of the dynamic characteristic are presented and discussed as a analysis results.

• 대한기계학회논문집B
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• 제41권1호
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• pp.53-62
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• 2017

이중냉각 환형핵연료 집합체를 위한 비틀림 혼합날개 지지격자의 강제대류열전달 성능을 실험적으로 평가하였다. 비틀림 혼합날개 지지격자는 부수로 간 혼합뿐 아니라 부수로 내 혼합을 동시에 증대시킬 수 있도록 설계되었다. 실험을 위한 이중냉각 환형핵연료 모의 집합체로, 봉 중심 간 거리와 봉 외경의 비가 1.08인 봉 간격이 좁은 $4{\times}4$ 정사각 배열의 봉다발을 준비하였다. 실험은 봉다발 유동의 축방향 평균속도가 1.5 m/s, 열유속은 $26kW/m^2$인 조건에서 수행하였다. 원주방향 온도 분포의 경우, 지지격자 상류에서는 부수로 중심 벽면에서, 하류에서는 비틀림 혼합날개 끝이 향하는 벽면에서 온도가 가장 낮게 나타났다. 축방향 온도 분포의 경우, 지지격자 하류 근처에서 온도가 급격하게 감소하는 것으로 측정되었고, 비틀림 혼합날개에 의해 누셀트 수는 최대 56 % 증대되는 것으로 나타났다. 본 실험결과를 토대로 봉 간격이 좁은 이중냉각 환형핵연료 집합체에서 비틀림 혼합날개 지지격자에 의해 강제대류열 전달 성능이 효과적으로 증대될 수 있음을 확인하였다.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.269-274
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• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2007년도 추계학술대회논문집
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• pp.281-286
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• 2007

Needs of fuel's performance evaluation for the dual-cooled fuel pellet (annular shape) necessitate the irradiation test in the test reactor. Irradiation test rig for the HARARO reactor, which is a special-purposed equipment used for material, irradiation and creep test, must satisfy the operational requirement on the hydraulic characteristics and structural integrity. In this paper, pressure drop and flow-induced vibration test for the newly developed non-instrumented test rig were carried out using FIVPET as a out-pile evaluation test. The test results show that the new test rig satisfy the HANARO operational requirement with sufficient margin. The spectral response characteristics of the flow-induced vibration of the test rid were also discussed.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.67-68
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• 2017

• 대한기계학회논문집A
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• 제35권4호
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• pp.417-424
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• 2011

핵연료집합체는 상단고정체, 하단고정체, 지지격자체, 제어관/계측관과 핵연료봉의 5 가지 주요 구성품으로 이루어져 있다. 여기서 상/하단고정체는 다른 구성품과 달리 ASME, Section III, Division 1 - Subsection NB 의 운전조건 A, B 에서의 stress intensity limits 를 만족하여야 한다. 이중냉각 핵연료집합체는 집합체당 출력을 증가시키기 위해서 기존의 가압경수로용 핵연료집합체에서 핵연료봉의 배열과 위치를 변화시켰는데 이로 인하여 핵연료봉 내/외부유로로 냉각수가 잘 흐를 수 있도록 상/하단고정체의 유로판의 형상을 수정하여야 한다. 본 논문에서는 설계하중조건에 대하여 수정된 상단고정체 유로판의 건전성평가를 위해 수행한 응력선형화과정을 자세히 설명하고 평가결과에 대하여 기술하였다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 추계학술대회 논문집
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• pp.204-205
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• 2012

• 대한기계학회논문집B
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• 제40권12호
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• pp.815-824
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• 2016

가압경수로에 장전되는 핵연료집합체는 연료 봉 다발과 지지격자 및 상하단 고정체로 구성되어 있다. 고온 고압의 냉각수는 원자로 하부로 유입되어 연료 봉 사이로 형성된 부수로를 따라 노심 상부로 흐른다. 경수로핵연료의 주요 열수력 성능인자는 정상운전시 압력강하 및 임계열속이며 사고시에는 급랭 시간이다. 한국원자력연구원에서는 경수로핵연료의 성능을 향상시키고 국산화를 위해 고성능 경수로핵연료, 이중냉각 핵연료 및 사고저항성 핵연료를 개발하였다. 경수로핵연료의 열수력 핵심기술을 개발하기 위해 압력강하 실험, 난류 유동혼합/열전달 실험, 임계열속 및 급랭 시험을 수행하였으며 전산유체역학 방법도 활용하였다. 더불어 사용후핵연료의 임시저장을 위한 건식저장 용기의 열유동에 대한 전산유체해석을 수행하였다. 한편, 경수로핵연료의 열수력 기반기술을 개발하고 실용화를 위해 대학 및 산업체와 협력연구도 진행하였다.