• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.219-224
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• 1996

하나로 핵연료인 uranium silicide 봉상 핵연료의 cladding은 핵연료 심재인 U$_3$Si-Al봉과 Al 1060 cladding 재의 접합이 잘 이루어지고, cladding 재인 Al이 완전하게 용접되어 cladding 층내에 결함이 없이 cladding 되는 최적의 온도는 51$0^{\circ}C$이며, 핵연료심의 직경이 감소되거나 변형되지 않고 핵연료심과 cladding 재가 잘 압착되는 nipple과 die 사이 거리는 0.9 - 1.5mm 이다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.175-180
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• 1998

중수로에 0.88 w/o 의 순환 핵연료를 사용하여 기존 중수로의 출력을 증강시키는 방안이 모색되었다. 기존 중수로와 양립하여야 하므로, 37봉 핵연료 다발과 CANFLEX 핵연료다발에 대한 격자 특성 계산과 노심 계산을 수행하였다. 열수력 여유도 증가와 고연소도 핵연료를 위하여 개발한 개량 핵연료 (CANFLEX)를 사용하면 원자로의 임계채널출력 (CCP)이 5 % 이상 증대하므로, 기존 원자로의 총 출력을 같은 열수력 한계 내에서 5 % 증가시킬 수 있다. 또한 개량 핵연료 다발에 순환우라늄을 사용하면 기존 월성 원자로의 구조 변화 없이 노심 출력분포의 재 분포에 의하여 15 % 까지 출력을 증강할 수 있다고 평가되었다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.137-142
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• 1998

이 연구의 목적은 혼합핵연료 장전노심에서와 같이 스펙트럼이 상이한 인접핵연료의 영향으로 나타나는 노드내 스펙트럼 이력효과를 보정해 주고자 하는 것이다. 이를 위해 이 연구에서는 노드내 13개 지역에서 국부적 미시연소를 수행하여 스펙트럼 이력이 각각 다른 13 개의 독립적인 핵단면적을 구하였고 이로부터 노드내 핵단면적의 분포를 다항식으로 근사하였다. 스펙트럼 이력구배 효과의 보정은 노드내 중성자속 가중평균 핵단면적과 노드내 핵단 면적의 분포에 따른 불연속인자로 보정하였다. 이 스펙트럼 이력구배 효과 보정방법을 혼합 핵연료와 우라늄핵연료가 Checkerboard 형으로 무한히 장전된 경우에 대하여 검증계산을 수행하여 참조해인 CASMO-3 결과와 비교하였다. 스펙트럼 이력분포가 고려되지 않은 경우는 연소도 40 MWD/kgHM에서 노심 반응도에서 약 0.38%, 봉출력에서 최대 11.2 %, 평균 4.3%의 오차를 보였으나 스펙트럼 이력분포를 반영함으로서 노심 반응도에서 0.12 %, 봉출력에서 최대 4.9%, 평균 1.3%의 오차를 보였다.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1996.04a
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• pp.100-104
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• 1996

신형경수로의 설계 대안으로서 기존 가압경수로와 가압중수로의 단점들을 극복하고, 장점들을 채택한 새로운 중수감속 경수로의 노심 설계를 제안하였다. 기존 가압중수로의 압력관내에 경수를 냉각제로 순환시키며 중수를 감속재로서 압력관 외부에 배치하였으며, 핵연료로서 농축우라늄을 사용하는 설계 개념은 많은 설계 장점을 갖는다. 본 연구에서는 시스템은 기존 CANDU의 설계를 입증기술로서 가능한 그대로 채택하고, 핵연료와 냉각재에 대해 핵설계를 수행하여 핵적 타당성을 검토하였다. 핵연료다발은 월성 2호기 사양을 그대로 사용하여 37봉 핵연료 다발로 하였으며, 농축도, 봉간간격, 핵연료다발간 간격들을 변형시켜 높은 연소도를 확보하면서 냉각재 온도계수와 감속재 온도계수가 음의 안전성을 갖는 원자로가 설계될 수 있음을 확인하였다.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.100-107
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• 1994

A statistical methodology is developed for calculating the nuclear fuel pod internal pressure of Korean PWR fuel in order to reduce over-conservatism of the current KAERI deterministic methodology. The developed statistical methodology employs the response surface method and Monte Carlo calculation. The simple regression equation for the rod internal pressure is derived by taking into account the various fuel fabrication-related and fuel performance model-related parameters. The validity of the regression equation is examined by the F-test, $R^2$-method and Cp-test The internal pressure predicted by the regression equation is in good agreement with that calculated by he computer code using the KAERI deterministic methodology. The distribution of the internal pressure from the Monte Carlo calculation is found to be normal. Comparison of the 95/95 rod internal pressure predicted by the developed statistical methodology with the maximum rod internal pressure by the deterministic methodology shows that the developed statistical methodology reduces significantly over-conservatism of the deterministic methodology.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.189-196
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• 1988

This study investigates the effect of fuel pellet eccentricity on fuel rod thermal performance under the steady state condition. The governing equations in the fuel pellet and the cladding region are set up in 2-dimensional cylindrical coordinate (r, $\theta$) and are solved by finite element method. The angular-dependent heat transfer coefficient in the gap region is used in order to account for the asymmetry of gap width. Material propeties are used as a function of temperature and volumetric heat generation as a function of radial position. The results show the increase of maximum local heat flux at the cladding outer surface and the decrease of maximum and average fuel temperatures due to eccentricity. The former is expected to affect the uncertainties in the minimum DNBR calculation. The latter two are expected to reduce the possibility of fuel melting and the fuel stored energy. Also, the fuel pellet eccentricity introduces asymmetry in fuel pellet temperature and movement of the location of maximum fuel pellet temperature.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.41 no.1
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• pp.53-62
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• 2017

The forced convection heat transfer performance of a twist-vane spacer grid for a dual-cooled annular fuel assembly was examined experimentally. The twist-vane spacer grid was uniquely designed to enhance mixing inside subchannels and mixing between adjacent subchannels. For testing, a $4{\times}4$ square-arrayed rod bundle with narrow gaps between rods was prepared as the dual-cooled annular fuel assembly to be simulated. The pitch-to-rod diameter ratio of simulated dual-cooled annular fuel assembly was 1.08. The experiments were performed under the following conditions: axial bulk velocity, 1.5 m/s and heat flux, $26kW/m^2$. With regard to the circumferential temperature distribution, the lowest rod-wall temperatures upstream and downstream were measured at the subchannel center and the position toward the tip of twist-vane, respectively. With regard to the axial temperature distribution, behind the twist-vane spacer grid, the rod-wall temperature decreased drastically, and the Nusselt number was enhanced by up to 56 %. The present measured data indicate that the twist-vane spacer grid can effectively improve the forced convection heat transfer in the dual-cooled annular fuel assembly with narrow gaps.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2000.11a
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• pp.639-642
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• 2000

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.614-620
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• 2003

KAERI has developed DUPIC nuclear fuel with the refabrication of spent PWR fuel discharged from domestic nuclear power plant by a dry process at M6 hot-cell in IMEF To verify the performance of DUPIC nuclear fuel, irradiation test at the operating conditions of commercial power plant is essential. Since the HANARO research reactor of KAERI does not have fuel test loop(FTL) for irradiating nuclear fuel under high temperature and high pressure conditions, DUPIC fuel cannot be irradiated in the FTL of HANARO. In the 13-th PRM among Korea, Canada, USA and IAEA, AECL proposed that KAERI fabricated DUPIC fuel can be irradiated in the FTL of the NRU research reactor without charge of neutrons. The transportation quantity of DUPIC fuel to Canada is 10 elements(about 6kg). This transportation package is classified as the 7-th class according to "recommendation on the transport of dangerous goods" made by the United Nations. In case of air shipment, until now, there is no proper air transportation cask for DUPIC fuel. In case of sea transportation is possible but requires very high cost.high cost.