• Nuclear Engineering and Technology
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• 제36권2호
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• pp.142-152
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• 2004

The liquid entrainment and vapor pull-through offtake model of RELAP5/MOD3 had been developed for SBLOCA (Small Break Loss of Coolant Accident). The RELAP5/MOD3 model for horizontal volumes accounts for the phase separation phenomena and computes the flux of mass and energy through a branch when stratified conditions occur in the horizontal pipe. In the case of CANDU reactor, this model should be used in the coolant flow of 95 feeders connected to the reactor header component under the horizontal stratification in header. The current RELAP5 model can treat the only 3 directions junctions; vertical upward, downward, and side oriented junctions, and thus improvements for the liquid entrainment and vapor pull-through model were needed for considering the exact angles. The RELAP5 off-take model was modified and generalized by considering the geometric effect of branching angles. Based on the previous experimental results, the critical height correlation was reconstructed by use of the branch line connection angle and validation analyses were also performed using SET. The new model can be applied to vertical upward, downward and angled branch, and the accuracy of the new correlations is more improved than that of RELAP5.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.57-61
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• 1997

A Canadian Deuterium Uranium (CANDU) reactor core calculation was performed using lattice parameters generated by HELIOS. The HELIOS-based lattice parameters were processed by TABGEN in a form suitable for the core analysis code RFSP. The core calculation was performed and the results were compared to those of the reference calculation which uses POWDERPUFS-V (PPV) for the lattice parameter generation. The characteristics of the core calculated based on the PPV and HELIOS lattice parameters match within 0.4%$\Delta$k and 7% for the excess reactivity and the channel power distribution, respectively.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.575-576
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.41-47
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• 1995

A parametric study bas been performed for the various refueling schemes of CANDU 6 reactor loaded with reference DUPIC fuel. The optimum discharge burnup was determined such that the peak bundle power is minimized for the equilibrium core. Based on the results of instantaneous core calculation using patterned random age distributions, it was decided to perform the refueling simulations only for 2-bundle and 4-bundle shift refueling schemes. The 600 FPD simulation has shown that the operational margins of the channel and bundle power to the license limits are 7.9% and 17.1%, respectively, for 2-bundle shift refueling scheme. The 4-bundle shift refueling scheme also satisfies the license limits and the operational margins of the channel and bundle power are 7.1% and 9.8%, respectively. The result of refueling simulation indicate the possibility of using reference DUPIC fuel in current CANDU 6 reactor.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.112-117
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• 1998

A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.570-577
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• 1994

The loss of water in the carbon steel balls and water region of the end shield for CANDU 6 reactor could lead to significant temperature gradient through the end shield structure which amy result in the excessive deformation. With an assumed end shield drained scenario, the heat deposition rates were calculated through the end shield associated with the central fuel channel during full power operation as an initial step to thermal stress analysis. The drained case was compared with that of water present normal case in therms of heat deposition rater and the total heating throughout the end shield regions. The compared results show that the heat deposition and the total heating remain almost the same between the two cases. It was found that the change of volume integrated flux in the end shield regions due to the loss of water contribute a negligible effect on the heat deposition in this region.

• Nuclear Engineering and Technology
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• 제34권2호
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• pp.176-185
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• 2002

A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuellin’g method change from the 8- to 4-bundle shift refuelling ill not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the FM lifetime.

• Nuclear Engineering and Technology
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• 제15권2호
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• pp.87-97
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• 1983

본 연구에서는 평형로심 특성을 계산하기 위하여 로심 해석용 코드인 WBURN (2-D, 2-group, coarse mesh)을 개발하여 월성 원자로의 평형 로심 특성을 해석하고 최종 안전성 보고서에 주어진 결과와 비교하였다. 그리고 최근에 입수된 설계상수들의 변화에 기인된 평형로심 특성변화를 조사하였으며, 월성 원자로의 가동 조건이 변화되어 핵연료의 연소도를 약 5% 높일 수 있음이 나타났다.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.638-644
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• 2017

The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.293-294
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• 2004