• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.133-133
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• 2002

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.192-192
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• 2008

As the lifetime of nuclear power plants (NPPs) reaches design life, the probability for fatal accidents increases. Most of accidents are known to be caused by degradation of mechanical components. Pressure tubes are the most important components in CANDU reactor. They are subjected to various aging mechanisms such as delayed hydride cracking (DHC), irradiation and corrosion, etc. Therefore, the integrity of pressure tube is key concern in CANDU reactor. Up to recently, conventional deterministic approaches have been utilized to evaluate the integrity of components. However, there are many uncertainties to prevent a rational evaluation. The objective of this paper is to assess the failure probability of pressure tube in CANDU. To do this, probability fracture mechanics (PFM) analysis based on the Genetic Algorithm (GA) is performed. For the verification of the analysis, a comparison of the PFM analysis using a commercial code and mathematical method is carried out.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.146-149
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• 1994

A nonlinear dynamic optimization has been performed for reactor power control system of CANDU 6 nuclear power plant considering xenon, fuel and moderator temperature feedback effects. Integral-of-Time-multiplied Absolute-Error (ITAE) criterion has been used as a performance index of the system behavior. Optimum controller gain are found by searching algorithm of Sequential Quadratic Programming (SQP). System models are referenced from most recent literatures. Signal flow network construction and optimization have been done by using commercial computer software package.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.95-105
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• 2014

The CUPID code has been developed for a transient, three-dimensional, two-phase flow analysis at a component scale. It has been validated against a wide range of two-phase flow experiments. Especially, to assess its applicability to single- and two-phase flow analyses in the Calandria vessel of a CANDU nuclear reactor, it was validated using the experimental data of the 1/4-scaled facility of a Calandria vessel at the STERN laboratory. In this study, a preliminary thermal-hydraulic analysis of the CANDU reactor moderator tank using the CUPID code is carried out, which is based on the results of the previous studies. The complicated internal structure of the Calandria vessel and the inlet nozzle was modeled in a simplified manner by using a porous media approach. One of the most important factors in the analysis was found to be the modeling of the tank inlet nozzle. A calculation with a simple inlet nozzle modeling resulted in thermal stratification by buoyance, leading to a boiling from the top of the Calandria tank. This is not realistic at all and may occur due to the lack of inlet flow momentum. To improve this, a new nozzle modeling was used, which can preserve both mass flow and momentum flow at the inlet nozzle. This resulted in a realistic temperature distribution in the tank. In conclusion, it was shown that the CUPID code is applicable to thermal-hydraulic analysis of the CANDU reactor moderator tank using the cost-effective porous media approach and that the inlet nozzle modeling is very important for the flow analysis in the tank.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.15-22
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• 2016

The heat transport system transients of Wolsong 2,3,4 nuclear power plants were analysed during abnormal operating conditions. The compliance with requirements of AECB Regulatory Document R-77 for CANDU reactor was estimated. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements. The effect of LRV that is one of the overpressure protection device was very minor.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.05a
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• pp.130-137
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• 1994

As a part of developing engineering simulator for CANDU 6 nuclear power plants, present paper gives the tentative simulation results of reactor and turbine power transients including reactor-follow-turbine operation. One point kinetics equations are used for neutron dynamics, iodine and xenon loads. To calculate time-dependent high and low pressure turbine powers and grid frequency deviation, simple first order differential equations are used. In addition, control logics (reactor regulating system, demand power routine, and unit power regulator) used in the plant's process computers have been referenced.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.98-109
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• 1983

A two-step alternating direction explicit method is developed to solve the space-dependent reactor kinetics equations in two space dimensions. As a special case in the general class of alternating direction implicit methods, this method is analysed for accuracy and stability. To test the validity of this method it is compared with the implicit-difference method used in the TWIGL program. It is shown that the two methods are closely related. The time dependent neutron fluxes of the pressurized water reactor (PWR), during control rod insertion, and, of the CANDU-PHW reactor, in case of postulated loss of coolant accident, are obtained from the numerical calculation results.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.479-490
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• 2005

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.358-373
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• 1995

The Heat Transport system loop stability of CANDU-6 reactor using the CANFLEX fuel bundle was studied. The Thermal-hydraulic behavior of CANFLEX fuel bundle is similar to the conventional 37-element fuel bundle since the reactor power and the frictional pressure drop through the fuel channel is almost the same each other, Mounter the CANFLEX fuel bundle gives higher critical channel power and more homogeneous enthalpy distributions in the subchannels than 37-element fuel bundle. The SOPHT modelling or the CANFLEX fuel bundle and the Reactor outlet Header(ROH) interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. Without the ROH interconnection line the Heat Transport system loop using 43-element fuel bundle is unstable like the current 37-element fuel bundle. With the ROH interconnection line, however, the Heat Transport system is stable within $\pm$1％ of nominal flow. In the Heat Transport system loop stability point of view for Wolsong-1 plant therefore, the CANFLEX fuel loading is considered to be acceptable.