• Nuclear Engineering and Technology
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• 제36권6호
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• 한국기계기술학회지
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• 제13권2호
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.499-504
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• 2001

A validation of a 3D CFD model for predicting local subcooling of moderator in the vicinity of calandria tubes in a CANDU reactor is performed. The small scale moderator experiments performed at Sheridan Park Experimental Laboratory(SPEL) in Ontario, Canada[1] is used for the validation. Also a comparison is made between previous CFD analyses based on 2DMOTH and PHOENICS, and the current model analysis for the same SPEL experiment. For the current model, a set of grid structures for the same geometry as the experimental test section is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The matrix of calandria tubes is simplified by the porous media approach. The standard $k-\varepsilon$ turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used and buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is a buoyancy-dominated flow, which is generated by the interaction between the dominant buoyancy force by heating and inertial momentum forces by the inlet jets. As a result, the current CFD moderator analysis model predicts the moderator temperature reasonably, and the maximum error against the experimental data is kept at less than $2.0^{\circ}C$ over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well.

• 소음진동
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• 제6권2호
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• pp.163-177
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• 1996

An investigation on thermohydraulic stability of flow oscillations in the CANada Deuterium Uranium-600(CANDU-6) heat transport system has been conducted. Flow oscillations in reactor coolant loops, comprising two heat sources and two heat sinks in series, are possibly caused by the response of the pressure to extraction of fluid in two-phase region. This response consists of two contributions, one arising from mass and another from enthalpy change in the two-phase region. The system computer code used in the investigation os SOPHT, which is capable of simulating steady states as well as transients with varying boundary conditions. The model was derived by linearizing and solving one-dimensional, homogeneous single- and two-phase flow conservation equations. The mass, energy and momentum equations with boundary conditions are set up throughout the system in matrix form based on a node-link structure. Loop stability was studied under full power conditions with interconnecting the two compressible two phase regions in the figure-of-eight circuit. The dominant function of the interconnecting pipe is the transfer of mass between the two-phase regions. Parametric survey of loop stability characteristics, i. e., damping ratio and period, has been made as a function of geometrical parameters of the interconnection line such as diameter, length, height and orifice flow coefficient. The stability characteristics with interconnection line has been clarified to provide a simple criterion to be used as a guide in scaling of the pipe.

• 대한기계학회논문집A
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• 제34권9호
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• pp.1273-1279
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• 2010

가압중수로 연료관은 CSA N 285.4 기술기준에 따라 주기적인 가동중검사가 수행되고 있다. 가동중검사시 발견된 결함이 CSA N 285.4 의 허용기준을 초과하는 경우, 결함 연료관의 계속 운전을 위해 가동적합성 평가를 허용하고 있다. 캐나다 COG(CANDU Owner's Group)를 중심으로 중수로 연료관의 결함 건전성 평가 기술기준인 CSA N285.8 이 개발되었다. 본 논문에서는 CSA N285.8 을 기반으로 연료관의 건전성 평가시스템 WIES(Wolsong In-service Evaluation System)를 개발 하였다. 중수로 연료관의 가동중검사시 결함이 발견되는 경우, 개발된 시스템은 신속하고 정확한 건전성 평가를 수행하여 계획예방 정비기간의 연장을 방지하여 원전 이용률 향상에 기여할 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.257-264
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• 2013

In Canada Deuterium Uranium (CANDU)-type nuclear power plants, the reactor is composed of 380 fuel channels and refueling is performed on one or two channels per day. At the time of refueling, the fluid force of the cooling water inside the channel is exploited. New fuel added upstream of the fuel channel is moved downstream by the fluid force of the cooling water, and the used fuel is pushed out. Through this process, refueling is completed. Among the 380 fuel channels, outer rows 1 and 2 (called the FARE channel) make the process of using only the internal fluid force impossible because of the low flow rate of the channel cooling water. Therefore, a Flow Assist Ram Extension (FARE) tool, a refueling aid, is used to refuel these channels in order to compensate for the insufficient fluid force. The FARE tool causes flow resistance, thus allowing the fuel to be moved down with the flow of cooling water. Although the existing FARE tool can perform refueling in Korean plants, the coolant flow rate is reduced to below 80% of the normal flow for some time during refueling. A Flow rate below 80% of the normal flow cause low flow rate alarm signal in the plant operation. A flow rate below 80% of the normal flow may cause difficulties in the plant operation because of the increase in the coolant temperature of the channel. A new and improved FARE tool is needed to address the limitations of the existing FARE tool. In this study, we identified the cause of the low flow phenomena of the existing FARE tool. A new and improved FARE tool has been designed and manufactured. The improved FARE tool has been tested many times using laboratory test apparatus and was redesigned until satisfactory results were obtained. In order to confirm the performance of the improved FARE tool in a real plant, the final design FARE tool was tested at Wolsong Nuclear Power Plant Unit 2. The test was carried out successfully and the low flow rate alarm signal was eliminated during refueling. Several additional improved FARE tools have been manufactured. These improved FARE tools are currently being used for Korean CANDU plant refueling.

• Journal of Radiation Protection and Research
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• 제29권3호
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• pp.187-196
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• 2004

경수로형인 한국형 표준원전과 CANDU형 중수로형 원자력 발전소의 가상 중대사고시 대기 중으로 방출되는 방사성 물질로 인한 인체 건강영향에 미치는 리스크를 평가하고 비교하였다. 두 발전소 모두 반경 80km 까지의 인구분포와 2단계 PSA의 결과로 주어지는 방사선원 방출군별 방출 분율과 노심 재고량을 이용하였으며 평가 도구로는 MACCS2를 이용하였다. 인체에 미치는 영향은 조기 사망과 암 사망을 선정하였으며 반경 10 마일 밖으로 소개가 이루어진다고 가정하고 평가 결과는 사고 발생빈도를 고려한 리스크를 CCDF 곡선군으로 나타냈다. 평가 결과에 의하면 경수로형 원전에 비해 중수로형 원전이 리스크가 적게 나타나는데 이는 중수로형 원전이 경수로형 원전에 비해 가상 중대사고로 인해 대기 중으로 방출되는 방사성 물질의 양이 적기 때문이다. 두 발전소 모두 최대 리스크를 보이는 방사선원 방출군의 대표적인 초기사건은 증기발생기 세관파손 사고로 나타났다. 따라서, 경수로형 및 중수로형 발전소 모두 사고로 인한 주변 주민 보호를 위해서는 증기발생기 세관파손 사고의 발생빈도와 이로 인한 대기 중으로의 방사성 물질의 방출을 감소시키기 위한 방안이 강구되어야 한다.