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The steady, isothermal-fluidic field of forced moderator streams in a ¼ scaled moderator circulation tank (MCT) of a CANDU-6 reactor is analysed. Particular focus is placed on characterizing a secondary jet formed by the collision of symmetric, counter flowing wall jets discharged upwards by nozzles installed at each side of the MCT’s horizontal axis. Two separate sets of experiments have been conducted employing water and air as the working fluids with different nozzle configurations (i.e., discrete dove-tail nozzles for water and two-dimensional slot jet nozzles for air). Then, cross validation of the fluidic data sets from these two different test facilities have been made where the fluidic similarity of the secondary jet generated by the inlet nozzles (particularly its rate of dissipation as it interacts with the calandria tubes (CTs)) has been established. Based on the established fluidic similarity, a series of numerical simulations have been performed according to the boundary conditions obtained extensively from the two-dimensional MCT experiments. Subsequently, the numerical results have been validated against the experimental data to assess the capability of a commercial code (ANSYS CFX V15.0) to resolve the moderator interaction with the CTs. It has been demonstrated that a monotonic decay of the secondary jet whose peak (having roughly 50% of the nozzle’s exit mean vertical velocity component) exists at the throat of the first row of the calandria tubes in the MCT takes place. After which, the momentum of the secondary jet is substantially reduced along its centreline. The rate of dissipation of the secondary jet is under-predicted by the default k-ω SST turbulence model, leading to a dramatic over-prediction of the penetration distance of the secondary jet into the core region. However, the accuracy of the numerical results can be substantially improved with turbulence model parameter tuning to achieve a closer fit with the reference experimental data.

This study experimentally characterizes mixed convection around calandria tubes (CTs) in a ¼ scale CANDU-6 moderator circulation tank (MCT) that uses air as the working fluid. In a full scale CANDU-6 reactor that undergoes a postulated dual failure with a loss-of-coolant accident without the emergency core cooling system available, mixed convection heat transfer occurs around the CTs. The cooling effectiveness of the moderator is diminished as an emergency heat sink if overheating eventually leads to film boiling. To prevent the onset of film boiling, local sub-cooling margins of the moderator needs to be maintained or else the critical heat flux should be increased. Circulating the moderator which interacts with the overheated CTs increases the heat transfer into the moderator which may suppress film boiling. The present experimental results demonstrate that the cooling effectiveness of the circulating moderator, in particular the secondary jet, is attenuated substantially as it is convected away from the inner wall towards the center of the MCT. The momentum of the secondary jet is diffused through the CTs. At a low jet Reynolds number, the secondary jet becomes ineffective so that some overheated CTs positioned in the other half of the MCT are cooled only by natural convection.

Severe reactor core damage may occur from fuel channel failure as a consequence of excessive heat emitted from calandria tubes (CTs) in a pressurised heavy water (D 2 O) reactor (CANDU). The heating of the CTs is caused by creep deformation of the pressure tubes (PTs), which may be ballooning or sagging depending on the internal pressure of the PTs. The deformation of the pressure tube is due to overheating as a result of a loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) failure. To prevent the exacerbation of the LOCA, circulating D 2 O in the moderator tank may be utilized by forming a secondary jet that externally cools the individual CTs. The buoyant plume develops around the CTs and interacts with the secondary jet at a certain oblique angle with respect to the gravitational axis, depending on the spatial location of the hot calandria tubes (or the hot reactor core region). This study reports on how the local and overall heat transfer characteristics on a calandria tube where the buoyant plume develops, are altered by the obliqueness of the external secondary jet (from a co-current jet to a counter-current jet) in a simplified configuration at the jet Reynolds number of Re j = 1500 for the Archimedes number of Ar D = 0.11 and Rayleigh number of Ra D = 1.6 × 10 6 (modified Rayleigh number of 3.0 × 10 7 ).

Particle image velocimetry (PIV) measurements of the flow patterns in a pressurized heavy water reactor (PHWR) were made in a 1/4 scaled-down model of the Canadian deuterium uranium (CANDU) reactor. Archimedes’ number (Ar) was introduced to evaluate the dynamic similarities between the isothermal (Ar = 0) and heating conditions. Small and large PIV fields of view were used to measure the overall flow field. Flow patterns were analyzed by measuring the spatial velocity distributions and the impingement positions of the inlet jet. The momentum-dominated flow patterns were measured under isothermal conditions (Ar = 0). The impingement position bias increased as Ar increased due to buoyancy effects. The overall moderator circulation characteristics were explored for various Ar values up to 0.157.

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult’s law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ± 6%. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m 2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

A LBLOCA scenario for an advanced pressurized water reactor, call APR1400, developed in Korea is analyzed in order to obtain an overall insight into a severe accident progression from an initiating event to the reactor vessel failure in detail by using the MELCOR computer code Versions 1.8.5 and 1.8.6. The present results (the amount of molten corium and vessel failure timing) would be used as input for the establishment of severe accident management strategies or for the design of a core catcher for the APR1400. The MELCOR results showed that the lower head instrumentation tube penetration failure model and internal structure in the reactor vessel had influence on the amount of corium ejected and the timing of reactor vessel failure.

Following the previous relevant works [1]-[3], a scaling analysis is performed to derive a set of scaling criteria which were thought to be suitable for reproducing the major thermal-hydraulic phenomena in a scaled-down CANDU moderator tank similar to that in a prototype power plant during a full power steady state condition. The objective of building a scaled-down moderator tank is to generate the experimental data necessary to validate the computer codes which are used to analyze the accident analysis of CANDU-6 plants. The major variables of interests in this paper are moderator flow velocity and temperature of the moderator which is D2O inside the moderator tank during a steady state and transient conditions. The reason is that the local subcooling of the moderator is found to be a critical parameter determining whether the stable film boiling can sustain on the outer surface of the calandria tube if the contact of overheated pressure tube and cold calandria tube should occur due to pressure tube ballooning during LBLOCA with ECC injection failure [4]. The key phenomena involved include the inlet jet development and impingement, buoyancy force driven by the moderator temperature gradient caused by non-uniform direct heating of the moderator, and the pressure drop due to viscous friction of the flow across the calandria tube array. In this paper, the previous researches are reviewed, some concerns or potential problems associated with them implied by comparing CFD analyses results between the CANDU-6 moderator tank and 1/4 scaled-down test facility are described, and as a way to examine the assumption of the scaling analysis is true an order-of-magnitude analyses are performed. Based on the results of these analyses the assumption of neglecting and terms cannot be justified for the power of 0.5 MW and 1.566 MW for the 1/4 scaled-down facility. Further investigation is thought to be necessary to confirm this result, i.e. if the scaling of the previous work1 is justifiable by some other independent analyses.

Korea Atomic Energy Research Institute (KAERI) has been carrying out a scaled-down moderator test program to simulate the CANDU-6 moderator circulation phenomena during steady state operation and accident conditions. The Moderator Circulation Test (MCT) facility was designed and constructed as 1/4 scale of the prototype of CANDU-6 reactor. In the present work spatial distribution of two-dimensional velocity in the MCT facility under isothermal condition was measured using planar Particle Image Velocimetry (PIV) system which consists of double pulsed laser, synchronizer, and high speed camera. TSI’s Insight TM 4G software was used to perform PIV image capturing, PIV calculation and post processing. 10 μm sized silver-coated hollow sphere particles were used as flow tracer particles. Multiple experiments were conducted to cover large area of the MCT facility with limited field of view of a single camera. Instantaneous and averaged velocity field were analyzed for each spatial position and flow rate. This research is capable of offering validation data for self-reliant CFD tools to predict moderator subcooling margin in CANDU-6 reactor.

We are planning to construct a scaled-down moderator facility to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to investigate the anticipated problems of the flow visualization and measurement in the planning scaled-down moderator facility. We shortly describe CFD analysis result for the 1/40 scaled-down test model and the flow measurement techniques used for this test facility under isothermal flow conditions. The Particle Image Velocimetry (PIV) method is used to visualize and measure the velocity field of water in a transparent Plexiglas tank. Planar Laser Induced Fluorescence (PLIF) technique is used to evaluate the feasibility of temperature field measurement in the range of of water temperature using an one-color method.

The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down. These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena.

At supercritical pressure condition, the thermal–hydraulics behavior of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal–physical properties across the pseudo-critical line. A coupling analysis of neutronics and thermal–hydraulics has become important for SCWR, because of the strong link between the water density and the neutron spectrum and subsequently the power distribution. The neutronics code Monte Carlo N-Particle code (MCNP) and the subchannel code Advanced Thermal–Hydraulics Analysis Subchannel (ATHAS) are used in a coupled way to better understand the design characteristics of a pressure tube type SCWR fuel channel. The results show that: the developed coupled code system can be used to analyze pressure tube type SCWR fuel bundles; improved radial fuel enrichment profile will optimize the coolant and cladding temperature distribution to meet the design criteria; smaller pressure tube pitch will result in more flatten axial power distribution and more uniform radial power distribution.

To form a licensing basis for a new methodology for a fuel channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown fuel channel analysis has been developed, and tested for a high temperature thermal–chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal–chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H 2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results for the fuel element simulators (FESs) for the three fuel rings and the pressure tube were reasonably good as proven by the simulation results. This is thought to be due to success in reproducing the initial steady state condition well, and due to the detailed modeling features of CATHENA code for the coupled conduction, convection, and radiation heat transfer in a complex fuel bundle geometry. However, one problem still remained unresolved, i.e. the inability to accurately predict the pressure tube temperature at the initial steady state condition, and an adjustment of the CO 2 gap conductivity necessary to match the measured pressure tube temperatures. However, this raises a question as to how the transient FES and the pressure tube temperature can be predicted so well in spite of an insufficient justification for using the “non-participating medium assumption” for the CO 2 gas gap. Through this study and the affiliated previous study for the steady state, it was found that the radiation heat transfer model of CATHENA among the FESs of the three rings and the pressure tube as well as the exothermic metal–water reaction model based on the Urbanic–Heidrick correlation are reasonably accurate and sound. Also it was found that an accurate prediction of the initial condition of the experiment is very important for an accurate prediction of the whole transient as it serves as the starting point for the transient.

To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20–25 °C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using “transparent” assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code’s inherent modeling feature is raised from this study. Based on a comparison of the prediction of the current CATHENA model and the experiment data, the steady state model is deemed to be adequate as a starting point for the following high temperature thermal–chemical experiment of a metal–water reaction.

A computational fluid dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal–chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal–chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (large break loss of coolant accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal–chemical behavior of the 28-element fuel channel, the measured temperatures of the fuel element simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions. In spite of some discrepancy between the measurement data and CFX results, it was found that the CFX-10 prediction based on the Urbanic–Heidrick correlation of the zirconium-steam reaction as well as the Discrete Transfer Model for a radiation heat transfer among the FES of three rings and the pressure tube are quite accurate and sound even for the offset fuel cluster of an aged fuel channel.

In this study a CFD simulation of a horizontal heater rods test (CS28-2) has been performed with the CFX-10 code and it has been compared with an experiment in order to develop a post-blowdown modeling of a CANDU fuel channel during a large break loss of coolant accident (LBLOCA). The CS28-2 experiment is one of three series of experiments to simulate the fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA of CANDU reactors. Since radiation heat transfer is dominant in the CS28-2 experiment, a benchmark problem of radiation heat transfer for the same geometry as the CS28-2 is carried out first to assess the radiation model by CFX-10, to confirm whether CFX-10 can be applied to a complex fuel geometry or not. And, the CFD study of the CS28-2 has been performed with the aim to simulate the steady-state condition of the CS28-2. These steady-state calculation results will be used for the initial conditions of the transient simulation.

Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a the national research and development programs from 2012. This research program includes the construction of the moderator circulation test (MCT) facility, production of the validation data for self-reliant computational fluid dynamics (CFD) tools, and development of optical measurement system using the particle image velocimetry (PIV). In the present paper we introduce the scaling analysis performed to extend the scaling criteria suitable for reproducing thermal-hydraulic phenomena in a scaled-down CANDU- (CANada Deuterium Uranium-) 6 moderator tank, a manufacturing status of the 1/4 scale moderator tank. Also, preliminary CFD analysis results for the full-size and scaled-down moderator tanks are carried out to check whether the moderator flow and temperature patterns of both the full-size reactor and scaled-down facility are identical.

Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production. As part of a computational fluid dynamics (CFD) simulation of the CS28-2 high-temperature experiment for this accident analysis, two Zircaloy/steam reaction models based on a parabolic rate law are implemented in a commercial CFD code (CFX-10) through a user FORTRAN. It is confirmed that the present oxidation models for the CFX-10 reproduce the results of each empirical correlation in the verification tests well. Then the CFX-10 predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation are compared with the results of the CS28-2 experiment. From these validation processes, it is shown that the Urbanic-Heidrick model, which is widely used in CANDU fuel channel codes, is also applicable to a CFX-10 simulation of Zircaloy/steam oxidation in a CANDU fuel channel.

A three-dimensional (3-D) computational fluid dynamics (CFD) model has been developed to analyze the liquid poison injection phenomenon of shutdown system 2 (SDS-2) of a Canada deuterium uranium (C...

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is ...

The compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactor (DUPIC) fuel with the existing 713-MW(electric) CANDU (CANDU-6) reactor has been analyzed for large-break loss-of-coolant-accident (LOCA) scenarios such as a 55% pump suction break, a 35% reactor inlet header break, and a 100% reactor outlet header break. The compatibility was assessed for the fuel integrity against the stored energy and the radiation environmental effect resulting from the large-break LOCA. The analysis showed that the stored energy of the DUPIC fuel was below the fuel breakup energy by 32%. The environmental effect was estimated for the personal and public doses using the radiation source term obtained from one-fourth of the fission product inventory in the fuel gap of the CANDU-6 reactor, being steadily operated at full power. The analyses have shown that both the personal and population doses are below the design limits even for a postulated dual failure such as a complete loss of containment building isolation logic.

The CATHENA code uses the finite element method (FEM) for the one-dimensional heat conduction model, which determines the temperature distribution from the fuel center to the cladding in the radial direction. However, it is shown that the finite element solutions in the heat source region such as a fuel pellet are converged to exact solutions with an increasing number of the mesh elements. Since the finite volume method (FVM) ensures local and global energy conservation due to an integral conservation over each control volume, the FVM is applied to the CATHENA wall conduction model to avoid the mesh size effect on the fuel temperature prediction. The accuracy and validity of the finite volume model in the CATHENA code are tested against two cases, a steady state and a transient heat conduction case, for which exact solutions are available. The constant temperature on the boundary surface and a uniform internal heat generation rate are assumed for the steady state problem. In the transient heat conduction problem, a cylinder is initially at a uniform temperature and suddenly its boundary surface is subjected to convection with a constant heat transfer coefficient into an ambient at a constant temperature. The steady state solutions by the FVM model are found to give almost the same results as the analytical solutions and consistent results with varying mesh sizes, while the original CATHENA with FEM over-predicts the center temperature with larger mesh sizes. The new model also closely follows the analytic solutions of the dimensionless transient heat conduction equation for a long cylinder.

Three-dimensional numerical calculations have been performed for a transient moderator circulation inside the CANDU (Canada Deuterium Uranium) calandria vessel of Wolsong Units 2/3/4. The porous media approach was applied for the core region containing 380 calandria tubes. An anisotropic hydraulic resistance model for the porous media has been developed based on the empirical pressure loss correlations. The selected event was the 35% RIH (Reactor Inlet Header) break with a loss of ECC (Emergency Core Cooling) injection, which has been known to give the largest heat load to the moderator among all the DBA's (Design Basis Accidents). The calculation has been successfully done until 1,200 s after the break, when most of the considerable heat transfer procedure has been completed. During this LOCA (Loss of Coolant Accident) transient, the local subcoolings in the vicinity of any PT/CT (Pressure Tube/Calandria Tube) contact does not drop below the experimentally derived subcooling threshold of 30°C. Because th...

Since 1992, the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the analysis of technical feasibility, the fabrication of DUPIC fuels for irradiation tests, and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, and the regional overpower trip setpoint. The physics analysis showed the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

A computational fluid dynamics (CFD) model for predicting the 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 internal heating is accounted for by Boussinesq approximation. The standard k-[curly epsilon] turbulence model associated 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 porous media, in which anisotropic hydraulic impedance is modeled using an empirical correlation of the frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA Technology. The CFD model has been successfully verified and validated against experimental data obtained at Stern Laboratories Inc. in Hamilton, Ontario, Canada.

As a CANDU6 reactor has a high pressure primary cooling system and an independently cooled moderator system, the moderator in the calandria would act as a supplementary heat sink during a loss of coolant accident (LOCA) if the primary cooling and emergency coolant injection systems fail to remove the decay heat from the fuel. For the safety concern it is required to predict the 3-dimensional velocity and temperature distribution of moderator fluid to confirm the effectiveness of moderator heat sink. Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions.

We are planning to construct a scaled-down moderator test facility to simulate the CANDU-6 moderator circulation phenomena during steady state operation and accident conditions at the Korea Atomic Energy Research Institute. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a Particle Image Velocimetry (PIV) technique under an isothermal state, and the temperature field is measured using a Laser Induced Fluorescence (LIF) technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment.Copyright © 2013 by ASME

The fuel element of KMRR (Korea Multi-purpose Research Reactor) has 8 longitudinal, rectangular fins to enhance the heat transfer performance. The existence of these fins makes it difficult to analyze the heat transfer phenomena within the fuel element using the conventional one-dimensional heat conduction model. As the uncertainty in the computation of the maximum sheath temperature significantly affects the core thermal margin, a computer code, called, TEMP2D, which is based on a two-dimensional heat conduction model has been developed to deal with the finned element and validated. This computer code TEMP2D has a fully implicit numerical scheme and can solve both the steady state and transient problems such as the changes in coolant thermal-hydraulic conditions and fuel pin power. The code accuracy, which proved to be an excellent one, was verified by comparing its results with those from two widely accepted computer codes, MARC and ADINA. The result of this code calculation has been used to compute the KMRR core thermal margin and to develop a correlation for the equivalent 1D heated diameter which can reproduce the maximum cladding temperature (or heat flux) at various steady states when used in the 1D heat conduction model.

A steady-state 3D simulation for predicting the local subcooling of the moderator in the vicinity of the calandria tubes in a CANDU-6 reactor is performed. For the current simulation, a set of grid structures with the same geometry as the CANDU-6 moderator tank, called 'calandria vessel', is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The standard k-{epsilon} turbulence model associated with logarithmic wall treatment is used to model turbulence generation and dissipation within the vessel. The moderator fluid is heavy water. Buoyancy forces are modeled using the Boussinesq approximation in which density is assumed to be a linear function of temperature. The matrix of the calandria tubes in the center region of the calandria vessel is simplified by the porous media approach. The anisotropic hydraulic impedance of the calandria tubes is modeled using the frictional pressure drop correlations suggested by Idelchik and Szymanski. The heat load in this steady-state simulation is conservatively set as 103 MW of 103% full power, consisting of 96.7 MW to the core region and 6.3 MW to the reflector region. The total volumetric flow rate through eight inlet nozzles is 940 L/s andmore » the outlet temperature is constantly 71.0 deg. C. The thermal boundary condition of the circumferential vessel wall is assumed a little heat flux out. As a result, the velocity field and temperature distribution of a CANDU-6 moderator in the operating condition are presented. The flow pattern identified in this simulation is the weak jet momentum-dominated flow, which is generated by the interaction between the buoyancy force by heating and the dominant momentum forces by inlet jets. The calculated maximum temperature of the moderator is 83.0 deg. C at the lower center region of the core, which corresponds to the minimum subcooling of 33.0 deg. C considering the boiling point increase due to the hydrostatic pressure change. (authors)« less

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate.Copyright © 2002 by ASME

A validation of a 3D CFD model for predicting local subcooling of the moderator in the vicinity of calandria tubes in a CANDU-6 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 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-e turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used. Buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is the 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°C over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well.Copyright © 2002 by ASME