Canada Deuterium Uranium Reactor Research Articles

Nb speciation and microstructural characterization of the oxide layer formed on neutron-irradiated Zr-2.5 Nb alloy in high-temperature water

It is crucial to understand the degradation mechanisms of the alloy Zr-2.5Nb, which make the pressure boundary in Canada Deuterium Uranium reactors. In the present study, Zr-2.5Nb coupons were exposed for 567 days at 325 °C with and without neutron-irradiation in simulated primary water chemistry. A combination of scanning/transmission electron microscopy with energy dispersive X-ray spectroscopy, and electron energy loss spectroscopy was used to investigate the base metal and oxide. The obtained results suggest that the distribution of nanopores and the variation in Nb oxidation state likely create a gradient in the oxygen concentration and electrochemical potential through the film.

Accelerated Formation of Oxide Layers on Zircaloy-4 Utilizing Air Oxidation and Comparison with Water-Corroded Oxide Layers.

For the dry storage of Canada Deuterium Uranium (CANDU) spent nuclear fuels, the integrity of Zircaloy-4 fuel cladding has to be verified. However, the formation of ~10 µm-thick oxide layers in typical CANDU reactor operating conditions takes several years, which makes sample preparation a slow process. To overcome such limitations, in this study, an accelerated formation of an oxide layer on Zircaloy-4 cladding tube was developed with a combination of high-temperature water corrosion (HT-WC) and air oxidation (AO). First, Zircaloy-4 tubes were corroded in oxygenated (2 ppm dissolved oxygen) high-temperature water (360 °C/19.5 MPa) for 500 h. Then, the tubes were air-oxidized at 500 °C for 30 h. Finally, the tubes were corroded again in HT-WC for 500 h to produce ~10 µm-thick oxide layers. The morphology and characteristics of the oxide layer in each step were analyzed using X-ray diffraction, scanning and transmission electron microscopy. The results showed that the oxide layer formed in the accelerated method was comparable to that formed in HT-WC in terms of morphology and oxide phases. Thus, the accelerated oxide formation method can be used to prepare an oxidized Zircaloy-4 cladding tube for CANDU fuel integrity analysis.

Study of Influences of the Level of Zr Oxidation and Volumetric Heat Generation on Heat Transfer and Crust Formation Within CANDU Corium

A full-station blackout at a nuclear power plant can lead to fuel failures and radiological release to the environment if there is a breach of the reactor vessel. For Canada Deuterium Uranium (CANDU) reactors, the expected failure mechanism is through thermal stress concentration at the calandria vessel wall, and is thus influenced by local heat flux values. The current study uses computational fluid dynamics to simulate heat transfer, fluid flow, and crust formation within a CANDU geometry. Sensitivity to critical parameters, including the volumetric decay heat generation rate and the percentage of Zr oxidation is explored. The results show that as the volumetric heat generation rate decreases, the crust is thicker, and the wall heat flux is lower. This suggests that the activation of mitigating measures that delay the accident progression result in more favorable outcomes. The percentage of Zr oxidation primarily influences the thermal conductivity, which impacts the crust formation and wall heat flux rates. Specifically, corium with a lower percentage of Zr oxidation has higher thermal conductivity, and thus lower heat transfer resistance. This results in lower corium temperatures, which reduces the radiation heat transfer from the top surface and also increases crust thickness. Higher rates of heat removal from the vessel wall thus occur. In contrast, a higher percentage of Zr oxidation results in lower thermal conductivity, which leads to lower wall heat flux and a thinner crust at the vessel wall. Overall, these findings highlight the importance of considering the effects of sensitivity parameters on the heat flux distribution in the event of severe accidents.

Coupled module of radioactive waste analysis for CANDU reactor based on the Monte Carlo method

When decommissioning nuclear power plants, it is necessary that various types of radioactive wastes are to be estimated and be analyzed accurately. The first decommissioning process of the commercial light water reactor (LWR) is expected to start soon in South Korea, however, lots of processes are still under development for CANada Deuterium Uranium (CANDU) reactor. Unlike the LWR type reactor, the CANDU has complicated and different shapes of components in the reactor, which makes it difficult to carry out activation analysis considering region-wise different irradiation histories. In this study, a coupled module has been developed with Monte Carlo full core analysis and the region-wise depletion analysis by equipped automatic input generation. Furthermore, the levels of radioactive wastes are determined for transportation including estimation of cask contents based on the coupled module results.

Iron hydrolysis and lithium uptake on mixed-bed ion exchange resin at alkaline pH

The use of ion exchange resins to remove ionic impurities from solution is prevalent in industrial process systems, including in the primary heat transport system (PHTS) purification circuit of nuclear power plants. Despite its extensive use in the nuclear industry, our general understanding of ion exchange cannot fully explain the complex chemistry in ion exchange beds, particularly when operated at or near their saturation limit. This work investigates the behaviour of mixed-bed ion exchange resin, saturated with species representative of corrosion products in a CANDU (Canadian Deuterium Uranium) reactor PHTS, particularly with respect to iron chemistry in the resin bed and the removal of lithium ions from solution. Experiments were performed under deaerated conditions, analogous to normal PHTS operation. The results show interesting iron chemistry, suggesting the hydrolysis of cation resin bound ferrous species and the subsequent formation of either a solid hydrolysis product or the soluble, anionic Fe(OH)3-.

Analysis of natural convection heat transfer and solidification within a corium simulant

In the event of a severe nuclear accident such as that at the Fukushima Daiichi Nuclear Power Plant (NPP) in 2011, it is essential to minimize the release of radioactive material into the environment. Studies on Canada Deuterium Uranium (CANDU) reactors showed that thermal stresses are a potential threat to vessel wall integrity. There is thus a need to analyse the heat transfer and fluid flow within the corium and determine the spatial distribution of heat flux at the calandria vessel wall to ensure the ex-vessel cooling is sufficient. Canadian Nuclear Laboratories (CNL) designed and conducted an experiment to analyse the heat transfer within the corium using a 1/5 scale CANDU calandria vessel with molten salts as the corium simulant. The current study used computational fluid dynamics (CFD) to simulate the heat transfer and crust formation in the experiment to provide insight into the flow pattern and heat transfer and to assess the adequacy of the CFD modelling. Unsteady Reynolds-Averaged Navier–Stokes (URANS) and enthalpy-based solidification models were used to simulate the experiment. The numerical results show that crust forms along the vessel wall up to an approximate polar angle of 42°. There are natural convection circulations close to the top surface, and boundary layers form along the vessel wall. The results predicted by the numerical model match closely with the experimental results.

Feeder Pipe Oxidation in the Presence of Steam During a Nuclear Reactor Accident

A Canada Deuterium Uranium (CANDU) reactor has hundreds of carbon steel feeder pipes that are connected at one end to the fuel channels at the reactor face and are connected at the other end to the reactor inlet and outlet headers above the reactor. The production of hydrogen gas from metal–water interaction at high temperatures is a major concern during a severe accident in a nuclear reactor. It is generally accepted that the main source of hydrogen gas during a severe accident is the chemical interaction of Zircaloy fuel cladding and the water coolant. However, it has recently been suggested that the amount of hydrogen produced by the oxidation of carbon steel located outside of a CANDU core could exceed that produced by zirconium oxidation. In this work, the linear and parabolic oxidation rate constants of CANDU feeder carbon steel, in the temperature range of 600–1,100°C, were measured to be W/t = 1.806 × 103 exp (− 126,337/RT) and W2/t = 1.879 × 105 exp (− 135,835/RT), respectively. For parabolic steam oxidation beginning at roughly 90 minutes after reaching steady-state heating temperatures, the latter equation should be used. These rates are about a factor of 5–10 greater than oxidation rates measured in 304L stainless steel and in zirconium specimens, meaning that the carbon steel oxidation can be significant in an accident.

Auto Sizing of CANDU Nuclear Reactor Fuel Channel Flaws from UT Scans.

The inspection of nuclear power plants is an essential process that occurs during plant outages. During this process, various systems are inspected, including the reactor's fuel channels to ensure that they are safe and reliable for the plant's operation. The inspection of Canada Deuterium Uranium (CANDU®) reactor pressure tubes, which are the core component of the fuel channels and house the reactor fuel bundles, is performed using Ultrasonic Testing (UT). Based on the current process that is followed by Canadian nuclear operators, the UT scans are manually examined by analysts to locate, measure, and characterize pressure tube flaws. This paper proposes solutions for the auto-detection and sizing of pressure tube flaws using two deterministic algorithms, the first uses segmented linear regression, while the second uses the average time of flight (ToF) within ±σ of µ. When compared against a manual analysis stream, the linear regression algorithm and the average ToF achieved an average depth difference of 0.0180 mm and 0.0206 mm, respectively. These results are very close to the depth difference of 0.0156 mm when comparing two manual streams. Therefore, the proposed algorithms can be adopted in production, which can lead to significant cost savings in terms of time and labor.

Microstructure-based polycrystalline finite element modeling of Inconel X-750 irradiated in a CANDU reactor

Aging of Inconel X-750 spacers in Canada Deuterium Uranium (CANDU) reactors is managed through a combined program of in-service inspection, component removal for material surveillance testing, research and development, and fitness-for-service evaluation. Crush tests of ex-service spacer coils have shown that the load carrying capacities of Inconel X-750 spacers decrease with operating time. The spacers also exhibit reduced ductility and inter-granular failure that is typical of helium (He)-embrittlement. Examination of ex-service spacers confirmed that material interfaces (mainly grain boundaries) were perforated by He-bubbles. Assuming that the grain boundary bubble coverage dictates the grain boundary strength and thus the failure load of the polycrystalline component, we have simulated the degradation and inter-granular failure of the spacer coils at low and high operating temperatures.The observed inter-granular fracture has been simulated by a cohesive zone model that governs the grain boundary de-bonding process. The grain boundary strength is related to the perforation by He-stabilised cavities. The model of grain boundary traction separation consists of three parts: (i) a rate-theory calculation to predict grain boundary coverage; (ii) a local material degradation model to quantify helium induced grain boundary strength reduction; (iii) a traction separation model to simulate grain boundary de-cohesion. The results show that the accumulation of He-stabilised cavities on grain boundaries causes material failure at lower stresses and strains that are dependent on the irradiation temperature and the neutron exposure. The load carrying capacity of the X-750 spacers decreases non-linearly with irradiation and the rate of change is decreasing with increasing dose.

Comparison of measured strain and thermal effects during reactor building leakage rate tests of a nuclear power plant

With nuclear reactor power plants, Reactor Building (RB) Leak-Rate Tests (LRTs) are conducted at regular intervals to confirm the air leak tightness of a containment structure. These LRTs provide an excellent opportunity to study the containment structure’s behaviour when subjected to internal pressure, thereby evaluating the structural performance of the containment structure. However, significant time-dependent fluctuations are observed in the measured strains due to the contribution of thermal strain induced by the surrounding environment. In order to accurately study the mechanical behavior of the containment structure, the thermal strain must be separated from the measured total strain. In this study, environmental data was collected during an RB LRT. Finite element analyses were carried out by utilizing the environmental conditions to predict the thermal strain of the containment structure. The predicted thermal strain was then subtracted from the measured total strain, so the underlying mechanical strain induced by the internal pressure could be approximated. The mechanical behavior of a containment structure during LRTs over an eighteen-year period is compared to understand the overall aging effects of a CANDU®11CANDU (CANada Deuterium Uranium Reactor) is a registered trademark of the Atomic Energy of Canada Limited (AECL), used under exclusive license by Candu Energy Inc. containment structure. Only a slight change in structural stiffness was observed over eighteen years though further study is suggested to confirm this. This paper outlines how field data collected during LRTs can be used to provide insight into understanding the long-term behavior of a containment structure.

Prediction of the in-reactor deformation of Zr-2.5wt%Nb pressure tubes using the crystal plasticity finite element method framework

To predict the irradiation induced deformation of the pressure tube material in Canada Deuterium Uranium (CANDU) reactors, a deformation model based on the crystal plasticity finite element method framework was developed in this study. Based on the principle of superimposition, three deformation mechanisms—thermal creep, irradiation growth, and irradiation creep—were incorporated into the model. The model parameters were optimized and verified using in-service inspection (ISI) data from Korean CANDU reactors by considering the irradiation effect on deformation mechanisms and the influence of incompatible grain interactions. We found that the model with optimized parameters could successfully describe the results of not only the pressure tube ISI data but also the simulated mini capsules experiments in a test reactor. By analyzing the optimized irradiation creep parameters and results, the possibility of diffusion-base mechanisms was elucidated and the anisotropy in the activation energy terms along the tube directions was confirmed. For predictions using the model, three strategies for extracting the optimized parameters from available measurement data were examined. As a result, we found that, the better prediction ability can be achieved when only the latest data measured at the same reactor before the predictions are used to optimize the model parameters (method 3).

Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Vertical Pipe

During a severe power reactor accident, the plant core can melt. The resulting mixture of molten nuclear fuel and other in-core materials is known as corium. For a Canada Deuterium Uranium (CANDU) reactor, the corium is expected to settle at the bottom of the calandria vessel, but there is a potential for some melt to flow through connecting piping and other penetrations. The flow of corium through these structures can be contained if melt solidification and thus plugging occur. A numerical model was created to simulate the flow of molten metal through an empty vertical pipe. This model was benchmarked to a previous analytical model and validated against experimental results with gallium metal (which is a metal with low melting temperature) as an alternative for corium. The numerical model predicted the penetration length of gallium with an average percent error of 10.3% when compared to the experimental penetration length results of gallium. The model was also updated to predict the corium penetration length in cooling pipes of the CANDU reactor during a severe accident.

Contribution of microstructure to yield strength of X-750 alloy used for garter spring in Canada deuterium uranium reactor

The role of microstructures on the overall yield strength of garter-spring coil X-750 alloy used for the fuel channel structure in the Canada deuterium uranium reactor is investigated in this study. We perform a comprehensive characterization of the residual microstructures in the garter-spring coil X-750 alloy and heat-treated plate X-750 alloys using electron microscopy. Simultaneously, we perform small-scale mechanical tests and bulk mechanical tests to evaluate the yield strength of the compact garter-spring coil component. We compare the mechanical property data obtained via the micro-bending measurement method and derive a correlation between each measured data point to assess the increase in yield strength. The γ′ ordering phase contribution is the most crucial factor in strengthening, followed by dislocation density and boundary-decorated precipitates in the garter-spring coil X-750 alloy. We quantitatively evaluate the effects of the γ′ ordering phase and initial dislocation density on the yield strength of the garter spring coil X-750 alloy because they significantly affect the change in the mechanical properties in the irradiation environment.

Study of natural uranium utilization in a heavy water moderated molten salt reactor

Heavy water moderated molten salt reactor (HWMSR) is a newly proposed molten salt reactor. It adopts heavy water as the moderator and molten salt dissolved with actinide elements as the fuel and coolant as well. These unique features endow HWMSR a high neutron economy as performed in CANDU (CANada Deuterium Uranium) pressure tube heavy water reactor and a capability to implement the processes of online refueling and reprocessing as done in the conventional MSRs, making natural uranium (NU) utilization and together with U-233 production become feasible. The obtained results in this study demonstrate that the fuel utilization efficiency of HWMSR with NU fuel cycle can achieve 1%, slightly better than that of the CANDU reactor at the same physical size and power level. Meanwhile, around 2662 kg U-233, which can start 3 new Th–U fuel cycle HWMSR cores, can be produced over 20 years’ operation by online refueling thorium and extracting Pa-233.

Rewetting Delay Time During Jet Impingement Quench Cooling of Hot Curved Surfaces

Rewetting a hot dry surface is the establishment of wet contact between the hot surface and a liquid at a lower temperature. Rewetting occurs after destabilizing a vapor film that exists between the hot surface and the liquid. Situations involving rewetting heat transfer are encountered in a number of postulated accidents in Canada Deuterium Uranium (CANDU) reactors, such as rewetting of a hot dry calandria tube in a critical break loss-of-coolant accident (LOCA). It is also encountered in improving metals’ mechanical properties in metallurgical industries. One of the important parameters in rewetting cooling is the rewetting delay time, which is the time interval from starting to cool the surface by the liquid to the establishment of the wet contact. Determining the rewetting delay time is very important for limiting the extent of core damage during the early stages of reactor severe accidents and is essential for predicting the period after which the coolant effectively cools an overheated core. If the rewetting delay time is relatively long, an escalation in the calandria surface temperature can occur, and if the temperature was not reduced by the establishment of the wet contact, this may lead to failure of the fuel channel. Although there is increasing interest in literature in estimating the rewetting delay time of hot flat surfaces, very limited studies exist on rewetting of curved surfaces, such as tubes. In this study, experimental tests were carried out to measure the rewetting delay time at the stagnation point of hot horizontal tubes cooled by a vertical rectangular water jet. The tubes were heated to initial temperatures between 400°C and 740°C, then rapidly cooled to the jet temperature. The two-phase flow behavior was visualized using high-speed imaging, and the moment at which the vapor film collapses was captured. In addition to studying the effect of initial surface temperature on the delay time, effects of water subcooling in the range 15°C to 80°C and jet velocity in the range 0.17 to 1.43 m/s were studied and a correlation for the delay time was developed and validated. The delay time was found to strongly increase by increasing initial surface temperature and surface curvature and by decreasing water subcooling and jet velocity. The effects of solid material and tube wall thickness were also studied.

Phase analysis of hydrogen contained Zr-2.5wt%Nb CANDU pressure tube using plasmon energy loss analysis

Canada Deuterium Uranium (CANDU) reactors use Zr-2.5Nb alloys as the pressure tube material. Ensuring optimal performance of Zr-2.5Nb pressure tubes is vital to the safety of CANDU-type heavy water reactors. Long-term use of the pressure tube induces hydrogen pickup and hydride precipitation, causing delayed hydride cracking. Therefore, to evaluate the mechanical integrity of the pressure tube, it is critical to determine the phase distributions inside the pressure tube during hydrogen charging. This study focuses on visually expressing the hydride phasesusingscanning transmission electron microscopy-electron energy loss spectroscopy (STEM-EELS).By using EELS, two phase characterization possibilities were noted at the phase boundary between α-Zr and δ-hydride: nucleated β-Zr and γ-hydride and a mechanical mixture of α-Zr and δ-hydride. The first characterization possibility was reported previously by Barrowet al., and the second possibility is described comprehensively in this paper.

Carbon Dioxide Contamination of Aqueous Morpholine Solutions and Effects on Secondary Coolant Chemistry Under CANDU Conditions

Exposure to air can cause amine solutions in CANada Deuterium Uranium (CANDU) reactor secondary coolant circuit feed tanks to absorb carbon dioxide (CO2). Likewise, carbon dioxide can be absorbed directly into the amine-containing secondary coolant by air ingress during shutdown, lay-up, and startup. Sampling operations, including transferring the sample to the laboratory and subsequent analyses, can also provide opportunities for CO2 contamination. This paper reports the results of laboratory and chemical modeling studies to examine the effects of CO2 contamination on aqueous morpholine solutions. The chemistry of CO2 uptake by feed tanks containing up to 50 wt% (11.5 mol·kg−1) morpholine at 25°C was modeled using the OLI Studio 9.5.2 chemical equilibrium model, and the speciation was confirmed by 13C nuclear magnetic resonance spectroscopy (NMR) measurements. The effects of CO2 contamination on the pH of the secondary coolant containing 60 ppm (0.006 wt%, 7.00 × 10−4 mol·kg−1) morpholine and the resulting effects on the solubility of magnetite and nickel oxide from 25°C and 250°C at steam saturation were modeled as a function of CO2 loading using the Electrical Power Research Institute chemical modeling software MULTEQ v.8. The chemical modeling calculations show that concentrated alkaline morpholine solutions at room temperature and pressure would be expected to have a strong tendency to absorb CO2 and have additional uptake abilities due to the formation of morpholine carbamates. For dilute morpholine solutions at room temperature and pressure, the solutions are still sufficiently alkaline to absorb enough CO2 to cause a measurable change in the pH of the secondary coolant. This effect was shown to be negligible under reactor operating conditions. The absorption of CO2 would potentially have the most effect on either unprotected feed tanks or during lay-up conditions in the steam generators, as it could depress the pH of the lay-up solution and adversely affect the rate of corrosion in the internal components of the steam generators (e.g., carbon steel materials). The 13C NMR measurements on samples of 50 wt% aqueous morpholine solutions from feed tanks at the Ontario Power Generation’s Pickering Nuclear Generating Station found that CO2 was below the 0.02 wt% detection limit, and suggest that the procedures used to avoid CO2 contamination in feed tanks are effective. The 13C NMR was shown to be an effective tool for monitoring CO2 uptake by morpholine solution in the feed tanks under conditions in which they may have undergone abnormal exposure to air.

On the Amount of Moderation in the CANDU Lattice

Throughout the years, various reports and training manuals on CANada Deuterium Uranium (CANDU) reactors have mentioned that the CANDU lattice is overmoderated. Overmoderation is not always defined in such documents but often appears associated with the positive void reactivity of the CANDU lattice. Some documents refer, logically, to overmoderation as meaning that the lattice pitch is larger than the pitch that maximizes the infinite-lattice multiplication constant but do not demonstrate this is the case for CANDU. We demonstrate that in fact, the CANDU lattice is undermoderated; that is, the current 28.575-cm lattice pitch is smaller than the pitch for which the infinite-lattice multiplication constant reaches its maximum. We hypothesize that the misconception of CANDU overmoderation may have originated from attributing the CANDU positive void reactivity to too much moderator by incorrectly equating the effect of losing heavy water coolant with the effect of losing heavy water moderator.

A Novel Assessment of Delayed Neutron Detector Data in CANDU Reactors

Abstract A common opportunity for nuclear power plant operators is ensuring that routinely collected data are fully leveraged. Exploiting data analytics can enable improvements in anomaly detection and condition monitoring by identifying previously unseen data trends and correlations without major financial investment. One such opportunity is in facilitating the detection of fuel defects by augmenting the delayed neutron (DN) monitoring system deployed in the majority of Canada deuterium uranium (CANDU) reactors. In this paper, we demonstrate using archive data that the detection of fuel defects can be accelerated using this system in combination with the use of a deeper historical dataset and the introduction of a smoothing algorithm. The current defect identification process relies on the analysis of data of high variance and is subject to the judgment of a domain expert, resulting in variable defect identification periods. The proposed approaches seek to mitigate this and alleviate the variable identification time. Initial results presented here show that for an initial batch of 30 defects, identification periods can be meaningfully reduced compared to the current process, with defects potentially visible on an average of 11.4 days earlier. By shortening this identification period, fuel containing defects can be scheduled for earlier removal, reducing the risk of statutory shutdown obligations, protecting personnel, and promoting industry best practice. Exploring a historical dataset identifies previously undocumented trends and we discuss the potential to produce correlations with other reactor parameters. The application of this knowledge can lead to opportunities in the use of machine learning algorithms and, ultimately, more accurate predictions.

Best-Estimate Plus Uncertainty Analysis of CANDU Fuel Reliability Using Manufacturing and Simulated Core Data

Abstract A novel method of assessing the reliability of 37-element Canada deuterium uranium (reactor) (CANDU) fuel bundle was explored. The method implements a “best-estimate plus uncertainty” (BEPU) approach where a probabilistic treatment of manufacturing and operating inputs is used to predict fuel performance. The fuel performance was predicted using the Canadian industry standard codes for fuel performance, ELESTRESS and ELOCA, which, respectively, model fuel behaviors during normal and transient conditions. The outputs of the codes were compared against failure criteria from industry norms to determine the probability of failure. A Monte Carlo simulation method was applied to analyze this problem. Probability distributions of manufacturing input variables were estimated from real data, which were then randomly sampled. The inputs for fuel burnup and power were simulated using core-following data generated using a three-dimensional diffusion code, the Reactor Fuelling Simulation Program (RFSP), which were also then randomly sampled. The results of the simulations predict significant improvements in margins to limits for all performance parameters. An average improvement of 500 °C in centerline temperature, 10 °C in sheath temperature, 12 MPa in element internal pressure, and 0.8% in pellet end sheath hoop strain was predicted for the highest-powered region of the core, during normal operations, in comparison with the limit-of-envelope (LOE) benchmark. An 80% reactor overhead break (ROH) transient simulation was also simulated, and an average improvement of 500 °C in centerline temperature, 150 °C in sheath temperature, 6.5 MPa in internal pressure, and 2% in sheath hoop strain was predicted.