CANDU Research Articles

BEPU analysis of a CANDU LBLOCA RD-14M experiment using RELAP/SCDAPSIM

A key element of the safety analysis is Loss of Coolant Analysis (LOCA) which must be performed using system thermal-hydraulic codes. These codes are extensively validated against separate effect and integral experiments.RELAP/SCDAPSIM is one such code that may be used to predict LBLOCA response in a CANDU reactor. The RD-14M experiment selected for the Best Estimate Plus Uncertainty study is a 44 mm (22.7%) inlet header break test with no Emergency Coolant Injection. This work has two objectives first is to simulate pipe break with RELAP and compare these results to those available from experiment and from comparable TRACE calculations. The second objective is to quantify uncertainty in the fuel element sheath (FES) temperature arising from model coefficient as well as input parameter uncertainties using Integrated Uncertainty Analysis package.RELAP calculated results are found to be in good agreement with those of TRACE and with those of experiments. The base case maximum FES temperature is 335.5 °C while that of 95% confidence 95th percentile is 407.41 °C for the first order Wilk's formula. The experimental measurements fall within the predicted band and the trends and sensitivities are similar to those reported for the TRACE code.

Tolerance of oxide wedging crack growth of manufacturing flaws in CANDU Zr-2.5Nb pressure tubes to cyclic operating stresses

In-service inspections of Zr-2.5Nb pressure tubes in CANDU nuclear reactors have detected inner diameter surface breaking flaw indications which are classified as manufacturing flaws. It is considered prudent to evaluate potential growth by means of a postulated oxide wedging crack growth mechanism. An oxide wedging crack growth model has previously been developed and applied to predict future flaw sizes. Fatigue crack growth due to cyclic operating stresses is calculated for these flaws, and this is added to the postulated oxide wedging crack growth for fitness-for-service evaluations. This established approach is overly conservative since fatigue crack growth cannot occur while the oxide wedging crack growth mechanism is active, due to the essentially displacement-controlled loading of the flaw by the oxide wedge, which causes the stress intensity factor for the flaw to be effectively constant under cyclic operating stresses. This paper addresses this conservatism by evaluating crack growth due to postulated oxide wedging, and demonstrating that fatigue crack growth is not possible under oxide wedge loading.

Comparison of existing and new optical fiber-based scintillation detectors for spent-fuel verification equipment

The Wolsung Nuclear Power Plant in Korea is under IAEA safeguards. During IAEA physical inventory verifications, the Optical Fiber radiation Probe System has been used since 2007 as the official instrument for verification of spent fuel. Its low sensitivity requires excessive time and labor to perform the IAEA safeguard activities and increases the operator load for preparing a dangerous and time-consuming secondary inspection. The objective of this paper is to develop a superior new optical fiber-based scintillation detector to deal with such issues. To increase the sensitivity of the instrument, a new optical fiber-based scintillation detector was developed by combining a p-terphenyl scintillator with a new type of optical fiber. A field test was carried out in the CANDU Wolsung Unit 1 to characterize the performance of the new detector. The field test showed that the OFPS using the new optical fiber-based scintillation detector has at least 100-times higher sensitivity and can also measure even very weak signals not measurable by the existing detector. From the experimental results, we can conclude that the newly developed OFPS detector would increase the effectiveness and efficiency of nuclear safeguards for verification of spent fuel in wet storage pools of CANDU reactors.

Modeling and Intelligent Controller Design for Reactor Regulating System of Advanced CANDU Reactor (ACR-700) in LabVIEW

In this research work, an Advanced CANDU Reactor of 700 MWe rating (ACR-700) is attempted for Reactor Regulating System (RRS) modeling and intelligent controller design. ACR-700 is a state-of-the-art advanced generation-III reactor. The reactor regulating system is modeled with special emphasis on internal and external reactivity devices. ACR-700 is designed with Liquid Zone Control (LZC) Model, Adjuster Banks Model, Absorber Banks Model and Shutdown Banks Model. The RRS model is a highly sophisticated model developed based on the principle of spatial nuclear reactor dynamics. The RRS model is a multivariable model. The spatial reactor dynamics is modeled based on five internal feedbacks and three feedbacks. The original controller design of RRS is comprised of a PID control algorithm. The control design is reattempted with an advanced intelligent algorithm in which the ANFIS controller is used as a modern control design tool. The new ANFIS controller is basically a multivariable controller. All the modeling of RRS is implemented in Visual Basic (VB) Software while the controller is configured in LabVIEW. A special toolkit is designed for the interfacing of Visual Basic and LabVIEW known as VBLAB. The optimization of intelligent controller parameters is carried out by Genetic Algorithm (GA). The GA-optimized intelligent controller is configured with the VB RRS model. All the variable trends are visualized in the VB environment. The proposed closed-loop ACR-700 RRS control system is tested for small perturbation Analysis and power ramp-down transient and found with excellent behavior well within the design limits. The performance of the suggested ACR-700 RRS controller is compared with the existing conventional controller. The performance of the suggested control scheme is marked with reduced oscillations and faster as compared to the existing control scheme.

Deuterium concentration profiles at the rolled joints of CANDU fuel channels

A diffusion model is proposed to explain high concentrations of hydrogen isotopes seen in some rolled joints of CANDU reactors. Predicted concentration profiles emanating from the rolled joints into the Zr-2.5Nb pressure tubes suggest an early ingress event happens within the first six months of operation in some tubes. This event is consistent with gettering of hydrogen from the 403 stainless steel end-fittings by the pressure tubes.

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.

Integrated analysis to investigate the viability of using Thorium-based fuel instead of traditional fuel in CANDU reactor

This work gives the nuclear operator an integrated insight into the possibility of operating a CANDU reactor with thorium-based fuel types. MCNPX code was used to design 3D model of CANDU bundle. Thorium mixed with 235U, 233U and rgPu were investigated as a fuels in separate fuel cycles and their results were compared with the traditional fuel (UO2). The main neutronic and safety parameters such as infinity multiplication factor (kinf), neutron generation time (Λ), breeding ratio, Doppler reactivity coefficient (DRC), effective delayed neutron fraction (βeff) and moderator temperature coefficient (MTC) have been studied for the investigated fuels. Due to the significant impact of the rgPu and minor actinides (MAs) on the environment, their concentrations have been investigated with burn-up for the proposed fuels to reach the optimum one from the neutronic and economic point of view.

Assessment of Filtered Containment Venting Strategies for Mitigating Severe Accident Consequences at a Generic Single-Unit CANDU

Abstract In many jurisdictions, filtered containment venting systems are an integral part of severe accident management guidelines but are normally only used as a last resort, to be manually activated when the containment pressure approaches the containment failure limit. This approach fails to utilize the capability of the filtered venting system to better control containment pressure throughout an accident and to create capacity for possible future pressurization based on the accident progression. In this paper, smart venting strategies are explored, using a generic single-unit Canada Deuterium Uranium (CANDU) (660 MWel) plant model in MAAP-CANDU, as a means to mitigate the consequences of a severe accident. These strategies are able to account for both current conditions as well as possible future pressurization. The use of strategic venting offers an improvement in maintaining containment integrity, as well as reducing I-131 releases to the environment by about 10% to 31% compared to last-resort venting while being feasible for manual operator actions. Overall, this comprehensive analysis demonstrates the capability of a smart venting strategy paired with a filtered containment venting system as an active mitigation tool.

Investigating various factors affecting neutron spectrum in the pressure tube in pressurized heavy water reactor design

The fuel bundles in the pressurized heavy water reactor design (PHWR), such as the CANDU (Canada Deuterium Uranium), are placed inside horizontal fuel channels. Each fuel channel consists of concentric tubes called the pressure tube (PT) and the calandria tube (CT). Due to continuous exposure to extreme neutronics and thermal hydraulics conditions, the PTs undergo physical deformations in both the axial and radial directions, as the reactor ages. The deformations include diametral creep, wall thinning, sag, and axial elongation. One of the contributing factors to these deformations is the neutron spectrum in the PT, especially neutrons with energy greater than 1 MeV. In this paper, the sensitivity of the neutron spectrum in the PT to various operating conditions of a PHWR is evaluated. Sensitivity assessments have been completed on several parameters namely the coolant purity, coolant temperature, moderator purity, moderator temperature, and pressure tube diametral creep (PTDC). Since the deformations are inevitable, identifying parameters that have the greatest impact on the population of neutrons with energy greater than 1 MeV in the PTs would be important for fine-tuning the operational conditions which are conducive to reducing the rate of deformation. From this study, it is concluded that the coolant temperature and PTDC are the two variables having the most impact on the fast energy neutron population in the pressure tubes.

Computational fluid dynamic investigations of flow bypass through an aged CANDU pressure tube

Single-phase, isothermal computational fluid dynamics simulations of a whole CANada Deuterium Uranium (CANDUTM) channel with twelve simplified CANDU 37M fuel bundles were performed with Hydra. Both as-received and aged pressure tubes were considered in this work. The aged pressure tube geometry considered asymmetric sag and diametral expansion profiles with a maximum sag of 57 mm and a maximum diametral expansion of 5.2%. The predicted fluid velocity and turbulent kinetic energy obtained from the two cases were compared in order to predict the impact of pressure tube deformation (sag and diametral expansion) on coolant flow. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. It has been determined that pressure tube deformation has a significant impact on coolant flow behaviour, whereby up to 27% of flow bypasses the fuel.

Emergency response implications of tritium in CANDU severe accidents

Tritium is produced in large quantities by heavy-water moderated CANDU reactors. Even though it is only a weak beta emitter, its dose impact during reactor accidents can be important under certain circumstances. Improved models for characterizing the transport of tritium as part of steam and water have been developed as an external model in the MAAP-CANDU severe accident code. Owing to the high mobility of tritiated steam, a large fraction of the radionuclide inventory can be released to the environment, where its contribution to inhalation doses can be significant. Moreover, after the accident, plant evapotranspiration can lead to continued elevated airborne tritium concentrations in contaminated areas, as shown with a simplified environmental transport model. Gamma measurement-based operational intervention levels (OILs) can miss this contribution, and it is proposed to use derived air concentration limits as an additional metric for making decisions about emergency protective actions like evacuation or relocation.

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).

Numerical studies of gas pull-through and liquid entrainment in gas–liquid flows through multi-branch manifolds

Flows of air–water and saturated steam–water mixtures in a horizontal cylindrical vessel with a single or several cylindrical outlets (branches) were investigated computationally using the volume of fluid (VOF) method to separate the phases and the delayed detached eddy simulation (DDES) model to simulate turbulence. The time-dependent gas–liquid interface was resolved and the vertical distance between the interface elevation and the branch inlet was determined. The computed critical elevations at the onsets of gas pull-through (OGP) and liquid entrainment (OLE) in single-branch vessels were found to be in fair agreement with experimental values under comparable conditions. Simulations of gas–liquid flows in manifolds with three branches also predicted successfully OGP and OLE in the different branches. As an example of the applicability of the present method, we investigated a problem of interest to the CANDU nuclear reactor safety analysis, namely, the discharge of fluid through a small break at the top of a header model with 12 other branches (feeders), focusing on its dependence on the inlet flow rates.

Mesh and turbulence model sensitivity analyses of computational fluid dynamic simulations of a 37M CANDU fuel bundle

Mesh and turbulence model sensitivity analyses have been performed on computational fluid dynamics simulations executed with Hydra and ANSYS Fluent for a single CANadian Deuterium Uranium (CANDU) 37M nuclear fuel bundle placed within a standard pressure tube. The goal of this work was to perform a methodical analysis to objectively determine an appropriate mesh and to gauge the sensitivity of different turbulence models for CANDU subchannel flow under isothermal conditions. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. Four meshes were generated with ANSYS Workbench Meshing, ranging from 22 to 84 million cells, and analyzed here to determine an appropriate level of mesh resolution and quality. Five turbulence models were compared in the turbulence model sensitivity analysis: standard k−ε, RNG k−ε, realizable k−ε, SST k−ω, and the Reynolds Stress Model. The intent of this work was to gain confidence in mesh generation and turbulence model selection of a single bundle to inform the decision making of subsequent investigations of an entire fuel channel containing a string of twelve bundles.

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.

Analysis of Dimensional Change in Zr-2.5%Nb CANDU Pressure Tube Material aged at 300-400°C using Neutron Diffraction

The primary boundary of the CANDU reactor consists of Zr-2.5%Nb alloy pressure tubes and SA106 alloy feeder pipes. The pressure tubes are exposed to fast neutron (E>1Mev) irradiation during reactor operation, and experience dimensional changes. The actual dimensional change in the pressure tube during in-service operation is greater than the design estimate. It is not fully understood what is responsible for the dimensional changes in the pressure tubes. In this study, to accelerate the aging effect on the pressure tube material, aging was performed at 300-400°C for up to 20,000 hours and neutron diffraction was used to track variation in lattice spacing, thereby identifying the reason for the dimensional change. The analysis result showed that α-Zr, the main phase of Zr-2.5%Nb, had little dimensional change at 350°C or lower. On the other hand, the aging treatment at 400°C resulted in anisotropic expansion in the α-Zr of the HCP crystal, of 0.02% and 0.08% in the (1010) and (0002) directions, respectively. It was confirmed that above 350oC the β-phase in the Zr-2.5%Nb alloy was decomposed to precipitate β-Nb . The driving force of the lattice expansion may be due to the diffusing out tendency of the super saturated Nb in α-Zr and due to the residual cold working effect applied during the manufacturing process. The enhancing effects of fast neutron irradiation on lattice diffusion are discussed in detail.

Development and application of the improved multicell method for the PWR fuel assemblies adjacent to reflector

The conventional “two-step” scheme based on assembly homogenization is widely applied in the modeling and simulation for PWRs. In the “two-step” scheme, the reflective boundary condition is applied in the two-dimensional calculation of fuel assembly, providing the few-group constants and ADFs (Assembly Discontinuity Factors) for the three-dimensional core calculation. However, the neutron-leakage phenomenon would occur to the fuel assemblies adjacent to the reflector, in which case the approximation of the reflective boundary condition applied to the single fuel assembly would no longer be suitable. In order to consider the neutron-leakage phenomenon, the multicell method originally developed for CANDU reactor has been improved and applied to PWRs for the fuel assemblies adjacent to reflector in this paper. Through the heterogeneous simulation of the multicell case, which are composed of fuel and reflector assemblies to characterize the neutron leakage from the fuel assembly, the multicell correction factors for the few-group constants and ADFs can be determined and applied to the core calculation. Applying the improved multicell method, our home-developed Bamboo-C code has been improved. For the method verifications, the baffle-reflector PWR and heavy-reflector PWR have been modeled with the improved Bamboo-C code, and the keff and power distributions are compared with reference values provided by the Monte-Carlo code. The numerical results show that the improved multicell method can reduce the power-distribution differences notably, which indicate the effectivity of this method for the fuel assemblies adjacent to both the baffle-reflector PWR and heavy-reflector PWR.

Modeling of flux-dependent fluence effect on in-reactor deformation in Zr-alloy tubes

Zr-alloy tubes in reactors suffer from harsh thermal-mechanical-irradiation coupling conditions. Abundant modeling efforts have been devoted to predicting in-reactor deformation but the flux-dependent fluence effect was rarely considered. In this work, a polycrystalline model which considers this effect is proposed, by introducing a flux-dependent factor α which characterizes the trend that higher flux results in more accumulated irradiation defects and therefore higher obstruction on dislocation movement. With this model, the flux-dependent anisotropic deformation behaviors as a function of fluence can be well captured. Particularly, a linear relationship between α and flux is revealed from the experimental results in CANDU reactors.

Development and Testing of TRACE/PARCS ECI Capability for Modelling CANDU Reactors with Reactor Regulating System Response

The use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal-hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE-PARCS in this area. In this work, the system thermal-hydraulic code TRACE_Mac1.0 is natively coupled with the core physic code PARCS_Mac1.0, and RRS control is implemented via the exterior communications interface (ECI) in TRACE. ECI is used for coupling the external codes to TRACE, including additional physical models and control system models. In this work, a Python interface to the TRACE ECI library is developed, along with an RRS model written in Python. This coupling was tested using a CANDU-6 IAEA code coupling benchmark and a 900 MW CANDU model for various transients. For the CANDU-6 benchmark, the transients did not include RRS response, however, the TRACE_Mac1.0/PARCS_Mac1.0 coupling and ECI script functionality was compared to the previous benchmark simulations, which utilized external coupling. For the 900 MW CANDU simulations, all aspects of the ECI module and RRS were included. The results from the CANDU-6 benchmark when using the built-in coupling are comparable to those previously achieved using external coupling between the two codes with coupled simulations taking 2x to 3x less execution time. The 900 MW CANDU simulations successfully demonstrate the RRS functionality for the loss of flow events, and the coupled solutions demonstrate adequate performance for figure-of-eight flow instability modeling.