Mississauga, Canada

Candu Energy Inc.

www.candu.com
Mississauga, Canada

Candu Energy Inc. is a Canadian wholly owned subsidiary of Montreal-based SNC-Lavalin Inc., specializing in the design and supply of nuclear reactors, as well as nuclear reactor products and services. Candu Energy Inc. was created in 2011 when parent company SNC-Lavalin purchased the commercial reactor division of Atomic Energy of Canada Limited , along with the development and marketing rights to CANDU reactor technology.Candu Energy Inc. is located in Mississauga, Ontario, Canada. Candu Energy lists its main business lines as: CANDU life extension CANDU maintenance and performance services CANDU new buildThe reactor products offered by Candu Energy Inc. are the CANDU 6 and Enhanced CANDU 6 reactors. Candu Energy Inc. also specializes in advanced fuel cycle technology that exploits the fuel cycle flexibility of the CANDU design, including fuels based on Recovered Uranium from Light Water Reactors and Mixed-Oxide fuel incorporating thorium or plutonium.In 2014, Preston Swafford was hired to lead the company as its Chief Nuclear Officer, President & CEO. Also in 2014, Candu Energy increased sharing of human resources with SNC-Lavalin. Wikipedia.

SEARCH FILTERS
Time filter
Source Type

Liu Y.,Candu Energy Inc. | Lu Z.,Nanjing Southeast University
Earthquake and Structures | Year: 2014

A method is presented in this paper to analyze the dynamic response behavior of suspended building structures. The effect of semi-rigid connections that link suspended floors with their supporting structure on structural performance is investigated. The connections, like the restrains in non-structural suspended components, are designed as semi-rigid to avoid pounding and as energy dissipation components to reduce structural response. Parametric study is conducted to assess the dynamic characteristics of suspended building structures with varying connection stiffness and suspended mass ratios. Modal analysis is applied to identify the two distinct sets of vibration modes, pendulum and bearing, of a suspended building structure. The cumulative modal mass is discussed to ensure the accuracy in applying the method of response spectrum analysis by SRSS or CQC modal combination. Case studies indicate that a suspended building having semi-rigid connections and proper suspended mass ratios can avoid local pounding failure and reduce seismic response. © 2014 Techno-Press, Ltd.


Kastanya D.,Candu Energy Inc.
Nuclear Engineering and Design | Year: 2015

In CANDU® reactors, the regional overpower protection (ROP) systems are designed to protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. In the CANDU® 600 MW (CANDU 6) design, there are two ROP systems in the core, each of which is connected to a fast-acting shutdown system. Each ROP system consists of a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal flux detector assemblies. The placement of these ROP detectors is a challenging discrete optimization problem. In the past few years, two algorithms, DETPLASA and ADORE, have been developed to optimize the detector layout for the ROP systems in CANDU reactors. These algorithms utilize the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The objective of the optimization process is typically either to maximize the TSP value for a given number of detectors in the system or to minimize the number of detectors in the system to obtain a target TSP value. One measure to determine the robustness of the optimized detector layout is to evaluate the maximum decrease (penalty) in TSP value when any single detector in the system fails. The smaller the penalty, the more robust the design is. Therefore, in order to ensure that the optimized detector layout is robust, the single detector failure (SDF) criterion has been incorporated as an additional constraint into the ADORE algorithm. Results from this study indicate that there is a significant reduction in the TSP penalty value of the optimized solution, as a result of incorporating SDF criterion during the optimization process. © 2015 Elsevier B.V. All rights reserved.


Zhang X.,Candu Energy Inc.
Canadian Nuclear Society - 35th Annual Conference of the Canadian Nuclear Society and 38th CNS/CNA Student Conference 2015 | Year: 2015

A curved pipe element, ELBOW290, became available in ANSYS® 12. This element was developed based on a simplified shell theory, and maintains the ability to capture cross-sectional deformations of elbows. Numerical testing on the applicability of this element for the flexibility calculation of the tight radius bends in CANDU® reactors is carried out to determine the usability of this element in completing stress analyses for feeder pipes. Comparisons are made between the ELBOW290 and the shell element for various feeder bend types found in domestic and overseas CANDU reactors. The comparisons show that the ELBOW290 element is suitable for calculating the flexibility of the tight radius bends. © Copyright 2015 by the Canadian Nuclear Society. All rights reserved.


Reinhardt W.,Candu Energy Inc.
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2015

An assessment of pressure retaining components based on ultimate load (or plastic instability load) is used in Section VIII Div. 2 and Section III Appendix F of the ASME Boiler and Pressure Vessel Code, as well as in various fitness-for-service standards and guidelines. The ultimate load analysis strives for a realistic prediction of the plastic behavior of a component up to the highest load or load combination that the component can support. The high level of applied load may cause significant deformations in the component, and for an accurate assessment the analysis must consider the effect of these deformations on the material as well as on equilibrium and the state of stress. This paper discusses briefly the basis of ultimate load analysis and considerations in performing such analysis with finite element analysis. The main focus is to validate the analysis approach by demonstrating close agreement between finite element analysis and analytical solutions and the results of component tests. Copyright © 2015 by ASME.


Duan X.,Candu Energy Inc.
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2015

Wall thinning by Flow Accelerated Corrosion (FAC), Erosion-Cavitation (E-C), and fretting has been observed in carbon steel piping. At some locations, the predicted end-of-life wall thickness could be below the design minimum thickness required by ASME B&PV Code Section III, which is an extremely conservative analysis method. To support the life extension without replacement or repair, it is essential to demonstrate the fitness for service of degraded piping components that satisfy the mandatory structural factors with uncertainties explicitly identified and addressed in an integrated manner. The present paper discusses the key technical basis for a sound assessment of small bore high energy Class 1 piping with lower than Code specified design minimum thickness, and proposes some future research activities to further enhance the technical basis. Copyright © 2015 by ASME.


Kastanya D.,Candu Energy Inc.
Annals of Nuclear Energy | Year: 2014

In introductory courses for nuclear engineering, the concept of critical dimension and critical mass are introduced. Students are usually taught that the geometrical shape which needs the smallest amount of fissionable material to reach criticality is a sphere. In this paper, this concept is explored further using MCNP code. Five different regular polyhedrons (i.e.; the Platonic solids) and a sphere have been examined to demonstrate that sphere is indeed the optimal geometrical shape to minimize the critical mass. For illustration purpose, the fissile isotope used in this study is 239Pu, with a nominal density of 19.8 g/cm3. © 2014 Elsevier Ltd. All rights reserved.


The key design quantities of the pressure-tube-based (PT-based) Super Critical Water Reactor (SCWR) core design are expected to be computed with the traditional core-analysis code which solves the two-group neutron-diffusion equation by using lattice-homogenized cross sections calculated with the lattice code. Two issues may affect the accuracy of these computed quantities for the SCWR core: one is the two-energy-group neutron-diffusion theory; the other is the generation of lattice-homogenized properties with the lattice code based on the single-lattice-cell model without considering the effects of the environment. It has been illustrated that the single-lattice-cell method is not sufficiently accurate for heterogeneous core configurations when adjacent channels experience significant spectrum interaction. To ensure the qualification of these computed quantities for the SCWR core, a 2-D SCWR benchmark problem was setup (with the reference solution provided by the continuous energy Monte-Carlo code SERPENT) to assess the traditional neutron-diffusion core-analysis method. The assessment shows that the traditional two-group neutron-diffusion theory with the single-lattice-cell- based lattice properties is not sufficient to capture either the spectral change or the environment effect for the SCWR core. The solution of the eight-group neutron-diffusion equation by using lattice-homogenized cross sections calculated with the multicell model is considered appropriate for the analysis of the PT-based SCWR core. © 2012 Elsevier Ltd. All rights reserved.


The regional overpower protection (ROP) systems protect CANDU® reactors against overpower in the fuel that could reduce the safety margin-to-dryout. The overpower could originate from a localized power peaking within the core or a general increase in the global core power level. The design of the detector layout for ROP systems is a challenging discrete optimization problem. In recent years, two algorithms have been developed to find a quasi optimal solution to this detector layout optimization problem. Both of these algorithms utilize the simulated annealing (SA) algorithm as their optimization engine. In the present paper, an alternative optimization algorithm, namely the genetic algorithm (GA), has been implemented as the optimization engine. The implementation is done within the ADORE algorithm. Results from evaluating the effects of using various mutation rates and crossover parameters are presented in this paper. It has been demonstrated that the algorithm is sufficiently robust in producing similar quality solutions. © 2012 Elsevier Ltd. All rights reserved.


Shen W.,Candu Energy Inc.
Annals of Nuclear Energy | Year: 2012

Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancelation of error) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems with three different fuel lattices (CANDU, HWR and PWR) were designed with the reference solution provided by the Monte-Carlo code SERPENT in this paper. The analysis of these benchmark problems confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the core-diffusion calculation. When lattice-homogenized cross sections are used in the core-diffusion calculation, it is appropriate to use the fine-mesh FDM for reactors (such as PWR) with uniformly-distributed fuel pins; however, it is inappropriate to use the fine-mesh FDM to analyze CANDU-type reactors with the cluster fuel geometry because the lattice-homogenized cross sections assigned to each sub-mesh are not physically meaningful. It is recommended to use the coarse-mesh (2 × 2 meshes per lattice) to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core-diffusion calculation. © 2012 Elsevier Ltd. All rights reserved.


Methods and systems for detecting an individual leaking fuel channel included in a reactor. One system includes a plurality of inlet lines and a plurality of outlet lines. Each of the plurality of inlet lines feeding annulus fluid in parallel to an annulus space of each of a first plurality of fuel channels included in the reactor, and each of the plurality of outlet lines collecting in parallel annulus fluid exiting an annulus space of each of a second plurality of fuel channels included in the reactor. In some embodiments, the system also includes a detector positioned at an outlet of each of the plurality of outlet lines configured to detect moisture in annulus fluid and identify a first position of an individual leaking fuel channel, and an isolation valve positioned at an inlet of each of the plurality of inlet lines operable to stop annulus fluid from circulating through one of the plurality of inlet lines and to identify a second position of the individual leaking fuel channel.

Loading Candu Energy Inc. collaborators
Loading Candu Energy Inc. collaborators