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[en] The Canadian Nuclear Society's 13th International Conference on CANDU Fuel was held in Kingston, Ontario, Canada on August 15-18, 2016, with the theme as 'CANDU Fuel : Evolution Towards Optimal Performance'. The conference brought together more than 100 leading experts, strong designers, engineers, manufacturers, researchers and modellers to discuss and share their knowledge and experience of CANDU fuel performance, safety and design. The following topics were examined and explored, along with topics related to fuel and fuel channel science and technology were encouraged: fuel performance, fuel safety, fuel design and development, fuel code development, fuel manufacturing, fuel management, fuel and fuel channel thermalhydraulics, spent fuel management, and advanced code development.
[en] In the last fifteen years or so there has been, within the Canadian nuclear industry, a significant amount of fuel safety criteria related activities. The primary objective of this presentation is to provide a regulatory perspective of those activities. A brief overview of the CNSC's regulatory oversight activities will therefore be provided. Given that I am currently the Chair of an IAEA Working Group, whose mandate is to write a technical document (TECDOC) on fuel safety criteria used in countries where pressurized heavy water reactors are operated, I will conclude the presentation with a brief description of this IAEA project.
[en] This article interprets how to locate failed fuel by Xe133 trend analysis. Xe133 trend is related to bundle burnup, bundle power history, bundle movement and other factors. Xe133 trend following fuel defect can help to find failed channel, Xe133 trend following bundle movement can help to confirm the bundle position in the channel. GFP handswitching can help to locate the half loop in which the failed fuel situated. (author)
[en] Canadian Nuclear Laboratories is mandated by the Government of Canada to conduct nuclear fuel research and development under its Energy and Safety and Security programs. The Energy Program positions Canada and the world to meet future nuclear energy needs, particularly in the area of advanced fuel cycles. The Safety and Security Program promotes nuclear emergency preparedness and response capability, and includes studies on fuel behaviour during abnormal and accident conditions. These programs position CNL to deliver on federal government needs for innovation in nuclear fuel science and technology (S&T) and to provide support to the nuclear industry on a commercial basis. Collaboration with other nuclear S&T institutions and universities, both domestic and international, is integral to CNL leveraging its S&T commitments and intellectual property to maximize benefits to Canada and the world.
[en] Following a LOCA event, the long term decay power removal is ensured by low pressure ECC pumped flow. As per design, a Site Design Earthquake (SDE) can occur at 24 h after the initial assumed event. The paper will present the main results related to broken loop fuel behaviour following such a sequence of events. (author)
[en] LOCA (loss of coolant accident) is one of the representative severe accident in the operation of nuclear reactors. One of the possible cases is the ballooning of pressure tube and the contact with the Calandria tube for the outer coolant to boil abruptly to saturate. For the moderator system of CANDU-6 reactor, the benchmark model of IAEA/ISCP has been setup to conduct a test and comparison with numerical models on the multi-physics of ballooning PT(pressure tube) and its contact with CT(Calandria tube). The graphite heater is modeled as a volumetric heat source for the given unsteady power of 150 kW, and the PT and CT are installed eccentrically in a given interval. In the first stage, the radiation heat transfer between Zircaloy surfaces of tubes is considered with the same structural deformation model as CATHENA code for the zero-gradient temperature outer boundary condition. A commercial code, COMSOL Multiphysics is used for simulations, and the Gaussian quadrature is applied for the integration of gradient for temperature. The two-dimensional result shows a good agreement for high-order numerical integration in the model with CATHENA one. (author)
[en] The performance PHWR fuel bundle depends on the quality of resistance weld joints and individual components of the bundle. Stringent quality inspection techniques are followed, which includes visual and dimensional inspection of all resistance welds and bundle components to avoid defective components being converted to final bundle form in production stream. An automatic vision based tube inspection system is now being introduced at NFC, which identifies and separates acceptable and defective fuel tubes by carrying out multiple quality checks on a single unit. The details of the inspection system, its reliability and advantages are described in this paper. (author)
[en] CANDU® fuel bundle deforms during operation and under postulated AOO and DBA conditions. Bundle deformations under high-temperature (HT) transient conditions are complex and require many new models to be developed, implemented and tested. Candu Energy has developed a computer program prototype to simulate bundle deformations under AOO and DBA conditions. Some examples of new models include those for dealing with transient conditions, HT thermal loads, inter-element contacts at locations other than spacer pads bearing pads (i.e., sheath-sheath contacts and sheath-pressure tube (PT) contacts), effects of pre-transient conditions (bundle deformations from normal operation prior to AOOs or DBAs), HT sheath creep, and HT material property changes. This paper describes an overview of Candu Energy's experience in developing a proprietary computer program, HT version of BOW code (called 'BOW-HT' thereafter in this paper) to model fuel bundle deformation under HT transients conditions. (author)
[en] Light water fuels operate in an environment that involves complex multiphysics phenomena, and the multiphysics behavior is often tightly coupled. A fuel performance code called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed to predict light water reactor fuel behavior. CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, and fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO2 fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide fuel (UO2), mixed oxide fuel ((Th0.9,U0.1)O2), and accident tolerant fuels, for example, enhanced thermal conductivity UO2-BeO, UO2-SiC and U3Si2 fuels, under normal operation conditions. These demonstrate that CAMPUS is a practical tool not only for the existing LWR fuel performance modeling but also for new fuel design, development and behavior prediction. (author)
[en] In the IEC-I-1 test, an internally heated CANDU fuel element simulator was subjected to five temperature transients: three transients to a sheath peak temperature of 600oC, followed by two further transients to a sheath temperature of 800oC. A detailed 3D finite element model of this experiment was created using the ANSYS finite element software package. The model includes fuel pellets and sheath as separate components that interact via contact elements. The study included a detailed sensitivity study of a number of model parameters, including pellet-to-sheath friction, contact penalty function, rigid body damping, time step size and the creep rate ratio limit. (author)