[en] Canada is developing a next generation (Gen IV) reactor concept, the Canadian Super-Critical Water-cooled Reactor (SCWR), which will meet the technology goals of the Generation-IV International Forum (GIF). These goals include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization (∼40% more efficient than current nuclear power generating stations), sustainable fuel cycle, and greater proliferation resistance than Gen III reactors. The Canadian SCWR concept is a pressure-tube type reactor that uses supercritical water as a coolant, a separate low-pressure heavy water moderator, and a direct steam power cycle. This paper presents the evolution of Canadian SCWR core concept, in particular the recently developed counter-flow fuel assembly and its impact on reactor core concept. Integration of the reactor core with the supporting safety systems is also described to address long term reactor heat removal at station blackout conditions. (author)

[en] This paper presents a study on thermal-mechanical behaviour of a fuel element proposed for the Canadian Supercritical Water Cooled Reactor (SCWR). In the Canadian SCWR, the coolant pressure is 25 MPa, and the temperature is 350^oC at the inlet and 625^oC at the outlet of the reactor core. Critical design decisions for fuel design will be the selection of the fuel sheath material and details of the fuel element design options (sheath thickness, pellet-clad gap, internal pressure, etc.). The analysis presented in this paper predicted temperature, stress and strain in the fuel element of the Canadian SCWR with a collapsible sheath using ANSYS. Typical conditions for the evaluation of the fuel behaviour, such as linear heat generation rate, coolant temperature and sheath surface heat transfer coefficient, were extracted from core and fuel channel designs. The temperature distribution in the fuel element is predicted by a thermal model and then the thermal model is coupled sequentially with a structural model to predict fuel sheath deformation under the predicted temperature distribution and external (coolant) pressure. Nonlinear thermo-mechanical simulations include nonlinear buckling with elastic-plastic deformation. Three sheath collapse phenomena are considered: (1) elastic collapse by buckling, (2) longitudinal ridging and (3) plastic collapse by yielding. The numerical models are validated against analytical and experimental data. The presented results show the temperature distribution, deformed shape, stress and strain of the fuel element, allowing the designers to select appropriate sheath material and element design options for the SCWR fuel element design. (author)

[en] The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

[en] The Canadian Super-Critical Water-cooled Reactor (SCWR) is a Generation IV reactor that is being designed as part of the international initiative to design next generation nuclear reactors known as the Generation-IV International Forum (GIF). It is a heavy-water moderated pressure-tube type reactor that uses supercritical water as coolant (which allows ∼40% higher thermodynamic efficiency than current light-water reactors), employs passive safety systems, uses an insulated fuel channel design, and burns thorium as fuel for sustainability and proliferation resistance. The pressure tube design feature provides more flexibility to optimize the reactor efficiency in addition to enhancing reactor safety in comparison with other reactor designs. The operating pressure (25 MPa) and temperatures (typically 450^oC to 625^oC) of SCW reactors are significantly higher than those in existing light-water reactors, presenting design challenges that require innovative solutions. This paper provides a summary of current status of the mechanical design of the Canadian SCWR, discusses some of the design challenges and proposes path forward for future R&D to deal with these challenges. Also the paper discusses a variety of technologies that may be employed to achieve optimized reactor efficiency and increased plant reliability as well as plant economics. (author)

[en] A simple criterion is proposed to estimate fretting-wear damage in heat exchanger tubes with clearance supports. The criterion is based on parameters such as vibration frequency, mid-span vibration amplitude, span length, tube mass and an empirical wear coefficient. It is generally accepted that fretting-wear damage is proportional to a parameter called work-rate. Work-rate is a measure of the dynamic interaction between a vibrating tube and its supports. Due to the complexity of the impact-sliding behavior at the clearance-supports, work-rate calculations for heat exchanger tubes require specialized non-linear finite element codes. These codes include contact models for various clearance-support geometries. Such non-linear finite element analyses are complex, expensive and time consuming. The proposed criterion uses the results of linear vibration analysis (i.e., vibration frequency and mid-span vibration amplitude due to turbulence) and does not require a non-linear analysis. It can be used by non-specialists for a quick evaluation of the expected work-rate, and hence, the fretting-wear damage of heat exchanger tubes. The proposed criterion was obtained from an extensive parametric study that was conducted using a non-linear finite element program. It is shown that, by using the proposed work-rate criteria, work-rate can be estimated within a factor of two. This result, however, requires further testing with more complicated flow patterns. (author)

[en] Degradation of steam generators (SGs) has a significant effect on CANDU heat transport system effectiveness and the overall efficiency of a nuclear power plant. SG degradation results in an increase in the reactor inlet header temperature (RIHT), which has to be kept below an upper limit to avoid dryout of the fuel. Therefore, a better understanding of SG degradation mechanisms is needed to avoid plant de-ratings. In this report, various SG degradation mechanisms are identified and their effect on the RIHT is quantified using field data and numerical modelling. Analysis shows that thermal performance of CANDU SGs has been mostly affected by a combination of divider plate leakage and tube-bundle fouling. The results also indicate that the rate of tube-bundle fouling in CANDU 6 steam generators is equal to 2.05*10^-6 m²·^oC/W per effective full power year (EFPY), of which the majority of this is attributed to primary-side fouling. (author)

[en] Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. The purpose of this paper is to review some of the recent findings in the area of flow-induced vibration and to discuss some of the remaining questions. Vibration excitation mechanisms and damping mechanisms are described with particular emphasis on fluidelastic instability and damping in two-phase flows. The need for a better understanding of two-phase flow regimes, particularly in cross flow, is outlined. The dynamic characteristics of nuclear structures are explained. The statistical nature of some parameters, in particular support conditions, is discussed. The prediction of fretting-wear damage is approached from several points of view. An energy approach to formulate fretting-wear damage is proposed. (author)

[en] Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington NGS (nuclear generating station) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGS's (nuclear generating stations). (author). 12 refs., 2 tabs., 11 figs

[en] The degradation of steam generators (SGs) has a significant effect on nuclear heat transport system effectiveness and the lifetime and overall efficiency of a nuclear power plant. Hence, quantification of the effects of degradation mechanisms is an integral part of a SG degradation management strategy. Numerical analysis tools such as THIRST, a 3-dimensional (3D) thermal hydraulics code for recirculating SGs; SLUDGE, a 3D sludge prediction code; CHECWORKS a flow-accelerated corrosion prediction code for nuclear piping, PIPO-FE, a SG tube vibration code; and VIBIC and H3DMAP, 3D non-linear finite-element codes to predict SG tube fretting wear can be used to assess the impacts of various maintenance activities on SG thermal performance. These tools are also found to be invaluable at the design stage to influence the design by determining margins or by helping the designers minimize or avoid known degradation mechanisms. In this paper, the aforementioned numerical tools and their application to degradation mechanisms in CANDU recirculating SGs are described. In addition, the following degradation mechanisms are identified and their effect on SG thermal efficiency and lifetime are quantified: primary-side fouling, secondary-side fouling, fretting wear, and flow-accelerated corrosion (FAC). Primary-side tube inner diameter fouling has been a major contributor to SG thermal degradation. Using the results of thermalhydraulic analysis and field data, fouling margins are calculated. Individual effects of primary- and secondary-side fouling are separated through analyses, which allow station operators to decide what type of maintenance activity to perform and when to perform the maintenance activity. Prediction of the fretting-wear rate of tubes allows designers to decide on the number and locations of support plates and U-bend supports. The prediction of FAC rates for SG internals allows designers to select proper materials, and allows operators to adjust the SG maintenance strategy. CANDU nuclear power plants are pressurized heavy-water reactors that differ in design from pressurized water reactors (PWRs). As a result of this difference, degradation mechanisms in PWRs might be somewhat different; for example, unlike CANDU systems, PWRs do not experience significant primary-side fouling. However, the methodologies presented in this paper are applicable to both CANDU and PWR SGs. (authors)

[en] To investigate pressure-tube reactor in-vessel flow blockage during a simulated Loss-Of-Coolant Accident (LOCA) long-term core cooling operation, numerical simulations were performed for a pressure-tube reactor to understand its post-LOCA ECC flow in each fuel channel and the driving force between inlet header and outlet header. In addition, experiments were performed using a prototypical pressure tube with two dummy fuel bundles installed downstream of a reduced-scale strainer test rig. An end-fitting with a liner tube was connected to the pressure tube to simulate the actual fuel channel in a pressure-tube reactor. Particulate and fibrous debris simulating post-LOCA conditions were added to the strainer test rig. In addition, chemical reactants were also added to the test rig to simulate post-LOCA sump chemistry and to investigate the head loss effects of possible chemical precipitates. A total of five tests were performed, which included two thin bed tests, one full debris load tests and two chemical effects tests with full debris load. Bypassed debris from the strainer surface was observed to enter the fuel channel and accumulate inside the liner tube and on the fuel bundle endplates. Pressure drop due to the debris and chemical precipitates blockage across the fuel channel were measured and compared with the available driving force to determine whether sufficient flow would enter the pressure tube to remove decay heat. Preliminary testing results showed that the pressure tube head losses were not significant for the test conditions considered. (author)