WorldWideScience

Sample records for candu fuel channel

  1. CANDU-PHW fuel channel replacement experience

    International Nuclear Information System (INIS)

    One of the main characteristics of the CANDU pressurized heavy water reactor is the use of pressure tubes rather than one large pressure vessel to contain the fuel and coolant. This provides an inherent design capability to permit their replacement in an expeditious manner, without seriously affecting the high capacity factors of the reactor units. Of th eight Ontario Hydro commercial nuclear generating units, the lifetime performance places seven of them (including two that have had some of their fuel channels replaced), in the top ten positions in the world's large nuclear-electric unit performance ranking. Pressure tube cracks in the rolled joint region have resulted in 70 fuel channels being replaced in three reactor units, the latest being at the Bruce Nuclear Generating Station 'A', Unit 2 in February 1982. The rolled joint design and rolling procedures have been modified to eliminate this problem on CANDU units subsequent to Bruce 'A'. This paper describes the CANDU pressure tube performance history and expectations, and the tooling and procedures used to carry out the fuel channel replacement

  2. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

    The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author)

  3. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of a CANDU fuel channel are reviewed and the processes necessary to achieve the objectives are identified. Performance of future fuel channels will be enhanced by reduced operating costs and increased safety margins to postulated accident conditions compared with those for current channels. The approaches to these issues are discussed briefly in this report. (author)

  4. Remotely operated inspection equipment for the Candu fuel channels

    International Nuclear Information System (INIS)

    Equipment is described which has been successfully used for the nondestructive inspection of fuel channel components within Ontario Hydro's CANDU nuclear reactors. By the use of automated systems, significant savings in personnel radiation exposure and unit outage duration have been realized, with improved quality and quantity of nondestructive examination information. (author)

  5. Development of defueling device for CANDU fuel channel (modeling)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Yu, K. H.; Yang, J. S.; Lee, H. S.; Chang, K. J.; Kim, Y. J. [CNEC Technical Office, Taejon (Korea, Republic of); Lee, S. K. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    Commercial CANDU reactors use D{sub 2}O for moderator and heat transfer material and also have Fueling Machines(F/M) and related system equipment in order to assist on-power refueling operation. A Defuelling Device(DFD) is developed for the proper defuelling of all fuels in all fuel channels during shutdown condition of plant. This device is considered more efficient in defuelling compared to the existing Fuel Grapple System for its use of existing D{sub 2}O flow in the fuel channel. In this study, computational fluid dynamic software is used for optimize and evaluation of the design for its applicability.

  6. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of the channel are reviewed and the processes necessary to achieve the objectives identified. Performance of future fuel channels will be enhanced by reduced operating costs, increased safety margins to postulated accident conditions, and reduced retubing costs compared to current channels. The approaches to these issues are discussed briefly in the paper. (author)

  7. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

  8. Research and development for CANDU fuel channels and fuel

    International Nuclear Information System (INIS)

    The CANDU nuclear reactor is distinctly different from BWR and PWR reactors in that it uses many small pressure tubes rather than one large pressure vessel to contain the fuel and coolant. To exploit the advantages of the natural uranium fuel, the pressure tubes, like other core components, are manufactured from zirconium alloys which have low neutron capture cross sections. Also, because natural uranium fuel only achieves a modest burnup, a simple and inexpensive fuel design has been developed. The present paper reviews the features and the research that have led to the very satisfactory performance of the pressure tubes and the fuel in CANDU reactors. Reference is made to current research and development that may lead to further economies in the design and operation of future power reactors. (author)

  9. CANDU 6 fuel channel stress analysis using ANSYS fatigue module

    International Nuclear Information System (INIS)

    Design reliability can be confirmed by the stress analysis, and its results become the basis of the structural integrity for components. The report presents the development of CANDU 6 fuel channel stress analysis methodology and procedure per ASME Code using the ANSYS fatigue module. Stress analysis was performed in accordance with the procedure developed on the basis of ASME Code Section III NB-3200. FORTRAN programs and ANSYS macros used in data processing were developed to systematized the analysis. Stresses were separately analyzed for mechanical and thermal load respectively, and then combined in the post-processing stage for the various conditions. Maximum stress intensity range was then calculated at selected nodes by using the ANSYS fatigue module for the sum of mechanical and thermal stress values. As a results, structural integrity of CANDU 6 fuel channel was proved in this report and analysis reliability for CANDU reactor was shown to be enhanced by the establishment of analysis procedure bases upon ASME Code. (Author) 11 refs., 11 figs., 6 tabs

  10. Thermalhydraulic characteristics for fuel channels using burnable poison in the CANDU reactor

    International Nuclear Information System (INIS)

    The power coefficient is one of the most important physics parameters governing nuclear reactor safety and operational stability, and its sign and magnitude have a significant effect on the safety and control characteristics of the power reactor. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. However, the previous study has mainly focused on the safety characteristics by evaluating the power coefficient for the fuel channel using BP in the CANDU reactor. Together with the safety characteristics, the economic performance is also important in order to apply the newly designed fuel channel to the power plant. In this study, the economic performance has been evaluated by analyzing the thermal hydraulic characteristics for the fuel channel using BP in the CANDU reactor

  11. Diagnostic technology for degradation of feeder pipes and fuel channels in CANDU reactor

    International Nuclear Information System (INIS)

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including detection and monitoring technology has raised its head. Because the feeder pipes and the fuel channels are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the improvement of CANDU reactor safety. To ensure the integrity of feeder pipes and fuel channels in CANDU nuclear plant, the following 3 research tasks were performed in the first stage. - Development of a model for prediction of feeder wall thinning - Development of RFEC detection technology - Development of ICFD noise signal analysis. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  12. Research on nondestructive examination methods for CANDU fuel channel inspection

    International Nuclear Information System (INIS)

    The requirements of the 1994 edition of CAN/CSA-N285.4 Periodic Inspection Standard, which address all known and postulated degradation mechanisms and introduce material surveillance demands, involve a growing need for improved nondestructive examination (NDE) methods and technologies. In order to have a proper technical support in its decisions concerning fuel channel inspections at Cernavoda NPP, the Romanian Power Authority (RENEL) initiated a Research Program regarding the nondestructive characterization of the fuel channels structural integrity. The paper presents the most significant results obtained on this Research Program: the ENDUS experimental system for Laboratory simulation of the fuel channel inspection, ultrasonic Rayleigh-Lamb waves technique for pressure tubes examination, phase analysis technique for near-surface flaws, influence of the metallurgical state of the pressure tube material on the eddy current defectoscopic signals, characterization of plastic deformation and fracture of zirconium alloys by acoustic emission. (author)

  13. A modal method for transient thermal analysis of CANDU fuel channel

    International Nuclear Information System (INIS)

    The classical modal expansion technique has been applied to predict transient fuel and coolant temperatures under on-power conditions in a CANDU fuel channel. The temperature profile across the fuel pellet is assumed to be parabolic and fuel and coolant temperatures are expanded with Fourier series. The coefficient derivatives are written in state space form and solved by the Runge-Kutta method of fifth order. To validate the present model, the calculated fuel temperatures for several sample cases were compared with HOTSPOT-II, which employs a more rigorous finite-difference model. The agreement was found to be reasonable for the operational transients simulated. The advantage of the modal method is the fast computation speed for application to the real-time system such as the CANDU simulator which is being currently developed at the Institute for Advanced Engineering (IAE). (author)

  14. Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

    International Nuclear Information System (INIS)

    Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

  15. Fuel Temperature Characteristics for Fuel Channels using Burnable Poison in the CANDU reactor

    International Nuclear Information System (INIS)

    Although the CANFLEX RU fuel bundle loaded 11.0 wt% Er2O3 are originally designed focused on the safety characteristics, the fuel temperature characteristics is revealed to be not deteriorated but rather is slightly enhanced by the decreased fuel temperature in the outer ring compared with that of standard 37 fuel bundle. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. In a view of safety, the fuel temperature coefficient (FTC) is an important safety parameter and it is dependent on the fuel temperature. For an accurate evaluation of the safety-related physics parameters including FTC, the fuel temperature distribution and its correlation with the coolant temperature should be accurately identified. Therefore, we have evaluated the fuel temperature distribution of a CANFLEX fuel bundle loaded with a burnable poison and compared the standard 37 element fuel bundle and CANFELX-NU fuel bundle

  16. The simulation of CANDU fuel channel behavior in thermal transient conditions

    International Nuclear Information System (INIS)

    In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

  17. CFD analysis of the 37-element fuel channel for CANDU6 reactor

    International Nuclear Information System (INIS)

    We analyzed the thermal-hydraulic behavior of coolant flow along fuel bundles with appendages of end support plate, spacer pad, and bearing pad, which are the CANDU6 characteristic design. The computer code used is a commercial CFD code, CFX-12. The present CFD analysis model calculates the conjugate heat transfer between the fuel and coolant. Using the same volumetric heat source as the O6 channel, the CFD predictions of the axial temperature distributions of the fuel element are compared with those by the CATHENA (one-dimensional safety analysis code for CANDU6 reactor). It is shown that CFX-12 predictions are in good agreement with those by the CATHENA code for the single liquid convection region (especially before the axial position of the first half of the channel length). However, the CFD analysis at the second half of the fuel channel, where the two-phase flow is expected to occur, over-predicts the fuel temperature, since the wall boiling model is not considered in the present CFD model. (author)

  18. CANDU fuel performance

    International Nuclear Information System (INIS)

    The paper presents a review of CANDU fuel performance including a 28-element bundle for Pickering reactors, a 37-element bundle for the Bruce and Darlington reactors, and a 37-element bundle for the CANDU-6 reactors. Special emphasis is given to the analysis of fuel defect formation and propagation and definition of fuel element operating thresholds for normal operation and accident conditions. (author)

  19. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

  20. ASSERT/NUCIRC commissioning for CANDU 6 fuel channel CCP analysis

    International Nuclear Information System (INIS)

    CANDU PHWR fuel channel pressure tubes will expand or creep under long-term (aging process) influence of temperature, pressure, and neutron flux. This diametral pressure tube creep will influence the critical channel power (CCP), or conditions that lead to dryout. In order to provide safety analysis models to quantify the effect of diametral pressure tube creep on CCP, a COG (AECL/NBP/HQ) project is underway to commission the ASSERT and NUCIRC codes to establish reliable production tools for the assessment of CANDU6 CCP in nominal (uncrept) and crept pressure tube fuel channels. This paper gives an overview of the background and objectives of the project along with a brief introduction into the subchannel analysis code ASSERT and the 1-D thermalhydraulics code NUCIRC. This project is a multistage endeavour, for which the first stage results are presented. A detailed cross-comparison of the 1-D (NUCIRC) and subchannel (ASSERT) models of pressure drop (ΔP) and critical heat flux (CHF) has been undertaken and has led to several enhancements and refinements to the respective models. These results are presented in addition to results of ASSERT commissioning against NUCIRC for a matrix of ΔP and dryout cases in a nominal pressure tube, which are based upon Gentilly 2 and Point Lepreau site area. Additionally, the initial results of an assessment, using ASSERT, of the effects of creep on ΔP are presented. In concluding, the status and future directions for ASSERT/NUCIRC CANDU 6 CCP analysis project are summarized. (author). 2 refs., 12 figs

  1. CANDU fuel cycle flexibility

    International Nuclear Information System (INIS)

    High neutron economy, on-power refuelling, and a simple bundle design provide a high degree of flexibility that enables CANDU (Canada Deuterium Uranium; registered trademark) reactors to be fuelled with a wide variety of fuel types. Near-term applications include the use of slightly enriched uranium (SEU), and recovered uranium (RU) from reprocessed spent Light Water Reactor (LWR) fuel. Plutonium and other actinides arising from various sources, including spent LWR fuel, can be accommodated, and weapons-origin plutonium could be destroyed by burning in CANDU. In the DUPIC fuel cycle, a dry processing method would convert spent Pressurized Water Reactor (PWR) fuel to CANDU fuel. The thorium cycle remains of strategic interest in CANDU to ensure long-term resource availability, and would be of specific interest to those countries possessing large thorium reserves, but limited uranium resources. (author). 21 refs

  2. Application of Shuttle Remote Manipulator System technology to the replacement of fuel channels in the Pickering CANDU reactor

    International Nuclear Information System (INIS)

    Spar Aerospace Limited of Toronto was the prime contractor to the National Research Council of Canada for the design and development of the Shuttle Remote Manipulator (SRMS). Spar is presently under contract to Ontario Hydro to design and build a Remote Manipulation Control System to replace the fuel channels in the Pickering A Nuclear Generating Station. The equipment may be used to replace the fuel channels in six other early generation CANDU reactors

  3. Technology Development of Integrity Evaluation of Fuel Bundles and Fuel Channel in a Two-phase Flow CANDU-6 Fuel Channel

    International Nuclear Information System (INIS)

    Two phase flow induces dynamic fluid force that causes structural vibration. Enormous vibration may result in failures of components due to the fretting wear and the fatigue, which increases the maintenance cost of the plant. From this consideration, KINS required that fuel bundles and fuel channels be evaluated to assure their integrities in high flow of more than 24 kg/s and two phase condition. Because out-reactor test loop for the simulation of two phase high flow is not available, the Wolsong CANDU-6 reactor which is in operation was utilized for the test. In-bay inspection system for the under water inspection and measurement of irradiated fuel was developed. 36 fresh fuels were measured prior to the irradiation and loaded in the fuel channel. Besides, improved method for early detection and evaluation of defect fuel was suggested

  4. Benchmark calculations of a radiation heat transfer for a CANDU fuel channel analysis using the CFD code

    International Nuclear Information System (INIS)

    To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with those solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer. (author)

  5. candu fuel bundle fabrication

    International Nuclear Information System (INIS)

    This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly

  6. Eddy current detection of spacers in the fuel channels of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Garter Spring (GS) spacers in the fuel channels of CANDU nuclear reactors maintain separation between the hot pressure tube and surrounding moderator cooled calandria tube. Eddy current detection of the four GSs provides assurance that spacers are at or close to design positions and are performing their intended function of maintaining a non-zero gap between pressure tube and calandria tube. Pressure tube constrictions, resulting from relatively less diametral creep at end-of-fuel bundle locations, also produce large eddy current signals. Large constrictions, present in higher service pressure tubes, can produce signals that are 10 times larger than GS signals, reducing GS detectability to 30% in standard GS-detect probes. The introduction of field-focussing elements into the design of the standard GS detection eddy current probe has been used to recover the detectability of GS spacers by increasing the signal amplitude obtained from GSs relative to that from constrictions by a factor of 10. The work presented here compares laboratory, modelling and in-reactor measurements of GS and constriction signals obtained from the standard probe with that obtained from field-focussed eddy current probe designs. (author)

  7. Candu fuel and fuel cycles

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous emissions are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. The technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy. The world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, CANDU reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuels which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the CANDU reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential CANDU fuel cycle developments can be accommodated in existing

  8. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without re-enrichment, the plutonium as conventional Mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  9. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without reenrichment, the plutonium as conventional mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  10. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  11. fuel management in candu reactors: RFSP code

    International Nuclear Information System (INIS)

    The objective of in-core fuel management is to determine the required refuelling strategies for safe and reliable operation of the reactor with minimum total energy cost. CANDU reactors use natural uranium fuel and rely on semi-continuous on-power refuelling. For the purpose of fuel management, the CANDU core with 380 fuel channels is modelled dividing into inner-and outer core. Refuelling rate in the CANDU reactors is evaluated in three periods for the whole operating life: 1)From the initial core to refuelling onset (100-150 EFPD), 2) the intermediate period (400-500 EFPD), and 3)the equilibrium period (approximately 30 years). A channel in the CANDU-6 reactor contains 12 bundles, in the refuelling operation some bundles do not discharged, but are shifted to other place in the same channel. One of the methods used for selection the channel and determination the bundles to be discharged is simulation method one of which is the RFSP (reactor fuelling simulating program). RFSP is a computer programme to do neutronic calculations for CANDU reactors. It can calculate both static and time-dependent neutron flux and power distributions in the core. It is a modular program containing a lot of modules. RFSP can perform fuel-management calculations and simulate a reactor operating history at specified intervals, taking burnup steps and channel refuelling into account

  12. Improved CANDU fuel performance

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU power reactors has been very low (0.06 percent) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness and wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

  13. SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mladin, M. [Reactor Physics and Nuclear Safety Department, Institute for Nuclear Research-Pitesti, P.O. Box 78, Campului No. 1, 115400 Mioveni, Arges (Romania)], E-mail: mirea_mladin@easynet.ro; Dupleac, D.; Prisecaru, I. [Power Engineering Department, University ' Politehnica' of Bucharest (Romania)

    2009-02-15

    Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR-BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions.

  14. Second International Conference on CANDU Fuel

    International Nuclear Information System (INIS)

    Thirty-four papers were presented at this conference in sessions dealing with international experience and programs relating to CANDU fuel; fuel manufacture; fuel behaviour; fuel handling, storage and disposal; and advanced CANDU fuel cycles. (L.L.)

  15. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor can 'burn' a wide range of fuels without modification to the reactor system, including natural uranium, slightly enriched uranium, mixed oxide and spent LWR fuels. The economic feasibility of the advanced fuel cycles requires consideration of their disposal costs. Preliminary cost analyses for the disposal of spent CANDU-SEU (Slightly Enriched Uranium) and CANDU-DUPIC (Direct Use of spent PWR fuel In CANDU) fuels have been performed and compared to the internationally published costs for the direct disposal of spent CANDU and LWR fuels. The analyses show significant economic advantages in the disposal costs of CANDU-SEU and CANDU-DUPIC fuels. (author)

  16. CANDU fuel bundle skin friction factor

    International Nuclear Information System (INIS)

    Single-phase, incompressible fluid flow skin friction factor correlations, primarily for CANDU 37-rod fuel bundles, were reviewed. The correlations originated from curve-fits to flow test data, mostly with new fuel bundles in new pressure tubes (flow tubes), without internal heating. Skin friction in tubes containing fuel bundles (noncircular flow geometry) was compared to that in equivalent diameter smooth circular tubes. At Reynolds numbers typical of normal flows in CANDU fuel channels, the skin friction in tubes containing bundles is 8 to 15% higher than in equivalent diameter smooth circular tubes. Since the correlations are based on scattered results from measurements, the skin friction with bundles may be even higher than indicated above. The information permits over- or under-prediction of the skin friction, or choosing an intermediate value of friction, with allowance for surface roughnesses, in thermal-hydraulic analyses of CANDU heat transport systems. (author) 9 refs., 2 figs

  17. Mechanistic modeling of bearing pad to pressure tube contact under localized high temperature conditions in a CANDU fuel channel

    International Nuclear Information System (INIS)

    During a postulated critical break LOCA (loss of coolant accident) in a CANDU reactor the coolant flow rates in the fuel channels of the flow pass of the reactor core downstream of the pipe break can rapidly reduce to very low values and remain very low for a period of tens of seconds following the break. Under the sustained low flow conditions, the fuel sheath (cladding) temperature in the affected channels rapidly increases and the coolant in the channels becomes significantly voided. The pressure tubes in the affected pass heat up under a combination of convection heat transfer from the low flows of superheated seam and thermal radiation heat transfer from the hot fuel. Additionally, hot spots may develop on the inner surface of pressure tubes at locations where the fuel bearing pads are in direct contact with the pressure tube. Localized thermal creep strain deformation at the hot spots is a potential pressure tube failure mechanism which could challenge fuel channel integrity. This paper evaluates the local thermal-mechanical deformation of a pressure tube in a CANDU reactor under critical break LOCA conditions tube using a coupled thermal-mechanical finite element COMSOL multi-physics model and investigates the conditions resulting in fuel channel failure due to localized contact between bearing pad and pressure. The mechanistic models are validated against data obtained from COG funded experiments performed at WRL (Whiteshell Research Laboratory). Multiphysics calculations are performed in which the heat transfer, thermal-mechanical and creep strain equations are solved, simultaneously. Heat conduction from bearing pads to the inner surface of the pressure tube is modeled with appropriate convective and radiation heat transfer boundary conditions. Thermal creep strain deformation of the Zr-2.5%Nb pressure tube is modeled using correlations derived from separate uniaxial tests that are reported in the literature. Contact conductance models based on

  18. CANDU: The fuel conserving reactor

    International Nuclear Information System (INIS)

    Because of their high neutron economy and unique design features, CANDU heavy water moderated reactors are the only established commercial reactors able to use directly low fissile content fuels such as natural uranium or uranium recovered from spent light water reactor fuel (RU). These features also help them to achieve the highest fuel utilization of all commercially available reactors, whether the fuel is based on natural uranium or mixed oxides of plutonium, uranium or thorium. As nuclear capacity growth increases demands on the world's finite uranium resources, AECL envisages near term use in CANDU reactors of a fuel incorporating RU and fuels containing thorium, with either plutonium or low enriched uranium (LEU) as the fissile 'driver' fuel. In the long term, AECL proposes the use of future 'Generation X' CANDU reactors with enhanced neutron economy to achieve a near-Self-Sufficient Equilibrium Thorium (SSET) fuel cycle. This CANDU SSET would have a conversion ratio of unity and be able to produce power indefinitely, with the need for little additional fissile material once equilibrium is reached (the amount of 233U needed in the fresh fuel is the same as is present in the discharged fuel, including processing losses.) This would also enable a CANDU-Fast Breeder Reactor (FBR) synergism that would allow each fuel-generating, though expensive, FBR to supply the initial fissile requirements of several less-expensive, CANDU SSET reactors operating on the thorium cycle. The closer the approach to an SSET that CANDUs can achieve, the higher the ratio of CANDUs to breeders in an economically optimized reactor fleet. CANDU reactors thereby become natural partners of both light water-cooled thermal reactors and fast breeder reactors, in both cases making optimum use of their spent fuel components and enhancing the overall sustainability of nuclear power. (authors)

  19. A general computing code devoted to the analysis of bending vibrations specific to the CANDU type fuel channel

    International Nuclear Information System (INIS)

    It is known that circulation of the coolant through the pressure tube of a CANDU type reactor initiates and maintains bending vibrations in: individual fuel elements, fuel cluster, cluster column and in the pressure tube. The driving forces are either aleatory, due to turbulent flow, or harmonical due to the pressure pulsations from the circulation pumps. The vibrations induced by laminar flow in case of excessive intensities may induce both a acceleration of the fretting wear phenomena in the fuel elements and pressure tubes and a premature aging of the latter. In these conditions an important problem in the cluster design is that of obtaining, based on knowledge of laminar flow frequency structure, the eigenfrequencies for the four categories of oscillatory systems mentioned above and thus to avoid by construction the resonance phenomenon or at least to diminish its impairing effects. An activity of comparative analysis in different fuel cluster types is underway at INR Pitesti, a special attention being of course directed toward their vibrational behavior. The paper presents a general computational code devoted to characterization of bending vibration for: individual fuel elements, fuel element cluster, pressure tube loaded or not with fuel clusters and filled or not with coolant; fuel channel. During the presentation of the work the computing code will be run for demonstration

  20. The CANDU 9 fuel transfer system

    International Nuclear Information System (INIS)

    The CANDU 9 fuel transfer system is based on the CANDU 6 and the Ontario Hydro Darlington NGD designs, modified to suit the CANDU 9 requirements. The CANDU 9 new fuel transfer system is very similar to the CANDU 6, with modifications to allow new fuel loading from outside containment, similar to Darlington. The CANDU 9 irradiated fuel transfer system is based on the Darlington irradiated fuel transfer system, with modifications to meet the more stringent containment requirements, improve performance, and match station layout. (author). 2 refs., 6 figs

  1. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 5500C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  2. Development of a CANDU fuel channel model to assess the effect of a pressure tube creep on the safety related parameters

    International Nuclear Information System (INIS)

    Recently the effect of pressure tube creep on the reactor safety in CANDUs emerges as an important issue of safety analysis due to a need for an extended operation. The accident analysis for the aged plants needs to incorporate major degradations of the plant performance in the safety analysis. In this paper, a CATHENA fuel channel model for studying the effects of the vertical offset of the fuel bundles in a crept pressure tube on the fuel and pressure tube cooling is developed. The current practice of the CANDU safety analysis assumes that the fuel bundles stay in a manner concentric to the pressure tube centerline even in the crept pressure tubes, whereas in reality the bundles sit at the bottom of the pressure tube. With this point in mind, 37-pin models with and without vertical offset of the bundle in the crept fuel channel are developed and tested for Reactor Outlet Header (ROH) 100% break LOCA accident, and results compared. As a result, it was found that the difference between the uncrept fuel channel model and the two crept fuel channel models, a concentric one and another vertically offset one, is quite significant, whereas the difference between the two crept fuel channel models is insignificant. Therefore it is concluded that the use of the concentric crept fuel channel model for the aged CANDU-6 safety analysis is justifiable for the first 200 sec into an accident. (author)

  3. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO2 fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly discuss our

  4. Fuel for CANDU pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Unique properties, performance and evolution of CANDU fuel are described. The manufacturing conditions, uranium requirements, and fuel costs are discussed. The in-service performance of the fuel has been excellent and defect mechanisms and operating criterion are described. Evolutionary improvements in CANDU fuel and new fuel cycles such as plutonium and thorium are being explored to insure that the CANDU reactor remains competitive in the future. (author)

  5. Thorium fuel-cycle studies for CANDU reactors

    International Nuclear Information System (INIS)

    The high neutron economy of the CANDU reactor, its ability to be refuelled while operating at full power, its fuel channel design, and its simple fuel bundle provide an evolutionary path for allowing full exploitation of the energy potential of thorium fuel cycles in existing reactors. AECL has done considerable work on many aspects of thorium fuel cycles, including fuel-cycle analysis, reactor physics measurements and analysis, fuel fabrication, irradiation and PIE studies, and waste management studies. Use of the thorium fuel cycle in CANDU reactors ensures long-term supplies of nuclear fuel, using a proven, reliable reactor technology. (author)

  6. Monitoring defective CANDU fuel bundles

    International Nuclear Information System (INIS)

    In 2005, it was proposed that a passive substance such as Nanocrystals could be used to monitor and locate defect fuel elements in-core. The experimental goal was to determine if Nanocrystals could be used for this application. Originally nanocrystals tagging was suggested for current operational CANDU-600 fuel. Other methods, including noble gas tagging, are also being investigated. Moreover, the scope of the project has been extended to include the identification of Dysprosium-doped fuel in the new ACR fuel design. The purpose of this paper is to discuss the experimental progress made at RMC on this project. (author)

  7. The development of a remote gauging and inspection capability for fuel channels in Candu reactors

    International Nuclear Information System (INIS)

    Equipment under development for the inspection and gauging of pressure tubes in CANDU (Canadian Deuterium Uranium) type reactors is described. A brief overview of the mechanical scanning system is presented followed by a detailed description of the measurement and data processing systems for the gauging of diameter and wall thickness, volumetric inspection of the tube wall and gauging of the annular gap between the pressure tube and the calandria tube. Experience of testing ultrasonic transducers in very high (106 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

  8. Synergistic CANDU-LWR fuel cycles

    International Nuclear Information System (INIS)

    CANDU is the most neutron-efficient reactor available commercially, allowing utilization of a range of fuel cycles. The flexibility of on-line refuelling allows fuel management to accommodate these different fuels. A synergism with light-water reactors (LWR) is possible through the use in CANDU of uranium and/or plutonium recovered from spent LWR fuel. In the TANDEM fuel cycle, the unseparated uranium and plutonium (1.5% fissile) would give a burnup in CANDU of about 25 MW.d/kg HE, producing four times more energy than that available from simply recycling the plutonium in an LWR. In another potential fuel cycle, uranium recovered from spent LWR fuel during conventional reprocessing is also recycled in CANDU, without re-enrichment. An average recovered uranium (RU) enrichment of 0.9% in U-235 results in a CANDU burnup of at least 13 MW.d/kg U, allowing twice as much energy to be extracted, compared with that from an LWR. The fuelling cost for RU in CANDU are about 35% lower than for natural uranium. Additionally, direct use of spent LWR fuel in CANDU is theoretically possible, but requires practical demonstration. AECL and KAERI are developing the CANFLEX (CANDU Flexible Fuelling) advanced fuel bundle as the optimal carrier for all extended burnup fuel cycles envisaged for CANDU

  9. A study of intermittent buoyancy induced flow phenomena in CANDU fuel channels

    International Nuclear Information System (INIS)

    The present work focuses on two-phase flow behavior called 'Intermittent Buoyancy Induced Flow' (IBIF) resulting from the loss of coolant circulation in a CANDU reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. This paper will describe the experiments conducted with a set up simulating a pressure tube in a CANDU reactor. The test section was a 9.0-m long, 101.6-mm ID horizontal acrylic tube. Two vertical tubes simulating the feeder pipes in the reactor were attached at the end sections of the horizontal tube. Each vertical tube was connected to an open top cylindrical water tank simulating the header tank. Experiments were conducted using air and water at atmospheric pressure to qualitatively examine the IBIF phenomena. Air bubbles injected into the horizontal tube were observed to rapidly rise towards the top forming a continuous layer rather than flowing as discrete bubbles. The front of the air-water interface then moved horizontally towards the end sections at constant speeds and vented into the feeder tubes forming a slug flow. At low air injection rates an oscillating periodic behavior in the void fraction was observed in the two feeder tubes as the air vented alternately through each of these tubes. Small sagging of the pressure tube in the middle induced the injected air to move towards the end faster reducing the venting time. (author)

  10. Thorium fuel cycles in CANDU

    International Nuclear Information System (INIS)

    In recent years, Atomic Energy of Canada Limited has been examining in detail the implications of using thorium-based fuels tn the CANDU reactor. Various cycles initiated and enriched either with fissile plutonium or with enriched uranium, and with effective conversion ratios ranging up to 1.0, have been evaluated. We have concluded that: 1. Substantial quantities of uranium can be saved by adoption of the thorium fuel cycle, and the long-term security of fissile supply both for the domestic and overseas market can be considerably enhanced. The amount saved will depend on the details of the fuel cycle and the anticipated growth of nuclear power in Canada. 2. The fuel cycle can be introduced into the basic CANDU design without major modifications and without compromising current safety standards. 3. The economic conditions that make thorium competitive with the once-through natural uranium cycle depend a the price of uranium and on the costs both to fabricate α and γ-emitting fuels and to either enrich uranium or to extract fissile material from spent fuel. While timing is difficult to predict, we believe that competitive economic conditions will prevail toward the end of this century. 4. A twenty-year technological development program will be required to establish commercial confidence in the fuel cycle. (author)

  11. Advanced fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The current natural uranium-fuelled CANDU system is a world leader, both in terms of overall performance and uranium utilization. Moreover, the CANDU reactor is capable of using many different advanced fuel cycles, with improved uranium utilization relative to the natural uranium one-through cycle. This versatility would enable CANDU to maintain its competitive edge in uranium utilization as improvements are made by the competition. Several CANDU fuel cycles are symbiotic with LWRs, providing an economical vehicle for the recycle of uranium and/or plutonium from discharges LWR fuel. The slightly enriched uranium (SEU) fuel cycle is economically attractive now, and this economic benefit will increase with anticipated increases in the cost of natural uranium, and decreases in the cost of fuel enrichment. The CANFLEX fuel bundle, an advanced 43-element design, will ensure that the full benefits of SEU, and other advanced fuel cycles, can be achieved in the CANDU reactor. 25 refs

  12. CANDU fuel compression tests at elevated temperatures

    International Nuclear Information System (INIS)

    An inlet header large break loss of coolant accident (LOCA) in CANDU reactors with fuelling against flow can cause the fuel to shift in the channels with a consequent reactivity insertion. This results in an increased fuel power transient, and a potential increase in the mialyzed consequences for such events. As the reactor's age and the channel axial gaps increase, the magnitude of the predicted power u-dmient increases. A design solution to reduce the power transient is to limit the amount of fuel movement by reducing the channel axial gap. This solution was implemented into Ontario Hydro's Bruce B and Darlington reactors. A consequence of a reduced channel axial gap is the potential for the fuel column axial expansion to become constrained by the channel end components in large break LOCAs. This experimental program investigated the effects of pellet cracking and elevated sheath temperatures on the ability of the fuel elements, of the 37-element bundle design, to sustain axial loads. The unirradiated fuel elements tested were either in the as-received condition or with the U02 fuel pellets cracked in a mechanical process to simulate the effect of inufflation. The load deformation characteristics demonstrated that, for a given amount of axial compression. the loads sustainable by the elements at elevated sheath temperatures were low. As a result. excess axial expansion would be easily accommodated without further challenge to pressure tube integrity. (author)

  13. Korea's CANDU fuel R and D program

    International Nuclear Information System (INIS)

    As the first R and D activity led to the nuclear fuel industrialization in Korea, KAERI had successfully developed the CANDU-6 fuel bundle in the period of 1981 to 1986 and has commercially produced more than 35,000 fuel bundles for the use in Wolsong Unit 1 since 1987. The commercial production of the CANDU-6 fuel in KAERI will be terminated on the end of 1997 and KNFC will take over the mission of CANDU-6 fuel production with a capacity of 400 tons of uranium per year form 1998. (author)

  14. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  15. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

  16. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  17. Fuel management simulation for CANFLEX-RU in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

  18. Container storage of CANDU fuel

    International Nuclear Information System (INIS)

    Each Ontario Hydro operating plant has bays or water pools large enough to handle about a decade's used fuel. For the long term, however, dry storage is a practical alternative, especially for CANDU fuel with its low decay heat. The economic attractiveness of dry storage has long been recognized. Prospects improve if the same container can be used for transporting the fuel to its final destination without any need for repackaging. There are many engineering, scientific and licensing problems to be dealt with, but Ontario Hydro has made a start with its Dry Storage Demonstration Program. The long term goal of this program is to develop a container that can be used for storage, transportation, and disposal of used fuel. This paper describes the stages in the development of Ontario Hydro's approach to dry storage. Efforts have converged on the concept of a Concrete Integrated Container. This is a self-contained, high-integrity structure designed to be loaded with fuel directly from the station pools, moved to a storage area and eventually transported off-site to a final disposal facility. 5 refs., 4 tabs., 8 figs

  19. CANDU fuel performance and development

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU (Canada Deuterium Uranium) reactors continues to be very low, 0.06% since 1972. The power ramp defects, which constituted the majority of the early defects, have been virtually eliminated by changed fuelling schemes and through the introduction of graphite CANLUB coatings on the inside of the sheath. Laboratory and loop irradiations have demonstrated that the graphite CANLUB layers increase the tolerance to power ramps, but to obtain the maximum benefit, coating parameters such as thickness, adhesion and wear resistance must be optimized. Siloxane CANLUB coated fuel offers greater tolerance to power ramps than most graphite coatings; quality control appears simpler and no instance of localized sheath hydriding has been seen with cured and irradiated coatings. Limited testing has shown that fuel with graphite discs between fuel pellets also has high tolerance to power ramps, but it is more costly and has lower burnup. The number of defects due to faulty components has been extremely small (0.00014%), but improved quality control and welding procedures can lower this number even further. Defects from causes external to the bundle have also been very few. (author)

  20. CANFLEX - an advanced fuel bundle for CANDU

    International Nuclear Information System (INIS)

    The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration

  1. Canadian CANDU fuel development programs and recent fuel operating experience

    International Nuclear Information System (INIS)

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements! This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as longer-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  2. Canadian CANDU fuel development programs and recent fuel operating experience

    International Nuclear Information System (INIS)

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements. This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as long-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  3. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

    CANDU reactor technology was originally developed in Canada as part of the original introduction of peaceful nuclear power in the 1960s and has been continuously evolving and improving ever since. The CANDU reactor system was defined with a requirement to be able to efficiently use natural uranium (NU) without the need for enrichment. This led to the adaptation of the pressure tube approach with heavy water coolant and moderator together with on-power fuelling, all of which contribute to excellent neutron efficiency. Since the beginning, CANDU reactors have used [NU] fuel as the fundamental basis of the design. The standard [NU] fuel bundle for CANDU is a very simple design and the simplicity of the fuel design adds to the cost effectiveness of CANDU fuelling because the fuel is relatively straightforward to manufacture and use. These characteristics -- excellent neutron efficiency and simple, readily-manufactured fuel -- together lead to the unique adaptability of CANDU to alternate fuel types, and advancements in fuel cycles. Europe has been an early pioneer in nuclear power; and over the years has accumulated various waste products from reactor fuelling and fuel reprocessing, all being stored safely but which with passing time and ever increasing stockpiles will become issues for both governments and utilities. Several European countries have also pioneered in fuel reprocessing and recycling (UK, France, Russia) in what can be viewed as a good neighbor policy to make most efficient use of fuel. The fact remains that CANDU is the most fuel efficient thermal reactor available today [NU] more efficient in MW per ton of U compared to LWR's and these same features of CANDU (on-power fuelling, D2O, etc) also enable flexibility to adapt to other fuel cycles, particularly recycling. Many years of research (including at ICN Pitesti) have shown CANDU capability: best at Thorium utilization; can use RU without re-enrichment; can readily use MOX. Our premise is that

  4. A study on CANDU model assessment of RELAP5/CANDU using RD-14M B9401 multi-channel RIH break experiment

    International Nuclear Information System (INIS)

    B9401 experiment, performed in RD-14M[1] multi-channel facility, was analyzed using RELAP5/MOD3 and RELAP5/CANDU and compared with experiment results. The RELAP5/CANDU code has been developed since 1998, based on RELAP5, in order to have auditing tool of CANDU NPP. The RELAP5/CANDU code is under developing and they have not been assessed much for a CANDU reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14M facility and to get the assessment results for RELAP5/CANDU code. The RD-14M test facility at Whiteshell Nuclear Research Establishment is a full-scale multi-channel pressurized-water loop. The RELAP5/MOD3 and RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the main phenomena occurred during the transient, in qualitative view. In quantitative view, the RELAP5/CANDU[4] predicted better than that of RELAP5. In the case of experiment that the stratification in fuel channel is dominant, it is expected that RELAP5/CANDU can give more accurate result than RELAP5

  5. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  6. CANDU improvement

    International Nuclear Information System (INIS)

    The evolution of the CANDU family of nuclear power plants is based on a continuous product development approach. Proven equipment and system concepts from operating stations are standardized and used in new products. Due to the modular nature of the CANDU reactor concept, product features developed for CANDU 9 can easily be incorporated in other CANDU products such as CANDU 6. Design concepts are being developed for advanced CANDU 6 or larger advanced CANDU, depending on the number of fuel channels and the fuel cycle selected. This paper provides a description of the design improvements being incorporated in CANDU 9 and further design enhancements being studied for future incorporation in CANDU 6 or larger advanced CANDU meeting the requirements of future CANDU owners. The design enhancement objectives are: To improve operational simplicity by applying modern information technology; to improve safety in a cost effective way; to improve system and component reliability and to increase plant life; to improve economics and to reduce owners' risks during all phases of a project using up-front licensing, an improved engineering process and project tools during design, construction and operation; to continue to exploit the neutron economy of CANDU with the development of advanced fuels and fuel cycles. (author)

  7. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  8. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  9. Effect of Candu Fuel Bundle Modeling on Sever Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I. [Power Plant Engineering Faculty, Politehnica University, 313 Splaiul Independentei, 060042, sect. 6, Bucharest (Romania); Mladin, M. [Institute for Nuclear Research, Pitesti-Mioveni, 115400 (Romania)

    2009-06-15

    of the project is the assessment and adaptation of the SCDAP/RELAP5 code to Candu 6 severe accidents analysis. The SCDAP/RELAP5 code is a detailed mechanistic code, originated from the merging of the SCDAP and the RELAP5 code, each of them focussing on a specific part of severe accident phenomenology. The SCDAP code models the core behavior during a severe accident. Treatment of the core includes fuel rod heatup, ballooning and rupture, fission product release, rapid oxidation, zircaloy melting, UO{sub 2} dissolution, ZrO{sub 2} breach, flow and freezing of molten fuel and cladding, and debris formation and behavior. The code also models control rod and flow shroud behavior. The standard Candu fuel bundle consists of 37 identical base elements, arranged in circular rings, with 1, 6, 12 and 18 respectively associated rods. Fuel elements associated to different rings have different power peaking factors. Although no specific model for Candu fuel bundle is provided in SCDAP code, starting from fuel rod component Candu fuel bundle models of different complexity can be build. Three models were considered in this study. The first, the simplest one, consist in assuming that all fuel elements of a bundle behave in the same manner, having the same power. Thus, in this representation, the Candu fuel bundle is modeled by one SCDAP fuel rod component. The main advantage of this model is the increased speed of the simulation runs. The second model, consider four types of SCDAP fuel rod components to model the Candu fuel bundle. Thus, the fuel elements of different rings are modeled separately with their actual power peaking factors. However, in this model, the fuel channel is modeled as a single pipe component; consequently all the SCDAP fuel rod components have the same thermal hydraulic boundary conditions. The third approach, consider the same model for the Candu fuel bundle as in the second model but the fuel channel is modeled as four pipe components linked by cross

  10. Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

    International Nuclear Information System (INIS)

    In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

  11. CANDU spent fuel dry storage interim technique

    International Nuclear Information System (INIS)

    CANDU heavy water reactor is developed by Atomic Energy of Canada (AECL) it has 40 years of design life. During operation, the reactor can discharge a lot of spent fuels by using natural uranium. The spent fuel interim storage should be considered because the spent fuel bay storage capacity is limited with 6 years inventory. Spent fuel wet interim storage technique was adopted by AECL before 1970s, but it is diseconomy and produced extra radiation waste. So based on CANDU smaller fuel bundle dimension, lighter weight, lower burn-up and no-critical risk, AECL developed spent fuel dry interim storage technique which was applied in many CANDU reactors. Spent fuel dry interim storage facility should be designed base on critical accident prevention, decay heat removal, radiation protection and fissionable material containment. According to this introduction, analysis spent fuel dry interim storage facility and equipment design feature, it can be concluded that spent fuel dry interim storage could be met with the design requirement. (author)

  12. CANDU: Shortest path to advanced fuel cycles

    International Nuclear Information System (INIS)

    Full text: The global nuclear renaissance exhibiting itself in the form of new reactor build programs is rapidly gaining momentum. Many countries are seeking to expand the use of economical and carbon-free nuclear energy to meet growing electricity demand and manage global climate change challenges. Nuclear power construction programs that are being proposed in many countries will dramatically increase the demand on uranium resources. The projected life-long uranium consumption rates for these reactors will surpass confirmed uranium reserves. Therefore, securing sufficient uranium resources and taking corresponding measures to ensure the availability of long-term and stable fuel resources for these nuclear power plants is a fundamental requirement for business success. Increasing the utilization of existing uranium fuel resources and implementing the use of alternate fuels in CANDU reactors is an important element to meet this challenge. The CANDU heavy water reactor has unequalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and thorium. This CANDU feature has not been used to date simply due to lack of commercial drivers. The capability is anchored around a versatile pressure tube design, simple fuel bundle, on-power refuelling, and high neutron economy of the CANDU concept. Atomic Energy of Canada Limited (AECL) has carried out theoretical and experimental investigations on various advanced fuel cycles, including thorium, over many years. Two fuels are selected as the subject of this paper: Natural Uranium Equivalent (NUE) and thorium. NUE fuel is developed by combining RU and depleted uranium (DU) in such a manner that the resulting NUE fuel is neutronically equivalent to NU fuel. RU is recovered from reprocessed light water reactor (LWR) fuel and has a nominal 235U concentration of approximately 0.9 wt%. This concentration is higher than NU used in CANDU reactors

  13. Temperature effect of DUPIC fuel in CANDU reactor

    International Nuclear Information System (INIS)

    The fuel temperature coefficient (FTC) of DUPIC fuel was calculated by WIMS-AECL with ENDF/B-V cross-section library. Compared to natural uranium CANDU fuel, the FTC of DUPIC fuel is less negative when fresh and is positive after 10,000 MWD/T of irradiation. The effect of FTC on the DUPIC core performance was analyzed using the pace-time kinetics module in RFSP for the refueling transient which occurs daily during normal operation of CANDU reactors. In this study, the motion of zoen controller units (ZCU) was modeled externally to describe the reactivity control during the refueling transient. Refueling operation was modeled as a linear function of time by changing the fuel burnup incrementally and the average fuel temperature was calculated based on the bundle power during the transient. The analysis showed that the core-wide FTC is negative and local positive FTC of the DUPIC fuel can be accommodated in the CANDU reactor because the FTC is very small, the refueling operation occurs slowly, and the channel-front-peaked axial power profile weakens the contribution of the positive FTC. (author). 11 refs., 31 tabs., 10 figs

  14. Feasibility study of CANDU-9 fuel handling system

    International Nuclear Information System (INIS)

    CANDU's combination of natural uranium, heavy water and on-power refuelling is unique in its design and remarkable for reliable power production. In order to offer more output, better site utilization, shorter construction time, improved station layout, safety enhancements and better control panel layout, CANDU-9 is now under development with design improvement added to all proven CANDU advantages or applicable technologies. One of its major improvement has been applied to fuel handling system. This system is very similar to that of CANDU-3, and some parts of the system are applied to those of the existing CANDU-6 or CANDU-9. Design concepts and design requirements of fuel handling system for CANDU-9 have been identified to compare with those of the existing CANDU and the design feasibilities have been evaluated. (author). 4 tabs., 13 figs., 9 refs

  15. Flow visualization study of two-phase flow in the horizontal annulus of the fuel-channel outlet end-fitting of a CANDU reactor

    International Nuclear Information System (INIS)

    In CANDU-6 reactors, the pressurized hightemperature coolant flows through 380 fuel channels passing horizontally through the core. In 1996, higher than expected rates of wall thinning of the outlet feeders were ascribed to flow-accelerated corrosion (FAC). Such corrosion is strongly influenced by the hydrodynamics of the coolant. Results of preliminary flow visualization and modelling studies have suggested that flow conditions in the end-fitting annulus upstream of the outlet feeder may influence the pattern of FAC. For a full-scale flow visualization, an acrylic test section was built to simulate the cylindrical end-fitting with its annulus flow path. The tests were performed with water and air at atmospheric pressure and room temperature. The phase distribution along the length of the annulus was recorded with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Significant effects on the flow patterns of spacer buttons in the annulus were observed. A commercial computational fluid dynamics (CFD) code-Fluent 6.1-was used to model the results. (authors)

  16. CANDU fuel : safe, reliable and flexible. 12th international conference on CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-07-01

    The Canadian Nuclear Society's 12th International Conference on CANDU Fuel was hosted by Royal Military College of Canada in Kingston, Ontario, Canada on September 15-18, 2013. The theme for the conference was 'CANDU Fuel : Safe, Reliable & Flexible' bringing together international experts of the nuclear fuel industry and academia involved in design, R and D, manufacturing, operation, modeling, safety analysis, and regulations. Over 100 delegates including representatives from other countries, including India, Romania, Argentina, Korea, United States, Austria, and Canada attended this truly successful international event. Although CANDU fuel has performed well a number of the presentations were on a modified design of the standard 37 element bundle called 37M which is now being loaded into the Darlington reactors. The renewed interest in thorium was also the focus of several presentations.

  17. Next generation CANDU plants

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water Reactors systems featuring horizontal fuel channels and heavy water moderator will continue to evolve, supported by AECL's strong commitment to comprehensive R and D programs. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety operation based on design feedback. Therefore, CANDU reactor products will continue to evolve by incorporating further improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. Progressive CANDU development will continue in AECL to enhance the medium size product - CANDU 6, and to evolve the larger size product - CANDU 9. The development of features for CANDU 6 and CANDU 9 is carried out in parallel. Developments completed for one reactor size can then be applied to the other design with minimum costs and risk. (author)

  18. Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

    International Nuclear Information System (INIS)

    The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO2. The model was initially tested and the average discharge burnup for natural UO2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

  19. CANDU fuel behaviour under transient conditions

    International Nuclear Information System (INIS)

    The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

  20. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    International Nuclear Information System (INIS)

    This is final report of the CANDU advanced fuel (CANFLEX fuel) verification test project. This report describes performance verification tests performed for the development of the CANFLEX-NU bundle. The test items described in the report are as follows. - Fuel channel pressure drop test, -Fuel strength tests, - Fuel impact test, - Fuel endurance test (vibration test), - Compatibility test with fueling machine, - Critical heat flux test. 58 tabs., 60 figs., 32 refs. (Author)

  1. Methodology Improvement of Reactor Physics Codes for CANDU Channels Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun; Choi, Geun Suk; Win, Naing; Aung, Tharndaing; Baek, Min Ho; Lim, Jae Yong [Kyunghee University, Seoul (Korea, Republic of)

    2010-04-15

    As the operational time increase, pressure tubes and calandria tubes in CANDU core encounter inevitably a geometrical deformation along the tube length. A pressure tube may be sagged downward within a calandria tube by creep from irradiation. This event can bring about a problem that is serious in integrity of pressure tube. A measurement of deflection state of in-service pressure tube is, therefore, very important for the safety of CANDU reactor. In this paper, evaluation of impacts on nuclear characteristic due to fuel channel deformation were aimed in order to improve nuclear design tools for concerning the local effects from abnormal deformations. It was known that sagged pressure tube can cause the eccentric configuration of fuel bundles in pressure tube by O.6cm maximum. In this case, adverse pin power distribution and reactivity balance can affect reactor safety under normal and accidental condition. Thermal and radiation-induced creep in pressure tube would expand a tube size. It was known that maximum expansion may be 5% in volume. In this case, more coolant make more moderation in the deformed channel resulting in the increase of reactivity. Sagging of pressure tube did not cause considerable change in K-inf values. However, expansion of the pressure tube made relatively large change in K-inf. Modeling of eccentric and enlarged configuration is not easy in preparation of input geometry at both HELlOS and MCNP. On the other hand, there is no way to consider this deformation in one-dimensional homogenization tool such as WIMS code. The way of handling this deformation was suggested as the correction method of expansion effect by adjusting the number density of coolant. The number density of heavy water coolant was set to be increased as the rate of expansion increase. This correction was done in the intact channel without changing geometry. It was found that this correction was very effective in the prediction of K-inf values. In this study, further

  2. Performance testing of CANDU MOX fuel

    International Nuclear Information System (INIS)

    CANDU fuel bundles containing 0.5 wt % plutonium in natural uranium were fabricated at Chalk River Laboratories and were successfully irradiated in the NRU reactor at powers up to 65 Min and to burnups ranging from 13 to 23 MW·d/kg HE. Two of the bundles experienced power histories that bound the normal powers and burnups of natural UO2 CANDU fuel (2 fuel. Significantly more grain growth was observed than that typically expected for UO2 fuel; however, this increase in grain growth had no apparent effect on the overall performance of the fuel. Pellet-centre columnar grain growth was accompanied by plutonium homogenization. Two other MOX bundles operated to extended burnups of 19 to 23 MW·d/kg HE. Burnup extension above 15 MW·d/kg HE had no apparent effect on sheath strain or grain growth, and only a small effect on FGR and the amount of oxide observed on the inner surface of the sheath. (author)

  3. Performance testing of CANDU MOX fuel

    International Nuclear Information System (INIS)

    CANDU fuel bundles containing 0.5 wt % plutonium in natural uranium were fabricated at Chalk River Laboratories and were successfully irradiated in the NRU reactor at powers up to 65 kW/m and to burnups ranging from 13 to 23 MW·d/kg HE. Two of the bundles experienced power histories that bound the normal powers and burnups of natural UO2 CANDU fuel (2 fuel. Significantly more grain growth was observed than that typically expected for UO2 fuel; however, this increase in grain growth had no apparent effect on the overall performance of the fuel. Pellet-centre columnar grain growth was accompanied by plutonium homogenization. Two other MOX bundles operated to extended burnups of 19 to 23 MW·d/kg HE. Burnup extension above 15 MW·d/kg HE had no apparent effect on sheath strain or grain growth, and only a small effect on FGR and the amount of oxide observed on the inner surface of the sheath. (author)

  4. Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

    International Nuclear Information System (INIS)

    The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO2-SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

  5. Behaviour of CANDU fuel during LOCA

    International Nuclear Information System (INIS)

    A large break loss-of-coolant accident (LOCA) in a CANDU nuclear reactor would result in a rapid increase of fuel and sheath temperatures. The temperature increase would, in turn, increase the gas pressure within the fuel and reduce the strength of the sheath material. Outside the fuel the loss of coolant from the primary heat transport system decreases pressure. The resulting pressure difference would cause deformation of the hot fuel sheath. Under certain circumstances, the deformation could be severe enough to fail the sheath thus releasing the fission products to the primary heat transport system. The computer code ELOCA-A is used to model the transient fuel behaviour following such an accident. ELOCA-A is a modified version of ELOCA.Mk 2 enabling us to consider the effects of axial variations in the microstructure of the sheath material caused by brazing of appendages to the sheath. The ELOCA-A code also features modelling of axial variations in neutron flux, pellet heat generation rate and heat transfer to the coolant. It predicts fuel pellet and sheath temperatures, sheath oxidation, sheath strain and probability of beryllium assisted cracking. A loss-of-coolant accident (LOCA) experiment was jointly sponsored by AECL and Ontario Hydro in the Power Burst Facility (PBF) at Idaho National Engineering Laboratories (INEL). This test was undertaken to provide an all-effects verification of the understanding of CANDU fuel behaviour during LOCA's. An extensive out-pile experimental program had provided single effects data which had been used for modelling such excursions. Integrated out-pile tests have confirmed our understanding of and accurate modelling of fuel under LOCA conditions. The integrated test in PBF provided the final proof that our understanding was complete and provided an experimental database for verification of transient fuel codes (4). The experiment was performed with modified CANDU fuel elements. The post-test measurements are compared with

  6. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  7. Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

  8. The thermalhydraulic behaviour of a CANDU channel during a channel flow blockage accident

    International Nuclear Information System (INIS)

    The aim of this paper is to perform an analysis of the thermal-hydraulic behavior of CANDU channel during a channel flow blockage accident and to present the final results. This event assumes a flow blockage in one of the reactor fuel channels, that leads to a reduction of the flow in the affected channel, and consequently to fuel cladding and fuel temperature increase. The main conclusions of this analysis are as follows: - For the complete range of flow blockage the effective shutdown is achieved and a good cooling is maintained in the unblocked channels of both primary circuit loops; - Behaviour of the blocked channel depends on the size of the blockage; - For the blockage which are not large enough to cause coolant superheating, the channel remains intact and there are no release to the containment; - Coolant superheating occurs only if the channel flow is reduced to less than 20 % of the normal flow; - Cooling in the intact loop is maintained by pumped circulation and heat removal by steam generators; - Thermal-hydraulic results are used in containment, fuel and fuel channel analysis. (authors)

  9. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  10. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  11. An analytical assessment of the longitudinal ridging of CANDU type fuel element

    International Nuclear Information System (INIS)

    There are 380 fuel channels in a CANDU-6 reactor, and twelve fuel bundles are loaded into each fuel channel. High-pressure, heavy water coolant passes through the fuel bundle string to remove heat generated from the fuel. Fuel sheath collapses down around the uranium dioxide pellet due to the coolant pressure when the fuel is loaded into the reactor. Longitudinal ridges may form in CANDU fuel element sheaths as a result of sheath collapse onto the pellets. A static analysis, finite-element (FE) model was developed to simulate the longitudinal ridging of the fuel element with use of the structural analysis computer code ABAQUS. Collapse pressures were calculated for the fifty-one cases for which test results of WCL in 1973 and 1975 are available. Calculation results under-predicted the critical collapse pressure but it showed significant relationship against test results

  12. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

  13. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

    The R and D program for CANDU advanced fuel cycles may be roughly divided into two components which have a near-and long-term focus, respectively. The near-term focus is on the technology to implement improved once-through cycles and mixed oxide (plutonium-uranium oxides) recycle in CANDU and on technologies to separate zirconium isotopes. Included is work on those technologies which would allow a CANDU-LWR strategy to be developed in a growing nuclear power system. For the longer-term, activities are focused on those technologies and fuel cycles which would be appropriate in a period when nuclear fuel demand significantly exceeds mined uranium supplies. Fuel cycles and systems under study are thorium recycle, CANDU fast breeder systems and electro-nuclear fissile breeders. The paper will discuss the rationale underlying these activities, together with a brief description of activities currently under way in each of the fuel cycle technology areas

  14. Overview of activities on CANDU fuel in Argentina

    International Nuclear Information System (INIS)

    This paper gives an outline of activities on CANDU fuel in Argentina. It discusses the nuclear activities and electricity production in Argentina, evolution of the activities in fuel engineering, fuel fabrication, fuel performance at Embalse nuclear power plant and spent fuel storage options.

  15. The application of the goal programming to CANDU fuel management optimization

    International Nuclear Information System (INIS)

    A Goal Programming formulation to CANDU fuel management optimization is proposed. Four objectives are considered, respectively : the feed rate, CPPF (Channel Power Peaking Factor), BPPF (Bundle Power Peaking Factor) and CPR (Critical Power Ratio). This problem is investigated using a numerical approach to optimization established on STepMethod and the use of loss matrix. The optimization technique developed is more adequate for fuel management analysis for fissile enriched fuel cycles, in which cases the relative importance of the objectives could be modified. Numerical results are presented for 0.93% SEU fuelled CANDU 6Mk1 core and weapons-grade plutonium burning in CANDU 6Mk1 core, using standard 37 rod fuel bundle. (author). 12 refs., 3 tabs., 7 figs

  16. Ninth international conference on CANDU fuel, 'fuelling a clean future'

    International Nuclear Information System (INIS)

    The Canadian Nuclear Society's 9th International Conference on CANDU fuel took place in Belleville, Ontario on September 18-21, 2005. The theme for this year's conference was 'Fuelling a Clean Future' bringing together over 80 delegates ranging from: designers, engineers, manufacturers, researchers, modellers, safety specialists and managers to share the wealth of their knowledge and experience. This international event took place at an important turning point of the CANDU technology when new fuel design is being developed for commercial application, the Advanced CANDU Reactor is being considered for projects and nuclear power is enjoying a renaissance as the source energy for our future. Most of the conference was devoted to the presentation of technical papers in four parallel sessions. The topics of these sessions were: Design and Development; Fuel Safety; Fuel Modelling; Fuel Performance; Fuel Manufacturing; Fuel Management; Thermalhydraulics; and, Spent Fuel Management and Criticalty

  17. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

    These proceedings contain full texts of all paper presented at the first International Conference on CANDU Fuel. The Conference was organized and hosted by the Chalk River Branch of the Canadian Nuclear Society and utilized Atomic Energy of Canada Limited's facilities at Chalk River Nuclear Laboratories. Previously, informal Fuel Information Meetings were used in Canada to allow the exchange of information and technology associated with CANDU. The Chalk River conference was the first open international forum devoted solely to CANDU and included representatives of overseas countries with current or potential CANDU programs, as well as Canadian participants. The keynote presentation was given by Dr. J.B. Slater, who noted the correlation between past successes in CANDU fuel cycle technology and the co-operation between researchers, fabricators and reactor owner/operators in all phases of the fuel cycle, and outlined the challenges facing the industry today. In the banquet address, Dr. R.E. Green described the newly restructured AECL Research Company and its mission which blends traditional R and D with commercial initiatives. Since this forum for fuel technology has proven to be valuable, a second International CANDU Fuel Conference is planned for the fall of 1989, again sponsored by the Canadian Nuclear Society

  18. Study of the end flux peaking for the Candu fuel bundle types by transport methods

    International Nuclear Information System (INIS)

    The region separating the Candu fuel in two adjoining bundles in a channel is called the end region. The end of the last pellet in the fuel stack adjacent to the end region is called the fuel end. In the end region of the bundle the thermal neutron flux is higher than at the axial mid-point, because the end region of the bundle is made up of very low neutron absorption material: coolant and Zircaloy-4. For accurate evaluation of fuel performance, it is important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle, including the end region. The work reported here had two objectives. First, calculation of the flux distributions (axial and radial) and the end flux peaking factors for some Candu fuel bundles. Second objective is a comparative analysis of the obtained results. The Candu fuel bundles considered in this paper are NU37 (Natural Uranium, 37 elements) and SEU43 (Slightly Enriched Uranium, 43 elements, with 1.1wt% enrichment). For realization of the proposed objectives, a methodology based on WIMS, PIJXYZ and LEGENTR codes is used in this paper. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the flux distribution in Candu fuel bundles it is used PIJXYZ and LEGENTR respectively codes. These codes are consistent with WIMS lattice-cell calculations and allow a good geometrical representation of the Candu bundle in three dimensions. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for Candu cell geometry. LEGENTR is a 3D SN transport code based on projectors technique and can be used for 3D cell and 3D core calculations. (author)

  19. Large scale replacement of fuel channels in the Pickering CANDU reactor using a man-in-the-loop remote control system

    International Nuclear Information System (INIS)

    Spar Aerospace Limited of Toronto is presently under contract to Ontario Hydro to design a Remote Manipulation and Control System (RMCS) to be used during the large scale replacement of the fuel channels in the Pickering A Nuclear Generating Station. The system is designed to support the replacement of all 390 fuel channels in each of the four reactors at the Pickering A station in a safe manner that minimizes worker radiation exposure and unit outage time

  20. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  1. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  2. INR Recent Contributions to Thorium-Based Fuel Using in CANDU Reactors

    International Nuclear Information System (INIS)

    The paper summarizes INR Pitesti contributions and latest developments to the Thorium-based fuel (TF) using in present CANDU nuclear reactors. Earlier studies performed in INR Pitesti revealed the CANDU design potential to use Recovered Uranium (RU) and Slightly Enriched Uranium (SEU) as alternative fuels in PHWRs. In this paper, we performed both lattice and CANDU core calculations using TF, revealing the main neutron physics parameters of interest: k-infinity, coolant void reactivity (CVR), channel and bundle power distributions over a CANDU 6 reactor core similar to that of Cernavoda, Unit 1. We modelled the so called Once Through Thorium (OTT) fuel cycle, using the 3D finite-differences DIREN code, developed in INR. The INR flexible SEU-43 bundle design was the candidate for TF carrying. Preliminary analysis regarding TF burning in CANDU reactors has been performed using the finite differences 3D code DIREN. TFs showed safety features improvement regarding lower CVRs in the case of fresh fuel use. Improvements added to the INR ELESIMTORIU- 1 computer code give the possibility to fairly simulate irradiation experiments in INR TRIGA research reactor. Efforts are still needed in order to get better accuracy and agreement of simulations to the experimental results. (author)

  3. Economic and system aspects of CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    It is somewhat a paradox that Canada, which ranks as one of the world's leading uranium producers and has large economic uranium resources, should also have developed the CANDU reactor. This reactor is the most fuel efficient of all reactor types which are commercially available at the present time. The explanation of the paradox is that the design basis of the CANDU was established three decades ago when the full extent of Canadian uranium resources was unknown, and an early transition to recycle fuelling was anticipated as being necessary to sustain a growing power generation system. Consequently, the objectives of fuel efficiency and flexibility in using a variety of uranium, plutonium and thorium fuels were established at an early stage. One result of this is the ability to use the current design of CANDU in an advanced converter role with very little change in reactor design or operating procedures. As a result, in projections of future power costs, all major uncertainty is focused on fuel cycle parameters since the capital and operating costs are well defined by current commercial experience. The paper will examine the economic and resource characteristics of CANDU in an advanced converter role, both in terms of stand-alone technology and as a partner in a CANDU-light-water-reactor and in a CANDU-fast-breeder-reactor system. The use of results to establish cost targets to guide the current research and development program will be discussed, together with considerations of deployment strategy. (author)

  4. Subchannel analysis of CANDU 37-element fuel bundles

    International Nuclear Information System (INIS)

    The subchannel analysis codes COBRA-IV and ASSERT-4 have been used to predict the mass and enthalpy imbalance within a CANDU 37-element fuel channel under various system conditions. The objective of this study was to assess the various capabilities of the ASSERT code and highlight areas where further validation or development may be needed. The investigation indicated that the ASSERT code has all the basic models required to accurately predict the flow and enthalpy imbalance for complex rod bundles. The study also showed that the code modelling of void drift and diffusion requires refinement to some coefficients and that further validation is needed at high flow rate and high void fraction conditions, where ASSERT and COBRA are shown to predict significantly different trends. The results of a recent refinement of ASSERT modelling are also discussed

  5. The back end of the fuel cycle and CANDU

    International Nuclear Information System (INIS)

    CANDU reactor operators have benefited from several advantages of the CANDU system and from AECL's experience, with regard to spent fuel handling, storage and disposal. AECL has over 20 years experience in development and application of medium-term storage and research and development on the disposal of used fuel. As a result of AECL's experience, short-term and medium-term storage and the associated handling of spent CANDU fuel are well proven and economic, with an extremely high degree of public and environmental protection. In fact, both short-term (water-pool) and medium-term (dry canister) storage of CANDU fuel are comparable or lower in cost per unit of energy than for PWRs. Both pool storage and dry spent fuel storage are fully proven, with many years of successful, safe operating experience. AECL's extensive R and D on the permanent disposal of spent-fuel has resulted in a defined concept for Canadian fuel disposal in crystalline rock. This concept was recently confirmed as ''technically acceptable'' by an independent environmental review panel. Thus, the Canadian program represents an international demonstration of the feasibility and safety of geological disposal of nuclear fuel waste. Much of the technology behind the Canadian concept can be adapted to permanent land-based disposal strategies chosen by other countries. In addition, the Canadian development has established a baseline for CANDU fuel permanent disposal costs. Canadian and international work has shown that the cost of permanent CANDU fuel disposal is similar to the cost of LWR fuel disposal per unit of electricity produced. (author)

  6. CANDU fuel elements behaviour in the load following tests

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Instiute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Palleck, Steve [Sheridan Park Research-AECL, Mississauga, ON (Canada). Fuel Deisgn Branch

    2011-08-15

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during test period. New LF tests are planed to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions. (orig.)

  7. Designing and calculating the pressure loses for different geometries of CANDU type fuel clusters

    International Nuclear Information System (INIS)

    It is well known that circulation of the coolant through the pressure tube of a CANDU type reactor must ensure, through its flow rate values, the optimal conditions of heat transfer from the fuel clusters towards the heavy water. The flow rate through fuel channels differs from one another (up to 24 kg/s) depending on the fuel element sheath temperature, the latter depending in turn one the channels/clusters positions in the calandria vessel. In these conditions, one of the main problem of design in the CANDU type reactor plants is related to the hydraulic resistance represented by the fuel clusters loading the pressure tube or, in other words, the problem of pressure losses (pressure drops) over the length of the fuel cluster column. More precisely, this hydraulic resistance should not exceed a given value imposed by the performance calculations for the pumps used. A sustained activity of analysing comparatively the different geometry types of the fuel clusters was developed at INR Pitesti, a special attention being paid to their behavior as hydraulic resistances. The paper presents a set of computation programs devoted on one hand to the design of fuel clusters of different types and to an estimating computation of the pressure losses resulting from loading these clusters into a specific fuel channel of the CANDU type reactor, on the other hand. During the presentation of the work, different computing codes will be run for demonstration

  8. Reducing the impact of used fuel by transmuting actinides in a CANDU reactor

    International Nuclear Information System (INIS)

    With world stockpiles of used nuclear fuel increasing, the need to address the long term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes in CANDU reactors to reduce the decay heat period. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle facilitates the fabrication and handling of active fuels. Online refueling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation in CANDU reactors, including both recent and past activities. The transmutation schemes that are presented reflect several different partitioning schemes and include both homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. (author)

  9. Conference proceedings of the 4. international conference on CANDU fuel. V. 1,2

    International Nuclear Information System (INIS)

    These proceedings contain the full texts of all 65 papers presented at the 4th International Conference on CANDU fuel. As such, they represent an update on the state-of-the-art in such important CANDU fuel topics as International Development Programs and Operating Experience with CANDU fuel, Performance Assessments and Fuel Behavior Modeling, Fuel Properties, Licensing and Accident Analyses for CANDU fuel, Design, Testing and Manufacturing, and Advanced Fuel Designs. The large number of papers required the use of parallel sessions for the first time at a CANDU Fuel Conference

  10. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. the technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy environment. the world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, Candu reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuel which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the Candu reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential Candu fuel cycle developments can be accommodated in existing

  11. CANDU fuel quality and how it is achieved

    International Nuclear Information System (INIS)

    In this three part presentation CANDU fuel quality is reviewed from the point of view of a designer/operator and a fabricator. In Part 'A' fuel performance and quality considerations are discussed from the point of view of a designer-operator. In Parts 'B' and 'C' fuel quality is reviewed from the point of view of a fabricator. The presentation was divided in this way to convey the 'team effort' attitude which exists in the Canadian program; the team effort which is an essential part of the CANDU story. (auth)

  12. Economic analysis of alternative options in CANDU fuel cycle

    International Nuclear Information System (INIS)

    In this study, fuel cycle options for CANDU reactor were studied. Three main options in a CANDU fuel cycle involve use of : (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option , including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed by using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. Cost estimations were carried out using specially-developed computer programs. Comparison of levelized costs for the fuel cycle options and sensitivity analysis for the cost components are also presented

  13. Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs

  14. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

    The use of recycled uranium (RU) fuel offers significant benefits to CANDU reactor operators particularly if used in conjunction with advanced fuel bundle designs that have enhanced performance characteristics. Furthermore, these benefits can be realised using existing fuel production technologies and practices and with almost negligible change to fuel receipt and handling procedures at the reactor. The paper will demonstrate that the supply of RU as a ceramic-grade UO2 powder will increasingly become available as a secure option to virgin natural uranium and slightly enriched uranium(SEU). In the context of RU use in Canadian CANDU reactors, existing national and international transport regulations and arrangements adequately allow all material movements between the reprocessor, RU powder supplier, Canadian CANDU fuel manufacturer and Canadian CANDU reactor operator. Studies have been undertaken of the impact on personnel dose during fuel manufacturing operations from the increased specific activity of the RU compared to natural uranium. These studies have shown that this impact can be readily minimised without significant cost penalty to the acceptable levels recognised in modem standards for fuel manufacturing operations. The successful and extensive use of RU, arising from spent Magnox fuel, in British Energy's Advanced Gas-Cooled reactors is cited as relevant practical commercial scale experience. The CANFLEX fuel bundle design has been developed by AECL (Canada) and KAERI (Korea) to facilitate the achievement of higher bum-ups and greater fuel performance margins necessary if the full economic potential of advanced CANDU fuel cycles are to be achieved. The manufacture of a CANFLEX fuel bundle containing RU pellets derived from irradiated PWR fuel reprocessed in the THORP plant of BNFL is described. This provided a very practical verification of dose modelling calculations and also demonstrated that the increase of external activity is unlikely to require any

  15. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

  16. Thorium fuel studies for CANDU reactors

    International Nuclear Information System (INIS)

    Applying the once-through Thorium (OTT) cycle in existing and advanced CANDU reactors might be seen as an evolved concept for the sustainable development both from the economic and waste management points of view. Using the Canadian proposed scheme - loading mixed ThO2-SEU CANFLEX bundles in CANDU 6 reactors - simulated at lattice cell level led to promising conclusions on higher burnup, lesser actinide inventory and proliferation resistance. The calculations were performed using the lattice codes WIMS and DRAGON (together with the corresponding nuclear data library based on ENDF/B-VII). (authors)

  17. CANDU 9 safety improvements

    International Nuclear Information System (INIS)

    The CANDU 9 is a family of single-unit Nuclear Power Plant designs based on proven CANDU concepts and equipment from operating CANDU plants capable of generating 900 MWe to 1300 MWe depending on the number of fuel channel used and the type of fuel, either natural uranium fuel or slightly enriched uranium fuel. The basic design, the CANDU 9 480/NU, uses the 480 fuel channel Darlington reactor and employs Natural Uranium (NU) fuel Darlington, the latest of the 900 MWe Class CANDU plants, consists of four integrated units with a total output of approximately 3740 MWe located in Ontario, Canada. AECL has completed the concept definition engineering for this design, and will be completing the design integration engineering by the end of 1996. AECL's design philosophy is to build-in product improvements in evolutionary from the initial prototype plants, NPD and Douglas Point, to today's operating CANDU's construction projects and advanced designs. CANDU 9 safety design follows the evolutionary path, including simple improvements based on existing well-proven CANDU safety concepts. The CANDU 9 builds on the experience base for the Darlington reference plant, and on AECL's extensive safety design experience with single unit CANDU 6 power plants. The latest CANDU 6 plants are being built in Korea by KEPCO at Wolsong 2,3 and 4. The Safety improvements for the CANDU 9 power plant are intended to provide the owner-operator with increased assurance of reliable, trouble-free operation, with greater safety margin, with improved public acceptance, and with ease of licensibility

  18. Fuel Management in Candu Reactors Using Tabu Search

    International Nuclear Information System (INIS)

    Meta-heuristic methods are perfectly suited to solve fuel management optimization problem in LWR. Indeed, they are originally designed for combinatorial or integer parameter problems which can represent the reloading pattern of the assemblies. For the Candu reactors the problem is however completely different. Indeed, this type of reactor is refueled online. Thus, for their design at fuel reloading equilibrium, the parameter to optimize is the average exit burnup of each fuel channel (which is related to the frequency at which each channel has to be reloaded). It is then a continuous variable that we have to deal with. Originally, this problem was solved using gradient methods. However, their major drawback is the potential local optimum into which they can be trapped. This makes the meta-heuristic methods interesting. In this paper, we have successfully implemented the Tabu Search (TS) method in the reactor diffusion code DONJON. The case of an ACR-700 using 7 burnup zones has been tested. The results have been compared to those we obtained previously with gradient methods. Both methods give equivalent results. This validates them both. The TS has however a major drawback concerning the computation time. A problem with the enrichment as an additional parameter has been tested. In this case, the feasible domain is very narrow, and the optimization process has encountered limitations. Actually, the TS method may not be suitable to find the exact solution of the fuel management problem, but it may be used in a hybrid method such as a TS to find the global optimum region coupled with a gradient method to converge faster on the exact solution. (authors)

  19. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  20. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

    Flow-induced fuel vibrations in the pressure tubes of CANDU reactors are of vital interest to designers because fretting damage may result. Computer simulation is being used to study how bundles vibrate and to identify bundle design features which will reduce vibration and hence fretting. (author)

  1. Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-15

    The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO{sub 2}UO{sub 2} and ThO{sub 2}UO{sub 2}-DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future.

  2. A study of coolant thermal mixing within CANDU fuel bundles using ASSERT-PV

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles. The approach taken in the present work is to identify the physical mechanisms contributing to coolant mixing, and to systematically assess the importance of each mechanism. Coupled effects were also considered by flow simulation with mixing mechanisms modelled simultaneously. For the limited range of operating conditions considered and when all mixing mechanisms were modelled simultaneously, the flow was found to be very close to fully mixed. A preliminary model of coolant mixing, suitable for use in the fuel and fuel channel code FACTAR, is also presented. (author)

  3. Integrity of spent CANDU fuel during and following dry storage

    International Nuclear Information System (INIS)

    This report examines the issue of CANDU fuel integrity at the back end of the fuel cycle and outlines a program designed to provide assurance that used CANDU fuel will retain its integrity over an extended period. In specific terms, the program is intended to provide assurance that during and following extended dry storage the fuel will remain fit to undergo, without loss of integrity, the handling, packaging and transportation operations that might be necessary until it is placed in disposal containers. The first step in the development of the program was a review of the available technical information on phenomena relevant to fuel integrity. The major conclusions from that review were the following: Under normal storage conditions it is unlikely that the spent fuel will suffer significant degradation during a one-hundred year period and it should be possible to retrieve, repackage and transport the fuel as required, using methods and systems similar to those used today. However, to provide increased confidence regarding the above conclusion, investigations should be conducted in areas where there is higher uncertainty in the prediction of fuel condition and on some degradation processes to which the fuel appears to present higher vulnerability. The proposed program includes, among other tasks, irradiated fuel tests, analytical studies on the most relevant fuel degradation processes and the development of a long-term fuel verification program. (Author)

  4. Study on advanced nuclear fuel cycle of PWR/CANDU synergism

    International Nuclear Information System (INIS)

    According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

  5. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  6. Third international conference on CANDU fuel

    International Nuclear Information System (INIS)

    These proceedings contain full texts of all 49 papers from the ten sessions and the banquet address. The sessions were on the following subjects: International experience and programs; Fuel behaviour and operating experience; Fuel modelling; Fuel design; Advanced fuel and fuel cycle technology; AECL's concept for the disposal of nuclear fuel waste. The individual papers have been abstracted separately

  7. Used CANDU fuel waste consumed and eliminated: environmentally responsible, economically sound, energetically enormous

    International Nuclear Information System (INIS)

    The 43,800 tonnes of currently stored CANDU nuclear fuel waste can all be consumed in fast-neutron reactors (FNRs) to reduce its long-term radioactive burden 100,000 times while extracting about 130 times more nuclear energy than the prodigious amounts that have already been gained from the fuel in CANDU reactors. The cost of processing CANDU fuel for use in FNRs plus the cost of recycling the FNR fuel is about 2.5 times less on a per kWh energy basis than the currently projected cost of disposal of 3.6 million used CANDU fuel bundles in a deep geological repository. (author)

  8. Post-irradiation neutron emissions from CANDU fuels and their nuclear forensics applications

    International Nuclear Information System (INIS)

    Fissile materials within a fuel pellet can be discerned rapidly and non-destructively via the analysis of post-irradiation delayed neutron emissions. The delayed neutron counting technique is well established within the Canadian nuclear industry, and uses include the detection of defective CANDU fuel. This work discusses these neutron emissions from CANDU fuel in the context of detection and attribution. Monte Carlo simulations of these emissions from current and proposed CANDU fuels (thoria and mix oxide based) have been performed. These simulations are compared to measurements of delayed neutron emissions from 233U and 235U, and the feasibility of CANDU fuel characterization is discussed. (author)

  9. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  10. Romanian progress in the advanced CANDU fuel manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Ohai, D.; Benga, D. [RAAN, Inst. for Nuclear Research, Pitesti- Mioveni (Romania)]. E-mail: dohai@nuclear.ro

    2005-07-01

    The initial concept in developing an advanced fuel compatible with CANDU 6 Reactor, using part of Nuclear Fuel Plant (FCN) Pitesti facilities [1] should be revised. New aspects were considered: working within FCN area, a technological transfer suspicion appears (inobservance of AECL-FCN confidentiality agreement), and the enriched Uranium use on FCN area is prohibited (IAEA requirement). Under these conditions, the Institute for Nuclear Research (ICN) decided to develop or modernize its own facilities for nuclear fuel (CANDU type) manufacturing. The intention was to cover the main technological steps in fuel manufacturing, beginning with powder manufacturing and ending up with fuel bundle assembling. The development or modernization of own facilities for the nuclear fuel manufacturing open the possibilities for the collaboration with other entities interested in advanced fuel development. Having a Research Reactor for material testing and a Post Irradiation+ Facility, ICN can complete the irradiation and post-irradiation services with experimental fuel elements manufacturing, the services being completed. This can be a possibility to eliminate the interstates transport of nuclear materials. The new international requirements for the transport of the nuclear materials are drastic and need a lot of time and money for obtaining authorizations and for transport. It is financially advantageous to manufacture experimental fuel elements on the same site with the irradiation and post-irradiation facilities. (author)

  11. Nanocrystal and noble gas tagging for monitoring defective CANDU fuel bundles

    International Nuclear Information System (INIS)

    The purpose of this paper is to discuss two possible defective fuel bundle tagging techniques that have been suggested for CANDU-6 nuclear reactors. The general design of a CANDU-6 reactor and fuel bundle is reviewed. Nanocrystal tagging is introduced. A current production method for CdTe nanocrystals and future experimental goals are outlined and noble gas tagging is reviewed. Considerations for the future implementation of these tagging methods for fuel in a CANDU-6 reactor is also discussed. (author)

  12. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    Energy Technology Data Exchange (ETDEWEB)

    Lau, J.H. [ed.

    1997-07-01

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference.

  13. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    International Nuclear Information System (INIS)

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference

  14. Recent advances in thorium fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The once-through thorium fuel cycle in CANDU reactors provides an evolutionary approach to exploiting the energy potential of thorium. In the 'mixed bundle' strategy, the central 8 elements in a CANFLEX fuel bundle contain thoria, while the outermost 35 elements contain slightly enriched uranium (SEU). Detailed full-core fuel-management simulations have shown that this approach can be successfully implemented in existing CANDU reactors. Uranium requirements are lower than for the natural uranium fuel cycle. Further energy can be derived from the thorium by recycling the irradiated thoria fuel elements, containing 233U, as-is without any processing, into the center of a new mixed bundle. There are several examples of such 'demountable' bundles. Recycle of the central 8 thoria elements results in an additional burnup of 20 MW·d/kgHE from the thoria elements, for each recycle. The reactivity of these thoria elements remains remarkably constant over irradiation for each recycle. The natural uranium requirements for the mixed bundle (which includes the natural uranium feed required for the outer SEU fuel elements), without recycle, is about 10% lower than for the natural uranium fuel cycle. After the first recycle, the uranium requirements are -35% lower than for the natural uranium cycle, and remain fairly constant with further recycling (the total uranium requirement averaged over a number of cycles is 30% lower than a natural uranium fuelled CANDU reactor). This thorium cycle strategy is a cost-effective means of reducing uranium requirements, while producing a stockpile of valuable 233U, safeguarded in the spent fuel, that can be recovered in the future when predicated by economic or resource considerations. (author)

  15. Silicon carbide TRIPLEX materials for CANDU fuel cladding and pressure tubes

    International Nuclear Information System (INIS)

    Ceramic Tubular Products has developed a superior silicon carbide (SiC) material TRIPLEX, which can be used for both fuel cladding and other zirconium alloy materials in light water reactor (LWR) and heavy water reactor (CANDU) systems. The fuel cladding can replace Zircaloy cladding and other zirconium based alloy materials in the reactor systems. It has the potential to provide higher fuel performance levels in currently operating natural UO2 (NEU) fuel design and in advanced fuel designs (UO2(SEU), MOX thoria) at higher burnups and power levels. In all the cases for fuel designs TRIPLEX has increased resistance to severe accident conditions. The interaction of SiC with steam and water does not produce an exothermic reaction to produce hydrogen as occurs with zirconium based alloys. In addition the absence of creep down eliminates clad ballooning during high temperature accidents which occurs with Zircaloy blocking water channels required to cool the fuel. (author)

  16. Luncheon address: Early days of CANDU fuel

    International Nuclear Information System (INIS)

    I will briefly describe how the original dimensions of the fuel bundle were defined and how that early designs of fuel evolved. I will also touch on some of the historical events of the materials and experiments which effected the fuel programme. Also how I became with Canada's Nuclear Fuel programme. (author)

  17. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    International Nuclear Information System (INIS)

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO2, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author)

  18. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  19. Development of CANDU Spent Fuel Sipping System

    International Nuclear Information System (INIS)

    As the tendency is toward radioactivity zero-leakage on the reactor core for the safe operation of nuclear power plants, the importance of detecting radioactivity leaking from fuel assemblies irradiated in the core is being on the rise. Nuclear fuel, even though it is designed and fabricated in terms of excellent thermal performance and mechanical integrity, can be damaged under unexpected circumstances. An excessive hydriding on fuel rods and pellet-to-clad interaction., etc. can result in failed fuel rod. It is, thus, considered that a inspection process is prerequisite procedure to identify causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water becomes contaminated by leaked fission products, and the power level of the plant has to be lowered or the operation to be stopped. In addition, the spent fuel that have been stored in a spent fuel storage pool for a long time is now transferring to a dry storage. To maximize the integrity of the dry storage, all the fuels transferring to a dry storage should be examined their integrities exactly and efficiently. Therefore, the ultimate purpose of this study is to develop a system capable of judging whether the long-term stored fuel in spent fuel storage pool is failed or not. In this study, a spent fuel sipping system with wet leakage detection technology is developed to make it possible

  20. Advanced operator interface design for CANDU-3 fuel handling system

    International Nuclear Information System (INIS)

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

  1. Developments in CANDU MOX fuel fabrication

    International Nuclear Information System (INIS)

    As a strategic component of its advanced fuel cycle program, AECL continues to implement the MOX fuel development program involving MOX fuel fabrication and characterization, irradiation testing, post-irradiation examination, as well as reactor physics and fuel management studies. AECL performs its MOX fuel fabrication activities in the Recycle Fuel Fabrication Laboratories (RFFL) located at the Chalk River site. The RFFL facility is designed to handle alpha-active fuel material and produce experimental quantities of MOX fuel for reactor physics tests and demonstration irradiations. From 1979 to 1988, several fabrication campaigns were conducted in the RFFL, producing close to two tonnes of MOX fuel with various compositions. RFFL operations were suspended from 1989 until 1994, at which time the facility was needed to fabricate MOX fuel for physics testing in the ZED-2 reactor. After completion of an extensive rehabilitation and re-commissioning of the RFFL, MOX operations were resumed in the facility in August 1996. An up-to-date description of the facility, including the fabrication process and the associated equipment, as well as the upgraded safety systems and laboratory services, is presented. Since the resumption of MOX operations in the RFFL in 1996, several MOX fuel fabrication campaigns have been conducted in the facility; increasing the total amount of MOX fuel fabricated to-date in the RFFL to about three tonnes of MOX fuel. An overview of each of the fabrication campaigns is discussed. The fabrication processes used to manufacture the fuel from the starting powders to the finished elements are summarized. The various fabrication campaigns involved different technical requirements mainly due to the different intended uses of the fuel, i.e., test irradiations in NRU, physics tests in ZED-2, and dissolution experiments in support of the waste management program. Fabrication data including production throughputs and typical inspection results are discussed

  2. CANDU 9 - Overview

    International Nuclear Information System (INIS)

    The CANDU 9 plants are single unit versions of the very successful four unit Bruce B design, incorporating relevant technical advances made in the CANDU 6 and the newer Dalington and CANDU 3 designs. The CANDU 9 plant described in this paper is the CANDU 9 480/SEU with a net electrical output in the range of 1050 MW. In this designation 480 refers to the number of fuel channels, and SEU refers to slightly enriched uranium. Emphasis is placed on evolutionary design and the use of well-proven design features to ensure minimum financial risk to utilities choosing a CANDU 9 plant by assuring regulatory licensability and reliable operation. In addition, the CANDU 9 power plants reflect the important lessons learned by utilities in the construction and operation of CANDU units and, indeed, relevant experience gained by the world nuclear community in its operation of over 400 reactors of a variety of types. As a results, the CANDU 9 plants offer a high level of investment security to the owner, together with relatively low energy costs. The latter results from reduced specific capital cost, reduced operation and maintenance cost, and reduced radiation exposure to plant staff. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9 plants; all key components (steam generators, heat transport pumps, pressure tubes, fuelling machines, etc.) are of the same design as those proven in-service on operating CANDU power stations. The CANDU 9 power plants are readily adaptable to the individual requirements of different utilities and are suitable for a range of site conditions. (author). 12 figs

  3. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

    CANDU spent fuel discharged from the reactor core contains Pu, so, a special attention must be focussed into two directions: tracing for the fuel reactivity in order to prevent critical mass formation and personnel protection during the spent fuel manipulation. Shielding analyses, an essential component of the nuclear safety, take into account the difficulties occurred during the manipulation, transport and storage of spent fuel bundles, both for personnel protection and impact on the environment. The main objective here consists in estimations on radiation doses in order to reduce them under specified limit values. In order to perform the shielding calculations for the spent fuel transport cask three different codes were used: XSDOSE code and MORSE-SGC code, both incorporated in the SCALE4.4a system, and PELSHIE-3 code, respectively. As source of radiation one spent standard CANDU fuel bundle was used. All the geometrical and material data, related to the transport casks, were considered according to the shipping cask type B model, whose prototype has been realized and tested in the Institute for Nuclear Research Pitesti. The radial gamma dose rates estimated to the cask wall and in air, at different distances from the cask, are presented together with a comparison between the dose rates values obtained by all three recipes of shielding calculations. (authors)

  4. Public health risks associated with the CANDU nuclear fuel cycle

    International Nuclear Information System (INIS)

    This report analyzes in a preliminary way the risks to the public posed by the CANDU nuclear fuel cycle. Part 1 considers radiological risks, while part 2 (published as INFO-0141-2) evaluates non-radiological risks. The report concludes that, for radiological risks, maximum individual risks to members of the public are less than 10-5 per year for postulated accidents, are less than 1 percent of regulatory limits for normal operation and that collective doses are small, less than 3 person-sieverts. It is also concluded that radiological risks are much smaller than the non-radiological risks posed by activities of the nuclear fuel cycle

  5. Considerations in recycling used natural uranium fuel from CANDU reactors in Canada

    International Nuclear Information System (INIS)

    This paper identifies the key factors that would affect the recycling of used natural uranium (NU) fuel from CANDU reactors which are in operation in Canada and in several other countries. There has been little analysis of those considerations over the past 25 years and this paper provides a framework for such analysis. In particular, the large energy potential of the plutonium in used CANDU NU fuel provides a driver for consideration of used-fuel recycling. There would be a long lead-time (at least 30 years) and a large investment required for establishing the infrastructure for used-fuel recycling. While this paper does not promote the recycling of used CANDU NU fuel in Canadian CANDU reactors, it does suggest that it is timely to start the analysis and to consider the key factors or circumstances that warrant the recycling of used CANDU NU fuel. (author)

  6. Fission Product Inventory in CANDU Fuel

    International Nuclear Information System (INIS)

    When the reactor is operated at power, fuel composition changes continuously. The fission reaction produces a large variety of fission fragments which are radioactive and decay into other isotopic species. For different accident analyses or operational events, detailed calculations of the fuel radioactive inventory (fission products and actinides) are needed. The present paper reviews two types of radioactive inventory calculations performed at Cernavoda NPP: one for determining the whole core inventory and one for determining the evolution of the inventory within fuel bundles stored in the Spent Fuel Bay. Two computer codes are currently used for radioactive inventory calculations: ORIGEN-S and ELESTRES-IST. The whole core inventory calculation was performed with both codes, the comparison showing that ELESTRES-IST gives a more conservative result. One of the challenges met during the analysis was to set a credible, yet conservative “image” of the in core fuel power/burnup distribution. Consequently, a statistical analysis was performed to find the best estimate plus uncertainties map for the power/burnup distribution of all in core fuel elements. For each power/burnup in the map, the fission product inventory was computed using a scaled irradiation history based on the Limiting Overpower Envelope. After the Fukushima accident, the problem of assessing the consequences of a loss of cooling event at the Spent Fuel Bay was raised. In order to estimate its impact, a calculation for determining the fission products inventory and decay heat evolution within the spent fuel bundles stored in the bay was performed. The calculation was done for a bay filled with fuel bundles up to its maximum capacity. The results obtained have provided a conservative estimation of the decay heat released and the expected evolution of the water temperature in the bay. This provided a technical basis for selecting the emergency actions required to cope with such events. (author)

  7. Integrity Assessment of CANDU Spent Fuel During Interim Dry Storage in MACSTOR

    International Nuclear Information System (INIS)

    This paper presents an assessment of the integrity of CANDU spent fuel during dry storage in MACSTOR. Based on review of the safety requirements for sheath integrity during dry storage, a fuel temperature limit for spent CANDU fuel stored in MACSTOR is specified. The spent fuel conditions prior to, and during dry storage are assessed. The safety margin for spent CANDU fuel stored in MACSTOR is assessed against various failure mechanisms using the probabilistic estimation approach derived from US LWR fuel data set. (author)

  8. Interim storage of CANDU spent fuel and safety performance

    International Nuclear Information System (INIS)

    'Full text:' Pickering Waste Management Facility (PWMF) is operational since November 1995 and safely storing spent fuel from Pickering 8 CANDU reactors. To date, equivalent to 22 reactor-years worth of spent fuel have been loaded, processed and stored in Dry Storage Containers (DSC). One DSC contains spent fuel from approximately one reactor-month of full power operation. The design life for the storage containers is 50 years. A Nuclear Waste Management Organization (NWMO) has been formed to advise on the long-term Canadian strategy for management of spent fuel. This paper will present the DSC processing steps, radiological hazard magnitude experienced during the DSC loading and processing for interim storage. A brief description of environmental and occupational safety performance will be presented. (author)

  9. Improved CANDU fuel performance. A summary of previous AECL publications

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU power reactors has been very low (0.06%) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding, increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness, wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

  10. Safe, permanent disposal of used CANDU fuel

    International Nuclear Information System (INIS)

    AECL's assessment of nuclear fuel waste disposal deep in plutonic rock of the Canadian Precambrian Shield is now well advanced. A comprehensive understanding has evolved of the chemical and physical processes controlling the containment of radionuclides in used fuel. The following conclusions have been reached: containers with outer shells of titanium and copper can be expected to isolate used fuel from contact with groundwater for at least 500 years, the period during which the hazard is greatest; uranium oxide fuel can be expected to dissolve at a rate less than 10-8 per day, resulting in uranium concentrations less that 1 μg/L, which is consistent with observations of uranium oxide deposits in the earth's crust; movement of dissolved radionuclides away from the containers can be delayed for thousands of years by placing a compacted bentonite-clay layer between the container and the rock mass; and, the granite plutons of interest consist of relatively large rock volumes of low permeability separated by relatively thin fracture zones, and the low permeability volumes are sufficiently large to accommodate a vault design that will ensure radionuclides do not reach the surface in unacceptable concentrations

  11. Economic potential of advanced fuel cycles in CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles in CANDU offer the potential of a many-fold increase in energy yield over that which can be obtained from uranium resources using the current once-through natural uranium cycle. This paper examines the associated economics of alternative once-through and recycle fuelling. Results indicate that these cycles will limit the impact of higher uranium prices and offer the potential of a period of stable constant-dollar generating costs that are only approximately 20% higher than current levels

  12. Procurement and supply of CANDU fuels

    International Nuclear Information System (INIS)

    In 1955 a decision was made to proceed with construction of a Nuclear Power Demonstration Station (NPD) near Rolfton, Ontario. This project, headed by Atomic Energy of Canada with major involvement of private industry, was the genesis for the development of nuclear electric generation in Canada. This paper reviews one aspect of the Canadian program: the evolution of fuel procurement and supply, which in itself has been a remarkable Canadian achievement. (author)

  13. A model for fission product distribution in CANDU fuel

    International Nuclear Information System (INIS)

    This paper describes a model to estimate the distribution of active fission products among the UO2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I131. The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

  14. The R and D program in support of advanced fuel cycles for CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles for CANDU reactors are well on their way to being implemented. The first step is slightly enriched uranium (SEU) which is economical today. A new fuel bundle is seen as the vehicle for all fuels in CANDU. CANDU fuel fabricated from uranium recovered from fuel discharged from light-water reactors (LWR) is also economical today and readily achievable technically. Future fuel cycles would utilize plutonium recovered from light-water reactors or CANDU's and eventually thorium. R and D in support of these cycles focuses on those topics that require a high degree of confidence in their implementation such as fuel fabrication and defect-free performance to high burnup. Reactor physics codes and nuclear data for advanced fuel cycles will be validated against experiments. (author). 8 refs

  15. Use of Utility Codes for Fuel Analysis during Off-Stagnation Feeder Break in CANDU

    International Nuclear Information System (INIS)

    Feeder break accident is regarded as one of the design basis accident in CANDU reactor which results in a fuel failure. For a particular range of inlet feeder break sizes, the flow in the channel is reduced sufficiently that the fuel and fuel channel integrity can be significantly affected to have damage in the affected channel, while the remainder of the core remains adequately cooled. The flow in the downstream channel can be more or less stagnated due to a balance between pressure at the break on the upstream side and the reverse driving pressure between the break and the downstream end. In the extreme, this can lead to rapid fuel heatup and fuel damage and failure of the fuel channel similar to that associated with a severe channel flow blockage. Such an inlet feeder break scenario is called a stagnation break. For an inlet feeder break which is slightly larger or smaller than that for the stagnation break case, the result is a channel flow which is low enough to result in fuel failure but high enough that the pressure tube remains intact. This event is identified as the off-stagnation break. In this report, the fuel analysis methodology and the usage of utility codes to evaluate the fission gas release during the off-stagnation feeder break are described. The accident was assumed to be occurred in the refurbished Wolsong unit 1 and the latest safety codes were used in the analysis. Fission product inventories during the steady state were calculated by using ELESTRES-IST 1.2 code. After starting the off-stagnation break, ELOCA code evaluated the timing of fuel failure and the following fission gas release due to the oxidation of the pellet are calculated by using several utility codes until the reactor trip. The calculated fission product releases are provided to the following dose assessment as a source term

  16. Development of the CANDU high-burnup fuel design/analysis technology

    International Nuclear Information System (INIS)

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

  17. Assessment of CANDU-6 reactivity devices for DUPIC fuel

    International Nuclear Information System (INIS)

    Reactivity device characteristics for a CANDU 6 reactor loaded with DUPIC fuel have been assessed. The lattice parameters were generated by WIMS-AECL code and the core calculations were performed by RFSP code with a 3-dimensional full core model. The reactivity devices studied are the zone controller, adjusters, mechanical control absorber and shutoff rods. For the zone controller system, damping capability for spatial oscillation was investigated. For the adjusters, the restart capability was investigated. For the adjusters, the restart capability was investigated. The shin operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster system. The mechanical control absorber was assessed for the function of compensating temperature reactivity feedback following a power reduction. And shutoff rods were also assessed to investigate the following a power reduction. And shutoff rods were also assessed to investigate the static reactivity worth. This study has shown that the current reactivity device system of CANDU-6 core with the DUPIC fuel. (author). 9 refs., 17 tabs., 7 figs

  18. Quality control of CANDU6 fuel element in fabrication process

    International Nuclear Information System (INIS)

    To enhance the fine control over all aspects of the production process, improve product quality, fuel element fabrication process for CANDU6 quality process control activities carried out by professional technical and management technology combined mode, the quality of the fuel elements formed around CANDU6 weak links - - end plug , and brazing processes and procedures associated with this aspect of strict control, in improving staff quality consciousness, strengthening equipment maintenance, improved tooling, fixtures, optimization process test, strengthen supervision, fine inspection operations, timely delivery carry out aspects of the quality of information and concerns the production environment, etc., to find the problem from the improvement of product quality and factors affecting the source, and resolved to form the active control, comprehensive and systematic analysis of the problem of the quality management concepts, effectively reducing the end plug weld microstructure after the failure times and number of defects zirconium alloys brazed, improved product quality, and created economic benefits expressly provided, while staff quality consciousness and attention to detail, collaboration department, communication has been greatly improved and achieved very good management effectiveness. (authors)

  19. Regulatory review of the CANDU fuel modification program in Canada

    International Nuclear Information System (INIS)

    Aging of a CANDU nuclear power plant affects various safety margins of the plant. Margin to fuel sheath dryout is one of the safety margins that have been detrimentally affected, leading to a reduced margin to dryout with time. If no proactive actions are taken, the plant will have to de-rate its operation at an earlier time. To postpone the de-rating, the Canadian nuclear Industry has taken multi-initiatives to restore, or partially restore the safety margins that have been eroded due to plant aging. One of the initiatives is modification/re-optimization of the current fuel design, in order to improve the fuel thermalhydraulic performance, i.e., to suppress fuel sheath dryout, whereby offset partially the erosion of margin to fuel sheath dryout. Several Canadian utilities have already proceeded with the fuel modification program and requested approval of the Canadian Nuclear Safety Commission (CNSC) to load the modified fuel into their reactors. This paper summarizes the CNSC requirements, the review processes, the current status, and the technical challenges associated with licensing review of the fuel modification in Canada. (author)

  20. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov

    2007-02-01

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼ 13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable.

  1. The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

  2. Systems for transporting used CANDU fuel by road, rail and water

    International Nuclear Information System (INIS)

    Ontario Hydro's CANDU nuclear power stations are situated on the shores of the Great Lakes and are accessible by road, rail and water. For the off-site shipment of used CANDU fuel from the stations to a disposal, reprocessing or central storage facility, all three modes are being considered. This paper presents Ontario Hydro's 'reference transportation systems' for the shipment of used CANDU fuel as developed for the Nuclear Fuel Waste Management Program (NFWMP) in Canada. These are workable systems developed by Ontario Hydro for the purpose of showing modal feasibility. The systems have not yet been optimized

  3. Safety assessment to support NUE fuel full core implementation in CANDU reactors

    International Nuclear Information System (INIS)

    The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

  4. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  5. Fuel management optimization in CANDU reactors cooled with light water

    International Nuclear Information System (INIS)

    This research has two main goals. First, we wanted to introduce optimization tools in the diffusion code DONJON, mostly for fuel management. The second objective is more practical. The optimization capabilities are applied to the fuel management problem for different CANDU reactors at refueling equilibrium state. Two kinds of approaches are considered and implemented in this study to solve optimization problems in the code DONJON. The first methods are based on gradients and on the quasi-linear mathematical programming. The method initially developed in the code OPTEX is implemented as a reference approach for the gradient based methods. However, this approach has a major drawback. Indeed, the starting point has to be a feasible point. Then, several approaches have been developed to be more general and not limited by the initial point choice. Among the different methods we developed, two were found to be very efficient: the multi-step method and the mixte method. The second kind of approach are the meta-heuristic methods. We implemented the tabu search method. Initially, it was designed to optimize combinatory variable problems. However, we successfully used it for continuous variables. The major advantage of the tabu method over the gradient methods is the capability to exit from local minima. Optimisation of the average exit burnup has been performed for CANDU-6 and ACR-700 reactors. The fresh fuel enrichment has also been optimized for ACR-700. Results match very well what the reactor physics can predict. Moreover, a comparison of the two totally different types of optimization methods validated the results we obtained. (author)

  6. Development of CANDU Spent Fuel Disposal Concepts for the Improvement of Disposal Efficiency

    International Nuclear Information System (INIS)

    There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over have been proposed. These new disposals have made it possible to introduce the concept of long term storage and retrievability and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

  7. CANDU fuel research and development in Korea: current status and future prospects

    International Nuclear Information System (INIS)

    The current status and future prospect of CANDU fuel R and D in Korea is always subjected to the consideration of domestic and international environments concerning nuclear safety, nuclear waste, nonproliferation and economy in favor of the arguments from public acceptance, international environments, and utilities. Considering that, at the end of 2000, the procurement of additional CANDU units at the Shin Wolsong site was decided not to proceed, the current and future CANDU fuel R and D would be oriented to the safety and economy of fuel and reactor operations rather than the national strategy of nuclear fuel cycle and reactor programs in Korea. Therefore, the current CANDU advanced fuel R and D programs such as CANFLEX-NU fuel industrialization, CANFLEX-0.9% SEU/RU fuel R and D, and DUPIC fuel cycle development in a laboratory-scale will be continued for the time being as it was. But the R and D of CANDU innovative fuels such as CANFLEX-1.2% ∼ 1.5 % SEU fuel, thorium oxide fuel and DUPIC fuel would have some difficulties to continue in the mid- and long-term if they would not have the justifications in the points of the nonproliferation, economic and safety views of fuel, fuel cycle and reactor. (author)

  8. Current status and future prospect of Candu fuel research and development in Korea

    International Nuclear Information System (INIS)

    The current status and future prospect of CANDU fuel R and D in Korea is always subjected to the consideration of domestic and international environments concerning nuclear safety, nuclear waste, non-proliferation and economy in favor of the arguments from public acceptance, international environments, and utilities. Considering that, at the end of 2000, the procurement of additional CANDU units at the Shin Wolsong site was decided not to proceed, the current and future CANDU fuel R and D would be oriented to the safety and economy of fuel and reactor operations rather than the national strategy of nuclear fuel cycle and reactor programs in Korea. Therefore, the current CANDU advanced fuel R and D programs such as CANFLEX-NU fuel industrialization, CANFLEX-0.9% SEU/RU fuel R and D, and DUPIC fuel cycle development in a laboratory-scale will be continued for the time being as it was. But the R and D of CANDU innovative fuels such as CANFLEX- 1.2% ∼ 1.5 % SEU fuel, thorium oxide fuel and DUPIC fuel would have some difficulties to continue in the mid-and long-term if they would not have the justifications in the points of the non-proliferation, economic and safety views of fuel, fuel cycle and reactor. (author)

  9. Canflex: A fuel bundle to facilitate the use of enrichment and fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    The neutron economy of the CANDU reactor results in it being an ideal host for a number of resource-conserving fuel cycles, as well as a number of potential ''symbiotic'' fuel cycles, in which fuel discharged from light-water cooled reactors is recycled to extract the maximum energy from the residual fissile material before it is sent for disposal. The resource conserving fuel cycles include the natural-uranium, slightly-enriched-uranium and thorium fuel cycles. The ''LWR-symbiotic'' cycles include recovered uranium and various options for the direct use of spent LWR fuel in CANDU reactors. However, to achieve the maximum economic potential of these fuel-cycle options requires irradiation to burnups higher than that possible with natural uranium. To provide a basis for the economic use of these fuel cycles, a program is underway to develop and demonstrate a CANDU fuel bundle capable of both higher burnups and greater operating margins. This new bundle design is being developed jointly by AECL and KAERI, and uses smaller-diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This allows operation to burnups greater than 21 MWd/KgU. A combination of this lower peak-element rating, plus development work underway at AECL to enhance the thermalhydraulic characteristics of the bundle (including both critical heat flux and bundle pressure drop), provides a greater operating margin for the bundle. This new bundle design is called CANFLEX, and the program for its development in Canada and Korea is described in this paper. (author). 19 refs, 5 figs

  10. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors

  11. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

  12. Next generation CANDU plants

    International Nuclear Information System (INIS)

    Future CANDU designs will continue to meet the emerging design and performance requirements expected by the operating utilities. The next generation CANDU products will integrate new technologies into both the product features as well as into the engineering and construction work processes associated with delivering the products. The timely incorporation of advanced design features is the approach adopted for the development of the next generation of CANDU. AECL's current products consist of 700MW Class CANDU 6 and 900 MW Class CANDU 9. Evolutionary improvements are continuing with our CANDU products to enhance their adaptability to meet customers ever increasing need for higher output. Our key product drivers are for improved safety, environmental protection and improved cost effectiveness. Towards these goals we have made excellent progress in Research and Development and our investments are continuing in areas such as fuel channels and passive safety. Our long term focus is utilizing the fuel cycle flexibility of CANDU reactors as part of the long term energy mix

  13. Five years of successful CANDU-6 fuel manufacturing in Romania

    International Nuclear Information System (INIS)

    This paper describes the evolution of CANDU-6 nuclear fuel manufacturing in Romania at FCN Pitesti, after the completion of the qualification in 1994. Commercial production was resumed early 1995 and fuel bundles produced were entirely delivered to Cernavoda Plant and charged in the reactor. More than 12,000 fuel bundles have been produced in the last five years and the fuel behaved very well. Defective bundles represents less than 0.06% from the total irradiated fuel, and the most defects are associated to the highest power positions. After qualification, FCN focused the effort to improve braze quality and also to maintain a low residual hydrogen content in graphite coated sheaths. The production capacity was increased especially for component manufacturing, appendages tack welding and brazing. A new graphite baking furnace with increased capacity, is under design. In the pelleting area, a rotating press will replace the older hydraulic presses used for pelleting. Plant development taken inter consideration the future demands for Cernavoda Unit 2. (author)

  14. CANDU advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    In the fast-growing economies of the Pacific Basin region, sustainability is an important requisite for new energy development. Many countries in this region have seen, and continue to see, very large increases in energy and electricity demand. The investment in any nuclear technology is large. Countries making that investment want to ensure that the technology can be sustained and that it can evolve in an ever-changing environment. Three key aspects in ensuring a sustainable energy future are: technological sustainability; economic sustainability; and environmental sustainability (including resource utilization). The fuel-cycle flexibility of the CANDU reactor provides a ready path to sustainable energy development in both the short and the long term. (author). 23 refs

  15. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

  16. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  17. Dry storage of irradiated CANDU fuel at Pickering NGS

    International Nuclear Information System (INIS)

    Ontario Hydro generates about 86 million MW-h/year from its 20 nuclear CANDU reactors. The combination of a large generating capacity and relatively low fuel burn-up means that Ontario Hydro must manage very large volume of its used fuel. Irradiated fuel bays at Pickering NGS will be full by mid 1995. Additional storage capacity will be required by this date for the station to continue operation. Several long term storage options to supplement existing on-site facilities were studied. The dry storage system, based on the modular storage container, as an option, was found to be economical and operationally simple. The dry storage facility at Pickering NGS is planned to provide additional on-site storage capacity for fuel generated from 1994 to the end of the station's operating life (year 2025). This paper describes the design and operation of the Dry Storage System at Ontario Hydro's Pickering Nuclear Power Generating Station. The facility is planned as a two phased project. Phase I will provide a storage space for 700 dry storage containers (268,800 fuel bundles) or about 12 years of station's operation. Phase II will have the capacity for additional 800 containers (307,200 fuel bundles) and will provide storage until station decommissioning in 2025. Seventy containers will be required annually to meet storage requirements of the station's operation at 100% capacity factor. Staff of six persons will be required to operate the facility. Normal operation includes activities such as receiving and commissioning new containers, loading them with 4 modules of used fuel in the bays, draining and drying the cavity, decontaminating the container surface and lid welding. A helium leak test is performed before the container is placed in the storage. (author) 6 refs., 6 tabs., 7 figs

  18. CANDU-6 fuel bundle fabrication and advanced fuels development in China

    International Nuclear Information System (INIS)

    In recent years, China North Nuclear Fuel Corporation (CNNFC) has introduced several modifications to the manufacturing processes and the production line equipment. This has been beneficial in achieving a very high level of quality in the production of fuel bundles. Since 2008 CNNFC has participated in a multi party project with the goal of developing advanced fuels for use in CANDU reactors. Other project team members include the Nuclear Power Institute of China (NPIC), Third Qinshan Nuclear Power Company (TQNPC) and Atomic Energy of Canada Ltd (AECL). This paper will present the improvements developed during the manufacture of natural fuel bundles and advanced fuels. (author)

  19. The R and D program in support of advanced fuel cycles for CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles for CANDU reactors are well on their way to being implemented. The first step is slightly enriched uranium (SEU) which is economical today. A new fuel bundle is seen as the vehicle for all fuels in CANDU. CANDU fuel fabricated from uranium recovered from fuel discharged from light-water reactors (LWR) is also economical today and readily achievable technically. Future fuel cycles would utilize plutonium recovered from light-water reactors or CANDUs and eventually thorium. R and D in support of these cycles focuses on those topics that require a high degree of confidence in their implementation such as fuel fabrication and defect-free performance to high burnup. Reactor physics codes and nuclear data for advanced fuel cycles will be validated against experiments. (author).

  20. AECL experience in fuel channel inspection

    International Nuclear Information System (INIS)

    Inspection of CANDU fuel channels (FC) is performed to ensure safe and economic reactor operation. CANDU reactor FCs have features that make them a unique non-destructive testing (NDT) challenge. The thin, 4 mm pressure-tube wall means flaws down to about 0.1 mm deep must be reliably detected and characterized. This is one to two orders of magnitude smaller than is usually considered of significant concern for steel piping and pressure vessels. A second unique feature is that inspection sensors must operate in the reactor core--often within 20 cm of highly radioactive fuel. Work on inspection of CANDU reactor FCs at AECL dates back over three decades. In that time, AECL staff have provided equipment and conducted or supervised in-service inspections in about 250 FCs, in addition to over 8000 pre-service FCs. These inspections took place at every existing CANDU reactor except those in India and Romania. Early FC inspections focussed on measurement of changes in dimensions (gauging) resulting from exposure to a combination of neutrons, stress and elevated temperature. Expansion of inspection activities to include volumetric inspection (for flaws) started in the mid-1970s with the discovery of delayed hydride cracking in Pickering 3 and 4 rolled joints. Recognition of other types of flaw mechanisms in the 1980s led to further expansion in both pre-service and in-service inspections. These growing requirements, to meet regulatory as well as economic needs, led to the development of a wide spectrum of inspection technology that now includes tests for hydrogen concentration, structural integrity of core components, flaws, and dimensional change. This paper reviews current CANDU reactor FC inspection requirements. The equipment and techniques developed to satisfy these requirements are also described. The paper concludes with a discussion of work in progress in AECL aimed at providing state-of-the-art FC inspection services. (author)

  1. Slightly enriched uranium in CANDU: An economic first step towards advanced fuel cycles

    International Nuclear Information System (INIS)

    The natural-uranium fuelled Canada Deuterium-Uranium (CANDU) nuclear reactor system has proven to be a safe, reliable and economical producer of electricity for over a quarter of a century. The CANDU system, however, is not restricted to the use of natural-uranium fuel; a wide range of advanced fuel cycles can be accommodated. In the short term, slightly enriched uranium (SEU) is the most promising of these advanced fuel cycles. SEU offers a reduction in the total fuel cycle cost of between 25 and 50% relative to natural-uranium fuel. Uranium consumption is decreased by 30 to 40%. In addition the volume of spent fuel is reduced by a factor of two to three, depending on the enrichment selected. SEU also offers greater flexibility in the design of future CANDU reactors. A variety of fuel management options can be employed in CANDU with slightly enriched fuels. Fuel performance is expected to be good for the burnups of interest, but further fuel testing is planned and is currently in progress in order to confirm this. Programs in place at Atomic Energy of Canada Limited (AECL) will lead to the demonstration and introduction of slightly enriched uranium in CANDU. Ontario Hydro, a Canadian utility with twenty CANDUs operating or under construction, is considering a program which could lead to the implementation of SEU in its nuclear generating stations. (author). 30 refs, 7 figs

  2. CANDU 300

    International Nuclear Information System (INIS)

    The CANDU nuclear power system is under continuous review by AECL in order to advance the CANDU concept in a manner that will assure competitiveness in both current and future markets. Over the past three years development effort has featured the CANDU 300, a CANDU nuclear generating station with a net output in the range of 320 MW9e) to 380 MW(e). At the outset AECL recognized that coal-fired power plants would be the primary competition for the CANDU advantages such as the use of natural uranium fuel and on-power refuelling, while enhancing capacity factor, reducing man-rem exposure, reducing capital cost, and minimizing construction schedules. AECL believes that the resulting CANDU 300 nuclear generating station will have substantial appeal to many utilities, in both developed and developing countries. The key features of the CANDU 300 are presented here, with particular attention to the station layout, construction methods, and construction schedules

  3. The Study of Nuclear Fuel Cycle Options Based On PWR and CANDU Reactors

    International Nuclear Information System (INIS)

    The study of nuclear fuel cycle options based on PWR and CANDU type reactors have been carried out. There are 5 cycle options based on PWR and CANDU reactors, i.e.: PWR-OT, PWR-OT, PWR-MOX, CANDU-OT, DUPIC, and PWR-CANDU-OT options. While parameters which assessed in this study are fuel requirement, generating waste and plutonium from each cycle options. From the study found that the amount of fuel in the DUPIC option needs relatively small compared the other options. From the view of total radioactive waste generated from the cycles, PWR-MOX generate the smallest amount of waste, but produce twice of high level waste than DUPIC option. For total plutonium generated from the cycle, PWR-MOX option generates smallest quantity, but for fissile plutonium, DUPIC options produce the smallest one. It means that the DUPIC option has some benefits in plutonium consumption aspects. (author)

  4. The next generation of CANDU technologies: profiling the potential for hydrogen fuel

    International Nuclear Information System (INIS)

    This report discusses the Next-generation CANDU Power Reactor technologies currently under development at AECL. The innovations introduced into proven CANDU technologies include a compact reactor core design, which reduces the size by a factor of one third for the same power output; improved thermal efficiency through higher-pressure steam turbines; reduced use of heavy water (one quarter of the heavy water required for existing plants), thus reducing the cost and eliminating many material handling concerns; use of slightly enriched uranium to extend fuel life to three times that of existing natural uranium fuel and additions to CANDU's inherent passive safety. With these advanced features, the capital cost of constructing the plant can be reduced by up to 40 per cent compared to existing designs. The clean, affordable CANDU-generated electricity can be used to produce hydrogen for fuel cells for the transportation sector, thereby reducing emissions from the transportation sector

  5. Fuel bundle geometry and composition influence on coolant void reactivity reduction in ACR and CANDU reactors

    International Nuclear Information System (INIS)

    It is very well known that the CANDU reactor has positive Coolant Void Reactivity (CVR), which is most important criticisms about CANDU. The most recent innovations based on using a thin absorbent Hafnium shell in the central bundle element were successfully been applied to the Advanced CANDU Reactor (ACR) project. The paper's objective is to analyze elementary lattice cell effects in applying such methods to reduce the CVR. Three basic fuel designs in their corresponding geometries were chosen to be compared: the ACR-1000TM, the RU-43 (developed in INR Pitesti) and the standard CANDU fuel. The bundle geometry influence on void effect was also evaluated. The WIMS calculations proved the Hafnium absorber suitability (in the latest 'shell design') to achieve the negative CVR target with great accuracy for the ACR-1000 fuel bundle design than for the other two projects. (authors)

  6. Computer simulation of the behaviour and performance of a CANDU fuel rod

    International Nuclear Information System (INIS)

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  7. Photon dose rates estimation for CANDU spent fuel transport and intermediate dry storage

    International Nuclear Information System (INIS)

    The nuclear energy world wide development is accompanied by huge quantities of spent nuclear fuel accumulation. Shielding analyses are an essential component of the nuclear safety, the estimations of radiation doses in order to reduce them under specified limit values being the main task here. According to IAEA data, more than 10 millions packages containing radioactive materials are annually world wide transported. The radioactive material transport safety must be carefully settled. Last decade, both for operating reactors and future reactor projects, a general trend to raise the discharge fuel burnup has been world wide registered. For CANDU type reactors, one of the most attractive solutions seems to be SEU fuel utilization. In the paper there are estimated the CANDU spent fuel photon dose rates at the shipping cask/ storage basket wall for two different fuel projects after a defined cooling period in the NPP pools. The CANDU fuel projects considered were the CANDU standard 37 rod fuel bundle with natural UO2 and SEU fuels. In order to obtain radionuclide inventory and irradiated fuel characteristics, ORIGEN-S code has been used. The spent fuel characteristics are presented, comparatively, for both types of CANDU fuels. By means of the same code the photon source profiles have been calculated. The shielding calculations both for spent fuel transport and intermediate storage have been performed by using Monte Carlo MORSE-SGC code. The ORIGEN-S and MORSE-SGC codes are both included in ORNL's SCALE 4.4a program package. A photon dose rates comparison between the two types of CANDU fuels has been also performed, both for spent fuel transport and intermediate dry storage. (authors)

  8. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  9. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  10. A feasible approach to implement a commercial scale CANDU fuel manufacturing plant in Egypt

    International Nuclear Information System (INIS)

    Many planning scenarios have been examined to assess and evaluate the economic estimates for implementing a commercial scale CANDU fuel manufacturing plant in Egypt. The cost estimates indicated strong influence of the annual capital costs on total fuel manufacturing cost; this is particularly evident in a small initial plant where the proposed design output is only sufficient to supply reload fuel for a single CANDU-6 reactor. A modular approach is investigated as a possible way, to reduce the capital costs for a small initial fuel plant. In this approach the plant would do fuel assembly operations only and the remainder of a plant would be constructed and equipped in the stages when high production volumes can justify the capital expenses. Such approach seems economically feasible for implementing a small scale CANDU fuel manufacturing plant in developing countries such as Egypt and further improvement could be achieved over the years of operation. (author)

  11. Estimation of radiation doses characterizing CANDU spent fuel transport and intermediate dry storage

    International Nuclear Information System (INIS)

    Shielding analyses are an essential component of the nuclear safety. The estimations of radiation doses in order to reduce them under specified limitation values is the main task here. In the last decade, a general trend to raise the discharge fuel burnup has been world wide registered for both operating reactors and future reactor projects. For CANDU type reactors, one of the most attractive solutions seems to be SEU fuels utilization. The goal of this paper is to estimate CANDU spent fuel photon dose rates at the shipping cask/storage basket wall and in air, at different distances from the cask/ basket, for two different fuel projects, after a defined cooling period in the NPP pools. Spent fuel inventories and photon source profiles are obtained by means of ORIGEN-S code. The shielding calculations have been performed by using Monte Carlo MORSE-SGC code. A comparison between the two types of CANDU fuels has been also performed. (authors)

  12. Feasible advanced fuel cycle options for CANDU reactors in the Republic of Korea

    International Nuclear Information System (INIS)

    Taking into account the view points on nuclear safety, nuclear waste, non-proliferation and economics from the public, international environment, and utilities, the SEU/RU and DUPIC fuel cycles would be feasible options of advanced fuel cycles for CANDU-PHWRs in the Republic of Korea in the mid- and long-terms, respectively. Comparing with NU fuel, 0.9 % or 1.2 % SEU fuel would increase fuel burnup and hence reduce the spent fuel arisings by a factor of 2 or 3, and also could reduce CANDU fuel cycle costs by 20 to 30%. RU offers similar benefits as 0.9% SEU and is very attractive due to the significantly improved fuel cycle economics, substantially increased burnups, large reduction in fuel requirements as well as in spent fuel arisings. For RU use in a CANDU reactor, re-enrichment is not required. There are 25,000 tes RU produced from reprocessing operations in Europe and Japan, which would theoretically provide sufficient fuel for 500 CANDU 6 reactor-years of operation. According to the physics, thermal-hydraulic and thermal-mechanical assessments of CANFLEX-0.9% RU fuel for a CANDU-6 reactor, the fuel could be introduced into the reactor in a straight-forward fashion. A series of assessments of CANFLEX-DUPIC physics on the compatibility of the fuel design in the existing CANDU 6 reactors has shown that the poisoning of the central element of DUPIC with, for example, natural dysprosium, reduces the void reactivity of the fuel, and that a 2 bundle shift refuelling scheme would be the most appropriate in-core fuel management scheme for a CANDU-6 reactor. The average discharge burnup is ∼15 MWd/kgHE. Although these results have shown promising results for the DUPIC fuel cycle, more in-depth studies are required in the areas of ROP system, large LOCA safety analyses, and so on. The recycling fuel cycles of RU and DUPIC for CANDU are expected to achieve the environmental 3R's (Reduce, Reuse, Recycle) as applied to global energy use in the short- and long

  13. Wet channel measurement of pressure tube to calandria tube spacing in CANDU reactors

    International Nuclear Information System (INIS)

    The pressure tube (PT) to calandria tube (CT) spacing in CANDU reactors is an important parameter that relates to the general condition of the fuel channels. The measurement system that was developed to measure this parameter during the wet channel inspections of Pickering Units 1 and 2 is described in this paper. A send-receive eddy current probe was designed which is primarily sensitive to variations in PT/CT spacing but is also affected by pressure tube wall thickness. A computer simulation showed that the phase angles of the response to these variables are similar for all usable frequencies, thus eliminating the possibility of multifrequency compensation. A marriage of technologies was proposed involving the ultrasonic measurement of wall thickness values which are then used to extract the spacing information from the eddy current signal. The accuracy of the system is approximately ±(30% +.1mm) which has been sufficient to determine if and where any of the pressure tubes have come in contact with their calandria tube. Field experience with the new system is discussed and areas for development are also outlined

  14. System study of CANDU/LWR synergy in advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    This report proposes a study that will evaluate the effects of advanced nuclear fuel cycles on resource utilisation, repository capacity, waste streams, economics, and proliferation resistance. The proposed fuel cycles are designed to exploit the unique synergy that exists between light water and CANDU reactors. Also, several fuel cycle simulation codes have been proposed to be used. (author)

  15. Burn up Analysis for Fuel Assembly Unit i n a Pressurized Heavy Water CANDU Reactor

    International Nuclear Information System (INIS)

    MCNPX code has been used for modeling a nd simulation of an assembly of CANDU Fuel bundle . The assembly is composed of a heterogeneous lattice of 37-element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (Keff ) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing t he concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found PWR

  16. RU-43 a new uranium fuel bundle design for using in CANDU type reactors

    International Nuclear Information System (INIS)

    A unique feature of the CANDU reactor design is its ability to use alternative fuel cycles other than natural uranium (NU), without requiring major modifications to the basic reactor design. These alternative fuel cycles, which are known as advanced fuel cycles, utilize a variety of fissile materials, including Slightly Enriched Uranium (SEU) from enrichment facilities, and Recovered Uranium (RU) obtained from the reprocessing of the spent fuel of light-water reactors (LWR). A fissile content in the RU of 0.9 to 1.0 % makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficient high neutron economy to use RU as fuel. RU from spent LWR fuel can be considered as a lower cost source of enrichment at the optimal enrichment level for CANDU fuel pellets. In Europe the feedstock of RU is approaching thousands tones and would provide sufficient fuel for hundreds CANDU-6 reactors years of operation. The use of RU fuel offers significant benefits to CANDU reactor operators. RU fuels improve fuel cycle economics by increasing the fuel burnup, which enables large cost reductions in fuel consumption and in spent fuel disposal. RU fuel offers enhanced operating margins that can be applied to increase reactor power. These benefits can be realized using existing fuel production technologies and practices, and with almost negligible changes to fuel receipt and handling procedures at the reactor. The application of RU fuel could be an important element in NPP Cernavoda from Romania. For this reason the Institute for Nuclear Research (INR), Pitesti has started a research programme aiming to develop a new fuel bundle RU-43 for extended burnup operation. The current version of the design is the result of a long process of analyses and improvements, in which successive preliminary design versions have been evaluated. The most relevant calculations performed on this fuel element design version are presented. Also, the stages of an experimental

  17. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

  18. The evolution of Candu fuel cycles and their potential contribution to world peace

    International Nuclear Information System (INIS)

    The Candu(r) reactor is the most versatile commercial power reactor in the world. It has the potential to extend resource utilization significantly, to allow countries with developing industrial infrastructures access to clean and abundant energy, and to destroy long-lived nuclear waste or surplus weapons plutonium. These benefits are available by choosing from an array of possible fuel cycles. Several factors, including Canada's early focus on heavy-water technology, limited heavy-industry infrastructure at the time, and a desire for both technological autonomy and energy self-sufficiency, contributed to the creation of the first Candu reactor in 1962. With the maturation of the CANDU industry, the unique design features of the now-familiar product - on-power refuelling, high neutron economy, and simple fuel design - make possible the realization of its potential fuel-cycle versatility. Several fuel-cycle options currently under development are described. (authors)

  19. R and D activities at INR pitesti related to safety and reliability of CANDU type fuel

    International Nuclear Information System (INIS)

    The focus of Nuclear Fuel R and D Program of Institute for Nuclear Research (INR) Pitesti is to maintain and improve the reliability, economics and safety of 37-element natural uranium CANDU fuel bundles in Cernavoda Nuclear Generating Station (CNGS). The second requirement is to improve the CANDU fuel design and to develop 43-element advanced fuel bundle that will reduce capital and fuelling cost, increase the operating and safety margins, improve natural - uranium utilization, and provide synergy with other reactor systems to improve resource utilization and spent fuel management. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history etc. has been obtained using in-pile measurements and PIE results of CANDU fuel elements irradiated in the TRIGA Material Testing Reactor (MTR) of INR Pitesti. In last time the data base was updated to include the results of Power Pulse Tests performed in TRIGA - Annular Core Pulse Reactor (ACPR) of INR Pitesti. One of the current research objective of our fuel bahaviour studies is to investigate the reliability behaviour of CANDU type fuel during power cycling operation condition. The INR research programme also include the out pile separate effects experiments to evaluate properties of the UO2 and cladding and development of computer models to describe sheath deformation and gas release processes. A program for LOCA simulating in-reactor tests is in progress at INR Pitesti to provide a database for verification of transient fuel performance codes and demonstrate that the significant fuel behaviour phenomena have all been included in the models.This data base is used extensively for the validation of the fuel behaviour codes. This paper summarizes R and D activities of INR Pitesti, related to safety and reliability of CANDU type fuel and presents some of the recent results obtained from in reactor tests. (author)

  20. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

    The CANDU 6 Pressurized Heavy Water Reactor system, featuring horizontal fuel channels and heavy water moderator continues to evolve, supported by AECL's strong commitment to comprehensive R and D programs. The initial CANDU 6 design started in the 1970's. The first plants went into service in 1983, and the latest version of the plant is under construction in China. With each plant the technology has evolved giving the dual advantages of proveness and modern technology. CANDU 6 delivers important advantages of the CANDU system with benefit to small and medium-sized grids. This technology has been successfully adopted by, and localized to varying extents in, each of the CANDU 6 markets. For example, all CANDU owners obtain their fuel from domestic suppliers. Progressive CANDU development continues at AECL to enhance this medium size product CANDU 6. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety. The CANDU 6 product is also enhanced by incorporating improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. (author)

  1. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  2. Proceedings of the 1. international conference on CANDU fuel handling systems

    International Nuclear Information System (INIS)

    Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately

  3. Depleting a CANDU-6 fuel assembly using detailed burnup data and reactionwise energy release

    International Nuclear Information System (INIS)

    Temporal behavior of reactor fuel assembly due to neutron exposure is an integral part of lattice analysis. It is important to estimate the production of actinides and fission products as a function of burnup so as to decide the quality of fuel for further energy production. It is also important from the point of view of post irradiation behavior of fuel. The information on heat production during and after irradiation helps in determining the amount of time a fuel assembly needs to be cooled before taking it up for storage or reprocessing. In the present study we have considered the CANDU-6 fuel assembly as reference. Lattice analysis has been performed using development version of code DRAGON. A total of 192 nuclides have been selected as part of the analysis, of which 19 are actinides, 151 are fission products and the rest are structural elements. The fission products have been treated explicitly. There is no pseudo fission product. Using DRAGR module, a multigroup microscopic cross section library in DRAGLIB format has been generated. An important aspect of this library is the explicit treatment of most neutron induced reactions. We have for the first time attempted to perform power normalization due to energy from various neutron induced reactions including (n, γ), (n, f), (n, 2n), (n, 3n), (n, 4n), (n, α), (n, p), (n, 2α), (n, np), (n, d), (n, t). Energy due to decay has also been considered explicitly. Even though the decay energy contributes very little relative to the neutron induced reactions, the information will be very useful for post irradiation behavior of fuel. It was observed that the maximum contributing reactions for the power normalization are (n, f), (n, γ) and (n, 2n). We have assessed the contribution of fission products and actinides towards power normalization as a function of burnup. We have also studied the pinwise contribution towards power normalization in each ring of CANDU-6 fuel. We have attempted to compare the effect of

  4. Assessment of neutron transport codes for application to CANDU fuel lattices analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    1999-08-01

    In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

  5. Simulation of CANDU Fuel Behaviour into In-Reactor LOCA Tests

    International Nuclear Information System (INIS)

    The purpose of this work is to simulate the behaviour of an instrumented, unirradiated, zircaloy sheathed UO2 fuel element assembly of CANDU type, subjected to a coolant depressurization transient in the X-2 pressurized water loop of the NRX reactor at the Chalk River Nuclear Laboratories in 1983. The high-temperature transient conditions are such as those associated with the onset of a loss of coolant accident (LOCA). The data and the information related to the experiment are those included in the OECD/NEA-IFPE Database (IFPE/CANDU-FIO-131 NEA-1783/01). As tool for this simulation is used the TRANSURANUS fuel performance code, developed at ITU, Germany, along with the corresponding fabrication and in-reactor operating conditions specific of the CANDU PHWR fuel. The results, analyzed versus the experimental ones, are encouraging and perfectible. (author)

  6. Destructive Examination of Experimental Candu Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post- irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  7. Post Irradiation Examination of Experomental CANDU Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  8. Optimized CANDU-6 cell and reactivity device supercell models for advanced fuels reactor database generation

    International Nuclear Information System (INIS)

    Highlights: • Propose an optimize 2-D model for CANDU lattice cell. • Propose a new 3-D simulation model for CANDU reactivity devices. • Implement other acceleration techniques for reactivity device simulations. • Reactivity device incremental cross sections for advanced CANDU fuels with thorium. - Abstract: Several 2D cell and 3D supercell models for reactivity device simulation have been proposed along the years for CANDU-6 reactors to generate 2-group cross section databases for finite core calculations in diffusion. Although these models are appropriate for natural uranium fuel, they are either too approximate or too expensive in terms of computer time to be used for optimization studies of advanced fuel cycles. Here we present a method to optimize the 2D spatial mesh to be used for a collision probability solution of the transport equation for CANDU cells. We also propose a technique to improve the modeling and accelerate the evaluation, in deterministic transport theory, of the incremental cross sections and diffusion coefficients associated with reactivity devices required for reactor calculations

  9. ORIGEN-S cross section libraries for CANDU used-fuel characterization

    International Nuclear Information System (INIS)

    A code system for producing burn-up dependent cross-section libraries for CANDU used-fuel characterization for use with the ORIGEN-S isotope generation and depletion code system is described. Benchmark results against experimental isotopic data for three CANDU-PHW reactor stations are presented. The code system couples the WIMS-AECL reactor physics analysis code with an ORIGEN-S depletion analysis to produce application-specific libraries that can be used in subsequent used-fuel analyses. 11 refs., 1 fig., 3 tabs

  10. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  11. Development of the advanced PHWR technology -Design and analysis of CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

  12. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G. [Inst. for Nuclear Research (INR), Pitesti (Romania); Palleck, S. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada); Ionescu, D. [Inst. for Nuclear Research (INR), Pitesti (Romania)

    2010-07-01

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  13. Romania Monte Carlo Methods Application to CANDU Spent Fuel Comparative Analysis

    International Nuclear Information System (INIS)

    Romania has a single NPP at Cernavoda with 5 PHWR reactors of CANDU6 type of 705 MW(e) each, with Cernavoda Unit1, operational starting from December 1996, Unit2 under construction while the remaining Unit3-5 is being conserved. The nuclear energy world wide development is accompanied by huge quantities of spent nuclear fuel accumulation. Having in view the possible impact upon population and environment, in all activities associated to nuclear fuel cycle, namely transportation, storage, reprocessing or disposal, the spent fuel characteristics must be well known. The paper aim is to apply Monte Carlo methods to CANDU spent fuel analysis, starting from the discharge moment, followed by spent fuel transport after a defined cooling period and finishing with the intermediate dry storage. As radiation source 3 CANDU fuels have been considered: standard 37 rods fuel bundle with natural UO2 and SEU fuels, and 43 rods fuel bundle with SEU fuel. After a criticality calculation using KENO-VI code, the criticality coefficient and the actinides and fission products concentrations are obtained. By using ORIGEN-S code, the photon source profiles are calculated and the spent fuel characteristics estimation is done. For the shielding calculations MORSE-SGC code has been used. Regarding to the spent fuel transport, the photon dose rates to the shipping cask wall and in air, at different distances from the cask, are estimated. The shielding calculation for the spent fuel intermediate dry storage is done and the photon dose rates at the storage basket wall (active element of the Cernavoda NPP intermediate dry storage) are obtained. A comparison between the 3 types of CANDU fuels is presented. (authors)

  14. Studies at INR-Pitesti for developing fuels of high burnup suitable to CANDU 6 reactor

    International Nuclear Information System (INIS)

    Increasing burnup allows the utility to get the same kWh output with a diminished tonnage of fissile material and provides a saving in the cost of fuel manufacturing as well as of spent fuel disposal. The RU, SEU, MOX, DUPIC fuel cycles and CANFLEX fuel bundles concept compatible with CANDU 6 reactor are presented. INR projects for developing SEU 43 fuel bundles supported by IAEA-Vienna are also presented. Particularly, one gives an overlook of standard CANDU and advanced SEU 43 nuclear fuel cycles. The paper presents also the current and future directions of studies implied by the research program in the nuclear fuel field of RAAN (The Autonomous Authority for Nuclear Activities). Among these, mentioned are: working out of the manual of physics of CANDU core with slightly enriched uranium; technological studies aiming at reducing the effects of limiting factors of the fuel lifetime and at burnup extension; obtaining new fuels as vectors of advanced cycles; off reactor tests of SEU 43 clusters; in-reactor tests of SEU 43 experimental fuel elements; developing computer codes for analysis of SEU, MOX and DUPIC fuel behavior; in-reactor tests of experimental MOX and DUPIC elements

  15. Study of advanced nuclear fuel cycles in Candu type power reactors

    International Nuclear Information System (INIS)

    The fuel burn up can be increased to a large extent, up to 14, 0000 MWD/te, by using the slightly enriched uranium or Pu mixed fuel in CANDU type power reactors. In the present study, the previous work was extended to compare the isotopic inventories and corresponding activities of important nuclides for different fuel cycles of a CANDU 600 type power reactor. The detail can be found in our studies. The calculations were performed using the computer code WIMSD4. The isotopic inventories and corresponding activities were calculated versus the fuel burn-up for the natural UO/sub 2/ fuel, 1.2 % enriched UO/sub 2/ fuel and 0.45 % PuO/sub 2/-UO/sub 2/ fuel. It was found that 1.2 % enriched uranium fuel has the lowest activity as compared to other two fuel cycles. It means that improvement in the fuel cycle technology of CANDU type power reactors can lead to high burn up which results in the reduction of actinide content in the spent fuel, and hence has a good environmental impact. (orig./A.B.)

  16. CANDU fuel cycle economic efficiency assessments using the IAEA-MESSAGE-V code

    International Nuclear Information System (INIS)

    The main goal of the paper is to evaluate different electricity generation costs in a CANDU Nuclear Power Plant (NPP) using different nuclear fuel cycles. The IAEA-MESSAGE code (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts) will be used to accomplish these assessments. This complex tool was supplied by International Atomic Energy Agency (IAEA) in 2002 at 'IAEA-Regional Training Course on Development and Evaluation of Alternative Energy Strategies in Support of Sustainable Development' held in Institute for Nuclear Research Pitesti. It is worthy to remind that the sustainable development requires satisfying the energy demand of present generations without compromising the possibility of future generations to meet their own needs. Based on the latest public information in the next 10-15 years four CANDU-6 based NPP could be in operation in Romania. Two of them will have some enhancements not clearly specified, yet. Therefore we consider being necessary to investigate possibility to enhance the economic efficiency of existing in-service CANDU-6 power reactors. The MESSAGE program can satisfy these requirements if appropriate input models will be built. As it is mentioned in the dedicated issues, a major inherent feature of CANDU is its fuel cycle flexibility. Keeping this in mind, some proposed CANDU fuel cycles will be analyzed in the paper: Natural Uranium (NU), Slightly Enriched Uranium (SEU), Recovered Uranium (RU) with and without reprocessing. Finally, based on optimization of the MESSAGE objective function an economic hierarchy of CANDU fuel cycles will be proposed. The authors used mainly public information on different costs required by analysis. (authors)

  17. Thermo-mechanical analysis of SEU 43 fuel element in CANDU reactor normal operational conditions

    International Nuclear Information System (INIS)

    The main direction of developing the CANDU type fuel is designing a new type of fuel cluster with the number of elements increased from 37 to 43. This work presents results of the research done in INR Pitesti on the new concept of SEU 43 fuel cluster designed for burnups as high as 25 Mw·day/kgU using slightly enriched uranium (up to 1.1% U235). By using ROFEM 1.0 code the behaviour of two types of SEU 43 fuel was analyzed in normal conditions of CANDU 6 reactor operation. The main performance parameters of SEU fuel were analyzed. These are: temperature distribution; the volume and pressure of fission gases; stresses in the fuel can; can deformations. Comparisons with the standard CANDU fuel are done. The results show the adequacy of the design solutions implemented for the SEU 43 fuel. The power-burnup history required by the ROFEM computations was obtained from the overpower envelope of the fuel cluster, with the radial power distribution on cluster taken into account. Evolution of the main performance parameters during irradiation is given

  18. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  19. Irradiation device for power cycling testing of CANDU type fuel elements

    International Nuclear Information System (INIS)

    At INR Pitesti an irradiation device (capsule-C9) was designed and realized for testing the fuel element behaviour at reactor power variations occurring during normal operation of CANDU reactors in load following regime. This device allows the study of the phenomena at which the fuel elements in CANDU reactor are subject in conditions of: - normal restarting after shutdown and reactor de-poisoning; - variations of reactor power within 50-100% rated power; - return to rated power after operation at reduced power to prevent xenon poisoning; - restart within 30 minutes from the shutdown to prevent xenon poisoning; - adjusting reactivity during the return to rated power after reactor operation at reduced power; - displacement of fuel clusters in the channel by reactor loading. The power cycling can entail failure mechanisms specific to reactor operation in load following regimes, such as: - deformation of fuel element can by fuel-can interaction; - stress crevice corrosion; - corrosion assisted can fatigue; - can thinning in the vicinity of cracked pellets; - can cracking due to relocation of pellet fragments. Power cycling on the fuel element subjected to irradiation in capsule C-9 is performed by displacing the tested section in the experimental channel. Displacing the tested section under flux allows obtaining the required power values on the tested fuel element while the capsule instrumentation allows the monitoring of irradiation parameters, namely: - the linear power on the fuel element; - instant neutron flux at the force tube level; - coolant pressure within the tested section; - coolant activity; - chemical characteristics of the coolant. The main thermal-hydraulic characteristics of the capsule C-9 are: - working fluid, demineralized and degassed water; - coolant pressure, 120 bar; - coolant temperature, 150-160 deg. C; - maximum temperature on fuel element can, 325 deg. C; - thermosyphon flow rate at the tested section level, 0.15 kg/s; - disposable maximum

  20. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  1. The parallex project: CANDU MOX fuel testing with weapons-derived plutonium

    International Nuclear Information System (INIS)

    The Parallex Project consists of a parallel experiment in which weapons-derived plutonium (WPu) from the United States and from the Russian Federation will be tested as mixed-oxide (MOX) CANDU fuel in the National Research Universal (NRU) reactor at the Chalk River Laboratories in Canada. Plutonium derived from excess weapons will be fabricated into CANDU MOX fuel at the A.A. Bochvar Institute in Moscow and at the Los Alamos National Laboratory in the United States. The MOX fuel will be transported to CRL, where it will be characterized, assembled into fuel bundles and then irradiated in the NRU reactor. Following irradiation, the fuel will be examined in hot cells to assess its irradiation performance. This paper describes the scope, rationale and current status of the Parallex Project. (author)

  2. The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core

    Energy Technology Data Exchange (ETDEWEB)

    Balaceanu, V. [Institute for Nuclear Research, Pitesti (Romania); Pavelescu, M. [Academy of Romanian Scientists, Bucharest (Romania)

    2010-05-15

    In order to reduce the total fuel costs for the CANDU reactors, mainly by extending the fuel burnup limits, some fuel bundle concepts have been developed in different CANDU owner countries. Therefore, in our Institute the SEU43 (Slightly Enriched Uranium with 43 fuel elements) project was started in early '90s. The neutronic behavior analysis of the CANDU core with SEU43 fuel was an important step in our project design. The objective of this paper is to highline an analysis of the neutronic behavior of the CANDU SEU43 core with the fuel burnup. More exactly, the study refers to the dependence of some global neutronic parameters, mainly the reactivity, on the fuel burnup. Two types of CANDU core were taken into consideration: reference core (without any reactivity devices) and perturbed core (with a strong reactivity system inserted). The considered reactivity system is the Mechanical Control Absorber (MCA) one. The performed parameters are: k{sub eff.} values, the MCA reactivity worth and flux distributions. The fuel bundles in the core are SEU43, with the fuel enrichment in U{sup 235} of 0.96% and at nominal power. For the fuel burnup the values are: 0.00 GWd/tU (fresh fuel); 8.00 GWd/tU and 25.00 GWd/tU. For reaching this objective, a global neutronic calculation system named WIMSPIJXYZ LEGENTR is used. Starting from a 69-groups ENDF/B-V based library, this system uses three transport codes: (1) the standard lattice-cell code WIMS, for generating macroscopic cross sections in supercell option and also for burnup calculations; (2) the PIJXYZ code for 3D simulation of the MCA reactivity devices and the 3D correction of the macroscopic cross sections; (3) the LEGENTR 3D transport code for estimating global neutronic parameters (CANDU core). The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON and DIREN codes. This comparison shows a good agreement between these results. (orig.)

  3. R and D activities on CANDU-type fuel in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, A.; Badruzzaman, M.; Latief, A. [Nuclear Fuel Element Centre, National Atomic Energy Agency of Indonesia (BATAN), Puspiptek, Serpong (Indonesia)

    1997-07-01

    The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

  4. International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the Embalse CANDU reactor

    International Nuclear Information System (INIS)

    In the last few years, Nucleoelectrica Argentina S.A. and Atomic Energy of Canada Limited have collaborated on a study of the technical feasibility of implementing Slightly Enriched Uranium (SEU) fuel in the Embalse CANDU reactor in Argentina. The successful conversion to SEU fuel of the other Argentine heavy-water reactor, Atucha 1, served as a good example. SEU presents an attractive incentive from the point of view of fuel utilization: if fuel enriched to 0.9% 235U were used in Embalse instead of natural uranium, the average fuel discharge burnup would increase significantly (by a factor of about 2), with consequent reduction in fuel requirements, leading to lower fuel-cycle costs and a large reduction in spent-fuel volume per unit energy produced. Another advantage is the change in the axial power shape: with SEU fuel, the maximum bundle power in a channel decreases and shifts towards the coolant inlet end, consequently increasing the thermalhydraulics safety margin. Two SEU fuel carriers, the traditional 37-element bundle and the 43-element CANFLEX bundle, which has enhanced thermalhydraulic characteristics as well as lower peak linear element ratings, have been examined. The feasibility study gave the organizations an excellent opportunity to perform cooperatively a large number of analyses, e.g., in reactor physics, thermalhydraulics, fuel performance, and safety. A Draft Plan for a Demonstration Irradiation of SEU fuel in Embalse was prepared. Safety analyses have been performed for a number of hypothetical accidents, such as Large Loss of Coolant, Loss of Reactivity Control, and an off-normal condition corresponding to introducing 8 SEU bundles in a channel (instead of 2 or 4 bundles). There are concrete safety improvements which result from the reduced maximum bundle powers and their shift towards the inlet end of the fuel channel. Further improvements in safety margins would accrue with CANFLEX. In conclusion, the analyses identified no issues that would

  5. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    International Nuclear Information System (INIS)

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles

  6. Fuel channel refilling

    International Nuclear Information System (INIS)

    Analysis of existing data on fuel channel refilling is presented. The analysis focuses on the data obtained using the Stern Laboratories Cold Water Injection Test (CWIT) Facility. The two-fluid model thermal-hydraulics computer code CATHENA is also used to simulate experimental results on fuel channel refilling from both the CWIT and RD-14 facilities. Conclusions related to single and double break tests, including the effect of non-condensible gases, are presented. A set of recommendations is given for further analysis and separate effect experiments. (67 figs., 5 tabs., 24 refs.)

  7. Development of CANDU spent fuel verification system using optical fiber scintillator

    International Nuclear Information System (INIS)

    In CANDU, spent fuels discharge 16∼24 bundles from the reactor core at everyday. Those are contained on the tray and the tray is stacked in the spent fuel. Currently, the Agency uses the CANDU Bundle Verifier for Stack (CBVS). It consists of a CZT gamma spectrometric probe which moves vertically along the space in between the columns of trays. Somewhat, spent fuel verification by non-destructive assay has been implemented for safeguards purpose using various radiation detectors such as a gas type detector, a semiconductor detector and so on. However, due to the severe circumstance of spent fuel storage such as high temperature, high radiation intensity and difficult to access area, the applicable radiation detectors and measurement techniques are very limited. An optical fiber scintillator has been known to have a good radiation hardness and physical properties for high temperature and humidity. In order to verify spent fuels stored in difficult to access area, KINAC designed and developed a prototype which was a spent fuel verification equipment using an optical fiber scintillator. The field test was performed at Wolsung NPP (Nuclear Power Plant) pond storage area. And this system will be made an entry for IAEA's verification equipment. For registration, KINAC would be followed the IAEA's QA procedure. At now, KINAC/IAEA developed the user, functional requirement and design specification for System, Hardware and Software separately. After finishing the procedure, it will be used for verification of spent fuel in lieu of CANDU Bundle Verifier for Baskets (CBVS). (author)

  8. Status of the parallex project--testing CANDU MOX fuel with weapons-derived plutonium

    International Nuclear Information System (INIS)

    The Parallex project is a parallel experiment, the purpose of which is to demonstrate the use of weapons-derived plutonium (WPu) from the United States (U.S.) and the Russian Federation (R.F.) in CANDU mixed-oxide (MOX) fuel elements. The scope of the project includes the fabrication of CANDU MOX fuel in the R.F. and the U.S., the transporting of the fuel to Canada, and the testing of the fuel in the NRU research reactor at the Chalk River Laboratories (CRL). A significant milestone in the project was achieved earlier this year, with the start of the irradiation testing of the MOX elements received from both the U.S. and the R.F. This paper presents the status of the project, and highlights the major activities that lead to the commencement of this historic experiment. (author)

  9. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Dragos; Pauna, Eduard [Institute for Nuclear Research (INR), Pitesti (Romania)

    2011-07-01

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the ME01 fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during the test period. Both LF tests were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets. This paper presents the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (orig.)

  10. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the ME01 fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during the test period. Both LF tests were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets. This paper presents the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (orig.)

  11. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  12. Experience in the manufacture and performance of CANDU fuel for KANUPP

    International Nuclear Information System (INIS)

    Karachi Nuclear Power Plant (KANUPP) a 137 MWe CANDU unit is In operation since 1971. Initially, it was fueled with Canadian fuel bundles. In July 1980 Pakistani manufactured fuel was introduced in the reactor core, irradiated to a burnup of about 7500 MWd-teU-1 and successfully discharged in May 1984. The core was progressively fuelled with Pakistani fuel and in August 1990 the reactor core contained all Pakistani made fuel. As of the present, 3 core equivalent Pakistani fuel bundles have been successfully discharged at an average bumup of 6500 MWd-teU-1. with a maximum burnup of ∼ 10,200 MWd-teU-1. No fuel failure of Pakistani bundles has been observed so far. This paper presents the indigenous efforts towards manufacture and operational aspects of KANUPP fuel and compares its behaviour with that of Canadian supplied fuel. The Pakistani fuel has performed well and is as good as the Canadian fuel. (author)

  13. Development of Romanian SEU-43 fuel bundle for CANDU type reactors

    International Nuclear Information System (INIS)

    SEU-43 fuel bundle is a CANDU type fuel consisting of two element sizes, to reduce element ratings, while maintaining the same bundle power, and an uranium content very close to the uranium content of a standard 37-element bundle. In order to reduce the detrimental effects of the life limiting factors at extended burnup a set of solution have been adopted for fuel element design. As a part of the design verification program, experimental bundles have been fabricated and utilized in typical out of reactor tests conducted at the laboratories of INR, Pitesti. These tests simulated current CANDU-6 reactor normal operating conditions of flow, temperature and pressure. The results are in accordance with the specified acceptance criteria. (author)

  14. Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU fuel

    International Nuclear Information System (INIS)

    The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. ((orig.))

  15. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  16. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

  17. Fuel cost analysis of CANDU-PHWR Wolsung Nuclear Power Plant unit 1

    International Nuclear Information System (INIS)

    Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1), currently under construction in Korea aiming at its completion in 1982. An attempt was also made for the sensitivity analysis of each fuel component; i.e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor. (author)

  18. Aging effect on the fuel behaviors for CANDU fuel safety analysis

    International Nuclear Information System (INIS)

    Because of the aging of heat transport system components, the reactor thermalhydraulic conditions can vary, which may affect the safety response. In a recent safety analysis for the refurbished Wolsong 1 NPP, various aging effects were incorporated into the hydraulic models of the components in the primary heat transport system (PHTS) for conservatism. The aging data of the thermal-hydraulic components for an 11 EFPY of Wolsong 1 were derived based on the site operation data and were modified to the appropriate input data for the thermal-hydraulic code for a safety analysis of a postulated accident. This paper deals with the aging effect of the PHTS of the CANDU reactor on the fuel performance during normal operation and transient period following a postulated accident such as a feeder stagnation break. (author)

  19. Aspects regarding the behaviour in operation of the fuel channel and spacer ring between the calandria tube and the pressure tube

    International Nuclear Information System (INIS)

    This work presents the applied solutions for achievement of pressure tube support, motivation for choosing solutions of fitting fuel channels in the case of CANDU-6 from Cernavoda NPP and some experimental test results. (Author) 5 Figs

  20. The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

    International Nuclear Information System (INIS)

    The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDUR* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

  1. The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Morreale, A. C.; Ball, M. R.; Novog, D. R.; Luxat, J. C. [Dept. of Engineering Physics, McMaster Univ., 1280 Main St. W, Hamilton, ON (Canada)

    2012-07-01

    The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

  2. Post irradiation tests on CANDU fuel irradiated in power ramp conditions

    International Nuclear Information System (INIS)

    Nuclear Research Branch Pitesti disposes of facilities, which allow the testing, manipulation and examination of nuclear fuel and of irradiated structure materials in CANDU reactor from Cernavoda NPP. These facilities imply the materials testing reactor TRIGA and the Post-Irradiation Examination Laboratory (PIEL). The purpose of this work is to determine by post-irradiate examination, the behavior of CANDU indigenous fuel, irradiated in 14 MW(th) TRIGA reactor into a multiple/various power ramp tests. The results of post-irradiate examination consist of: - Visual inspection and photography of the outer appearance of sheath; - Profilometry (diameter, bending, ovalization) and length measuring; - Determination of axial and radial distribution of the fission products activity by gamma scanning and tomography; - measurement of pressure, volume and isotopic composition of fission gas; - Microstructural characterization by metallographic and ceramographic analyzes; - Isotopic composition and burn-up determination; - Mechanical properties determination. The obtained data from the post-irradiate examinations are used, on one hand, in order to confirm the security, reliability and nuclear fuel performance, and on the other hand, for further optimization of the CANDU fuel. (authors)

  3. In-situ verification of CANDU spent fuel by the Cherenkov technique

    International Nuclear Information System (INIS)

    Multilayered and densely stacked irradiated CANDU fuel bundles in storage ponds make direct viewing of bundles in order to observe Cherenkov glow practically impossible. Ability to defect the source of Cherenkov glow by visual observation invariably suffers from subjective judgment. In the case of CANDU-type storage geometry, the difficulty in drawing conclusions is even greater for a number of reasons including the near neighbour effect. In this paper, the first results of Cherenkov photographic procedure without the isolation of individual trays are presented in which a new model of the Hungarian underwater telescope in combination with a lightmeter for Cherenkov intensity measurements has been used. It is demonstrated by this technique that photographs of bundles with cooling time of up to 2 a provide a satisfactory record for conclusive attribute verification result for irradiated fuel bundles stacked in multilayers. A distinct glow, with a brightness of higher intensity between the rod of a bundle compared to the surroundings of the bundle, is clearly shown by the pictures. Based on the results of the glow intensity measurements, the use of this photographic method for fuel bundles with longer cooling time of up to 15 a or more would require considerably longer exposure times or more sensitive film. Possible impact on IAEA safeguards of CANDU spent fuel bays by a system, which offers simultaneous item counting and NDA attribute test capabilities in a relatively low intrusive manner, is discussed. The limitations are also considered. (author)

  4. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D2O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

  5. Neutron measurements for CANDU-type fuel characterization at TRIGA-ACPR

    International Nuclear Information System (INIS)

    In order to measure the parameters for a CANDU cell it is important to determine thermal flux distribution in the cell, spectral indices (absolute values and distributions). If it has to be done in a critical assembly, the small flux value poses problems to the measurements. Also the measurements performed with detectors placed into the bundle have to be treated with care. In order to test the methods related to such measurements we decided to perform the most important of them using a CANDU bundle placed on the experiment loading tube of TRIGA-ACPR. The reactor was operated in stationary regime to give the necessary thermal flux at the experiment position. A set of activation and fissionable foil detectors was used and measurements for absolute reaction rate determinations, thermal flux distributions and spectral indices absolute values were performed. Also the sensitivity to heterogeneous poisoning of the bundle was measured in the same configuration. The sensitivity of the ACPR to heterogeneous poisoning of a CANDU - bundle placed in central hole is also determined. In conclusion a set of methods for neutronic measurements on CANDU type fuel were tested in TRIGA-ACPR, in spectral conditions which can be considered worse than in a D2O lattice. The source of error was investigated in detail. One can conclude that these methods will work in measurements upon a D2O - natural uranium lattice

  6. Study of CANDU thorium-based fuel cycles by deterministic and Monte Carlo methods

    International Nuclear Information System (INIS)

    In the framework of the Generation IV forum, there is a renewal of interest in self-sustainable thorium fuel cycles applied to various concepts such as Molten Salt Reactors [1, 2] or High Temperature Reactors [3, 4]. Precise evaluations of the U-233 production potential relying on existing reactors such as PWRs [5] or CANDUs [6] are hence necessary. As a consequence of its design (online refueling and D2O moderator in a thermal spectrum), the CANDU reactor has moreover an excellent neutron economy and consequently a high fissile conversion ratio [7]. For these reasons, we try here, with a shorter term view, to re-evaluate the economic competitiveness of once-through thorium-based fuel cycles in CANDU [8]. Two simulation tools are used: the deterministic Canadian cell code DRAGON [9] and MURE [10], a C++ tool for reactor evolution calculations based on the Monte Carlo code MCNP [11]. (authors)

  7. CANDU fuel attribution through the analysis of delayed neutron temporal behaviour

    International Nuclear Information System (INIS)

    Delayed Neutron Counting (DNC) is an established technique in the Canadian nuclear industry as it is used for the detection of defective fuel in several CANDU reactors and the assay of uranium in geological samples. This paper describes the possible expansion of DNC to the discipline of nuclear forensics analysis. The temporal behaviour of experimentally measured delayed neutron spectra were used to determine the relative contributions of 233U and 235U to the overall fissile content present in mixtures with average absolute errors of ±4 %. The characterization of fissile content in current and proposed CANDU fuels (natural UO2, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations. (author)

  8. Measurement of gap and grain-boundary inventories of 129I in used CANDU fuels

    International Nuclear Information System (INIS)

    Combined gap and grain-boundary inventories of 129I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of 128I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of 44 kW/m), the 129I values were considerably smaller than expected. The combined gap and grain-boundary inventories of 129I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 ± 2.4% which suggests that the average value of 8.1 ± 1% used in safety assessment calculations overestimates the instant release fraction for 129I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of 129I release. Results could be fitted tentatively to half-order reaction kinetics, implying that 129I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of 129I from used CANDU fuel

  9. Analytical and experimental assessment of CANDU fuel sheath integrity under post dryout conditions

    International Nuclear Information System (INIS)

    The experiments that investigated the CANDU fuel sheath behavior under different pressures, temperatures, oxidizing environment, material structure (as-received or thermally treated to attach appendages), and heating rates were reviewed and assessed to determine the limits of post-dryout duration, sheath temperature, and pressure difference across the sheath required to ensure the fuel sheath integrity. A number of burst curves at different heating rates were studied. Time-at- temperature fuel sheath failure maps were developed based on temperature ramp and isothermal experiments for the 28-element fuel bundle. Analytical time-at-temperature fuel sheath failure maps were also developed for both of 28- and 37-element fuel bundles using the ELOCA fuel analysis computer code and were compared to the experimental time-at-temperature sheath failure maps. Time-at-temperature sheath failure maps could be used as a simple and effective screening tool to demonstrate fuel sheath integrity during postulated design basis accident. (author)

  10. Fuel management simulations for 0.9% SEU in CANDU 6 reactors

    International Nuclear Information System (INIS)

    Slightly Enriched Uranium (SEU) of 0.9 weight % 235U enrichment is a promising fuel cycle option for CANDU reactors. An important component of the investigation of this option is the demonstration of the feasibility of on-line refuelling with this fuel type in reactor physics fuel-management simulations. Two fuel-management schemes have been investigated in detail during 500-day core-follow simulations, these were a 2-bundle-shift and a 4-bundle-shift axial refuelling scheme. The 43-element CANFLEX fuel design has been used in these studies because of its improved fuel performance characteristics in this application. The results of the studies are discussed in detail in this paper. The most significant conclusion of this study was that both 2- and 4-bundle-shift refuelling schemes with CANFLEX fuel result in bundle power and bundle power boost envelopes that meet current fuel-performance requirements. (author)

  11. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K., E-mail: Kyle.Gamble@rmc.ca [Royal Military College of Ontario, Kingston, Ontario (Canada); Williams, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  12. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    International Nuclear Information System (INIS)

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  13. A finite element model for static strength analysis of CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.V. [Institute for Nuclear Research, Pitesti (Romania)

    2006-08-15

    A static strength analysis finite-element model has been developed using the ANSYS computer code in order to simulate the axial compression in CANDU type fuel bundle subject to hydraulic drag loads, deflection of fuel elements and stresses and displacements in the end plates. The validation of the finite-element model has been done by comparison with the out-reactor strength test results. Comparison of model predictions with the experimental results showed very good agreement. The comparative assessment reveals that SEU43 and SEU43L fuel bundles are able to withstand high flow rate without showing a significant geometric instability. (orig.)

  14. Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Serbanel, M. [Institute for Nuclear Research, Pitesti (Romania); Diaconu, C.

    2012-11-15

    Presently, INR Pitesti is developing an advanced fuel design RU-43LV (recovered uranium fuel bundle with 43 elements and low void reactivity feature) based on recovered uranium from LWR. Compared with the current design of 37 natural uranium element (NU-37) fuel bundle, RU-43LV will have higher power capability and higher burn-up potential in CANDU reactors of Cernavoda-Romania Nuclear Power Plant (NPP). Fuel burn-up of RU-43LV fuel will be about two times the burn-up usually achieved in CANDU reactors fuelled with natural uranium fuel. The effect of the design changes of RU-43LV bundle on the reactor safety has been analyzed and the results are presented in this paper. As part of the conceptual design study, the performance of the RU-43LV fuelled core during a large loss-of-coolant accident (LLOCA) was assessed with the use of several computer codes. The most relevant calculations performed regarding RU-43 RV fuel safety are presented in this paper. Also, the stages of an experimental program aiming to study RU-43LV fuel behaviour in high temperature transients are briefly described. (orig.)

  15. Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Patrulescu, I. [Inst. for Nuclear Research, Pitesti (Romania). Physics

    2008-03-15

    Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. A fissile content in the RU of 0.9 to 1.0% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. The Institute for Nuclear Research (INR) Pitesti has analyzed the feasibility of using RU fuel with 0.9-1.1 w% {sup 235}U in the CANDU-6 reactors of the Cernavoda Nuclear Power Plant (Cernavoda NPP). Using RU fuel would produce a significant increase in the fuel discharge burnup, from 170 MWh/kgU currently achieves with natural-uranium (NU) fuel to about 355 MWh/kgU. This would lead to reduced fuel-cycle cost and a large reduction in spent-fuel volume per full-power-year of operation. The RU fuel bundle design with recovered uranium fuel, known as RU-43, is being developed by the INR Pitesti and is now at the stage of final design verification. Early work has been concentrated on RU-43 fuel bundle design optimization, safety and reactor physics assessment. The changes in fuel element and fuel bundle design contribute to the many advantages offered by the RU-43 bundle. Verification of the design of the RU-43 fuel bundle is performed in a way that shows that design criteria are met, and is mostly covered by proof tests such as flow and irradiation tests. The most relevant calculations performed on this fuel bundle design version are presented. Also, the stages of an experimental program aiming to verify the operating performance are briefly described in this paper. (orig.)

  16. Fuel channel replacement experience at Wolsong Unit 1

    International Nuclear Information System (INIS)

    The Wolsong Nuclear Power Station is located on the shore of the Eastern Sea near the ancient capital city of Kyongju in the Republic of Korea. Wolsong Unit 1 is the only operating CANDU 6 Pressurized Heavy Water Reactor at the four unit station owned and operated by the Korea Electric Power Corporation (KEPCO). The other three CANDU 6 units are under various stages of construction and all three are scheduled to be in-service during the 1997-1999 period. The Wolsong 1 fuel channels were subjected to periodic inspections in accordance with the regulatory requirements and the inspection findings revealed acceptable indications until 1990. The debris induced wear marks identified during pressure tube inspections in 1990, 1992, and 1994 led to the decision to replace fuel channels E11, M11, and O08 during the scheduled plant outage in 1994. The three fuel channels were replaced in 26.5 days (8.8 days per channel) and a total of 75.8 Rem (25.3 Rem per channel) dose was expended. Unshielded radiation fields on reactor face ranged from 450 mRem/hour on contact tithe the end fitting face to 2000 mRem/hour on contact with the feeder ports. The work was performed on two twelve hour shifts involving 20 technicians and engineers from AECL; and 4-4 engineers, technicians, and tradesmen from KEPCO and KPS. (author)

  17. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - II: DUPIC Fuel-Handling Cost

    International Nuclear Information System (INIS)

    The Direct Use of spent Pressurized water reactor fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel-handling technique has been investigated through a conceptual design study to estimate the unit cost that can be used for the DUPIC fuel cycle cost calculation. The conceptual design study has shown that fresh DUPIC fuel can be transferred to the core following the existing spent-fuel discharge route, provided that new fuel-handling equipment, such as the manipulator, opening/sealing tool of shipping casks, new fuel magazine, new fuel ram, dryer, gamma-ray detector, etc., are installed. The reverse path loading option is known to minimize the number of additional pieces of equipment for fuel handling, because it utilizes the existing spent-fuel handling equipment, and the discharge of spent DUPIC fuel can be done through the existing spent-fuel handling system without any modification. However, because the decay heat of spent DUPIC fuel is much higher than that of spent natural uranium fuel, the extra cooling capacity should be supplemented in the spent-fuel storage bay. Based on the conceptual design study, the capital cost for DUPIC fuel handling and extra storage cooling capacity was estimated to be $3 750 000 (as of December 1999) per CANDU plant. The levelized unit cost of DUPIC fuel handling was then obtained by considering the amount of fuel that will be required during the lifetime of a plant, which is 5.13 $/kg heavy metal. Compared with the other unit costs of the fuel cycle components, it is expected that DUPIC fuel handling has only a minor effect on the overall fuel cycle cost

  18. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  19. Investigations on flow induced vibration of simulated CANDU fuel bundles in a pipe

    International Nuclear Information System (INIS)

    In this paper, vibration of a two-bundle string consisting of simulated CANDU fuel bundles subjected to turbulent liquid flow is investigated through numerical simulations and experiments. Large eddy simulation is used to solve the three-dimensional turbulent flow surrounding the fuel bundles for determining fluid excitations. The CFD model includes pipe flow, flow through the inlet fuel bundle along with its two endplates, half of the second bundle and its upstream endplate. The fluid excitation obtained from the fluid model is subsequently fed into a fuel bundle vibration code written in FORTRAN. Fluid structure interaction terms for the fuel elements are approximated using the slender body theory. Simulation results are compared to measurements conducted on the simulated fuel bundles in a testing hydraulic loop. (author)

  20. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  1. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  2. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Dobrea, D.; Parvan, M.; Stefan, V. [Institute for Nuclear Research, Pitesti (Romania)

    2009-04-15

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g{sup -1} UO{sub 2}. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  3. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.; Olteanu, G. [Inst. for Nuclear Research, Pitesti (Romania)

    2008-07-01

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g{sup -1} UO{sub 2}. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  4. Quantification of factors affecting thermally-induced bow in a CANDU fuel element simulator

    International Nuclear Information System (INIS)

    Thermally induced bow, caused by a circumferential temperature distribution around a fuel element, was investigated in this study using a fuel element simulator. The objective was to identify the factors affecting CANDU fuel element bow induced by dryout as a result of some predicted reactor transients in which the maximum fuel temperature reaches 600 deg C. The results showed that circumferential temperature distribution, pellet-to-sheath mechanical interaction and creep were the major factors affecting bow. Transient bow increased with increasing diametral sheath temperature difference and with mechanical interaction between the pellet and the sheath. Permanent bow of the fuel element was observed in some tests which was the result of creep. Mechanical interaction between the sheath and pellet produced the stresses necessary for creep deformation. A simplified ABAQUS model was developed to explain the experimental findings and could be used to predict the bow behaviour of fuel elements during reactor transients, where the dry patches are of different sizes. (author)

  5. Subchannel Analysis for enhancing the fuel performance in CANDU reactor

    International Nuclear Information System (INIS)

    The effect of the fuel rod geometry in a fuel bundle using the subchannel code ASSERT has been evaluated to design the fuel bundle having the advanced fuel performance. Based on the configuration of standard 37-element fuel bundle, the element diameter of fuel rods in each ring has been changed while that of fuel rods in other rings is kept as the original size. The dryout power of each element in a fuel bundle has been obtained for the modified fuel bundle and compared with that of a standard fuel bundle. From the calculated mixture enthalpy and void fraction of each subchannel, it was found that the modification of element diameter largely affects to the thermal characteristics of the subchannel on the upper region of a modified element by the buoyancy drift effect. The optimized geometry in a fuel bundle has been suggested from the consideration of the change of void reactivity as well as the dryout power of a bundle. The dependency of the transverse interchange model on the present results has been checked by examining the dryout power of a bundle for the different mixing coefficient and buoyancy drift model

  6. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  7. Thermal-hydraulics performance optimization of Candu fuel using Assert subchannel code

    International Nuclear Information System (INIS)

    An optimization of fuel bundle geometry using the subchannel code ASSERT is performed in support of Candu fuel design to enhance the thermohydraulics performance. The new bundle design is based on a reference CANFLEX bundle with changes to the centre and inner-ring element diameters and pitch-circle diameters (PCDs) of various element rings. Different methods of varying the PCDs for reaching the optimized geometry are considered in an attempt to minimize the optimization effort. The optimized geometry in the present analysis is the one that maximizes the dryout power and that has simultaneous CHF (critical heat flux) initiation involving more than one subchannel rings. (authors)

  8. Fission gas release of (Th, Pu)O{sub 2} CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Karam, M.; Dimayuga, F.C.; Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2008-07-01

    The use of thorium was identified, as early as the 1950s, as a promising fuel cycle in the CANDU development program because of its expected improved fuel performance (e.g., reduced fission gas release (FGR) due to thoria's enhanced thermal and chemical properties) and the relative abundance of thorium. AECL maintains an ongoing R and D program on thorium within the Advanced Fuel Cycles Program, which covers various aspects of the thorium fuel cycle including fabrication, irradiation testing, post-irradiation examination (PIE), and assessments of fuel performance. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type 37-element bundles fuelled with (Th, Pu)O{sub 2} pellets. Fuel fabrication was conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of mixed oxide thoria fuel. The fuel pellets contained 1.53 wt. % Pu in (Th, Pu)O{sub 2}. The six bundles were irradiated in the NRU reactor loops under CANDU normal operating conditions to burnups ranging from about 450 MWh/kgHE (19 MWd/kgHE) to 1181 MWh/kgHE (49 MWd/kgHE) with peak element linear power ratings ranging from 52 kW/m to 73 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are currently being analyzed. FGR of bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE) ranged from 1% to 5%. These FGR values are significantly lower than those observed for CANDU UO{sub 2} and (U, Pu)O{sub 2} fuel with similar power histories. The low FGR is attributed to low operating fuel temperatures resulting from high thermal conductivity of thoria. This demonstrates excellent performance of (Th, Pu)O{sub 2} fuels compared to UO{sub 2}. This paper focuses on FGR results for (Th, Pu)O{sub 2} fuel irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE). (author)

  9. FEAT4.1: modeling of sheath oxidation and heat flow in CANDU fuel elements

    International Nuclear Information System (INIS)

    This paper describes recent developments in the AECL-developed computer program, FEAT (Finite Element Analysis for Temperature), which is used to assess the thermal integrity of CANDU ® fuel elements. The FEAT code is used to calculate temperatures in the fuel pellet and in the Zircaloy sheath of a CANDU fuel element under normal operating conditions (NOC), as well as the temperature peaking due to end flux peaking during a transient such as a postulated loss of coolant accident (LOCA). For normal operation of high burnup fuel, the Zircaloy oxidation effect on fuel temperatures needs to be considered due to the long residence time in the reactor. The oxide layer on the coolant side of the fuel sheath has a lower thermal conductivity than that of Zircaloy. Therefore, the heat flow from the fuel element to coolant will be reduced resulting in increased fuel pellet and sheath temperatures. To ensure that the FEAT code is suitable for application in analysis of advanced fuels such as the ACR®-1000 fuel, a number of model developments and code improvements were conducted based on the existing version FEAT 4.0, including the modeling of sheath oxidation and its effect on heat flow in the fuel element, time-dependence of end flux peaking during the postulated LOCA (Loss Of Coolant Accident) conditions, pre-processing and postprocessing of analysis data. This paper describes the theories for the models, as well as other improvements, and verification and validation of the new FEAT version (i.e., FEAT 4.1). (author)

  10. Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

    International Nuclear Information System (INIS)

    The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new

  11. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    International Nuclear Information System (INIS)

    The spent fuel bay cooling and purification system for Wolsong Nuclear Power Plant (NPP) Units 2, 3 and 4 was designed to remove heat from the spent fuel bay generated by 10 years accumulation of spent fuel at an 80% capacity factor refueling rate plus an emergency discharge of one-half the core fuel inventory over a 20-day period for 25.5 .deg. C of the cooling sea water temperature. The heat load in the spent fuel bay depends on the capacity factor refueling rate and the amount of spent fuel accumulated at the spent fuel bay. An 80% capacity factor refueling rate was considered as a design condition, but the highest capacity factor refueling rate of 93.75% for Wolsong NPPs was calculated based on nine (9) years of operating experience from 2000 to 2008. For the abnormal operating condition, the operating temperature of spent fuel bay does not meet with the acceptance criterion of 49 .deg. C for the conditions of the capacity factor refueling rate of 93.75%. These operating modes are not recommended for the abnormal operating condition

  12. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Mann; Jang, Ho Cheol; Jang, Jin A.; Kim, Eun Kee [KEPCO Engineering and Construction Company, Daejeon (Korea, Republic of); Park, WanGyu [KHNP, Uljingun (Korea, Republic of)

    2015-05-15

    The spent fuel bay cooling and purification system for Wolsong Nuclear Power Plant (NPP) Units 2, 3 and 4 was designed to remove heat from the spent fuel bay generated by 10 years accumulation of spent fuel at an 80% capacity factor refueling rate plus an emergency discharge of one-half the core fuel inventory over a 20-day period for 25.5 .deg. C of the cooling sea water temperature. The heat load in the spent fuel bay depends on the capacity factor refueling rate and the amount of spent fuel accumulated at the spent fuel bay. An 80% capacity factor refueling rate was considered as a design condition, but the highest capacity factor refueling rate of 93.75% for Wolsong NPPs was calculated based on nine (9) years of operating experience from 2000 to 2008. For the abnormal operating condition, the operating temperature of spent fuel bay does not meet with the acceptance criterion of 49 .deg. C for the conditions of the capacity factor refueling rate of 93.75%. These operating modes are not recommended for the abnormal operating condition.

  13. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    International Nuclear Information System (INIS)

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  14. Understanding CANDU fuel bowing in dryout: an industry approach

    International Nuclear Information System (INIS)

    Fuel element bow induced by dryout could potentially perturb the coolant flow distribution and heat transfer from the fuel element to the coolant. Some accident scenarios leading to dryout of the fuel element are: loss of power regulation pump trip, pump seizure, small and large break loss of coolant accidents. In these accidents, it is desirable to show with confidence that the fuel remains sufficiently cooled to maintain its geometry, even if it is in dryout. This can be demonstrated if fuel elements are separated from each other and from the pressure tube, with a sufficient (and stable) gap. Therefore, the prediction of the amount of bow, and its effect on heat transfer conditions is required for the assessments. The utilities have joined force in launching an experimental investigation at Stern Laboratories to characterize the bowing phenomena. This program will investigate the amount of deflection, transient and permanent, that results from accident conditions which cause a dry patch on one side of the sheath. This is expected to bound the consequences of fuel bowing due to dryout. Since the accident transients begin at full power and high coolant pressure (about 10 MPa) they generate sharp thermal gradients (dry patch) and it is necessary to develop a simulation with representative dry fuel sheath conditions initiated from a normal full power and coolant state. The amount of bow is driven by thermal gradients in both the fuel pellets and the sheath, therefore, the thermal gradients should be representative. This program is structured in a series of tests progressing from simple representation to complex simulation. It is divided into 3 experimental phases: Phase 1 - Thermalhydraulic simulation of fuel element bow by a heated tube; Phase 2 - Thermal and mechanical bow with a simulator which accounts for pellet / fuel sheath interaction with internal pellet temperature distributions; and Phase 3 - Fuel element bow with a simulator using Zircaloy-4 fuel sheath

  15. Enrichment effects on CANDU-SEU spent fuel Monte Carlo shielding analysis

    International Nuclear Information System (INIS)

    Shielding analyses are an essential component of the nuclear safety, the estimations of radiation doses in order to reduce them under specified limitation values being the main task here. According to IAEA data, more than 10 millions packages containing radioactive materials are annually transported world wide. All the problems arisen from the safe radioactive materials transport assurance must be carefully settled. Last decade, both for operating reactors and future reactor projects, a general trend to raise the discharge fuel burnup has been recorded world wide. For CANDU type reactors, the most attractive solution seems to be SEU and RU fuels utilization. The basic tasks accomplished by the shielding calculations in a nuclear safety analysis consist in dose rates calculation, to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper aims to study the effects induced by fuel enrichment variation on CANDU-SEU spent fuel photon dose rates for a Monte Carlo shielding analysis applied to spent fuel transport after a defined cooling period in the NPP pools. The fuel bundles projects considered here have 43 Zircaloy rods, filled with SEU fuel pellets, the fuel having different enrichment in U-235. All the geometrical and material data related on the cask were considered according to the shipping cask type B model. After a photon source profile calculation by using ORIGEN-S code, in order to perform the shielding calculations, Monte Carlo MORSE-SGC code has been used, both codes being included in the ORNL's SCALE 5 system. The photon dose rates to the shipping cask wall and in air, at different distances from the cask, have been estimated. Finally, a photon dose rates comparison for different fuel enrichments has been performed. (author)

  16. Enhanced CANDU 6 (EC6): a proven mid-sized reactor with fuel cycle capability

    International Nuclear Information System (INIS)

    Atomic Energy of Canada (AECL) is finalizing development of the Enhanced CANDU 6 (EC6), which incorporates the CANDU 6's well-proven features, and enhancements that make the reactor even more safe and easier to operate. The EC6 is the only mid-sized reactor (700 MWe class) with a proven pedigree that meets modern reactor expectations and regulatory standards. It is sized for smaller grids and also has outstanding fuel-cycle capability. Changes are incremental and consistent with the CANDU 6 project approach. The EC6 utilizes modern computers and a distributed control system housed in an advanced control room which, along with automated testing and on-line diagnostics, make the plant easier and safer to operate, with minimal operator intervention. Containment and seismic capability are upgraded to meet modern standards. The first deployment of the EC6 is anticipated in Canada; international markets are also being pursued. AECL is performing a comprehensive review of the EC6 design in the wake of the Fukushima accident, will review lessons learned, and incorporate any necessary improvements into new build design. (author)

  17. 'CANDU-fueling machine head tests' at the Institute for Nuclear Research - Pitesti

    International Nuclear Information System (INIS)

    The Fueling Machine (F/M) Head is the most complex equipment of the Fuel Handling System in the CANDU reactor and performs the change of the nuclear fuel during the reactor operation. Before the installation of the F/M Head at the Nuclear Power Plant, it was required to test its technical performances, to ensure that the equipment is ready for operation. Testing of the F/M Head at the Institute for Nuclear Research - Pitesti is a part of the overall program to assimilate in Romania the CANDU technology. There was an economic contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 untested. To perform testing of these machines at the Institute for Nuclear Research - Pitesti, a special testing rig was built and is available for this goal. Both the testing rig and staff have been successfully assessed by the AECL representatives during two visits, dated on December 2001 and March 2002. In 2003 the testing of the F/M Head no. 4 (RAM 5) was successfully completed. Today, in 2004, the functional test of the F/M Head no. 5 (RAM 6) is already performing. (authors)

  18. Data base for a CANDU-PHW operating on a once-through, natural uranium fuel cycle

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

  19. CANDU passive shutdown systems

    International Nuclear Information System (INIS)

    CANDU incorporates two diverse, passive shutdown systems (Shutdown System No. 1 and Shutdown System No. 2) which are independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels; the shutdown systems do not penetrate the heat transport system pressure boundary. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive

  20. Assessment of Welding System Modification of The Candu and PWR Fuel Element Types end Plug

    International Nuclear Information System (INIS)

    To anticipate future possibility of a nuclear fuel element industry in Indonesia, research on other types of nuclear fuel element beside Cirene type has to be done. It can be accomplished, one of them, by modifying the already available equipment. Based on the sheath material, the sheath dimension and the welding process parameters such as welding current and welding cycles, the available Magnetic Force Welding can be used for welding end plug of Candu nuclear fuel element by modifying some of its components (tube clamp, plug clamp, etc). The available Pellet drying and element filling furnace with its supporting system with includes helium gas filling, welding chamber, argon gas supply, vacuum system, sheath clamp and sheath driving system can be used for welding end plug with sheath of PWR nuclear fuel element by adding og Tungsten inert Gas (TIG) welding machine in the welding chamber and modifying a few components (seal clamp, sheath clamp)

  1. An experimental investigation of the temperature behavior of a CANDU 37-element spent fuel bundle with air backfill

    International Nuclear Information System (INIS)

    As part of the thermal analysis of a CANDU spent fuel dry storage system, a series of experiment has been conducted using a thermal mock-up of a simulated CANDU spent fuel bundle in a dry storage basket. The experimental system was designed to obtain the maximum fuel rod temperature along with the radial and axial temperature distributions within the fuel bundle. The main purpose of these experiments was to characterize the relevant heat transfer mechanisms in a dry, vertically oriented CANDU spent fuel bundle, and to verify the MAXROT code developed for the thermal analysis of a CANDU spent fuel bundle in a dry storage basket. A total of 48 runs were made with 8 different power inputs to the 37-element heater rod bundle ranging from 5 to 40 W, while using 6 different band heaters power inputs from 0 to 250 W to maintain the basket wall at a desired boundary condition temperature at the steady state. The temperature distribution in a heater rod bundle was measured and recorded at the saturated condition for each set of heater rod power and band heaters power. To characterize the heat transfer mechanism involved, the experimental data were corrected analytically for radiation heat transfer and presented as a Nusselt number correlation in terms of the Rayleigh number of the heater rod bundle. The results show that the Nusselt number remains nearly constant and all the experimental dada fall within a conduction regime. The experimental data were compared with the predictions of the MAXROT code to examine the code's accuracy and validity of assumptions used in the code. The MAXROT code explicitly models each representative fuel rod in a CANDU fuel bundle and couples the conductive and radiative heat transfer of the internal gas between rods. Comparisons between the measured and predicted maximum fuel rod temperatures of the simulated CANDU 37-element spent fuel bundle for all 48 tests show that the MAXROT code slightly over-predicts and the agreement is within 2

  2. Public health risks associated with the CANDU nuclear fuel cycle

    International Nuclear Information System (INIS)

    This report has been prepared in the hope that it will calculate, apparently for the first time, the non-radiological risks associated with the use of nuclear fuels. The specific risks identified and evaluated in this work should be balanced against the benefits resulting from the use of nuclear fuels or against the risks inherent in other fuels. Due to lack of sufficient data in certain areas the results obtained are subject to a large degree of uncertainty and therefore the results indicate an order of magnitude rather than exact values of hazard. The total hazard can be expressed as 6.0 ± 4.8 x 10-3 fatalities and 4.8 ± 0.7 x l0-2 injuries per 1 GWy of electricity produced

  3. The CANDU irradiated fuel safeguards sealing system at the threshold of implementation

    International Nuclear Information System (INIS)

    The development of a safeguards containment and surveillance system for the irradiated fuel discharged from CANDU nuclear generating stations has inspired the development of three different sealing technologies. Each seal type utilizes a random seal identity of different design. The AECL Random Coil (ARC) Seal combines the identity and integrity elements in the ultrasonic signature of a wire coil. Two variants of an optical seal have been developed which features identity elements of crystalline zirconium and aluminum. The sealed cap-seal uses a conventional IAEA 'Type X Seal' (wire seal). The essential features and relative merits of each seal design are described

  4. ITER SAFETY TASK NID-10A:CANDU occupational exposure experience: ORE for ITER fuel cycle and cooling systems

    International Nuclear Information System (INIS)

    This report contains information on TRITIUM Occupational Exposure (Internal Dose) from typical CANDU Nuclear Generating Stations. In addition to dose, airborne tritium levels are provided, as these strongly influence operational exposure. The exposure dose data presented in this report cover a period of five years of operation and maintenance experience from four CANDU Reactors and are considered representative of other CANDU reactors. The data are broken down according to occupational function ( Operators, Maintenance and Support Service etc.). The referenced systems are mainly centered on CANDU Hear Transport System, Moderator System, Tritium Removal Facility and Heavy Water (D20) Upgrading System. These systems contain the bulk part of tritium contamination in the CANDU Reactor. Because of certain similarities between ITER and CANDU systems, this data can be used as the most relevant TRITIUM OCCUPATIONAL DOSE information for ITER COOLING and FUEL CYCLE systems dose assessment purpose, if similar design and operation principles as described in the report are adopted. (author). 16 refs., 8 tabs., 13 figs

  5. Neutronic performance of ({sup Reprocessed}U/Th)O{sub 2} fuel in CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gholamzadeh, Z. [Talca Univ. (Chile). Dept. of Energy; Mirvakili, S.M. [Nuclear Science and Technology Research Institute, Reactor Research School, Tehran (Iran, Islamic Republic of); Feghhi, S.A.H. [Shahid Beheshti Univ., Dept. of Radiation Application, Tehran (Iran, Islamic Republic of)

    2015-07-15

    Utilization of thorium-based fuel in different reactors has been under investigation for several decades. In fact, excellent breeding features, rather flattened distribution of power as well as proliferation resistance of such fuel cycle draws the attention towards utilization of this type of fuel in nuclear power technology. In the present study, the neutronic performance of a typical thorium core loading is addressed. In this configuration, a mixed uranium and thorium oxide is loaded in CANDU 6 reactor. The obtained results determine a total peaking factor of 2.73 for the proposed configuration. The values obtained for the β and the β{sub eff} are 332 and 303 pcm respectively. The core reactivity coefficients were more negative comparing the CANDU 6 loaded with {sup nat}UO{sub 2}. The initial fissile material loaded in the core increased by a factor of 1.5 after 730 GWd burnup. The obtained burn-up results show the core reactivity variations were highly positive after 6 and 12 h shut down because of considerably high buildup of {sup 233}Pa after 1-year core operation at 2000 MW power.

  6. Demonstrating the compatibility of Canflex fuel bundles with a CANDU 6 fuelling machine

    International Nuclear Information System (INIS)

    CANFLEX is a new 43-element fuel bundle, designed for high operating margins. It has many small-diameter elements in its two outer rings, and large-diameter elements in its centre rings. By this means, the linear heat ratings are lower than those of standard 37-element bundles for similar power outputs. A necessary part of the out-reactor qualification program for the CANFLEX fuel bundle design, is a demonstration of the bundle's compatibility with the mechanical components in a CANDU 6 Fuelling Machine (FM) under typical conditions of pressure, flow and temperature. The diameter of the CANFLEX bundle is the same as that of a 37-element bundle, but the smaller-diameter elements in the outer ring result in a slightly larger end-plate diameter. Therefore, to minimize any risk of unanticipated damage to the CANDU 6 FM sidestops, a series of measurements and static laboratory tests were undertaken prior to the fuelling machine tests. The tests and measurements showed that; a) the CANFLEX bundle end plate is compatible with the FM sidestops, b) all the dimensions of the CANFLEX fuel bundle are within the specified limits. (author). 3 tabs., 3 figs

  7. The economics of advanced fuel cycles in CANDU (PHW) reactors

    International Nuclear Information System (INIS)

    The economic assessments of advanced fuel cycles performed within Ontario Hydro are collated and summarized. The results of the analyses are presented in a manner designed to provide a broad perspective of the economic issues regarding the advanced cycles. The enriched uranium fuel cycle is shown to be close to competitive at today's uranium prices, and its relative position vis-a-vis the natural uranium cycle will improve as uranium prices continue to rise. In the longer term, the plutonium-topped thorium cycle is identified as being the most economically desirable. It is suggested that this cycle may not be commercially attractive until the second or third decade of the next century. (auth)

  8. Evolution of procurement and supply conditions for CANDU fuels

    International Nuclear Information System (INIS)

    In 1955 a decision was made to proceed with construction of a Nuclear Power Demonstration Station (NPD) near Rolphton, Ontario. This project, headed by Atomic Energy of Canada with major involvement of private industry, was the genesis for the development of nuclear electric generation in Canada. This paper reviews one aspect of the Canadian program: the evolution of fuel procurement and supply, which in itself has been a remarkable Canadian achievement. (author)

  9. The vault model for the disposal of used CANDU fuel

    International Nuclear Information System (INIS)

    The Vault Model has been developed to assess the performance of engineered barriers in a conceptual disposal vault for used nuclear fuel. The disposal concept being assessed is that of a sealed vault mined at a depth of 500 to 1000 m in plutonic rock in the Canadian Shield. This report documents the conceptual and mathematical framework of the Vault Model. The model represents (i) failure modes for titanium-based containers, including short-term failures due to undetected manufacturing defects and long-term failures due to uniform and local corrosion; (ii) release and radionuclides from used fuel, including relatively fast release of soluble fission products from gap and grain boundaries, and slow, congruent release controlled by the dissolution of the fuel matrix itself; and (iii) mass transport of released radionuclides through the clay-based sealing materials surrounding the waste container, including diffusion, convection, retardation and radioactive decay effects. In addition, this report presents the results of preliminary scoping calculations carried out using the Vault Model. These calculations provide insight into the model and produce test cases for comparison with simple analytical estimates and with similar computer codes, as they become available. The analytic estimate generally support the Vault Model results and thus enhance our confidence in the accuracy of the Vault Model calculations

  10. Bruce 'A' Unit 4 fuel channel feeder coupling leakage

    International Nuclear Information System (INIS)

    CANDU reactor fuel channels are connected to the primary heat transport system by mechanical joints known as feeder couplings. In 1991, three feeder coupling leakages were discovered in Bruce 4. These leakages required the unit to be shut down for repair. An investigation showed that the leakage was caused by a small 'separation' at the couplings. The assessment was that this was unlikely to be a generic issue, because the later reactors have stronger capscrews. At the time of the conference, further improvements were being considered for the seal ring and capscrew materials, and more accurate capscrew preload measurements. 3 tabs., 10 figs

  11. Progress in developing an on-line fuel-failure monitoring tool for CANDU reactors

    International Nuclear Information System (INIS)

    This paper describes the continued development of an on-line defected fuel diagnostic tool for CANDU reactors. One of the key capabilities of this tool is the ability to estimate the power and number of defects in the core based on the Gaseous Fission Product Monitoring System (GFP), and grab sample data. To perform this analysis, a clear understanding of the empirical diffusion coefficient D' [s-1] is required. This paper examines two existing models for D' and presents a new model based on 133Xe release data from commercial reactor experience. The new model is successfully applied to commercial data to demonstrate a novel technique for extracting defected fuel element power from GFP data during a reactor power change. The on-line defected fuel diagnostic tool is in a developmental stage, and this paper reports the latest enhancements. (author)

  12. A catalogue of advanced fuel cycles in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    A catalogue raisonne is presented of various advanced fuel cycle options which have the potential of substantially improving the uranium utilization for CANDU-PHW reactors. Three categories of cycles are: once-through cycles without recovery of fissile materials, cycles that depend on the recovery and recycle of fissile materials in thorium or uranium, cycles that depend primarily on the production of fissile material in a fertile blanket by means of an intense neutron source other than fission, such as an accelerator breeder. Detailed tables are given of the isotopic compositions of the feed and discharge fuels, the logistics of materials and processes required to sustain each of the cycles, and tables of fuel cycle costs based on a method of continuous discounting of cash flow

  13. Measurement of krypton grain-boundary inventories in CANDU fuel

    International Nuclear Information System (INIS)

    A technique for measuring the Kr-85 grain-boundary inventory in irradiated fuel based on the conversion of UO2 to U3O8 at low temperatures has been improved. The improvements include: 1) the use of a tracer isotope to account for release from the matrix during measurement of the grain-boundary inventory and 2) the cutting of samples from known locations. With these improvements it is possible to measure radial variations in the grain-boundary inventory. The measurements of Kr-85 grain-boundary inventory can be combined with gamma mapping and ceramography to allow investigation of the connection between microstructure and fission-product distribution. (author)

  14. Fuel and fuel channel behaviour in loss of coolant accident without the availability of the emergency coolant injection system

    International Nuclear Information System (INIS)

    Safety Analysis of CANDU reactors assesses fuel and fuel channel behaviour under high temperature transient accident conditions. The basic purpose of the analysis is to establish the channel integrity (a sufficient, but not necessary condition) even when the Emergency Cooling Injection System is presumed to be unavailable. For such severe accident conditions, the channel is heated to temperatures where it deforms and creates a heat removal path from the fuel through the pressure tube and calandria tube to the moderator. The moderator in CANDU reactor is a separate system and can provide heat removal for the heat produced within the channel. This occurs through pressure tube deformation either by circumferential strain or sag whereby the pressure tube contacts the calandria tube and allowing heat to be conducted directly to the moderator. It is found that the heat generated within the channel is transported to the moderator, and that the implied modes of channel (pressure tube) deformation are physically possible, and do not lead to failure of the pressure tube (i.e. of the pressure boundary). This paper considers the fuel and channel thermal and mechanical behaviour at very high temperatures. It discusses modelling of fission product release from fuel, deformation of the pressure tube and calandria tube, and hydrogen production insofar as it affects the fuel analysis and the containment analysis. (author)

  15. The effects of actinide based fuels on incremental cross sections in a Candu reactor

    International Nuclear Information System (INIS)

    The reprocessing of spent fuel such as the extraction of actinide materials for use in mixed oxide fuels is a key component of reducing the end waste from nuclear power plant operations. Using recycled spent fuels in current reactors is becoming a popular option to help close the fuel cycle. In order to ensure safe and consistent operations in existing facilities, the properties of these fuels must be compatible with current reactor designs. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU reactor. Specifically, the effect of this fuel design on the incremental cross sections related to the use of adjuster rods is investigated. The actinide concentrations studied in this work were based on extraction from thirty year cooled spent fuel and mixed with natural uranium to yield a MOX fuel of 4.75% actinide by weight. The incremental cross sections were calculated using the DRAGON neutron transport code. The results for the actinide fuel were compared to those for standard natural uranium fuel and for a slightly enriched (1% U-235) fuel designed to reduce void reactivity. Adjuster reactivity effect calculations and void reactivity simulations were also performed. The impact of the adjuster on reactivity decreased by as much as 56% with TRUMOX fuel while the CVR was reduced by 71% due to the addition of central burnable poison. The incremental cross sections were largely affected by the use of the TRUMOX fuel primarily due to its increased level of fissile material (five times that of NU). The largest effects are in the thermal neutron group where the ΣT value is increased by 46.7%, the Σny) values increased by 13.0% and 9.9%. The value associated with thermal fission, υΣf, increased by 496.6% over regular natural uranium which is expected due to the much higher reactivity of the fuel. (author)

  16. Simulator for candu600 fuel handling system. the experimental model

    International Nuclear Information System (INIS)

    A main way to increase the nuclear plant safety is related to selection and continuous training of the operation staff. In this order, the computer programs for training, testing and evaluation of the knowledge get, or training simulators including the advanced analytical models of the technological systems are using. The Institute for Nuclear Research from Pitesti, Romania intend to design and build an Fuel Handling Simulator at his F/M Head Test Rig facility, that will be used for training of operating personnel. This paper presents simulated system, advantages to use the simulator, and the experimental model of simulator, that has been built to allows setting of the requirements and fabrication details, especially for the software kit that will be designed and implement on main simulator. (authors)

  17. Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

    International Nuclear Information System (INIS)

    A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

  18. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    International Nuclear Information System (INIS)

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  19. LONGER: a computer program for longitudinal ridging and axial collapse assessment of CANDU fuel

    International Nuclear Information System (INIS)

    CANDU® fuel element sheath is designed to be thin and flexible for the benefit of enhanced heat transfer from the pellet to the coolant through the sheath. The flexibility of the sheath may allow the formation of longitudinal ridges on the sheath or collapse of the sheath into an axial gap under certain conditions. For both cases of deformations, the sheath may experience significant strains, and may result in sheath failure. To ensure the sheath mechanical integrity, the fuel element design needs to be assessed to preclude the conditions for longitudinal ridging and sheath collapse into the axial gap. The AECL developed LONGER computer program is used in fuel design analysis for such purpose. The LONGER code contains a number of models derived based on measurements (empirical models) and based on analytical equations, to predict the following parameters related to the deformations of CANDU nuclear fuel element sheaths. For longitudinal ridging: The critical diametral clearance for sheath longitudinal ridging, and The critical pressure for longitudinal ridging of the sheath. For axial collapse: The critical pressure for instantaneous sheath collapse into an axial gap. For circumferential collapse: The critical pressure for elastic collapse of the sheath, and The effective circumferential collapse pressure of the sheath by taking into account the axial and radial loads and the ovality of the sheath. The LONGER code has been qualified in accordance with the CSA standard N286.7-99 compliant AECL Software Quality Assurance (SQA) program. This paper describes the features and capabilities of the LONGER code that are used in CANDU fuel design analysis. (author)

  20. Leaching of used CANDU fuel: Results from a 19-year leach test under oxidizing conditions

    International Nuclear Information System (INIS)

    A fuel leaching experiment has been in progress since 1977 to study the dissolution behavior of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH2O, ∼2 mg/L carbonate) and tapwater (∼50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of 137Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of 137Cs and 90Sr leached are slightly larger in tapwater than in DDH2O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH2O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH2O is not prevalent, and in tapwater appears to be limited to the outer ∼0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution

  1. Post-irradiation examination of CANDU MOX fuel bundle containing weapons grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Dimayuga, F.C.; Karam, M.; Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2008-07-01

    The Parallex Project is an experiment designed to demonstrate the feasibility of dispositioning US and Russian weapons grade plutonium (WPu) in CANDU reactors as a mixed-oxide (MOX) fuel. The Parallex Project involved the fabrication, irradiation testing, and post-irradiation examination (PIE) of three experimental CANDU MOX fuel bundles containing WPu fuel elements that were manufactured in the US and Russia. Some of the bundles contained MOX fuel fabricated at Chalk River Laboratories (CRL) from civilian plutonium (CivPu). This paper will describe the irradiation testing and post-irradiation examination of the second Parallex bundle. The second Parallex bundle is a 37-element bundle with its centre element removed to accommodate its irradiation in the National Research Universal (NRU) reactor. The bundle was assembled at CRL using intermediate and inner elements containing WPu MOX fuel pellets fabricated by the Bochvar Institute (Russia) and CivPu MOX pellets fabricated by AECL. The 18 outer elements were fuelled with natural uranium oxide fuel pellets containing dysprosia (to reduce the neutron flux that the Pu-bearing elements would be exposed to). Half of the intermediate and inner elements contained MOX fuel pellets fabricated with depleted uranium containing 4.6 wt% WPu. The other half of the intermediate and inner elements contained MOX fuel pellets fabricated with depleted uranium containing 5.3 wt% CivPu. The irradiation testing of the second bundle was completed in NRU. The intermediate MOX elements experienced linear powers up to 49 kW/m and achieved a burnup of 294 MWh/kgHE (12 MWd/kgHE). The inner MOX elements experienced linear powers up to 23 kW/m and achieved a burnup of 130 MWh/kgHE (5 Wd/kgHE). There was a significant difference between the performance of AECL-made MOX fuel containing CivPu and Russian MOX fuel containing WPu in terms of fission gas release (FGR). This is attributed to the different fabrication processes used to manufacture the

  2. Radiological assessment of 36Cl in the disposal of used CANDU fuel

    International Nuclear Information System (INIS)

    An assessment of the potential radiological impact of 36Cl in the disposal of used CANDU fuel has been performed. The assessment was based on new data on chlorine impurity levels in used fuel. Data bases for the vault, geosphere, and biosphere models used in the EIS postclosure assessment case study (Goodwin et al. 1994) were modified to include the necessary 36Cl data. The resulting safety analysis shows that estimated radiological risks from 36Cl are forty times lower than from 129I at 104 a; this, incorporation of 36Cl into the models does not change the overall conclusions of the study of Goodwin et al. (1994a). For human intrusion scenarios, an analysis using the methodology of Wuschke (1992) showed that the maximum risk is unaffected by the inclusion of 36Cl. (author). 51 refs., 5 tabs., 15 figs

  3. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.

    2012-03-15

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  4. Intercomparison of safety evaluations for CANDU or LWR burned fuel disposed in a salt massive

    International Nuclear Information System (INIS)

    A safety analysis of a generic vault, located into a salt massive, was carried out for for spent fuel from CANDU or LWR NPPs. Three scenarios were considered for the evolution of the system: - sub-erosion, as normal evolution, - a combined process of water intrusion through an anhydrous vein and from brine bags (remaining undetected in the vicinity of the repository areas), - human intrusion. As key parameter in evaluation of long-term repository's safety the biosphere exposure (dose) was chosen. For the first scenario considered the maximum dose, due mainly to U-234, was found below the German standard value of 3 x 10-4 Sv/y. The effects of sub-erosion rate and salt concentration in ground water on the maximum dose were calculated and found rather serious. Although, having in view the rather excessive conservative assumptions adopted (the barrier effect of the geosphere was neglected) more conclusive results should be based upon a more realistic approach of the issue. In the case of human intrusion scenario the maximum exposure would be 8 x 10-5 Sv/y for CANDU fuel and 1.3 x 10-4 Sv/y for LWR fuel, as due mainly to Np-237. The analyses of local sensitivity carried out to investigate the influence of input parameters upon the release in geosphere took into consideration four parameters: a. the solubility limits; b. the diffusion coefficients; c. reference convergence rate, Kr; d. the maximum brine pressure pmax. Due to the low probability of the human intrusion scenario the effects appear to be acceptable. For the case of combined scenario the maximum doses,were found to be 9 x 10-6 Sv/y and 1 x 10-7 Sv/y for CANDU and LWR, respectively, mainly due to I-129 and Ra-225, and to I-129 and Cs-135, respectively. The effects of brine bags upon the temperature in the repository and the radiological consequences are presented for the two types of spent fuels

  5. Some aspects on security and safety in a potential transport of a CANDU spent nuclear fuel bundle, in Romania

    International Nuclear Information System (INIS)

    Each Member States (MS) is responsible for the security and safety of radioactive material during transport, since radioactive material is most vulnerable during transport. The paper presents some aspects on security and safety related to the potential transport of a CANDU Spent Nuclear Fuel (SNF) bundle from NPP CANDU Cernavoda to INR Pitesti. The possible environmental impact and radiological consequences following a potential event during transportation is analyzed, since the protection of the people and the environment is the essential goal to be achieved. Some testing for the package to be used for transportation will be also given. (author)

  6. Some aspects on security and safety in a potential transport of a CANDU spent nuclear fuel bundle, in Romania

    Energy Technology Data Exchange (ETDEWEB)

    Vieru, G., E-mail: gheorghe.vieru@nuclear.ro [Inst. for Nuclear Research, Pitesti (Romania)

    2010-07-01

    Each Member States (MS) is responsible for the security and safety of radioactive material during transport, since radioactive material is most vulnerable during transport. The paper presents some aspects on security and safety related to the potential transport of a CANDU Spent Nuclear Fuel (SNF) bundle from NPP CANDU Cernavoda to INR Pitesti. The possible environmental impact and radiological consequences following a potential event during transportation is analyzed, since the protection of the people and the environment is the essential goal to be achieved. Some testing for the package to be used for transportation will be also given. (author)

  7. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

  8. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  9. Advanced CFD simulations of turbulent flows around appendages in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Computational Fluid Dynamics (CFD) was used to simulate the coolant flow in a modified 37-element CANDU fuel bundle, in order to investigate the effects of the appendages on the flow field. First, a subchannel model was created to qualitatively analyze the capabilities of different turbulence models such as k.ε, Reynolds Normalization Group (RNG), Shear Stress Transport (SST) and Large Eddy Simulation (LES). Then, the turbulence model with the acceptable quality was used to investigate the effects of positioning appendages, normally used in CANDU 37-element Critical Heat Flux (CHF) experiments, on the flow field. It was concluded that the RNG and SST models both show improvements over the k.ε method by predicting cross flow rates closer to those predicted by the LES model. Also the turbulence effects in the k.ε model dissipate quickly downstream of the appendages, while in the RNG and SST models appear at longer distances similar to the LES model. The RNG method simulation time was relatively feasible and as a result was chosen for the bundle model simulations. In the bundle model simulations it was shown that the tunnel spacers and leaf springs, used to position the bundles inside the pressure tubes in the experiments, have no measureable dominant effects on the flow field. The flow disturbances are localized and disappear at relatively short streamwise distances. (author)

  10. Optimization of the self-sufficient thorium fuel cycle for CANDU power reactors

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available The results of optimization calculations for CANDU reactors operating in the thorium cycle are presented in this paper. Calculations were performed to validate the feasibility of operating a heavy-water thermal neutron power reactor in a self-sufficient thorium cycle. Two modes of operation were considered in the paper: the mode of preliminary accumulation of 233U in the reactor itself and the mode of operation in a self-sufficient cycle. For the mode of accumulation of 233U, it was assumed that enriched uranium or plutonium was used as additional fissile material to provide neutrons for 233U production. In the self-sufficient mode of operation, the mass and isotopic composition of heavy nuclei unloaded from the reactor should provide (after the removal of fission products the value of the multiplication factor of the cell in the following cycle K>1. Additionally, the task was to determine the geometry and composition of the cell for an acceptable burn up of 233U. The results obtained demonstrate that the realization of a self-sufficient thorium mode for a CANDU reactor is possible without using new technologies. The main features of the reactor ensuring a self-sufficient mode of operation are a good neutron balance and moving of fuel through the active core.

  11. CANDU spent fuel shielding analysis during intermediate dry storage by using Monte Carlo methodology

    International Nuclear Information System (INIS)

    Almost all the countries that operate or construct nuclear power plants have r and d programs for spent nuclear fuel and radioactive waste management. In these programs, optimal solutions for nuclear fuel cycle management are to be identified, geological disposal being one of the main goals here. Romania did not yet adopt a decision for final disposal. Nevertheless, researches and studies are in progress in order to select and characterize the geological formation for spent fuel final disposal. Currently, although there is no comprehensive EU policy in the field of safe spent fuel and radioactive waste management it is desirable to bring and keep the safety of radioactive waste management on a uniform high level among the Member States and the accession countries. The Romanian Cernavoda NPP, of CANDU type, has the following spent fuel management facilities: a spent fuel bay (for spent fuel wet storage) and a spent fuel interim dry storage facility. The dry storage technology is based on MACSTOR system consisting of storage modules located outdoors in the storage site, and equipment operated at the spent fuel storage bay for preparing the spent fuel for dry storage. The spent fuel is transferred from the preparation area to the storage site in a transfer flask. The concrete storage modules have two sealed barriers for storing the spent fuel: a seal welded stainless steel basket containing 60 spent fuel bundles and a seal welded cylinder containing 10 baskets. Twenty storage cylinders are in one storage module for a total capacity of 12,000 bundles per module. In 2003, the first storage module has become operational. The paper has the following contents: Introduction; The paper objectives; Theoretical model Set-Up; Results; Conclusions. In conclusions one notifies that SEU fuel leads to higher burnup degrees associated both with spent fuel and actinides mass reduction for 1 MWh generated electric power (from 7100 MWD/tU for UNAT to 20000 MWD/tU for SEU43

  12. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    International Nuclear Information System (INIS)

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  13. Release of 14C from the gap and grain-boundary regions of used CANDU fuels to aqueous solutions

    International Nuclear Information System (INIS)

    This study was undertaken as part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), to measure 14C inventories of used CANDU fuel. Other objectives were to measure the fraction of the total 14C inventory that would be instantly released to solution from used CANDU fuels upon sheath failure and to determine if the assumptions made in safety assessment calculations of used fuel waste disposal regarding instant release of 14C were correct. Results showed that the measured 14C inventories were a factor of 11.5 ± 3.9 lower than the estimated 14C inventory values used in safety assessment calculations. Measured instant release values for 14C ranged from 0.06 to 5.04% (of total 14C inventories) with an average of 2.7 ± 1.6%, indicating that instant release fractions for 14C used in safety assessment calculations (1.2--25%) were overestimated

  14. Conceptual Study on Dismantling of CANDU Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    In this paper, we reviewed 3D design model of the CANDU type reactor and suggested feasible cutting scheme. The structure of CANDU nuclear reactor, the calandria assembly was reviewed using 3-D CAD model for future decommissioning. Through the schematic diagram of CANDU nuclear power plant, we identified the differences between PWR and CANDU reactor assembly. Method of dismantling the fuel channels from the calandria assembly was suggested. Custom made cutter is recommended to cut all the fuel channels. The calandria vessel is recommended to be cut by band saw or plasma torch. After removal of the fuel channels, it was assumed that radiation level near the calandria vessel is not very high. For cutting of the end shields, various methods such as band saw, plasma torch, CAMC could be used. The choice of a specific method is largely dependent on radiological environment. Finally, method of cutting the embedment rings is considered. As we assume that operators could cut the rings without much radiation exposure, various industrial cutting methods are suggested to be applied. From the above reviews, we could conclude that decommissioning of CANDU reactor is relatively easy compared to that of PWR reactor. Technologies developed from PWR reactor decommissioning could be applied to CANDU reactor dismantling.

  15. Exporting technology for CANDU fuel manufacturing to the People's Republic of China - a stimulating experience for the Romanian nuclear fuel plant

    International Nuclear Information System (INIS)

    Adopting CANDU type reactors to produce nuclear-generated electricity, Romania has also developed his capability to produce nuclear fuel. Since 1995, FCN Pitesti is the unique nuclear fuel supplier for Cernavoda CANDU Power Station. Fuel plant upgrading and qualification was achieved in co-operation with AECL and Zircatec Precision Industries Inc. The fuel bundles manufactured at FCN Pitesti proved to be of excellent quality, operating with a very low defect rate, all defected fuel being reported in the first period of the reactor operation. It is a fact now that FCN has the capability to solve a wide variety of aspects one of the most significant being the development of new equipment and the increase of the capacity in order to cover the future nuclear fuel needs. On this basis FCN was invited to contribute with his potential to a supplying contract with China National Nuclear Corporation - 202 Plant, for CANDU nuclear fuel technology. Following an offer including several categories of equipment and technology, the option was for beryllium coaters and coating technology and training for end cap manufacturing. The arrangements consider Romanian company as a sub-supplier, this option ensuring the consistence with the largest part of the supply for CANDU fuel technology, offered by Zircatec. Two pieces of beryllium coaters have been produced and tested in Romania and the operating demonstration was made in the presence of Zircatec staff and Chinese delegates. The Chinese delegated were trained for complete operating modes and their ability to handle the equipment was certified accordingly. They also have been trained in the end cap technology and related quality inspection. The paper includes a short presentation of the equipment and associated work to fit the specified needs. The involvement of the Romanian fuel plant in this contract could be considered as an extension of the previous co-operation with the Canadian partners on CANDU nuclear fuel and finally

  16. Assessment of the performance of used CANDU fuel under disposal conditions

    International Nuclear Information System (INIS)

    Results of the work conducted since 1991 on determining gap and/or grain-boundary inventories for several important radionuclides such as 137Cs , 129I, 14C, 90Sr , 99Tc and 36Cl in used CANDU fuel and investigation of effect of parameters such as fuel power and burnup on their release rates is summarized in the report. Since the great majority of radionuclides are contained within the grains of the fuel pellets, the long-term release rate is governed by the dissolution rate of the uranium oxide matrix. Although UO2 is highly insoluble, the solubility of uranium increases by many orders of magnitude under oxidizing conditions. The rate of UO2 dissolution, and thus release of fission products from the fuel, is most sensitive to vault redox conditions, radiation field, groundwater composition and temperature and these factors have been investigated if justifiable assurances are to be given that radionuclide releases from a waste vault will be very limited. The redox conditions within a waste vault will evolve with time from initially oxidizing to eventually non-oxidizing as oxygen, trapped within the vault on sealing, is consumed and radiation fields, which can produce oxidants by the radiolysis of water, decay. Effective containment of the fuel should prevent its contact with groundwater until this redox evolution is complete. (author)

  17. Three Dimensional Finite Element Modelling of a CANDU Fuel Pin Using the ANSYS Finite Element Package

    International Nuclear Information System (INIS)

    The ANSYS finite element modelling package has been used to construct a three-dimensional, thermomechanical model of a CANDU fuel pin. The model includes individual UO2 pellets with end dishes and chamfers, and a Zircaloy-4 fuel cladding with end caps. Twenty node brick elements are used with both mechanical and thermal degrees of freedom, allowing for a full coupling between the thermal and mechanical solutions under both steady state and transient conditions. Each fuel pellet is modelled as a separate entity that interacts both thermally and mechanically with the cladding and other pellets via contact elements. The heat transfer between the pellets and cladding is dependent on both the interface pressure and temperature, and all material properties of both the pellets and the sheath are temperature dependant. Spatially and temporally varying boundary conditions for heat generation and convective cooling can be readily applied to the model. The model naturally exhibits phenomena such as pellet hour glassing and ridging of the cladding at the Pellet to pellet interfaces, allowing for the prediction of localized sheath stresses. The model also allows for the prediction of fuel pin bowing due to asymmetric thermal loads and fuel pin sagging due to overheating of the cladding, which may occur under accident conditions. (author)

  18. CANDU development

    International Nuclear Information System (INIS)

    Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 320C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

  19. Metallographic examination of a CANDU fuel bundle heated under severe accident conditions

    International Nuclear Information System (INIS)

    Post-test metallographic examination of bundle cross sections of a 19-element modified CANDU fuel bundle was carried out. The bundle, HTBS-004, had been subjected to a severe temperature excursion to 1900 degrees Celsius in superheated steam. For this study, quantitative image analysis, Auger analysis and SEM-EDX techniques were applied. A significantly large quantity of molten (Zr, U, O) alloy was relocated in the bundle section 50 mm from the upstream end, whereas the 377-mm section showed little relocated material except at the inner element junctions. These variations in the molten material generation and relocation have been correlated with the corresponding axial and radial variations in the heatup rates

  20. Thermal analysis of dry concrete canister storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

  1. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    International Nuclear Information System (INIS)

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  2. Optimizing in-bay fuel inspection capability to meet the needs of today's CANDU fleet

    International Nuclear Information System (INIS)

    With the recent return to service of many CANDU units, aging of all others, increasingly competitive energy market and aging hot cell infrastructure - there exists now a greater need for timely, cost-effective and reliable collection of irradiated fuel performance information from fuel bay inspections. The recent development of simple in-bay tools, used in combination with standardized technical specifications, inspection databases and assessment techniques, allows utilities to characterize the condition of irradiated fuel and any debris lodged in the bundle in a more timely fashion and more economically than ever. Use of these tools and 'advanced' techniques permits timely engineering review and disposition of emerging issues to support reliable operation of the CANDU fleet. (author)

  3. Investigation of the grain-boundary chemistry in used CANDU fuel by x-ray photoelectron spectroscopy (XPS)

    International Nuclear Information System (INIS)

    The grain-boundary chemistry of used CANDU fuel is being systematically investigated by X-ray photoelectron spectroscopy (XPS) using a McPherson ESCA-36 instrument that has been adapted for routine studies of highly radioactive materials. Initial stages of fuel corrosion under various storage and disposal conditions can be identified from chemical-shift effects for uranium. For example, pervasive but highly selective grain-boundary oxidation has been revealed in CANDU fuels exposed to moist air at 150 deg. C for extended periods, suggesting aggressive radiolytic processes operating in a thin film of adsorbed water. Pronounced segregation of a number of fission products to cracks and grain boundaries in used CANDU fuels has been explicitly demonstrated by XPS as well. Model calculations and composition depth profiles are indicative of near monolayer films. Some correlations between fuel power history and fission-product distributions have been established and possible evidence of migration during moist-air exposure has been obtained. The key advantages and limitations of XPS in this context are discussed and illustrated with selected results. (author). 23 refs, 8 figs, 1 tab

  4. PIE of a CANDU fuel element irradiated for a load following test in the INR TRIGA reactor

    International Nuclear Information System (INIS)

    As part of the collaboration under the Romania - Canada Memorandum for co-operation in research and development of nuclear energy and technology, a load following test has been devised to demonstrate the load following capability of CANDU-6 fuel within the established design envelope for operating powers. A 37-element CANDU-6 fuel bundle element fabricated by AECL was irradiated in the TRIGA 14 MW(th) material testing reactor at the Institute for Nuclear Research (INR) in Pitesti, Romania. The load following cycle consisted of 200 daily cycles from 100% power to 50% power within the reference overpower envelope for fuel in a CANDU-6 reactor. Full power operation was 57 kW/m Element Linear Power. The paper provides the results obtained by post-irradiation examination of the fuel element in the INR hot cells. The following techniques were used: - Visual inspection and photography by periscope; - Profilometry; - Axial gamma scanning; - Fuel element puncturing and fission gas analysis; - Metallographic and ceramographic examinations by optical microscopy; - Burn-up measurement by mass spectrometry using the 235U depletion method. (authors)

  5. Formation of Corrosive Deposits and Their Impact on Operational Safety of Fuel Elements in Candu Reactor

    International Nuclear Information System (INIS)

    Interaction between fuel element cladding and water coolant plays an important role in normal operation, can have a dominant role in accidental situations and can lead to failure of fuel rods and activity release. For the future, the tendency will be to increase the coolant temperature, extend fuel residence time in the reactor core (for higher burnup) and increase the heat flux. This can lead to increased probability of fuel failures due to waterside corrosion, corrosion products accumulation and deposition. In order to prevent cladding failures, the coolant chemistry must be monitored and controlled in order to reduce the amount of deposited crud and the oxygen potential. Corrosive deposits together with aqueous corrosion influence the performance of fuel elements by increase of temperature on cladding surface or changes in the coolant chemistry (increase of water pH), phenomena which lead to cladding failures. The process of corrosion products formation on zircaloy-4 fuel cladding surface and their consequences was evidenced by performing of experiments in: autoclaves circuits assembled in a by-pass loop of a CANDU-6 Reactor at NPP Cernavoda; irradiation loop of the TRIGA Reactor, and in laboratory static autoclaves. The determination of corrosion and the characterization of crud deposits on the zircaloy-4 surfaces were performed using gravimetric method, metallographic and electronic microscopy, and gamma spectrometry analysis and impedance electrochemical spectroscopy (EIS) determinations. The experimental results showed that the composition, thickness and evolution of corrosive deposits on fuel assembly surfaces depend very much on operational conditions, such as steady state operation, water chemistry conditions (pH and oxygen concentration) and different oxidation conditions of cladding surface. (author)

  6. Development of a CATHENA Fuel Channel Analysis Model for a Fuel Channel with Axial Variation of Radial Pressure Tube Creep in a Stratified Two-Phase Flow Condition

    International Nuclear Information System (INIS)

    A two-phase heat transfer phenomena in the fuel bundle strings located in a horizontal pressure tube with an axial variation of the radial creep, especially under a low stratified two-phase flow condition such as encountered in the CANDU reactor under the later stage of the blowdown phase of a LBLOCA, involves a complex heat transfer nature. This includes the conduction in the fuel rods, pressure tube, convection in the vapor and liquid regions, and radiation between the fuel rods exposed in the steam and the pressure tube, pressure tube and calandria tube. As these three modes of heat transfer has to be treated in a combined way, modeling the heat transfer phenomena inside the fuel bundle under the stratified flow during the later stage of LBLOCA blowdown has been one of the most challenging tasks in the CANDU safety analyses. The main reason for this hot attention is that it closely related to the integrity of the pressure tube. In this study a heat transfer model for handling this situation is developed, implemented and under preliminary testing of the analysis results. The analysis result up to now is encouraging and the validation of the model developed is ongoing. The major motivation of this study is to evaluate the conservatism of the current CANDU safety analysis methodology for a fuel channel with an axial variation of the radial creep of the pressure tube as easily experienced in the aged CANDU plant as it assumes the centerline of the fuel bundle string is the same as that of the pressure tube

  7. Microchemical study of high-burnup CANDU fuel by imaging-XPS

    Energy Technology Data Exchange (ETDEWEB)

    Do, Than [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)], E-mail: dot@aecl.ca; Irving, Karen G. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)], E-mail: irvingk@aecl.ca; Hocking, William H. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)], E-mail: hockingw@aecl.ca

    2008-12-15

    An advanced facility for characterization of highly radioactive materials by Imaging X-ray Photoelectron Spectroscopy (XPS) has been developed at the Chalk River Laboratories (CRL), based upon over a decade of prior experience with a prototype system. Auxiliary electron and ion guns provide additional in situ capabilities for scanning electron microscopy (SEM), scanning Auger microscopy (SAM) and composition depth profiling. The application of this facility to the characterization of irradiated fuel materials will be illustrated with selected results taken from a detailed study of the microchemistry at the fuel-sheath interface in a CANDU fuel element that was irradiated to extended burnup in the NRU (National Research Universal) reactor at CRL. Inside surfaces of the end caps and the welds between the sheath and the end caps as well as the thin-walled Zircaloy-4 sheath were analyzed. The in situ SEM capability was essential for selecting different areas on each sample, such as sheath locations with and without a visible retained CANLUB graphite layer, for XPS analysis. Effective infiltration of segregated fission products, especially cesium, into the graphite was demonstrated by depth profiling. A richer chemistry of segregated fission products was found on the end caps than on the sheath with elevated levels of barium, strontium, tellurium, iodine and cadmium as well as cesium. The results are consistent with current understanding of the primary migration route for fission products to the sheath and also indicate that the CANLUB layer functions as a chemical rather than a physical barrier to segregated fission products.

  8. Drop Test of the Candu Spent Fuel Storage Basket in MACSTOR/KN-400

    International Nuclear Information System (INIS)

    The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded which have the same weight of real spent fuel bundles. On the external surface of the basket, 8 strain gauges and 4 accelerometers were attached for the data acquisition. In order to measure the velocity when a basket impacts, three different devices were utilized. And the impact velocity results were compared and cross-checked. After the dropping tests, helium leak tests were conducted to evaluate the leakage rate. (authors)

  9. Drop Test of the Candu Spent Fuel Storage Basket in MACSTOR/KN-400

    Energy Technology Data Exchange (ETDEWEB)

    Choi, W.S.; Jeon, J.Y.; Seo, K.S. [KAERI, 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Park, J.E.; Yoo, G.S.; Park, W.G. [Korea Hydro Nuclear Power - KHNP (Korea, Republic of)

    2009-06-15

    The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded which have the same weight of real spent fuel bundles. On the external surface of the basket, 8 strain gauges and 4 accelerometers were attached for the data acquisition. In order to measure the velocity when a basket impacts, three different devices were utilized. And the impact velocity results were compared and cross-checked. After the dropping tests, helium leak tests were conducted to evaluate the leakage rate. (authors)

  10. Utilization of fluorescent uranium x-rays as verification tool for irradiated CANDU fuel bundles

    International Nuclear Information System (INIS)

    The use of fluorescent uranium x-rays for in-situ safeguards verification of irradiated CANDU fuel bundles is described. Room temperature CdZnTe (supergrade) semiconductor detector of low sensitivity coupled to charge sensitive pre-amplifier is used. This detector is characterized by moderate resolving power in the low energy region around 100 keV. It as such allows the separation of uranium x-rays in the close proximity of tungsten x-rays emanating from the shielding/collimator assembly. On account of strong attenuation, the detection of low energy x-rays requires the shielding to be of an optimized thickness. Further, in view of high intensity of this radiation the use of small volume detector is warranted. In dealing with the subject, this paper therefore presents an assessment, not only of the detector but also the shield-collimator assembly for the required verification of short cooling time fuel bundles. Results of the associated optimization measurements with respect to collimator aperture and detector sensitivity are consequently included. The future course of work from the viewpoint of development of a suitable x-ray spectrometer specifically for the purpose of verifying extremely short (< 1 month old) cooling time fuel bundles is moreover identified. (author)

  11. Seismic Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Cho, Chun Hyung; Lee, Heung Young [Korea Hydro and Nuclear Power Co., Ltd., Taejon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su; Kim, Jong Soo [KONES Co., Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside of the concrete module are built 40 storage cylinders accommodating ten 60- bundle dry storage baskets, which are suspended from the top slab and eventually constrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module is by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants except for local geologic characteristics. As per USNRC SRP Section 3.7.2 and current US practices, Soil-Structure Interaction (SSI) effect shall be considered for all structures not supported by a rock or rock-like soil foundation materials. An SSI is a very complicated phenomenon of the structure coupled with the soil medium that is usually semi-infinite in extent and highly nonlinear in its behavior. And the effect of the SSI is noticeable especially for stiff and massive structures resting on relatively soft ground. Thus the SSI effect has to be considered in the seismic design of MACSTOR/KN-400 module resting on soil medium. The scope of the this paper is to carry out a seismic SSI analysis of the MACSTOR/KN-400 module, in order to show how much the SSI gives an effect on the structural responses by comparing with the fixed-base analysis.

  12. The next generation of CANDU reactor: evolutionary economics

    International Nuclear Information System (INIS)

    AECL has developed the design for a next generation of CANDUR plants by applying a set of enabling technologies to well-established successful CANDU features from the CANDU 6 Reactors in service and the design of the CANDU 9. Advances made in the construction of the Wolsong reactors have been built upon in the current project in China. The basis for the new design is to evolve from the current CANDU units by replicating or adapting existing components for a new core design. Using slightly enriched uranium fuel, a core with light water coolant, and heavy water moderator and reflector has been defined, based on the existing CANDU fuel channel module. This paper summarizes the main features and characteristics of the reference next-generation CANDU design. The progress of the next generation of CANDU design program in meeting challenging cost, schedule and performance targets is described. AECL's cost reduction methodology is summarized as an integral part of the design optimization process. Examples of cost reduction features are given, together with enhancement of design margins

  13. Micro-focus x-ray inspection of the bearing pad welded by laser for CANDU fuel element

    International Nuclear Information System (INIS)

    To attach the bearing pads on the surface of CANDU fuel element, laser welding technique has been reviewed to replace brazing technology which is complicate process and makes use of the toxic beryllium. In this study, to evaluate the soundness of the weld of the bearing pad of CANDU fuel element, a precise X-ray inspection system was developed using a micro-focus X-ray generator with an image intensifier and a real time camera system. The weld of the bearing pad welded by Nd:YAG laser has been inspected by the developed inspection system. Image processing technique has been applied to reduce random noise and to enhance the contrast of the X-ray image. A few defects on the weld of the bearing pads have been detected by the X-ray inspection process

  14. Modeling coupled bending, axial, and torsional vibrations of a CANDU fuel rod subjected to multiple frictional contact constraints

    International Nuclear Information System (INIS)

    In this paper, a finite element based dynamic model is presented for bending, axial, and torsional vibrations of an outer CANDU fuel element subjected to multiple unilateral frictional contact (MUFC) constraints. The Bozzak-Newmark relaxation-integration scheme is used to discretize the equations of motion in the time domain. At a time step, equations of state of the fuel element with MUFC constraints reduce to a linear complementarity problem (LCP). Results are compared with those available in the literature. Good agreement is achieved. The 2D sliding and stiction motion of a fuel element at points of contact is obtained for harmonic excitations. (author)

  15. Investigation of the CANLUB/sheath interface in CANDU fuel at extended burnup by XPS and SEM/WDX

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Behnke, R.; Duclos, A.M.; Gerwing, A.F. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Chan, P.K. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    1997-07-01

    A systematic investigation of the fuel-sheath interface in CANDU fuel as a function of extended burnup has been undertaken by XPS and SEM/WDX analysis. Adherent deposits of UO{sub 2} and fission products, including Cs, Ba, Rb, I, Te, Cd and possibly Ru, have been routinely identified on CANLUB coated and bare Zircaloy surfaces. Some trends in the distribution and chemistry of key fission products have begun to emerge. Several potential mechanisms for degradation of the CANLUB graphite layer at high burnup have been practically excluded. New evidence of carbon relocation within the fuel element and limited reaction with excess oxygen has also been obtained. (author)

  16. Development of a spent fuel bay operator support system for PHWR-CANDU nuclear power plant

    International Nuclear Information System (INIS)

    The safety advantages of the CANDU 600 NPPs could be further enhanced by the supplementation of the Canadian experience on plant operation with Computerised Operator Support System (COSSs) adjusted to the actual plant configuration, the idiosyncrasies of the given equipment, the errors during engineering and construction phases and the psychological features of the operation personnel. One of the most relevant systems with an important decision component that involves relatively complex procedures - thereby related human errors - is Cernavoda NPP's Spent Fuel Bay Cooling and Purification System. Strong motivations to consider a flexible COSS, both for operation and for intervention purpose, are given by the global activity stored in the systems bays and the location of the bays outside the containment. A Spent Fuel Bay Operator Support System (SFOSS) is at a research-grade at the Institute of Atomic Physics and at the Center of Technology and Engineering for Nuclear Projects in compliance with the general principles of the expert systems and under a IAEA Co-ordinate Research Program. The paper illustrates a generic description of the system and also of the SFOSS structure. (Author) 10 Refs

  17. Romanian-Canadian joint program for qualification of FCN as a CANDU fuel supplier

    International Nuclear Information System (INIS)

    RENEL (Romania Power Authority), the co-ordinator of Romanian Nuclear Program, have decided to improve, starting 1990 the existing capability to produce CANDU nuclear fuel at FCN Pitesti. The objective of the program was defined with AAC (AECL - ANSALDO Consortium) for the qualification of FCN fuel plant according to Canadian Z299.2 standard. The Qualification Program was performed under AAC Work Order C-003. The co-ordination was assumed by AECL, as overall Design Authority. ZPI (Zircatec Precision Industries Inc., Canada), were designated to supply technical assistance, equipments and know how where necessary. After a preliminary verification of the FCN fuel plant, including the processes and system investigation, performed under AECL and ZPI assistance, the Qualification Program was defined in all details. The upgrading of documentation on all aspects required by Z299.2 was performed. Few processes needed to be reconsidered and equipment was delivered by ZPI or other suppliers. This includes mainly welding equipments and special inspection equipments. Health Physics was practically fully reconsidered. New equipment and practice were adapted to provide adequate control on health conditions. Every manufacturing and inspection process was checked to determine their performance during a Qualification Run based on acceptance criteria which have been established in the Qualification Plan. Manufacturing Demonstration Run was an important step to prove that all plant functions have been accomplished during the fabrication of 200 fuel bundles. These bundles have been fully accepted and 66 of them have been loaded in the first charge of Unit 1 Cemavoda NPS. The surveillance and audit actions made by AECL and ZPI during this period confirmed the FCN capability to operate an adequate system meeting the to required quality assurance standard. The very open attitude of AECL, Zircatec and FCN staff have stimulated the progress of the project and a successful achievement of the

  18. Thorium utilization in Candu reactors

    International Nuclear Information System (INIS)

    In this study, means of thorium utilization in a CANDU reactor are considered. A once through thorium-DUPIC cycle is analyzed in detail. CANDU has the best neutron economy among the commercially available power reactors, which makes it suitable for many different fuel cycle options. A review of the available fuel cycles is also done in the scope of this study to select an economically viable cycle which does not impose profound changes in the neutronic properties of the core that require remodeling of core and related systems. To create a good model ot the CANDU core for the necessary calculations, the steady state properties of CANDU reactor are analyzed. It is assumed that approximation ot refueling as moving the bundles at a constant velocity is valid. This approximation leads to a corollary; The average cross sections of two adjacent bidirectionally refueled channels are independent of axial location. This is also veritied. A result of this corollary the CANDU core can be modeled only in radial direction in cylindirical geometry. The steady state CANDU core model is prepared using the actual power values and these values are sought in the results. The control systems which effect the neutron flux shape are introduced into the model later in the form of additional absorption cross section and lower diffusion coefficient. The results are in good agreement with the actual values. Several different thorium-DUPIC fuel bundle configurations are considered and the one with 12 Th02 elements in the third ring is found to have similar burnup dependent cross-sections and location infinite multiplication factors. Using the model created, the bundle is tested also in the tull core model and it is tound that this bundle configuration complies with the current refueling scheme. That is, no changes are necessary in the refuelind rate or the control systems. A higher conversion ratio of 0.82 is attained, while the excess reactivity of the core is found to decrease by 0.01 Ak

  19. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    A relational database management system with a query language, Reference 1, has been used to develop a Boiling Water Reactor (BWR) fuel channel tracking system on a microcomputer. The software system developed implements channel vendor and Nuclear Regulatory Commission recommendations for in-core channel movements between reactor operating cycles. A BWR Fuel channel encloses the fuel bundle and is typically fabricated using Ziracoly-4. The channel serves three functions: (1) it provides a barrier to separate two parallel flow paths, one inside the fuel assembly and the other in the bypass region outside the fuel assembly and between channels; (2) it guides the control rod as it moves between fuel assemblies and provides a bearing surface for the blades; and (3) it provides rigidity for the fuel bundle. All of these functions are necessary in typical BWR core designs. Fuel channels are not part of typical Pressurized Water Reactor (PWR) core designs

  20. Separating rings detection in fuel channels of Embalse NPP

    International Nuclear Information System (INIS)

    The design specifications of Embalse Nuclear Power Plants (CANDU Type Reactor 600Mw) define the positions to be taken by 4 separating rings of the fuel channels. Experience has demonstrated the displacement possibility of the above mentioned rings. It means a risk of contact between pressure tube and calandria tube. In order to determine the position of separating rings, an inspection system based on Eddy Currents technique was developed by CNEA personnel. Detection is performed through two special probes operating according the ''emitter-receiver'' principle. Obtained signals and its relative position are recorded in a video tape and registered in paper. The probe is telecommanded by an automatic equipment. In this paper the construction and calibration of the detection equipment is described, as well as the propulsion. Final results are also outlined in the inspection carried out in November 1986 when an effective displacement of separating rings was verified from its design position in most of the inspected tubes

  1. An elasto-plastic model for mechanical contact between the pellets and sheath in CANDU nuclear fuel elements

    International Nuclear Information System (INIS)

    During high-temperature transients, increased mechanical contact can occur between the fuel stack and the sheath in the axial and/or radial direction. As well, there is an axial linear power gradient and an axial gradient in mechanical properties of the sheath specific to a CANDU-type fuel element. This requires a code with the capability to treat multiple axial segments. This paper describes a contact model that allows the elasto-plastic mechanical contact in radial/axial direction for multiple axial segments. (14 figs., 5 refs.)

  2. The volumes of wastes resulting from the direct disposal or recycling of CANDU used fuel

    International Nuclear Information System (INIS)

    This report summarizes and compares the volumes of wastes that would be generated for disposal in the cases of the direct disposal of CANDU used fuel and the disposal of reprocessing wastes if the fuel were recycled. It is shown that the best estimates of these waste volumes are as follows: From direct disposal of fuel: 649 m3/a; From recycle with used U as waste: 1039 m3/a; From recycle with used U as resource: 854 m3/a. Modifications to the procedures in the reference cases are discussed, based on known or feasible technology, with the object of reducing the volumes of waste. The estimates of waste volumes that would result from reprocessing are compared with information on the waste arisings in the U.S., U.K. and French programs, together with a multi-program estimate derived by the International Atomic Energy Agency. The volumes are adjusted for the differences in burnup in the different programs and for the different levels of fission-product loading in the high-level waste glass. Most of the estimates are within a factor of 3 of each other, the exception being the arisings from the transuranic and low-level wastes. Th conclusions that can be drawn from this study are the following: 1 The volumes of wastes arising from the disposal of used fuel without recycle and the volumes of wastes arising from the reprocessing of used fuel, internationally and in the Canadian reference cases, are much the same when expressed on a common basis; 2. Any absolute differences in waste volumes are a consequence of different burnups, or of the choices of how a particular recycle process is operated; 3. Modification and optimization of the processes considered in reference Canadian programs, both for the direct disposal of used fuel and for the disposal of reprocessing wastes, could bring about a reduction in disposal volumes of factors of between 2 and 3. (author). 45 refs., 8 tabs., 8 figs

  3. Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

    International Nuclear Information System (INIS)

    The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

  4. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    AECL has developed a suite of technologies for CanduR reactors that enable the next step in the evolution of the Candu family of heavy-water-moderated fuel-channel reactors. These technologies have been combined in the design for the Advanced Candu Reactor TM1 (ACRTM), AECL's next generation Candu power plant. The ACR design builds extensively on the existing Candu experience base, but includes innovations, in design and in delivery technology, that provide very substantial reductions in capital cost and in project schedules. In this paper, main features of next generation design and delivery are summarized, to provide the background basis for the cost and schedule reductions that have been achieved. In particular the paper outlines the impact of the innovative design steps for ACR: - Selection of slightly enriched fuel bundle design; - Use of light water coolant in place of traditional Candu heavy water coolant; - Compact core design with unique reactor physics benefits; - Optimized coolant and turbine system conditions. In addition to the direct cost benefits arising from efficiency improvement, and from the reduction in heavy water, the next generation Candu configuration results in numerous additional indirect cost benefits, including: - Reduction in number and complexity of reactivity mechanisms; - Reduction in number of heavy water auxiliary systems; - Simplification in heat transport and its support systems; - Simplified human-machine interface. The paper also describes the ACR approach to design for constructability. The application of module assembly and open-top construction techniques, based on Candu and other worldwide experience, has been proven to generate savings in both schedule durations and overall project cost, by reducing premium on-site activities, and by improving efficiency of system and subsystem assembly. AECL's up-to-date experience in the use of 3-D CADDS and related engineering tools has also been proven to reduce both engineering and

  5. Some physics aspects of the in-core fuel management analysis for CANDU-PHW type reactors

    International Nuclear Information System (INIS)

    The primary objective of the ''in-core fuel management'' studies for the CANDU core is to determine fuel loading and fuel replacement strategies which will result in minimum total unit energy cost while operating the reactor in a safe and reliable fashion. Two types of calculations are mainly required in fuel management analyses: a) those used to determine the nominal power and burnup distributions, and b) those used to determine instantaneous distributions which include the time varying fine structure of the power distribution. A method for equilibrium power and burnup distributions determination is presented for the first type of calculations, based on computing the macroscopic cross-sections from the bundle power and burnup history. The computation model presented was programmed into the SERA 3-D code, which was developed at INPR. A series of results for the second type of calculations are presented, which were obtained by applying the random age approximation and the autorefuel methods in determining the instantaneous power distributions. Some improvements are proposed for these models on the basis of the above mentioned results. For the sake of numerical illustration a CANDU slightly enriched uranium core configuration is presented, the physics parameters of which were evaluated on the basis of fuel management analyses. (author). 10 refs, 7 figs, 3 tabs

  6. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    International Nuclear Information System (INIS)

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  7. Proceedings of the fourth international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance

  8. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

  9. Simulation of LOCA type accident for CANDU fuel in TRIGA materials testing reactor and its associated facilities at INR-Pitesti

    International Nuclear Information System (INIS)

    The specific objective of the experiment regards the simulation of a LOCA type accident in an irradiation facility in order to characterize the behaviour of a CANDU fuel element with respect to fuel-cladding interaction and fission product release in the case of clad failure occurrence. The work belongs to 'Nuclear Safety Program' contributing to computer codes qualification used for safety assessment of Cernavoda NPP. The experimental results of these tests will be used, among other input reference data, for evaluating the realistic safety limits for CANDU fuel element in case of anticipated transients. (Author)

  10. Thorium utilization in ACR (Advanced CANDU) and CANDU-6 reactors

    International Nuclear Information System (INIS)

    It is the main objective of this study to investigate fuel composition options for CANDU type of reactors that are capable of using a mixture of U-Th as fuel. A homogenous mixture of (U-Th)O2 was used in all elements of fuel bundles. The core of CANDU-6 and ACR (Advanced CANDU) were modeled using MCNP5. In equilibrium core, using MONTEBURNS2 code (coupled with MCNP5 and ORIGENS) for once-through uranium and once-through uranium-thorium fuel cycle of CANDU-6 and ACR, discharge burnups and spent fuel compositions were computed. For various enrichments of uranium and different fractions of thorium in a uranium-thorium fuel mixture, performing burnup calculations, relevant relations were derived; in addition, conversion ratio, fuel requirement, uranium resource utilization, and natural uranium savings were determined, and their changes with burnup were observed. Appropriate fuel compositions were discussed.

  11. Design research for accident prevention in CANDU reactor

    International Nuclear Information System (INIS)

    Study of PHWR Candu Design under severe accident has been done. Severe accident is defined as one in which the fuel is not removed by the coolant in the primary heat transport system. A severe accident could only result if a process system failed and the appropriate protective system was simultaneous unavailable. Severe accidents of the Candu reactor relevant to severe accident are set first by the inherent properties of the design. With the system sufficiently independent, the frequencies of a severe accident could be made acceptable low. This paper discussed that the separately cooled moderator in a Candu provides an effective heat sink in the event of a loss of coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintain the integrity of the fuel channels, therefore terminating this severe accidents short of severe core damage

  12. CANDU 9 design

    International Nuclear Information System (INIS)

    AECL has made significant design improvements in the latest CANDU nuclear power plant (NPP) - the CANDU 9. The CANDU 9 operates with the energy efficient heavy water moderated reactor and natural uranium fuel and utilizes proven technology. The CANDU 9 NPP design is similar to the world leading CANDU 6 but is based upon the single unit adaptation of the 900 MWe class reactors currently operating in Canada as in integrated four-unit configurations. The evolution of the CANDU family of heavy water reactors (HAIR) is based on a continuous product improvement approach. Proven equipment and systems from operating stations are standardized and used in new products. As a result of the flexibility of the technology, evolution of the current design will ensure that any new requirements can be met, and there is no need to change the basic concept. This paper will provide an overview for some of the key features of the CANDU 9 NPP such as nuclear systems and equipment, advanced control and computer systems, safety design and protection features, and plant layout. The safety enhancements and operability improvements implemented in this design are described and some of the advantages that can be expected by the operating utility are highlighted. (author)

  13. Wear inspection for fuel channels in HANARO

    International Nuclear Information System (INIS)

    It has been observed that fuel assemblies have mechanical damage on some components due to the flow-induced vibration in the fuel channels. The major damage was the fretting-wear on the bottom end plates and spacer plates. Both the components are aluminum but cause the fretting-wear on the corresponding surfaces of the zirconium fuel channels. Also the bottom guide arms and top springs of the fuel are considered to be major parts that cause wear on the fuel channels. Therefore, the inspection of the inner surfaces of the fuel channels is required from a lifetime point of view. It is very difficult and time-consuming work to remove and install the fuel channels because of their inherent characteristics and the physical interference of other components in the reactor core. Thus we developed special tools for the inspection of the fuel channels by using an impression material without the removal of the reactor components. The impression material is a compound to replicate the damage of the fuel channels within a limited working time considering the hardening time as well as the radiation effect. The wear-inspection tool is a mechanical tool to press the impression compound against the inner walls of the fuel channels where the fuel components contact. The inspection tool has 36 molding cups that are operated in radial directions by turning the central rod with a fuel-handling tool from the pool top 12m above the reactor core. The same concept but a more compact design is applied to the cylindrical fuel channel. The wear inspection was successfully accomplished for a few fuel channels. The results show visible wear marks on a hexagonal fuel channel at the positions corresponding to fuel components such as the bottom guide arms and top springs. The wear damage is slight, approximately 0.2mm depth, in comparison with the thickness (1.6mm) of the fuel channel. No visible wear mark has been found in the cylindrical flow tubes so far. The wear inspection is being continued

  14. The evolution of the CANDU energy system - ready for Europe's energy future

    International Nuclear Information System (INIS)

    As air quality and climate change issues receive increasing attention, the opportunity for nuclear to play a larger role in the coming decades also increases. The good performance of the current fleet of nuclear plants is crucial evidence of nuclear's potential. The excellent record of Cernavoda-1 is an important part of this, and demonstrates the maturity of the Romanian program and of the CANDU design approach. However, the emerging energy market also presents a stringent economic challenge. Current NPP designs, while established as reliable electricity producers, are seen as limited by high capital costs. In some cases, the response to the economic challenge is to consider radical changes to new design concepts, with attendant development risks from lack of provenness. Because of the flexibility of the CANDU system, it is possible to significantly extend the mid-size CANDU design, creating a Next Generation product, without sacrificing the extensive design, delivery and operations information base for CANDU. This enables a design with superior safety characteristics while at the same time meeting the economic challenge of emerging markets. The Romanian nuclear program has progressed successfully forward, leading to the successful operation of Cernavoda-1, and the project to bring Cernavoda-2 to commercial operation. The Romanian nuclear industry has become a full-fledged member of the CANDU community, with all areas of nuclear technology well established and benefiting from international cooperation with other CANDU organizations. AECL is an active partner with Romanian nuclear organizations, both through cooperative development programs, commercial contracts, and also through the activities of the CANDU owners' Group (COG). The Cernavoda project is part of the CANDU 6 family of nuclear power plants developed by AECL. The modular fuel channel reactor concept can be modified extensively, through a series of incremental changes, to improve economics, safety

  15. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. 95 refs, 3 tabs

  16. Measuring vibrations in fuel channels CNE

    International Nuclear Information System (INIS)

    This paper present a description of implementation and execution of vibration measurements made at the request of NUCLEOELECTRICA ARGENTINA S.A. on the ends of the reactor fuel channels of Embalse Nuclear Power Plant to explore possible differences between the dynamic behavior of empty fuel channel and with full charge of fuel elements inside. (author)

  17. CANDU development: the next 25 years

    International Nuclear Information System (INIS)

    CANDU Pressurized Heavy Water Reactors have three main characteristics that ensure viability for the very long term. First, great care has been taken in designing the CANDU reactor core so that relatively few neutrons produced in the fission process are absorbed by structural or moderator materials. The result is a reactor with high neutron economy that can burn natural uranium and a core that operates with 2-3 times less fissile content than other, similarly-sized reactors. In addition to neutron economy, the use of a simple bundle design and on-power fuelling augment the ability of CANDU reactors to burn a variety of fuels with relatively low fissile content with high efficiency. This ensures that fuel supply will not limit the applicability of the technology over the long term. Second, the presence of large water reservoirs ensures that even the severest postulated accidents are mitigated by passive means. For example, the presence of the heavy water moderator, which operates at low pressure and temperature, acts as a passive heat sink for many postulated accidents. Third, the modular nature of the core (e.g., fuel channels) means that components can be relatively easily replaced for plant life extension and upgrading. Since these factors all influence the long-term sustainability of CANDU nuclear technology, it is logical to build on this base and to add improvements to CANDU reactors using an evolutionary approach. This paper reviews AECL's product development directions and shows how the above characteristics are being exploited to improve economics, enhance safety, and ensure fuel cycle flexibility for sustainable development. (author). 21 refs., 9 figs

  18. R and R programmes to advance CANDU technology

    International Nuclear Information System (INIS)

    A key characteristic of the CANDU reactor design is the ability to meet future requirements via incremental modifications as opposed to revolutionary design changes. The main objectives for advancing CANDU technology are 1) to reduce capital and operating costs, 2) to increase capacity factors, 3) to increase passive safety, and 4) to enhance fuel/fuel cycle flexibility. These objectives are being met by performing research and development in 6 key areas: fuel channels, fuel/fuel cycle technology, safety, heavy water production, plant systems and components, and information technology. Fuel channel improvements are gained through the elucidation and application of basic materials science for life extension. Fuel and fuel cycle work is focusing on advanced cycles, and on the development of a bundle to act as a carrier for advanced fuels that improves burnup and economics. In safety, the inherent features of CANDU are used to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of low temperature fuels. Heavy water processes are being developed that can be used with existing hydrogen sources (such as electrolytic hydrogen or steam reformers), or that can be used in a stand-alone mode. Plant systems and components work includes improvements to plant components such as steam generators, and the application of advanced control centre technology. Information technology is being developed to improve all aspects of CANDU design, construction, and operation. This paper gives an overview of some of the R and D in these areas that is supporting the incremental improvement of current and advanced CANDU designs. (author). 7 figs

  19. Analysis of Fuel Temperature Reactivity Coefficients According to Burn-up and Pu-239 Production in CANDU Reactor

    International Nuclear Information System (INIS)

    The resonances for some kinds of nuclides such as U-238 and Pu-239 are not easy to be accurately processed. In addition, the Pu-239 productions from burnup are also significant in CANDU, where the natural uranium is used as a fuel. In this study, the FTCs were analyzed from the viewpoints of the resonance self-shielding methodology and Pu-239 build-up. The lattice burnup calculations were performed using the TRITON module in the SCALE6 code system, and the BONAMI module was executed to obtain self-shielded cross sections using the Bondarenko approach. Two libraries, ENDF/B-VI.8 and ENDF/B-VII.0, were used to compare the Pu-239 effect on FTC, since the ENDF/B-VII has updated the Pu-239 cross section data. The FTCs of the CANDU reactor were newly analyzed using the TRITON module in the SCALE6 code system, and the BONAMI module was executed to apply the Bondarenko approach for self-shielded cross sections. When compared with some reactor physics codes resulting in slightly positive FTC in the specific region, the FTCs evaluated in this study showed a clear negativity over the entire fuel temperature range on fresh/equilibrium fuel. In addition, the FTCs at 960.15 K were slightly negative during the entire burnup. The effects on FTCs from the library difference between ENDF/B-VI.8 and ENDF/B-VII.0 are recognized to not be large; however, they appear more positive when more Pu-239 productions with burnup are considered. This feasibility study needs an additional benchmark evaluation for FTC calculations, but it can be used as a reference for a new FTC analysis in CANDU reactors

  20. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    CANDU PHWRs have demonstrated generic benefits which will be continued in future designs. These include economic benefits due to low operating costs, business potential, strategic benefits due to fuel cycle flexibility and operational benefits. These benefits have been realized in Korea through the operation of Wolsong 1, resulting in further construction of PHWRs at the same site. The principal benefit, low electricity cost, is due to the high capacity factor and the low fuel cost for CANDU. The CANDU plant at Wolsong has proven to be a safe, reliable and economical electricity producer. The ability of PHWR to burn natural uranium ensures security of fuel supply. Following successful Technology Transfer via the Wolsong 2,3 and 4 project, future opportunity exists between Korea and Canada for continuing co-operation in research and development to improve the technology base, for product development partnerships, and business opportunities in marketing and building PHWR plants in third countries. High reliability, through excellent design, well-controlled operation, efficient maintenance and low operating costs is critical to the economic viability of nuclear plants. CANDU plants have an excellent performance record. The four operating CANDU 6 plants, operated by four utilities in three countries, are world performance leaders. The CANDU 9 design, with higher output capacity, will help to achieve better site utilization and lower electricity costs. Being an evolutionary design, CANDU 9 assures high performance by utilizing proven systems, and component designs adapted from operating CANDU plants (Bruce B, Darlington and CANDU 6). All system and operating parameters are within the operating proven range of current plants. KAERI and AECL have an agreement to perform joint studies on future PHWR development. The objective of the joint studies is to establish the requirements for the design of future advanced CANDU PHWR including the utility need for design improvements

  1. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    We describe the development philosophy, direction and concepts that are being utilized by AECL to refine the CANDU reactor to meet the needs of current and future competitive energy markets. The technology development path for CANDU reactors is based on the optimization of the pressure tube concept. Because of the inherent modularity and flexibility of this basis for the core design, it is possible to provide a seamless and continuous evolution of the reactor design and performance. There is no need for a drastic shift in concept, in technology or in fuel. By continual refinement of the flow and materials conditions in the channels, the basic reactor can be thermally and operationally efficient, highly competitive and economic, and highly flexible in application. Thus, the design can build on the successful construction and operating experience of the existing plants, and no step changes in development direction are needed. This approach minimizes investor, operator and development risk but still provides technological, safety and performance advances. In today's world energy markets, major drivers for the technology development are: (a) reduced capital cost; (b) improved operation; (c) enhanced safety; and (d) fuel cycle flexibility. The drivers provide specific numerical targets. Meeting these drivers ensures that the concept meets and exceeds the customer economic, performance, safety and resource use goals and requirements, including the suitable national and international standards. This logical development of the CANDU concept leads naturally to the 'Next Generation' of CANDU reactors. The major features under development include an optimized lattice for SEU (slightly enriched uranium) fuel, light water cooling coupled with heavy water moderation, advanced fuel channels and CANFLEX fuel, optimization of plant performance, enhanced thermal and BOP (balance of plant) efficiency, and the adoption of layout and construction technology adapted from successful on

  2. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    CANDU is a relatively young technology which has demonstrated many achievements as an electrical power generation system. These achievements include an unsurpassed safety record, high annual and lifetime capacity factors, low electricity cost and a broad range of other performance strengths which together indicate that the CANDU technology is fundamentally sound. Known capabilities not yet fully exploited, such as advanced fuel cycle options, indicate that CANDU technology will continue to pay strong dividends on research, development and design investment. This provides a strong incentive for the improvement of CANDU on a continuing basis

  3. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  4. Explicit core-follow simulations for a CANDU 6 reactor fuelled with recovered-uranium CANFLEX bundles

    International Nuclear Information System (INIS)

    Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. After fission products and plutonium (Pu) have been removed from spent LWR fuel, RU is left. A fissile content in the RU of 0.9 to 1.0% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. Explicit core-follow simulations were run to analyse the viability of RU as a fuel for existing CANDU 6 cores. The core follow was performed with RFSP, using WIMS-AECL lattice properties. During the core follow, channel powers and bundle powers were tracked to determine the operating envelope for RU in a CANFLEX bundle. The results show that RU fits the operating criteria of a generic CANDU 6 core and is a viable fuel option in CANDU reactors. (author)

  5. The clearance potential index and hazard factors of CANDU fuel bundle and a comparison of experimental-calculated inventories

    International Nuclear Information System (INIS)

    In the field of radioactive waste management, the radiotoxicity can be characterized by two different approaches: 1) IAEA, 2004 RS-G-1.7 clearance concept and 2) US, 10CFR20 radioactivity concentration guides in terms of ingestion / inhalation hazard expressed in m3 of water/air. A comparison between the two existing safety concepts was made in the paper. The modeled case was a CANDU natural uranium, 37 elements fuel bundle with a reference burnup of 685 GJ/kgU (7928.24 MWd/tU). The radiotoxicity of the light nuclide inventories, actinide, and fission-products was calculated in the paper. The calculation was made using the ORIGEN-S from ORIGEN4.4a in conjunction with the activation-burnup library and an updated decay data library with clearance levels data in ORIGEN format produced by WIMS-AECL/SCALENEA-1 code system. Both the radioactivity concentration expressed in Curie and Becquerel, and the clearance index and ingestion / inhalation hazard were calculated for the radionuclides contained in 1 kg of irradiated fuel element at shutdown and for 1, 50, 1500 years cooling time. This study required a complex activity that consisted of various phases such us: the acquisition, setting up, validation and application of procedures, codes and libraries. For the validation phase of the study, the objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from a Pickering CANDU reactor with inventories predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with the time dependent cross sections library, CANDU 28.lib, produced by the sequence SAS2H of SCALE 4.4a. In this way, the procedures, codes and libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors are being qualified and validated, in support for the safety management of the radioactive wastes

  6. Coolability of severely degraded CANDU cores. Revised

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  7. The effect of fuel power on the leaching of cesium and iodine from used CANDU fuel

    International Nuclear Information System (INIS)

    The safety assessment of the concept of geological disposal of used fuel requires a source term for the instantaneous release of long-lived radionuclides from used fuel. Preferential release of the gap inventories of Cs-137 and I-129 from used fuel with a variety of linear power ratings (LPR) was studied. A one-to-one and ten-to-one correlation exists between measured gap inventories of stable xenon and the amount of Cs-137 and I-129 in the gap, for high-LPR and low-LPR fuels, respectively, as obtained from 5-day leaching experiments. These differences in release patterns for high- and low-LPR fuels can be explained by differences in the microstructure, and disappear when longer leaching times (i.e. 3 months) are used. These results imply that, on a geological time scale, the entire inventory of the long-lived isotopes of cesium and iodine must be considered as part of the instantaneous release source term. Attempts to quantify inventories of cesium and technetium at grain boundaries by comparing the short-term leaching behaviour of oxidized and non-oxidized fuel indicated that either very insoluble cesium-uranium compounds may exist at the grain boundaries, or that cesium and stable xenon grain boundary inventories are not similar. More research is needed before a firm conclusion can be reached as to whether the entire grain boundary inventories of Cs-137 (and Tc-99), as estimated from power histories, should be part of the instantaneous release source term

  8. On the domestic fuel channel for BWR

    International Nuclear Information System (INIS)

    Kobe Steel Ltd. started the domestic manufacture of fuel channel boxes for BWRs in 1967, and entered the actual production stage four years after that. Since 1976, the mass production system was adopted with the increase of the demand. The requirements about the surface contamination and the dimensional accuracy over whole length are very strict in the fuel channel boxes, moreover, special consideration must be given so as to prevent the deformation in use. The unique working methods such as electron beam welding, high temperature press forming and so on are employed in Kobe Steel Ltd. to satisfy such strict requirements, therefore the quality of the produced fuel channel boxes is superior to imported ones. At present, the fuel channel boxes domestically made by Kobe Steel Ltd. are used for almost all BWRs in Japan. The functions of fuel channel boxes are to flow boiling coolant uniformly upward, to guide control rods, and to increase the rigidity of fuel assembly. The fuel channel boxes are the square tubes of zircaloy 4 of 134.06 mm inside width, 2.03 mm thickness, and 4118 or 4239 mm length. The progress of the development and the features of the fuel channel boxes and the manufacturing processes are described. Zircaloy plates are formed into channels, and two channels are electron beam-welded after the edge preparation, to make a box. Ultrasonic examination and stress relief treatment are applied, and clips and spacers are welded. (Kako, I.)

  9. The evolution of Candu fuel cycles and their potential contribution to world peace

    International Nuclear Information System (INIS)

    This paper describes how several factors, including Canada's early focus on heavy-water reactor technology, limited heavy-industry infrastructure, and desire for both technological autonomy and energy self- sufficiency, contributed to the creation of the first Candu reactor in 1962. (author)

  10. Fuel channel life limiting factors that dictate fuel channel maintenance requirements

    International Nuclear Information System (INIS)

    CANDU reactors have been operating for 33 years. The Nuclear Power Demonstration (NPD) Unit started up in 1962 and the prototype of CANDU, Douglas Point, started in 1967. The first commercial reactors, Pickering Units 1 and 2 both went into service in 1971 closely followed by Units 3 and 4 in 1972 and 1973 respectively. Operating commercial reactor experience represents over 10,000 pressure tubes, not including the replaced channels in all the Pickering A Units, and nearly 130,000 pressure tube operating years. No pressure tube has yet operated for its 30 year design lifetime of 210 KEFPH at 80% capacity factor. The longest operating time for pressure tubes to-date is about 120 KEFPH in Pickering Unit 4. Many lessons have been learned regarding pressure tube life limiting factors from the early CANDU units and these, together with the information obtained from an extensive pressure tube R and D program, have resulted in many design changes and improvements in material properties, mainly from manufacturing route changes. Reactors built recently are expected to achieve their 30 year design life. The development of Periodic and In-service Inspection programs and equipment, assessment methodologies and acceptance criteria, and the development of maintenance tooling and procedures are enabling the life limiting factors to be addressed in the currently operating units. The life limiting factors in currently operating Units are reviewed in relation to the experience gained from the early units, the R and D programs and the inspection and maintenance performed to date. (author)

  11. An integrated CANDU system

    International Nuclear Information System (INIS)

    Twenty years of experience have shown that the early choices of heavy water as moderator and natural uranium as fuel imposed a discipline on CANDU design that has led to outstanding performance. The integrated structure of the industry in Canada, incorporating development, design, supply, manufacturing, and operation functions, has reinforced this performance and has provided a basis on which to continue development in the future. These same fundamental characteristics of the CANDU program open up propsects for further improvements in economy and resource utilization through increased reactor size and the development of the thorium fuel cycle

  12. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    This presentation reviews some of the main factors that will govern the design and operation of reactors in remote Northern Canadian communities, as applied to a small CANDU-type power plant. The central advantage of the CANDU is the fact that it is modular at the level of a single fuel channel. Examining each of the main features of this SMR plant on a hypothetical site in the Canadian Arctic reveals some of the unique characteristics that will be either desirable or mandatory for any such power plant applied to service in this remote region. (author)

  13. A comparative assessment result of B9401 multi-channel RIH break experiment with Canadian test facility RD-14M using RELAP5/MOD3 and RELAP5/CANDU+

    International Nuclear Information System (INIS)

    The experiment, B9401, performed in RD-14M multi-channel experimental facility, was preliminary analyzed using RELAP5/MOD3 and RELAP5/CANDU+ and compared with experimental results. The RELAP5 code has been developed for best-estimate transient simulation of pressurized water reactors and associated system, but the RELAP5/CANDU+ code has been developed since 1998 in order to have auditing tool of CANDU NPP. The RELAP5/CANDU+ code is under developing and they have not been assessed much for a CANDU reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14M facility and to get the assessment results for RELAP5/CANDU+ code. The RD-14M test facility at Whiteshell Nuclear Research Establishment is a full-scale multi-channel pressurized-water loop. The RD-14M is not a 'scale' model of any particular CANDU reactor. It possesses many geometric features of a CANDU reactor heat transport system, and is capable of operating at conditions similar to those expected to occur in a reactor under normal operation and some postulated accident conditions. As preliminary results, the RELAP5/MOD3 and RELAP5/MOD+ analyses demonstrate the code's capability to predict reasonably the main phenomena occuring in the transient, in qualitative view. In quantative view, the RELAP5/CANDU+ predicted better than that of RELAP5. However, some discrepancies after emergency coolant injection, the behaviors of the ECI mass flow rate and the sheath temperatures were observed commonly

  14. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium: 1.2 weight percent U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed. (author)

  15. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle (AECL-6594)

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium - 1.2 weight % U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed

  16. The approach to enhancing CANDU safety

    International Nuclear Information System (INIS)

    Enhancements to CANDU safety have evolved in both design and operation, beginning with the operating experience with heavy-water research reactors in the 1950s. The CANDU safety philosophy has always placed great emphasis on quantitative reliability targets for the special safety systems, and on their separation from the normal operating (process) systems. The desire to minimize the risk of serious accidents was reflected in an early design requirement that CANDU must withstand a process system failure coupled with the unavailability of any special safety system, without resulting in significant core damage. For example, a loss-of-coolant plus a complete absence of emergency core cooling does not lead to a core melt; the decay heat is removed relatively passively from the fuel channels by the moderator. Confidence in CANDU safety design continues to be enhanced through more thorough safety analysis, more detailed probabilistic safety assessments, and continuing research and development. Prevention of accidents through quality-assured engineering has been improved; current projects use a comprehensive CADDS information management system which provides evidence that safety requirements are met, from design through construction, commissioning and operation

  17. Proceedings of the third international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately

  18. Neutronics and thermalhydraulics characteristics of the CANDU core fueled with slightly enriched uranium 0.9% U235

    International Nuclear Information System (INIS)

    The interest concerning the slightly enriched uranium (SEU) fuel cycle is due to the possibility to adapt (to convert) the current reactor design using natural uranium fuel to this cycle. Preliminary evaluations based on discharged fuel burnup estimates versus enrichment and on Canadian experience in fuel irradiation suggest that for a 0.93% U-235 enrichment no design modifications are required, not even for the fuel bundle. The purpose of this paper is to resume the results of the studies carried on in order to clarify this problem. The calculation methodology used in reactor physics and thermal-hydraulics analyses that were performed adapted and developed the AECL suggested methodology. In order to prove the possibility to use the SEU 0.93% without any design modification, all the main elements from the CANDU Reactor Physics Design Manual were studied. Also, some thermal-hydraulics analyses were performed to ensure that the operating and safety parameters were respected. The estimations sustain the assumption that the current reactor and fuel bundle design is compatible to the using of the SEU 0.93% fuel. (author)

  19. The post-irradiation examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA Reactor

    International Nuclear Information System (INIS)

    The INR hot cells have 10 years of practice in post-irradiation examination (PIE) on experimental nuclear fuel elements and structure materials. This paper summarises the result of a typical PIE work carried out on an experimental CANDU type fuel element irradiated in an assembly of six rods in a power ramp test in the TRIGA 14 MV (th) materials testing reactor. The fuel element has attained practically a burnup of 188.4 MWh/kg U (10% accuracy) as determined by nondestructive gamma scanning method, and of 194.3 MWh/kg U (3 % accuracy) as determined by destructive mass spectrometry method. These results determined by nondestructive and destructive methods are in agreement. The eddy current control for clad integrity has revealed the integrity of the fuel element, a fact also confirmed by the fuel puncture for internal gas pressure measurement. The metallography control of the cladding has revealed good quality welding and an acceptable quality brazing of a bearing pad. The ceramographic control of the fuel revealed an expected two-zone structure, except one end of the fuel element where a three-zone structure was found, due to the higher thermal rating induced by the flux peak. The results are presented in measurement worksheets and are accompanied by diagrams and pictures. (author)

  20. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    operating plants. The EC6 will utilize modern computers and control systems housed in an Advanced Control Room which along with automated testing will make the plants much easier to operate with minimal operator intervention. Improvements to the fire protection system and enhanced security features will further protect the assets. And given the CANDU 6's inherent ability to utilize different fuel cycles and the Advanced MACSTOR module to store spent dry fuel further add to the attractiveness of the EC6 product. This paper describes the various enhancements that are being made to the EC6 and explains these features in more detail. (authors)

  1. The next generation CANDU 6

    International Nuclear Information System (INIS)

    AECL's product line of CANDU 6 and CANDU 9 nuclear power plants are adapted to respond to changing market conditions, experience feedback and technological development by a continuous improvement process of design evolution. The CANDU 6 Nuclear Power Plant design is a successful family of nuclear units, with the first four units entering service in 1983, and the most recent entering service this year. A further four CANDU 6 units are under construction. Starting in 1996, a focused forward-looking development program is under way at AECL to incorporate a series of individual improvements and integrate them into the CANDU 6, leading to the evolutionary development of the next-generation enhanced CANDU 6. The CANDU 6 improvements program includes all aspects of an NPP project, including engineering tools improvements, design for improved constructability, scheduling for faster, more streamlined commissioning, and improved operating performance. This enhanced CANDU 6 product will combine the benefits of design provenness (drawing on the more than 70 reactor-years experience of the seven operating CANDU 6 units), with the advantages of an evolutionary next-generation design. Features of the enhanced CANDU 6 design include: Advanced Human Machine Interface - built around the Advanced CANDU Control Centre; Advanced fuel design - using the newly demonstrated CANFLEX fuel bundle; Improved Efficiency based on improved utilization of waste heat; Streamlined System Design - including simplifications to improve performance and safety system reliability; Advanced Engineering Tools, -- featuring linked electronic databases from 3D CADDS, equipment specification and material management; Advanced Construction Techniques - based on open top equipment installation and the use of small skid mounted modules; Options defined for Passive Heat Sink capability and low-enrichment core optimization. (author)

  2. Radionuclide Release after Channel Flow Blockage Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon; Jun, Hwang Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The channel flow blockage accident is one of the in core loss of coolant accidents, the release path of radionuclide is very different from conventional loss of coolant accidents. The large amount of radionuclide released from broken channel is being washed during it passes through the moderator in Calandria. The objective of containment behavior analysis for channel flow blockage event is to assess the amount of radionuclide release to the ambient atmosphere. Radionuclide release rates in case of channel flow blockage with all safety system available, that is containment building is intact, as well as with containment system impairment are analyzed with GOTHIC and SMART code

  3. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book Canada Enters the Nuclear Age. The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  4. New selection criteria for channel refueling of a Candu-6 reactor: introduction to floppy rules

    International Nuclear Information System (INIS)

    A revised set of rules is in use at Gentilly-2 NGS for the selection of channels for refuelling. Traditional hard channel rejection rules (of go/no-go type) have been replaced by a more efficient set of soft evaluation rules based on concepts borrowed to the Fuzzy Logic. New evaluation rules, labelled as 'Floppy Rules', enable to assess and rate the channel suitability for refuelling by using a smooth and natural continuum of values qualifying excellent, good, fair and poor choices. Global channel suitability for refuelling is measured by combining separate ratings obtained from individual evaluation rules. Each evaluation rule is based on a specific control parameter related to local or lumped core properties. Two new software codes (NEWRULES and REFUEL) designed around the concept of Floppy Rules enable to perform a very efficient selection of optimized channel refuelling sequences either in manual and automatic mode. (author)

  5. Modeling CANDU-type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

    International Nuclear Information System (INIS)

    The high burnup database for CANDU fuel includes several cases from both power station and experimental reactor irradiations, with achieved burnups of up to 800 MW.h/kgU. The power history for each of these cases is different, encompassing low steady-state, declining, and power-ramps. This variety offers a good opportunity to check the models of fuel behaviour, and to identify areas for improvement. The main parameters for comparing calculated versus measured data are the fission gas release and the sheath hoop strain. Good agreement of calculated values of these two parameters with experimental data indicates that the global behaviour of the fuel element is adequately simulated by our codes. The ELESIM computer code was used as the simulation tool. The models for fission gas release, swelling and for fuel pellet expansion were thoroughly analysed. Changes were proposed for both models. The fuel pellet expansion model was modified to account for gaseous swelling, which becomes very important at high burnups. As well, the mathematics of the fission gas release model was upgraded for the diffusional release of fission gas atoms to the grain boundaries. A revised version of the ELESIM computer code was used to simulate the cases from the high burnup database. Satisfactory agreement was found for most cases. The discrepancies are discussed in view of alternative mechanisms that can operate and be enhanced at high burnup. These include stoichiometry changes with burnup that affects fission gas release, and also outer pellet rim fission gas release by a grain boundary diffusion process. The main conclusion of this study is that the revised version of the ELESIM code is able to simulate with reasonable accuracy high burnup as well as low burnup CANDU fuel. This includes irradiations of steady-state, declining, or ramped fuel power histories with a prolonged hold at high power. However, future improvements to ELESIM are needed to model fuel power histories with short dwell

  6. CANDU market prospects

    International Nuclear Information System (INIS)

    This 1994 survey of prospective markets for CANDU reactors discusses prospects in Turkey, Thailand, the Philippines, Korea, Indonesia, China and Egypt, and other opportunities, such as in fuel cycles and nuclear safety. It was concluded that foreign partners would be needed to help with financing

  7. Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

    2015-03-15

    The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

  8. Fuel channel in-service inspection programs program design for maximum cost effectiveness

    International Nuclear Information System (INIS)

    Inspection is an integral part of fuel channel life management strategy. Inspection data is used to assess the state of reactor core integrity and provide the information necessary to optimize long term maintenance programs. This paper will provide an overview of the structured approach to developing fuel channel inspection programs within OHN. The inspection programs are designed to balance the resources utilized (cost, outage time, and dose expenditure) with the benefits provided by the inspection data obtained (improved knowledge of component status, degradation mechanisms and rates, etc..). The CANDU community has yet to have a fuel channel operate for a full 30 year design life. Since research programs can not fully simulate reactor operating conditions, inspections become an essential feature of the life management strategy as the components age. Inspection programs often include activities designed to develop predictive capability for long term fuel channel behaviour and provide early warning of changes in behaviour. It should be noted that although this paper addresses the design of fuel channel inspection programs, the basic principles presented can be applied to the design of inspection programs for any major power plant component or system. (author)

  9. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  10. The enhanced CANDU 6 reactor - Generation III CANDU medium size global reactor

    International Nuclear Information System (INIS)

    Full text: The Enhanced CANDU 6TM (EC6TM) is a Generation III 700 class, heavy water moderated pressure tube reactor, designed to provide safe, reliable, nuclear power. The EC6TM has evolved from the proven CANDU 6 plants licensed and operating in five countries (four continents) with over 150 reactor years of safe operation around the world. In recent years. this global CANDU 6 fleet, with over 92% average gross capacity factor has ranked in the world's top performing reactors. The EC6 reactor builds on this success of the CANDU 6 fleet by using the operation, experience and project feedback to upgrade the design and construction techniques. A key objective of the EC6 has been to review and incorporate design improvements in the CANDU 6 to meet current safety standards. The key characteristics of the highly successful CANDU 6 reactor design include: - Powered by natural Uranium; - Ease of installation with modular, horizontal fuel channel core; - Separate low-temperature, low-pressure moderator providing inherently passive heat sinks; Reactor vault filled with light water surrounding the core; - Two independent safety shutdown systems; - On-power fuelling; - The CANDU 6 plant has a highly automated control system, with plant control computers that adjust and maintain the reactor power for plant stability (which is particularly beneficial in less developed power grids-where fluctuations occur regularly and capacities are limited). The major improvements incorporated in the EC6 design include, - More robust containment and increased passive features e.g., thicker walls, steel liner; - Enhanced severe accident management with additional emergency heat removal systems; - Improved shutdown performance for improved Large LOCA margins; - Upgraded fire protection systems to meet current Canadian and International standards; - Additional design features to improve environmental protection for workers and public- ALARA principle; - Automated and unitized back-up standby

  11. A study on the radioactive waste management for DUPIC fuel cycle - A study on the direct use of spent PWR fuel in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyeon Soo; Chun, Kwan Sik; Kim, Jong Ho; Cho, Yeong Hyeon; Paek, Seung Uh; Kang, Hee Seok; Na, Jeong Won; Choi, Jong Won; Lee, Hu Keun; Park, Keun Il; Yang, Seung Yeong [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The characteristics of the radioactive wastes coming from the DUPIC fuel manufacturing process were analyzed and evaluated. The release amounts of nuclides were estimated, and the characteristics of the oxides of semi-volatile nuclides such as Ru and Cs were analyzed and also these trapping and fixation technologies were assessed. The conceptual methods of experiment including apparatus and procedure were completed to define the vaporization behavior of nuclides. The gross {alpha}-activity and {alpha}-, {gamma}-spectrum of irradiated zircaloy specimens from KORI unit 1 were analyzed and the press to test the compaction of hulls was manufactured. Ingestion hazard index for the disposal of spent DUPIC fuel were calculated using ORIGEN 2 code and compared with the disposal of once-through spent PWR and CANDU fuels, and then the index per kwh for DUPIC cycle was 1/8 {approx} 1/4 lower than that for once-through cycle. (Author).

  12. A study on the radioactive waste management for DUPIC fuel cycle -A study on the direct use of spent PWR fuel in CANDU-

    International Nuclear Information System (INIS)

    The characteristics of the radioactive wastes coming from the DUPIC fuel manufacturing process were analyzed and evaluated. The release amounts of nuclides were estimated, and the characteristics of the oxides of semi-volatile nuclides such as Ru and Cs were analyzed and also these trapping and fixation technologies were assessed. The conceptual methods of experiment including apparatus and procedure were completed to define the vaporization behavior of nuclides. The gross α-activity and α-, γ-spectrum of irradiated zircaloy specimens from KORI unit 1 were analyzed and the press to test the compaction of hulls was manufactured. Ingestion hazard index for the disposal of spent DUPIC fuel were calculated using ORIGEN 2 code and compared with the disposal of once-through spent PWR and CANDU fuels, and then the index per kwh for DUPIC cycle was 1/8 ∼ 1/4 lower than that for once-through cycle. (Author)

  13. Safety Evaluation Report for CANDU Spent Fuel Dry Storage Basket Drop Test

    International Nuclear Information System (INIS)

    This is technical report about the safety demonstration test for the CANDU S/F storage basket. Two demonstration tests were performed such as one case of dropping a basket into the bottom of the cylinder from 7.5 m height and another case of dropping a basket onto the other basket loaded in a cylinder. The existing basket was improved to satisfy the performance requirement. The leak test showed that the enhanced basket had no leak after all the drop tests. The dropped basket was able to be withdrawn through top of the cylinder by a handling tool, which proved the maintenance of retrievability after drop accident

  14. CANDU 600 fueling machine non-linear finite element modelling and modal testing

    International Nuclear Information System (INIS)

    This paper describes the development of the non-linear finite element model of and explains the significant non-linearities in the CANDU 600 fuelling machine system. Also discussed are the features and functions of a special purpose non-linear finite element code and the full scale modal testing of the fuelling machine test rig in the laboratory environment. The results for frequency response analysis, free vibration and forced vibration cases are briefly described for a simple test model using the non-linear code

  15. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  16. Status of the demonstration irradiation of the CANDU new fuel bundle CANFLEX-NU in Korea

    International Nuclear Information System (INIS)

    A demonstration irradiation (DI) of 24 KNFC made CANFLEX-NU fuel bundles in the Wolsong Power Generation Station-i has been conducted jointly by KEPRI/KHNP/KAERI since July 10, 2002. By selecting the Q07 (high power) and L21(low power) channels, the total 24 and 16 CANFLEX bundles were respectively loaded into and discharged from the reactor by 2003 August, and the final discharge of the other 8 CANFLEX bundles is expected on around February 2004. Tracking the reactor operation data, it is noted that the reactor has been stably operated during the DI. One CANFLEX bundle irradiated in the Q07 channel had a typical history of high power and high burnup, having the outer element power rating of ∼ 41 kW/m at the fuelling, ∼ 42 kW/m as a maximum power rating at the burnup of ∼ 50 MWh/kgU, and ∼ 35 kW/m at the discharge burnup of ∼ 210 MWh/kgU. While, another CANFLEX bundle also irradiated in the Q07 channel had a typical history of power ramping, having a outer element power rating of ∼ 7 kW/m from the fuelling to the burnup of ∼ 48 MWh/kgU at which the element powers were ramped to a ∼ 35 kW/m maximum element power rating, and ∼ 30 kW/m at the discharge burnup of 188 MWh/kgU. An unusual performance and integrity of the CANFLEX elements could not be found in the ELESTRES predictions. By looking at the discharged CANFLEX bundles in the bay, all the bundles were intact, free of defects and appeared to be in good condition. A detailed in-bay visual examinations and dimensional measurements of the discharged CANFLEX bundles will be made at the end of 2003. (author)

  17. Resolving wall loss in Candu outlet feeder pipes

    International Nuclear Information System (INIS)

    The primary circuit carbon steel feeder piping of CANDU reactors with Cr content in the range 0.02-0.04 wt% has been found to show wall thinning at the exit of the fuel channels as a result of a flow accelerated corrosion mechanism. The observed wall thinning correlates well with the coolant velocity in the fuel channels and is dependent on the pHa of the coolant. The pHa is the room temperature pH reading of a heavy water sample measured using a glass electrode calibrated using light water standards. Based on the results of recent low velocity corrosion tests, operation at the low end of the original pHa specification will allow the feeder pipes in existing CANDU reactors to achieve their design life. For reactors under construction and for future CANDU reactors, the feeder pipe material is now specified to have a minimum Cr content of 0.20 wt%, and available information indicates that the presence of Cr at these concentrations will eliminate concerns regarding flow accelerated corrosion of CANDU outlet feeder piping. (authors)

  18. A complete wood-fuel channel

    Energy Technology Data Exchange (ETDEWEB)

    Alfonsi, F. [Agence de l`Environnement et de la Maitrise de l`Energie, 75 - Paris (France)

    1993-12-31

    In Corsica, eight heating plants are supplied by a Mixed Economy Company (SEM). A complete wood fuel channel has been developed, but for wood fuel to be competitive, the Corsican experimentation proves that some conditions have to be obtained. (TEC). 1 fig.

  19. ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

    International Nuclear Information System (INIS)

    1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture

  20. Enhancing the seismic capability of the on-power refueling system of the CANDU reactor

    International Nuclear Information System (INIS)

    The CANDU reactor assembly includes several hundred horizontal fuel channels, each containing twelve fuel bundles, arranged in a square lattice, and supported by the reactor structures. CANDU operates on natural uranium or other low fissile content fuel, and is refueled on-power, with either four or eight fuel bundles in a channel being replaced during each refueling operation. The fueling machines clamp onto the opposite ends of the fuel channel to be refueled. The seismic capacity of this refueling system is evaluated in terms of its dynamic response during an earthquake. This paper describes the approach adopted to enhance the seismic capability of the fueling machine and calandria assembly for earthquakes of O.3g ground acceleration covering a broad range of soil conditions ranging from soft to hard. A detailed, 3-D finite element seismic model of the fueling machine and calandria assembly system is developed to calculate the seismic responses of the structure. Some relatively simple hardware design changes have been considered to increase the seismic capacity of the CANDU 6 reactor. These changes in the fueling machine and calandria assembly of the CANDU 6 reactor are briefly described. They have been incorporated into the finite element seismic model of the system. Most of these design changes have already been considered and implemented in other CANDU reactor projects. The current CANDU 6 reactor design fully meets the requirements of seismic qualification for sites with potential for O.2g ground acceleration where the seismic loads need to be combined with the other design loads for the support and pressure boundary components to demonstrate compliance with the applicable Code requirements. In the present study it is demonstrated that, with relatively simple hardware changes, the fueling machine and calandria assembly of the CANDU 6 reactor can withstand earthquakes of O.3g ground acceleration. Based on the current study and some preliminary analysis of the

  1. 6. CNS international conference on CANDU maintenance. Proceedings

    International Nuclear Information System (INIS)

    The 6th CNS International Conference on CANDU Maintenance took place in Toronto, Ontario on November 16-18, 2003. The theme for the conference was 'Maintenance for Life'. About 270 delegates attended the conference held by the Canadian Nuclear Society. The conference consisted of four parallel sessions, a pattern that continued throughout the conference. Papers were grouped under the following headings: Fuel Channels and End Fittings - Assessments; Fuel Channels and End Fittings - Inspections; Fuel Channels and End Fittings - Maintenance; Fuel Channels and End Fittings - Universal Delivery Machine; Water Upgrading; Performance and Plant Life Improvement; Steam Generator Life Management; Steam Generator Modifications; Steam Generators - Inspections; Steam Generators - Assessments; Maintenance Programs; Feeder Inspections; Feeder Assessment and Mitigation; Valve Maintenance; Instrumentation and Control; Inspection Technology; and Fuel Handling

  2. Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes

    International Nuclear Information System (INIS)

    Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-235U)O2 on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of 241Am, 243Am, 235U and 232Th respectively. 132.8 kg of 233U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of 241/243Am, 235U and 232Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional 233U production.

  3. SLARette operations at CANDU 6 stations

    International Nuclear Information System (INIS)

    The Objective of the SLARette system as an ongoing maintenance device was to design a system which could be installed in reactor, commissioned and operational in less than 48 hours and removed in less than 24 hours, thus allowing a selected number of fuel channels to be SLARed during the normal yearly maintenance outage. This required a relatively simple design that does not require any changes to the Station Fuel Handling Systems and is portable, reliable and easily maintained. Over the last 10 years virtually every part of the SLARette system has been enhanced to further reduce man-hour and man-rem expenditure and it has now been used successfully on four CANDU 6 stations and many fuel channels. (author)

  4. Fission product inventory for CANFLEX-NU and CANFLEX-RU in the limiting channel of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2007-09-15

    The fission product inventory was evaluated for the 37-element fuel, CANFLEX-NU and CANFLEX-RU fuels which irradiated in the assumed limiting O6{sub m}od channel under a normal operating condition. Generally, the assessment of fission product release is divided into 3 steps which are a calculation of FGR during steady state, a calculation of a prompt release at the time of an accident, and a calculation of delayed release after the accident. Among them the first step was described in this report. The analysis was performed by using the ELESTRES code which is a computer code used in a performance evaluation of the fuel bundle during the normal irradiation. The calculation results of the fission product inventory for the 37-element fuel, CANFLEX-NU and CANFLEX-RU were compared to each other.

  5. Applying operating experience to design the CANDU 3 process

    International Nuclear Information System (INIS)

    personnel in the early design stages of a project. This has been accomplished in a variety of ways for the CANDU 3 design, including visits to nuclear utilities, exchange of staff with utilities, and hiring of utility staff as consultants. Input from equipment manufacturers is also a key item, particularly in the area of the technical specifications that are produced by designers. CANDU 3 approached this by requesting comments on technical specifications from traditional or established suppliers and by pre-selecting major equipment suppliers in advance - to permit early discussions on equipment details, and by giving consulting contracts to suppliers to establish major design parameters. Design reviews have long been a traditional part of the design process. For the CANDU 3 participation of utility representatives has been actively solicited in order to incorporate their comments at an early stage. Some obvious benefits of this approach were in the areas of operational testing, maintainability, and that of the fuel channel design. The application of a more formal human factors program to CANDU 3 design has led to the development of a functional design methodology which requires designers in key areas to evolve their designs from operating procedures, and thus produce designs which are easier for operators and maintainers. This is the reverse of the traditional approach where operating procedures produced after design completion have had to adapt to design features. The paper includes an outline of the application of the functional design methodology to the design of an advanced control centre for CANDU 3. It also shows how operating and maintainability considerations have been factored in at all design stages based on inputs from operating utilities

  6. Installation of an irradiated fuel bundle discharge counter at Bruce NGS-B 3 000 MW(e) CANDU power station

    International Nuclear Information System (INIS)

    Design, manufacture and installation of an irradiated fuel bundle discharge counter for the multi-unit CANDU Bruce NGS-B Generating Station involved contributions from the International Atomic Energy Agency (Agency), designers (AECL), contractors, manufacturers, utility and the regulatory agency. The installation at Bruce NGS-B was the first made by the Agency as a retrofit to a multi-unit CANDU reactor approaching its fist critical operation, where the whole project was the responsibility of the Agency and where the original design of the reactor had not had provision for the Agency equipment. The scheduling and integration of the installation into the normal activities involved in starting up a 3 000 MW(e) multi-unit generating station were successfully achieved. The Agency has demonstrated the capability and performance of the fuel discharge counter

  7. Life Assurance Strategy for CANDU NPP

    International Nuclear Information System (INIS)

    Nuclear Plants have a nominal 'design life' that forms the basis of equipment specification, economic evaluations and licensing for some jurisdictions. Some component in the plant may require replacement, refurbishing and or rehabilitation during the plant 'design life'. Components which are extremely difficult or economically impossible to replace will place a limit on plant life. Rehabilitation programs completed to date on older CANDU plants to improve reliability of plant components, coupled with R and D programs, experimental data and advanced analytical methods form the basis for CANDU plant component life assurance. Life assurance is verified during plant operation by comprehensive in-service inspection programs and laboratory examinations. The paper provides an overview of the experiences to date on Refurbishment and Rehabilitation programs and some Canadian approaches on the main activities involved in scoping and managing nuclear plant life assurance. A number of proactive programs are underway to anticipate, detect and mitigate potential aging degradation at an early stage to ensure plant safety and reliability. Some of these programs include; systematic plant condition assessment, refurbishment and upgrading programs, environmental qualification programs and a program of examination of components from decommissioned reactors. These programs are part of an overall nuclear power plant maintenance strategy. Beyond life assurance, a longer term approach would be geared towards life extension as a viable option for the future. Recent CANDU designs have benefited from the early CANDU experience and are expected to require less rehabilitation. Examples of changes in CANDU 6 include fuel channel design and adopting a closed component cooling water system. New designs are based on 'design life' longer than that used for economic evaluations. The approach is to design for easy replace ability for components that can be economically replaced. Specific examples

  8. Materials performance in CANDU reactors: The first 30 years and the prognosis for life extension and new designs

    International Nuclear Information System (INIS)

    A number of CANDU reactors have now been in-service for more than 30 years, and several are planning life extensions. This paper summarizes the major corrosion degradation operating experience of various out-of-core (i.e., excluding fuel channels and fuel) materials in-service in currently operating CANDU reactors. Also discussed are the decisions that need to be made for life extension of replaceable and non-replaceable components such as feeders and steam generators, and materials choices for new designs, such as the advanced CANDU reactor (ACR) and enhanced CANDU-6. The basis for these choices, including a brief summary of the R and D necessary to support such decisions is provided. Finally we briefly discuss the materials and R and D needs beyond the immediate future, including new concepts to improve plant operability and component reliability

  9. Absorber materials in CANDU PHWRs

    International Nuclear Information System (INIS)

    In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in the relatively benign environment of low pressure, low temperature heavy water between neighbouring rows or columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a redesigned back-fit resolved the problem. (author). 3 refs, 8

  10. A theoretical model of the mean grain size evolution during the liquid phase sintering in CANDU advanced fuels

    International Nuclear Information System (INIS)

    In order to enhance the fuel burnup in CANDU reactors, the microstructure of the fuel pellet must by modified, in order to accommodate the fission gas released during irradiation. A large grain size structure is able to retain a large fraction of the released gases. The addition of some oxides to the UO2 powder leads to the liquid phase ocarina during sintering. Starting from the basic theory of the grain growth in solid and liquid phase sintering and from the experimental microstructure characterization, a model of the mean size evolution in the presence of very small liquid fraction is proposed. As a function of dopant concentration, solubility limits and liquid phase composition, this model explicitly takes account of the continuous distribution of the additive into the matrix during grain growth process of the wetting properties of the liquid phase and of all transfer processes governing the microstructure evolution: grain boundary diffusion, surface reaction and diffusion through the liquid phase. The model validation was done for UO2-Nb2O3 and UO2-TiO2 systems, in isothermal treatment at 1700 deg. C. The model predictions are confirmed by the qualitative and quantitative experimental results, given by SEM, X-ray diffraction investigations. (authors)

  11. Modelling the release of volatile fission product cesium from CANDU fuel under severe accident conditions using artificial neural networks

    International Nuclear Information System (INIS)

    An artificial neural network (ANN) model has been developed to predict the release of volatile fission products from CANDU fuel under severe accident conditions. The model was based on data for the release Of 134Cs measured during three annealing experiments (Hot Cell Experiments 1 and 2, or HCE- 1, HCE-2 and Metallurgical Cell Experiment 1, or MCE- 1) at Chalk River Laboratories. These experiments were comprised of a total of 30 separate tests. The ANN established a correlation among 14 separate input variables and predicted the cumulative fractional release for a set of 386 data points drawn from 29 tests to a normalized error, En, of 0.104 and an average absolute error, Eabs, of 0.064. Predictions for a blind validation set (test HCE2-CM6) had an En of 0.064 and an Eabs of 0.054. A methodology is presented for deploying the ANN model by providing the connection weights. Finally, the performance of an ANN model was compared to a fuel oxidation model developed by Lewis et al. and to the U.S. Nuclear Regulatory Commission's CORSOR-M. (author)

  12. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    International Nuclear Information System (INIS)

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129I, 85Kr and 14C. (author). 104 refs., 9 tabs., 5 figs

  13. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  14. ELESTRES 2.1 computer code for high burnup CANDU fuel performance analysis

    International Nuclear Information System (INIS)

    The ELESTRES (ELEment Simulation and sTRESses) computer code models the thermal, mechanical and micro structural behaviours of CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains in fuel element design analysis and assessments. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. ELESTRES 2.1 was developed for high burnup fuel application, based on an industry standard tool version of the code, through the implementation or modification to code models such as fission gas release, fuel pellet densification, flux depression (radial power distribution in the fuel pellet), fuel pellet thermal conductivity, fuel sheath creep, fuel sheath yield strength, fuel sheath oxidation, two dimensional heat transfer between the fuel pellet and the fuel sheath; and an automatic finite element meshing capability to handle various fuel pellet shapes. The ELESTRES 2.1 code design and development was planned, implemented, verified, validated, and documented in accordance with the AECL software quality assurance program, which meets the requirements of the Canadian Standards Association standard for software quality assurance CSA N286.7-99. This paper presents an overview of the ELESTRES 2.1 code with descriptions of the code's theoretical background, solution methodologies, application range, input data, and interface with other analytical tools. Code verification and validation results, which are also discussed in the paper, have confirmed that ELESTRES 2.1 is capable of modelling important fuel phenomena and the code can be used in the design assessment and the verification of high burnup fuels. (author)

  15. Pre-licensing of the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  16. Advanced CANDU reactor technology: competitive design for the nuclear renaissance

    International Nuclear Information System (INIS)

    AECL has developed the design for a new generation of CANDU nuclear power plants, the Advance CANDU Reactor or ACR. The ACR combines a set of underlying enabling technologies with well-established successful CANDU features in an optimized design with significantly lower costs. By adopting slightly enriched uranium fuel, an optimized core design with light water coolant, heavy water moderator and reflector has been defined based on the existing CANDU fuel channel module. The basic design for the complete reference ACR power plant has now been completed. This paper summarizes the main features and characteristics of the reference ACR-700 power plant design. The progress of the ACR design program in meeting challenging cost, schedule and performance targets is described. AECL's cost reduction methodology is summarized as an integral part of the design optimization process. Examples are given of cost reduction features together with the enhancement of design margins. AECL expects the detailed design and testing of ACR to be complete and pre-project licensing evaluation carried out to enable regulatory endorsement in key markets by the middle of the decade. (authors)

  17. Fuel bundle loss of cooling inside the fuelling machine at CANDU6 PHWR

    International Nuclear Information System (INIS)

    This article describes the that loss of forced circulation cooling flow of induce spent fuel bundle loss of cooling and fission product releasing, analyzes the effect of reactor building and environment due to the fuel bundle rupturing, discusses the countermeasure to deal with the event of loss of cooling of spent fuel bundle. (authors)

  18. Measurements of instant-release source terms for 137Cs, 90Sr, 99Tc, 129I and 14C in used CANDU fuels

    International Nuclear Information System (INIS)

    Combined gap and grain-boundary inventories of 137Cs, 129I, 90Sr, 99Tc and 14C were measured in 15 used CANDU fuel elements by leaching crushed fuel samples. A good correlation between the combined gap and grain-boundary inventories of 137Cs and 129I was found, suggesting that these fission products exhibit similar behavior in CANDU fuel. The expected correlation between combined gap and grain-boundary inventories of 137Cs and 129I with calculated fission-gas release to the gap and grain boundaries could only be confirmed for lower power fuels (90Sr were higher than expected and showed no correlation with calculated fission-gas release. No values for the combined gap and grain-boundary inventories of 99Tc were obtained because 99Tc in used fuel samples is very insoluble and appears to require oxidation prior to dissolution. Combined gap and grain-boundary inventories of 14C appeared to be independent of fuel power or burnup. (orig.)

  19. Demonstration Drop Test and Design Enhancement of the CANDU Spent Fuel Storage Basket in MACSTOR/KN-40

    International Nuclear Information System (INIS)

    A dry interim storage facility named MACSTOR/KN-400 has been constructed at the Wolsung power plant in Korea. The MACSTOR/KN-400 has the separated 7 modules. There are 400 long slender cylinders in one module. In one cylinder, ten baskets where CANDU spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facility, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. In a demonstration test, one of test basket models did not satisfy one of the safety-related requirements. So, the revised basket designs were generated by the structural evaluation based on the finite element analyses and specimen tests. Among these revised designs, one design was chosen as the final revised design of the basket. The final revised design is the one that makes the largest reduction of plastic strain in the upper welding region. And it also is the one that makes the smallest design change from the previous basket design and the one easy to adopt the design change. Drop tests and leak tests of the final revised basket were performed and it satisfied all the performance requirements. (author)

  20. Safeguards seals for underwater spent-fuel storage: the principles and the progress towards a practical system for CANDU reactors

    International Nuclear Information System (INIS)

    While seals are in widespread use for safeguards, the development of a completely satisfactory underwater seal for irradiated fuel has not been straightforward. One aspect of the problem is that, to produce a seal which is (ideally) impossible to duplicate requires a non-reproducible identity. Equally challenging is to provide a simple means of accurately comparing an underwater seal with a reference. The Canadian seal programme is producing seals using two different inspection technologies, ultrasonics and optics. Ultrasonics lends itself to relatively simple measuring tools, but accurate comparative measurements are difficult. The design now proposed for the CANDU ultrasonic cap seal uses a stainless-steel wire coil as the identity element. Tools to install and read the coil seal have been produced and the experimental results obtained with laboratory equipment have confirmed that the seal has real potential for IAEA use. Optical inspection techniques are well developed, but simple tools have yet to be demonstrated for the application. Progress on the optical seal has been in the area of selecting an identity which cannot be duplicated but can be readily examined. Millimetre-sized crystals have been grown in pure zirconium. These are irregular in shape, brightly coloured and the patterns can be readily matched. (author)

  1. Seismic Structure-Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Kim, Sung Hwan; Yang, Ke Hyung; Lee, Heung Young; Cho, Chun Hyung [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su [KONES Corporation, Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside the concrete module consists of 40 storage cylinders accommodating ten 60-bundle dry storage baskets, which are suspended from the top slab and eventually restrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module shall be by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants, except for local site characteristics required for soilstructure interaction (SSI) analysis. It is required for the structural integrity to fulfill the licensing requirements. As per USNRC SRP Section 3.7.2, it shall be reviewed how to consider the phenomenon of coupling of the dynamic response of adjacent structures through the soil, which is referred to as structure-soil-structure interaction (SSSI). The presence of closely spaced multiple structural foundations creates coupling between the foundations of individual structures . Some observations of the actual seismic response of structures have indicated that SSSI effects do exist, but they are generally secondary for the overall structural response motions. SSSI effects, however, may be important for a relatively small structure which is to be close to a relatively large structure, while they may be generally neglected for overall structural response of a large massive structure, such as nuclear power plant. As such the scope of the present paper is to carry out a seismic SSSI analysis in case of the MACSTOR/KN- 400 module, in order to investigate whether or not SSSI effect shall be included in the overall seismic

  2. The Enhanced CANDU 6TM Reactor - Generation III CANDU Medium Size Global Reactor

    International Nuclear Information System (INIS)

    The Enhanced CANDU 6TM (EC6TM) is a 740 MWe class heavy water moderated pressure tube reactor, designed to provide safe, reliable, nuclear power. The EC6TM has evolved from the proven eleven (11) CANDU 6 plants licensed and operating in five countries (four continents) with over 150 reactor years of safe operation around the world. In recent years, this global CANDU 6 fleet has ranked in the world's top performing reactors. The EC6 reactor builds on this success of the CANDU 6 fleet by using the operation, experience and project feedback to upgrade the design and incorporate design improvements to meet current safety standards.The key characteristics of the highly successful CANDU 6 reactor design include: Powered by natural Uranium; Ease of installation with modular, horizontal fuel channel core; Separate low-temperature, low-pressure moderator providing inherently passive heat sinks; Reactor vault filled with light water surrounding the core; Two independent safety shutdown systems; On-power fuelling; The CANDU 6 plant has a highly automated control system, with plant control computers that adjust and maintain the reactor power for plant stability (which is particularly beneficial in less developed power grids-where fluctuations occur regularly and capacities are limited). The major improvements incorporated in the EC6 design include: More robust containment and increased passive features e.g., thicker walls, steel liner; Enhanced severe accident management with additional emergency heat removal systems; Improved shutdown performance for improved Large LOCA margins; Upgraded fire protection systems to meet current Canadian and International standards; Additional design features to improve environmental protection for workers and public-ALARA principle; Automated and unitized back-up standby power and water systems; Other improvements to meet higher safety goals consistent with Canadian and International standards based on PSA studies; Additional reactor trip

  3. Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

    International Nuclear Information System (INIS)

    AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt. % Th and 1.53 wt. % Pu in (Th, Pu)O2. The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O2 fuel performance characteristics were superior to UO2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

  4. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    The CANDU program in Romania is now well established. The Cernavoda Nuclear Station presently under construction will consist of 5-CANDU 600 MWE Units and another similar size station is planned to be in operation in the next decade. Progress on the multi-unit station at Cernavoda was stalled for 18 months in 1982/83 as the Canadian Export Development Corporation had suspended their loan disbursements while the Romanian National debt was being rescheduled. Since resumption of the financing in August 1983 contracts worth almost 200M dollars have been placed with Canadian Companies for the supply of major equipment for the first two units. The Canadian design is that which was used in the latest 600 MWE CANDU station at Wolsong, Korea. The vast construction site is now well developed with the cooling water systems/channels and service buildings at an advanced stage of completion. The perimeter walls of the first two reactor buildings are already complete and slip-forming for the 3rd Unit is imminent. Many Romanian organizations are involved in the infrastructure which has been established to handle the design, manufacture, construction and operation of the CANDU stations. The Romanian manufacturing industry has made extensive preparations for the supply of CANDU equipment and components, and although a major portion of the first two units will come from Canada their intentions are to become largely self-supporting for the ensuing CANDU program. Quality assurance programs have been prepared already for many of the facilities

  5. Development of analysis system and analysis on reactor physics for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Gi; Bae, Chang Joon; Kwon, Oh Sun [Korea Power Electric Corporation, Taejon (Korea, Republic of)

    1997-07-01

    characteristics of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU= fuel core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. (Author) 32 refs., 25 tabs., 79 figs.

  6. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

  7. Zirconium ignition in exposed fuel channel

    Energy Technology Data Exchange (ETDEWEB)

    Elias, E., E-mail: merezra@technion.ac.il; Hasan, D.; Nekhamkin, Y.

    2015-05-15

    Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

  8. Zirconium ignition in exposed fuel channel

    International Nuclear Information System (INIS)

    Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions

  9. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310oC, and exits at ∼410oC. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants (

  10. CANDU 9 nuclear power plant design description

    International Nuclear Information System (INIS)

    Atomic Energy of Canada limited (AECL) has make significant design improvements in the latest CANDU nuclear power plant (NPP)-the CANDU 9. The CANDU 9 operates with the energy efficient heavy water moderated reactor and natural uranium fuel and utilizes proven technology. The CANDU 9 NPP design is similar to the world leading CANDU 6 but is based upon the single unit adaptation of the 900 MWe class reactors currently operating in Canada (as multiunit configurations). The CANDU 9 NPP was developed as part of the comprehensive AECL product development program which addresses all aspects of CANDU technology including such disciplines as safety, reactor systems and components, constructability, instrumentation and control, health and environment, fuel and fuel cycles and heavy water systems. This paper will provide an overview for some of the key features of the CANDU 9 NPP such as plant layout, safety enhancements and operability improvements implemented in this design as well as outlining some of the advantages that can be expected by the operating utility

  11. Gothic simulation of single-channel fuel heatup following a loss of forced flow

    International Nuclear Information System (INIS)

    GOTHIC v7.2 was used to develop a computer model for the simulation of 28- and 37-element fuel heat-up at a loss of forced flow. The model has accounted for the non-uniformity of both axial and radial power distributions along the fuel channel for a typical CANDU reactor. In addition, the model has also accounted for the fuel rods, end-fittings, feeders and headers. Experimental test conditions for both 28- and 37-element bundles at either low or high powers were used for model validation. GOTHIC predictions of the rod and/or pressure-tube temperatures at a variety of test locations were compared with the corresponding experimental measurements. It is found that the numerical results agree well with the experimental measurements for most of the test locations. Results have also shown that the channel venting time is sensitive to the initial temperature distribution in the feeders and headers. An imposed temperature asymmetry at the beginning will cause the channel flow to vent earlier. (author)

  12. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

    International Nuclear Information System (INIS)

    The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

  13. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  14. Specifications for reactor physics experiments on CANFLEX-RU fuel

    International Nuclear Information System (INIS)

    This is to describe reactor physics experiments to be performed in the ZED-2 reactor to study CANFLEX-RU fuel bundles in CANDU-type fuel channels. The experiments are to provide benchmark quality validation data for the computer codes and associated nuclear databases used for physics calculations, in particular WIMS-AECL. Such validation data is likely to be a requirement by the regulator as condition for licensing a CANDU reactor based on an enriched fuel cycle

  15. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    operating plants. The EC6 will utilize modern computers and control systems housed in an Advanced Control Room which, along with automated testing, will make the plants easier to operate, with minimal operator intervention. Improvements to the fire protection system and enhanced security features will further protect the assets. The CANDU 6s inherent ability to utilize different fuel cycles and the Advanced MACSTOR module to store spent dry fuel further add to the attractiveness of the EC6 product. This paper describes the various enhancements that are being made to the EC6 and explains these features in more detail

  16. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  17. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    International Nuclear Information System (INIS)

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  18. The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)

    International Nuclear Information System (INIS)

    The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)

  19. On-line power monitoring in CANDU using PMFD

    International Nuclear Information System (INIS)

    The use of digital computers in the on-line control of the spatial power distribution has been well established in CANDU reactors. The instantaneous spatial power distribution in a CANDU reactor is calculated on-line, once every two minutes, by the Flux mapping program. This program synthesizes the global reactor power distribution using a least-squares fit to a set of measured flux detector readings to find amplitudes for a subsequent expansion of a precalculated set of flux harmonics. The reactor control program compares the mapped power shape with the reference power shape. The difference between these two power shapes is minimized by the appropriate deployment of the zone control system. The higher fissile content and the higher fuel burnup of SEU reactors can result in higher power ripples than those in the CANDU natural uranium reactors. Also, concerns of fuel performance at high burnup require accurate on-line monitoring of individual rippled channel and bundle powers. The present Flux Mapping program cannot meet the requirements of Highly Advanced CANDU reactors. The power mapping finite difference (PMFD) program is specifically designed to supplement or replace the Flux Mapping program in the spatial control system of CANDU SEU reactors. It solves the system of three-dimensional, two-energy group neutron diffusion equations on-line. A unique feature of PMFD is its ability to use measured detector fluxes as internal boundary conditions in the flux solution. The resulting flux shape satisfies both the distribution of material properties and the measured flux values

  20. Thermal behavior of the CANDU type spent fuel dry-storage concrete canister

    International Nuclear Information System (INIS)

    This paper describes a simple model developed for calculation of the temperature distribution and thermal behavior analysis of the spent fuel dry-storage concrete canister. The model takes into account the relevant heat transfer processes and the cylindrical geometry of the concrete canister. (author)

  1. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

    The CANDU Owners Group (COG) was formed in 1984 by the Canadian CANDU owning utilities and Atomic Energy of Canada limited (AECL). Participation was subsequently extended to all CANDU owners world-wide. The mandate of the COG organization is to provide a framework for co-operation, mutual assistance and exchange of information for the successful support, development, operation, maintenance and economics of CANDU nuclear electric generating stations. To meet these objectives COG established co-operative programs in two areas: 1. Station Support. 2. Research and Development. In addition, joint projects are administered by COG on a case by case basis where CANDU owners can benefit from sharing of costs

  2. Build your own Candu reactor

    International Nuclear Information System (INIS)

    The author discusses the marketing of Candu reactors, particularly the export trade. Future sales will probably be of the nuclear side of a station only, thus striking a compromise between licensing and 'turnkey' sales. It is suggested that AECL might have made more money in the past had it not given the right to manufacture Candu fuel away to Canadian industry. Future sales to certain potential customers may be limited by the requirement of strict safeguards, which will almost certainly never be relaxed. (N.D.H.)

  3. Assessing CANDU requirements for irradiation - Research facilities

    International Nuclear Information System (INIS)

    The Canadian nuclear program needs ongoing access to irradiation-research facilities to support the safe operation of existing CANDU reactors and the evolutionary development of CANDU components and design features. The irradiation-research program must facilitate the testing of unique CANDU technology such as the fuel bundle, on-power refueling, the pressure tube, and the heavy-water coolant and moderator. Since 1957, NRU has operated as Canada's principal irradiation facility; however, it has become clear that NRU needs costly refurbishing if its lifetime is to be significantly extended. Accordingly, AECL has reviewed the requirements for CANDU irradiation research and is presently assessing alternatives for providing the necessary future access to irradiation-research facilities. Various options are under consideration, including renting space in existing research reactors, performing irradiations in CANDU power reactors, and building a new indigenous materials testing reactor specifically to meet essential program requirements

  4. CANDU-BLW-250

    International Nuclear Information System (INIS)

    The plant 'La Centrale nucleaire de Gentilly' is located between Montreal and Quebec City on the south shore of the St. Lawrence River and start-up is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. his reactor utilizes a heavy water moderator, natural uranium oxide fuel, and a boiling light water coolant. To be economic, this type of plant must have a minimum light water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low-coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (1) Heat Transfer: It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. (ii) Hydrodynamic Stability: Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control: This plant does have a positive power coefficient and the transient performance with various disturbances are detailed. (iv) Safety: The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analysis are presented. (author)

  5. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  6. A copper container corrosion model for the in-room emplacement of used CANDU fuel

    International Nuclear Information System (INIS)

    Copper containers in a Canadian nuclear fuel waste disposal vault are expected to undergo uniform corrosion and, possibly, pitting. The corrosion behaviour of the containers will be dictated by the evolution of environmental conditions within the disposal vault. The environment will evolve from an early warm, oxidizing phase, during which fast uniform corrosion and pitting may occur, to an indefinite period of cool, anoxic conditions, during which the container will only be susceptible to slow uniform corrosion. The results of corrosion and electrochemical studies of the uniform corrosion of Cu in O2-containing Cl- solutions are discussed and a detailed reaction mechanism presented. The relevant literature on pitting corrosion is briefly reviewed and models for the prediction of pit depth discussed. The potential for microbially influenced corrosion and stress-corrosion cracking is discussed, as are vapour-phase corrosion and the effects of β-radiation. The use of natural analogues for justifying long-term corrosion predictions is also considered. Finally, a model for uniform corrosion and pitting is presented and container lifetimes predicted. Copper containers having a minimum wall thickness of 25.4 mm are not predicted to fail by corrosion in periods 6 a. Thus, despite the assumption of poor rock quality made here, the safety of the entire disposal concept can be assured by the use of a long-lived container. (author). 125 refs., 1 tab., 24 figs

  7. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  8. Controlled beta-quenching of fuel channels using inert gas

    International Nuclear Information System (INIS)

    The trend towards higher fuel assembly discharge burnups poses new challenges for fuel channels in terms of their dimensional behavior and corrosion resistance. This led AREVA NP to develop a new technique for beta quenching of fuel channels that combines the effect of beta-quenching with the optimization of the microstructure. The first set of fuel channels with these optimized material properties have been placed in the core of a German boiling water reactor (BWR) nuclear power plant in spring of 2004. Some more channels have been sited in the core of a Scandinavian BWR in fall of 2007 to broaden the in-pile experience with these channels. Dimensional stability is the major requirement that is applied to fuel channels. High corrosion resistance and low hydrogen pickup are certainly required as well. However, corrosion and hydrogen pickup are usually not life limiting factors due to the large wall thickness of the material. Since thick layers of oxide may spall off extensively at high burnup and cause increase of the dose rate for the personnel, high corrosion resistance of fuel channels is mandatory. The fuel channels which surround BWR fuel assemblies are exposed to neutron irradiation as well as to loads induced by the reactor coolant flowing through them. These service conditions induce material growth and creep which cause permanent changes in the dimensions of the channels. Especially, fuel channel bow is of certain interest as increased channel bow may lead to some friction with control blades. Fuel channel bow is mainly induced by fluence gradients. However, there may be additional influences such as oxidation and hydrogen uptake to cause increased channel bow. The effect of hydrogen is currently discussed in the nuclear community to explain the unexpected high fuel channel bow that has been observed in some nuclear power plants. (orig.)

  9. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

  10. Review of fuel and fuel safety analysis verification

    International Nuclear Information System (INIS)

    The CANDU industry in Canada is collaborating to prepare validation documentation to address the outstanding CANDU safety analysis issues in the validation of high temperature fuel and fuel channel behavior during a LOCA. The main issues being addressed are fuel channel integrity, fuel behavior during a power pulse, the source term for single channel events, flow blockage, accidents in the irradiated fuel port, and equipment qualification following a failure of emergency core cooling. The CANDU industry is combining its' plans for both fuel and fuel channels in a technical basis document, validation matrix, code validation plans, and code validation reports. Documentation, listed below, has been provided to the contractor for review with special emphasis on the validation of fuel behavior during a power pulse and pressure tube integrity during a large LOCA. The documents are: Draft Fuel and Fuel Channel Validation Matrix Report, 1996, June. SMARTT Validation Plan. Draft SMARTT Code Validation: Simulation of Pressure Tube Circumferential Temperature Distribution Experiments RSOAD report N-31 100-955015 September 1995. FACTAR Validation Plan. Draft FACTAR55 Module Validation Report. FACTAR 2.0 Module Validation, RSOAD report N-06631.01 P FACTAR. Validation QA Qualification Plan for Mini-SMARTT, Version 1.814. 'Fuel Channel Integrity Assessment' of Ontario Hydro RSOAD report N-03503.7-955045 R0, 1994 April. Emphasis on Appendix III. Draft SOURCE 2 Development Plan, 1996, February. (author)

  11. Application of Be-free Zr-based amorphous sputter coatings as a brazing filler metal in CANDU fuel bundle manufacture

    International Nuclear Information System (INIS)

    Amorphous sputter coatings of Be-free multi-component Zr-based alloys were applied as a novel brazing filler metal for Zircaloy-4 brazing. By applying the homogeneous and amorphous-structured layers coated by sputtering the crystalline targets, the highly reliable joints were obtained with the formation of predominantly grown α-Zr grains owing to a complete isothermal solidification, exhibiting high tensile and fatigue strengths as well as excellent corrosion resistance, which were comparable to those of Zircaloy-4 base metal. The present investigation showed that Be-free and Zr-based multi-component amorphous sputter coatings can offer great potential for brazing Zr alloys and manufacturing fuel rods in CANDU fuel bundle system. (author)

  12. Upgrade of Dhruva fuel channel flow instrumentation

    International Nuclear Information System (INIS)

    Dhruva, a 100 MW Heavy Water moderated and cooled, vertical tank-type Research Reactor, using metallic natural Uranium fuel has flow instrumentation for all the 144 fuel channels, consisting of venturi and triplicate DP gauges for each fuel channel. These gauges provide contacts for generation of reactor trip on low flow through fuel channel. These DP gauges were facing numerous generic and ageing related failures over the years and was also difficult to maintain owing to obsolescence. While considering an upgrade for these DP gauges, it was also planned to replace the existing Coolant Low Flow Trip (CLFT) system with a computer based Reactor Trip Logic System (RTLS). Being a retrofit job, the existing panels for mounting the gauges, cable layout, impulse tubing layout, etc. were retained, thereby simplifying the site execution, reducing reactor down time and also reducing person-milli-Sievert consumption. A customized Electronic DP Indicating Switch (EDPIS) was conceptualized for achieving these objectives. Such a design, utilizing a standard DP transmitter with customized electronic circuitry, was developed, evaluated and finalized after a series of factory trials, field trials and prototyping. The instrument design included contact input for existing CLFT system and also provision for 4-20 mA current output for the proposed computer based RTLS. The display and form factor of the instrument remained identical to older one and ensures familiarity of O and M personnel. Since EDPIS is classified as Safety Class IA, stringent type tests, hardware FMEA and V and V of the micro-controller software were carried out as per the requirements laid down by relevant standards for qualification of these instruments. Being a customized instrument, the manufacturing process was closely monitored and was followed by stringent QA plan and acceptance tests. A total of 396 gauges were replaced in a phased manner during scheduled fuelling outages and thereby did not affect reactor

  13. CANDU refurbishment - managing the life cycle

    International Nuclear Information System (INIS)

    All utilities that operate a nuclear power plant have an integrated plan for managing the condition of the plant systems, structures and components. With a sound plant life management program, after about 25 years of operation, replacement of certain reactor core components can give an additional 25 to 30 years of operation. This demonstrates the long-term economic strength of CANDU technology and justifies a long-term commitment to nuclear power. Indeed, replacement of pressure tubes and feeders with the most recent technology will also lead to increased capacity factors - due to reduced requirements for feeder inspections and repair, and eliminating the need for fuel channel spacer relocation which have caused additional and longer maintenance outages. Continuing the operation of CANDU units parallels the successful life extensions of reactors in other countries and provides the benefits of ongoing reliable operation, at an existing plant location, with the continued support of the host community. The key factors for successful, optimum management of the life cycle are: ongoing, effective plant life management programs; careful development of refurbishment scope, taking into account system condition assessments and a systematic safety review; and, a well-planned and well-executed retubing and refurbishment outage, where safety and risk management is paramount to ensure a successful project The paper will describe: the benefits of extended plant life; the outlook for refurbishment; the life management and refurbishment program; preparations for retubing of the reactor core; and, enhanced performance post-retubing. Given the potential magnitude of the program over the next 10 years, AECL will maintain a lead role providing overall support for retubing and plant Life Cycle Management programs and the CANDU Owners Group will provide a framework for collaboration among its Members. (author)

  14. Monitoring parameters for fuel channel safety in AHWR - an overview

    International Nuclear Information System (INIS)

    Full text: Advanced heavy water reactor (AHWR) is a vertical pressure tube type reactor having boiling light water coolant circulating by thermosyphon in reactor coolant channels. The fuel channel monitoring system instrumentation becomes very important for maintaining channel safety and integrity. The various monitoring parameters for fuel channel monitoring system can be classified as (a) continuous monitoring and (b) sample monitoring. The continuous monitoring parameters for fuel channel safety and operation are i) channel inlet pressure ii) channel temperature iii) channel flow and iv) channel power. Coolant activity and Annulus gas monitoring are sample-monitoring parameters for ensuring channel integrity. The inlet pressure to the core being same for all the channels and AHWR being a pressurized boiling water system, pressure and temperature measurement cannot be considered as monitoring parameter for channel safety. Hence the continuously monitoring parameter for fuel channel safety becomes flow and power monitoring only. Since channel DT is negligible, the void fraction measurement is required for getting the channel power. Hence in AHWR the measurement of flow at the single-phase inlet section and measurement of two-phase flow/void fraction at the tail pipe section becomes important for estimating channel flow and channel power for reactor safety and operation. Measurement of two-phase flow in natural circulation system is complex and various attempts were made to understand and develop instrumentation so that an effective channel monitoring can be carried out. The channel integrity is monitored by sampling methods namely annulus gas monitoring and activity monitoring. Since the annulus gap is partially open, the moisture/activity in the annulus air need to be measured accurately to identify and quantify the leak rate and in turn the pressure tube/calandria tube integrity. This paper describes in detail the various aspects of fuel channel monitoring

  15. Investigation of flow blockage in a fuel channel with the ASSERT subchannel code

    International Nuclear Information System (INIS)

    On behalf of New Brunswick Power, a study was undertaken to determine if safe operation of a CANDU-6 reactor can be maintained at low reactor powers with the presence of debris in the fuel channels. In particular, the concern was to address if a small blockage due to the presence of debris would cause a significant reduction in dryout powers, and hence, to determine the safe operation power level to maintain dryout margins. In this work the NUCIRC(1,2), ASSERT-IV(3), and ASSERT-PV(3) computer codes are used in conjunction with a pool boiling model to determine the safe operation power level which maintains dryout safety margins. NUCIRC is used to provide channel boundary conditions for the ASSERTcodes and to select a representative channel for analysis. This pool boiling model is provided as a limiting lower bound analysis. As expected, the ASSERT results predict higher CHF ratios than the pool boiling model. In general, the ASSERT results show that as the model comes closer to modelling a complete blockage it reduces toward, but does not reach the pool boiling model. (author)

  16. Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

    International Nuclear Information System (INIS)

    Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

  17. A new fuel channel for Bruce NGS 'A'

    International Nuclear Information System (INIS)

    This paper describes the design and development of a new Bruce A fuel channel, designed for the Large Scale Fuel Channel Replacement Program. The original fuel channels were not capable of achieving their design life, because the effects of radiation on zirconium alloys were imperfectly understood at the time they were designed. The reason for the redesign is to produce a fuel channel that will survive the remaining life of the station, and to eliminate as many of the channel-related design problems as practical. In addition, a design that is quick and easy to install is wanted, to reduce radiation exposure to personnel and the duration of the outage. To achieve these goals, a dedicated design team comprising both fuel channel designers and installation tooling designers was assembled. 5 figs

  18. Assessment of Inner Channel Blockage on the Annular Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; In, W. K.; Oh, D. S.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. An inner channel blockage is principal one of technical issues of the annular fuel rod. The inner channel in an annular fuel is isolated from the neighbor channels unlike the outer channels. The inner channel will be faced with a DNB accident by the partial blockage. In this paper, the largest fractional channel blockage was assessed by subchannel analysis code MATRA-AF and an end plug design to complement inlet blockage of inner channel was estimated by CFD code, CFD-ACE

  19. The CANDU Reactor System: An Appropriate Technology.

    Science.gov (United States)

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  20. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century

  1. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

  2. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Z., E-mail: chengz@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

  3. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  4. CANDU nuclear power system

    International Nuclear Information System (INIS)

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

  5. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  6. Dimensional Behavior of Fuel Channels - Recent Experience and Consequences

    International Nuclear Information System (INIS)

    Fuel channels in boiling-water reactors (BWR) undergo distortions like bow, bulge, and twist due to their operating conditions. These distortions may adversely impact planned operating strategy, and therefore need to be adequately addressed during various stages of fuel channel design and manufacturing, core design and operation monitoring. Fuel channel distortion may lead to interference between the fuel channel and adjacent control blade. If severe, such interference can impair both positioning of control blades during normal operations and rapid control blade insertion during a reactor scram. During the last five years, unexpectedly high fuel channel distortions leading to problems in control blade operations have been observed in some C- and S-lattice BWR plants in the U.S. operating on 18 - 24 month cycles. As a result, U.S. operators have implemented costly surveillance programs to detect the onset of high distortions and have declared control blades inoperable when unacceptable control blade operation occurs. This unusual fuel channel distortion has not been observed with AREVA NP fuel supplied in Europe in this scale. Nevertheless fuel channel distortion-related problems were recently observed in reactors outside the U.S. with early control blade operation. The mechanisms causing this unexpected fuel-channel distortion and the influencing factors are not completely understood worldwide for the time being. Therefore, a prediction of channels which will exhibit high bow is very challenging. A status is given from the AREVA NP perspective on: - The existing fuel channel distortion database, - The understanding of the phenomenon, - Measures to gather further information and improve existing tools, materials, and designs, and - Customer actions to reduce potential high channel bow and associated control blade issues. (authors)

  7. U.S. utilities may seek CANDU reactors

    International Nuclear Information System (INIS)

    The economic advantage of the CANDU fuel cycle, using natural uranium, over the LWR cycle, using enriched uranium, is outlined. It is argued that President Carter's decision to shelve the fast breeder and plutonium recycle should make CANDU more attractive to American utilities. Possible effects of American licensing on the Canadian industry are discussed. Another possibility would be American assistance in developing the CANDU OCR thorium near-breeder. (N.D.H.)

  8. Moving towards sustainable thorium fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor has an unsurpassed degree of fuel-cycle flexibility as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle design. These features facilitate the introduction and full exploitation of thorium fuel cycles in CANDU reactors in an evolutionary fashion. Thoria (ThO2) based fuel offers both fuel performance and safety advantages over urania (UO2) based fuel, due its higher thermal conductivity which results in lower fuel-operating temperatures at similar linear element powers. Thoria fuel has demonstrated lower fission gas release than UO2 under similar operating powers during test irradiations. In addition, thoria has a higher melting point than urania and is far less reactive in hypothetical accident scenarios owing to the fact that it has only one oxidation state. This paper examines one possible strategy for the introduction of thorium fuel cycles into CANDU reactors. In the short term, the initial fissile material would be provided in a heterogeneous bundle of low-enriched uranium and thorium. The medium term scenario uses homogeneous Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. In the long term, the full energy potential from thorium would be realized through the recycle of the U-233 in the used fuel. With U-233 recycle in CANDU reactors, plutonium would then only be required to top up the fissile content to achieve the desired burnup. (author)

  9. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gheorghe [National Agency for Radioactive Waste, 1, Campului Str., 115400 Mioveni (Romania)], E-mail: gheorghe.negut@andrad.ro; Catana, Alexandru [Institute for Nuclear Research Pitesti, 1, Campului Str., Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie; Dupleac, Daniel [Politehnica University Bucharest, 313, Splaiul Independentei, Sect. 6, 060042 Bucharest (Romania)

    2009-09-15

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  10. Use of depleted fuel as a mitigation method for reducing FAC in selected channels

    International Nuclear Information System (INIS)

    Feeder pipes located at the outlet of fuel channels are known to corrode at a higher rate than expected. The degradation mechanism is identified as flow-assisted corrosion (FAC) and seems to affect all CANDU plants. The extent of feeder thinning was discovered in 1995 after the mechanical cleaning of steam generators. The transport mechanism of magnetite is dissolution of irons at the outlet feeder and deposition on the cold leg of steam generator tubes. Multiple factors affect the thinning rates and pH is one that was quickly readjusted in order to minimize iron solubility. Feeder inspection started rapidly at Gentilly-2 and thickness measurements were performed at every outage since 1996. Data collected over the years showed a certain pattern of thinning of the feeders most affected by FAC. The feeders that would need early dispositions are 2-inch size pipes with a compound bend (730 and 700). A closer look at this feeder outlet geometry showed some correlation between thinning rate and boiling rate. One observed tendency is a reduction of thinning rate following the cleaning of steam generators in 1999 - thus linking boiling to thinning rate. Although the possible correlation between boiling and thinning rate is based on a limited amount of data, Gentilly-2 decided to conduct an experiment to see if reducing boiling in selected channels may slow down the thinning rate. The experiment consisted of cuffing boiling by using a number of depleted fuel bundles in specific channels to maintain the outlet temperature below saturation temperature. Two channels were refueled with depleted fuel in Nov. and Dec. 2000. Close monitoring of these two channels (D5 and C6) and their twin (symmetric) channels (C17 and D18) would permit a conclusion regarding the effectiveness of controlling the power of selected channels for minimizing the thinning rate. Results obtained in 2002 show promising results but a longer assessment period is required to eliminate any doubt. Longer term

  11. Development of CANDU phenomena identification and ranking table for MARS-KS validation

    International Nuclear Information System (INIS)

    Korea Institute of Nuclear Safety (KINS) is establishing a REactor Thermal-hydraulic Analysis System (KINS-RETAS), where the best-estimate codes are organized to simulate the core, reactor coolant system (RCS) and containment thermal-hydraulic (T-H) response to an initiating event. The MARS code is a pivotal code in the system to predict T-H behavior in the core and RCS. MARS qualification was requested for accident analyses of CANDU6 as well as PWRs. After investigating the plausible important phenomena in core and RCS for accident scenarios in domestic PWRs, the CANDU PIRT development has been initiated in the same manner, while focusing on the core with horizontal channels and headers. Because there have been less effects on code qualification for CANDU than for PWR and is no available PIRT results for CANDU6 except for ACR-700, the phenomena was investigated for the representative scenario of design basis accident of emergency coolant injection system performance, e.g., a 35% reactor inlet header break. Some plausible phenomena were also investigated for complete flow blockage of a single fuel channel. The important phenomena and processes were identified and tabulated by systems and components and by phases. Rationales were discussed to support the assigned importance-rank. (author)

  12. Mechanistic modeling of thermal-mechanical deformation of CANDU pressure tube under localized high temperature condition

    International Nuclear Information System (INIS)

    Thermal strain deformation is a pressure tube failure mechanism. The main objective of this paper is to develop mechanistic models to evaluate local thermal-mechanical deformation of a pressure tube in CANDU reactor and to investigate fuel channel integrity under localized contact between fuel elements and pressure tube. The consequence of concern is potential creep strain failure of a pressure tube and calandria tube. The initial focus will be on the case where a fuel rod contacts the pressure tube at full power with highly cooling condition

  13. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  14. An analytic study on LBLOCA for CANDU type reactor using MARS-KS/CANDU

    International Nuclear Information System (INIS)

    This study provides the simulation results using MARS-KS/CANDU code for the Large Break LOCA of CANDU type reactor. The purpose of the study is to evaluate the capability of MARS-KS/CANDU for simulating the actual plants (Wolsong 2/3/4). The steady state and the transient analysis results were provided. After the sensitivity study depend on break size, the case that 35% of the inlet header known as the accident that has the most limiting effect on the temperature of the fuel sheath was calculated. In order to evaluate the results, the results were compared with those of CATHENA simulation. (author)

  15. Stress analysis of screws in the fuel channel fastener assembly

    International Nuclear Information System (INIS)

    The function of fuel channel fastener assembly is to keep enough clearance between fuel channels, allowing the insertion of control rod and fixing the channel on the fuel bundle. The assembly device is not safety related component, however, in case of the screw breaking, it may cause loose parts, which might adversely affect the normal operation of inserting and pulling fuel assemblies, and/or the movement of the control rods. In this paper, the possible loading conditions applied to the fuel channel fastener assembly are considered to analyze the stress state in screw. In order to assess the improper positioning of fuel channel, explicit finite element procedures is employed to simulate the complex contact/impact behaviors occurring between the fastener assembly and the neighboring fuel channel or the fuel rack, in which the effects of dynamic impact on the screw and initial contact speed are the main concern. The analysis results reveal that the reduced neck close to the screw head has the highest stress. If the external loads drive the stress up to the yielding limit, crack initiation will occur on the screw neck and thereby, under the tensile loadings and reactor core environment, initiating intergranular stress corrosion cracking (IGSCC) on the screw

  16. Future trends in the design of CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. Engineering work is well underway to refine the design of the CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule for CANDUs of the 1990s. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components, but with an innovative new plant layout that makes it cost competitive with coal-fired plants. For the long term, work on advanced fuel cycles and major system improvements are underway ensuring that CANDU plants will stay competitive well into the next century

  17. Assessment of RELAP5/CANDU+ code for regulatory auditing analysis of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Kim, Hho Jung; Yang, Chae Yong

    2001-12-15

    The objectives of this study are to undertake the verification and validation of RELAP5/CANDU+ code, which is developed in this project, by simulating the B8711 test of RD-14 facility, and to examine the properties of this code by doing the sensitivity analysis for experimental prediction modes about thermal-hydraulics phenomena in CANDU reactor systems added to this code. The B8711 test was an experiment of a 45% ROH break for simulating large LOCA. Also, in this study, the methods for making input cards related to CANDU options are described, so that some users can use the RELAP5/CANDU+ code with easy. RELAP/CANDU+ code can choose the options of Henry-Fauske mode, Ransom-Trapp model, and Moody model for prediction of the critical mass flow. It is examined that Henry-Fauske model and Ransom-Trapp model are considered properly, but Moody model is still required to be improved. Heat transfer correlations available in RELAP5/CANDU+ code for CANDU-type reactors are a horizontal stratified model, a fuel heat-up model and D2O/H2O CHF correlations, and these models take an important role to improve the predictability of the experimental procedures. It is concluded that RELAP5/CANDU+ code is useful for the auditing of the accident analysis of CANDU reactors, and the results of the sensitivity analysis for thermal-hydraulic models examined in this study are valuable for the actual auditing of real CANDU-type power plants.

  18. On the feasibility of a CANDU PHWR actinide burner

    International Nuclear Information System (INIS)

    In this work a review of the current solutions to burn the actinide i.e. the spallation method, LWR, FBR, Siemens proposal and inert matrix is presented. Finally, a proposal is made to use the CANDU PHWR for this purpose, taking into account the techniques envisaged for LWR and the prospect of the advanced fuel cycle in CANDU system. (Author) 5 Refs

  19. Structural evaluation of Siemens advanced fuel channel under accident loadings

    International Nuclear Information System (INIS)

    As a part of an effort to develop an advanced BWR fuel channel design, Siemens Power Corporation (SPC) and the Siemens AG Power Generation Group (KWU) performed structural analyses to verify the acceptability of the fuel channel design under combined seismic/LOCA (Loss Of. Coolant Accident) loadings. The results of the analyses give some interesting insights into the problem: 1) fluid-structure interaction (FSI) effects are significant and should be considered, 2) the problem may simplified by using a linear analysis despite non-linear features (gaps) between interfacing components, and 3) sufficient accuracy may be obtained by using only the first mode of vibration. The channeled fuel assembly can be considered to be a beam where the flexural stiffness is primarily determined by the fuel channel and the mass is given by the fuel assembly. The results from the analyses show the advanced fuel channel design meets applicable design criteria with adequate margins while at the same time exhibiting superior nuclear performance compared to a conventional BWR fuel channel. (author)

  20. Review of Ontario Hydro program to reduce fuel channel gap

    International Nuclear Information System (INIS)

    The consultant was contracted to review the 'Pressure Tube Fretting and Power Pulse Program' undertaken by Ontario Hydro. They were requested to report primarily on the adequacy of quality assurance activities in the project management, engineering, and manufacturing parts of the project, but in addition were asked to review and comment on the technical adequacy of the methods. The program involved three potential solutions or approaches. Each was designed to prevent fuel bundles shifting into the reactor core if there is fuel channel flow reversal. Each approach was also intended to reduce or eliminate fuel vibration and the consequential fretting observed in fuel channels with significant creep. (author). 4 figs

  1. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  2. CANDU-3: Features of next generation CANDU

    International Nuclear Information System (INIS)

    CANDU 3, with a net electrical output of 450 MW, is the latest and smallest version of the CANDU power system. Significant innovation built on proven CANDU reactor technology is the basis of the CANDU 3 design. This, coupled with the commitment to reduce plant cost, increase performance capacity factor, enhance safety features and incorporate technological improvements, makes CANDU 3 an advanced, world class product. This paper describes the following CANDU 3 features: A station layout to provide a flexible construction sequence, good system separation and ease of maintenance and operation. An up-front engineering and licensing process prior to beginning construction. Enhancement and simplification of safety features with extended time scales for response, thus limiting reliance on operator action for accident mitigation. Enhanced design capabilities through the use of the latest Computer Aided Design and Drafting (CADD) technology. A 38 month construction schedule achieved by using modularization and open-top construction and installation techniques. A more passive containment system incorporating a steel liner and eliminating the need for active spray. A grouping and separation philosophy for maximum protection of redundant safety systems. Ease of equipment qualification and maximum protection of critical components. Replacement of centralized control and monitoring computers with a redundant distributed control system and modern plant display system. A consistent, logical approach to control room design founded on human factors, automation and event management. (author). 6 refs, 3 figs

  3. Economics of CANDU

    International Nuclear Information System (INIS)

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

  4. Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

    International Nuclear Information System (INIS)

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

  5. The disposal of Canada's nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock volume 1: summary

    International Nuclear Information System (INIS)

    The concept for disposal of Canada's nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced and sealed within a vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The case for the acceptability of the concept as a means of safely disposing of Canada's nuclear fuel waste is presented in an Environmental Impact Statement (EIS) The disposal concept permits a choice of methods, materials, site locations and designs. The EIS presents a case study of the long-term (i.e., postclosure) performance of a hypothetical implementation of the concept, referred to in this report as the reference disposal system. The reference disposal system is based on borehole emplacement of used CANDU fuel in Grade-2 titanium alloy containers in low-permeability, sparsely fractured plutonic rock of the Canadian Shield. We evaluate the long-term performance of another hypothetical implementation of the concept based on in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. The geological characteristics of the geosphere assumed for this study result in short groundwater travel times from the disposal vault to the surface. In the present study, the principal barrier to the movement of contaminants is the long-lasting copper container. We show that the long-lasting container can effectively compensate for a permeable host rock which results in an unfavourable groundwater flow condition. These studies illustrate the flexibility of AECL's disposal concept to take advantage of the retention, delay, dispersion, dilution and radioactive decay of contaminants in a system of natural barriers provided by the geosphere and hydrosphere and of engineered barriers provided by the waste form, container, buffer, backfills, other vault seals and grouts. In an actual implementation, the engineered system would be designed for the geological conditions encountered at the host site. 34 refs., 2 tabs., 11 figs

  6. The manufacturing role in fuel performance

    International Nuclear Information System (INIS)

    Manufacturing companies have been involved in the CANDU fuel industry for more than 40 years. Early manufacturing contributions were the development of materials and processes used to fabricate the CANDU fuel bundle. As CANDU reactors were commissioned, the manufacturing contribution has been to produce economical, high quality fuel for the CANDU market. (author)

  7. Enhancing the moderator effectiveness as a heat sink during loss of coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    International Nuclear Information System (INIS)

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 deg. C. (author)

  8. Control of helium activity in the fuel reactor channels

    International Nuclear Information System (INIS)

    The objective of this task was to study the possibility of detecting a damaged fuel channel, and to introduce automated procedure for continuous control of reactor channels during operation. The existing control systems at the RA reactor (permanent control of heavy water and helium activity, radiation monitoring of heavy water and helium system, measurements of fire damp gas percent) are not sufficient for fast detection of fuel element failures. Since a 'hot' fuel channel cannot be removed from the core because it should be cooled in the core by heavy water circulation, it is not possible to prevent contamination of heavy water by fission products. It is concluded that it is not indispensable to detect the failed fuel element promptly, i.e. that tome is not a critical issue

  9. The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

    International Nuclear Information System (INIS)

    A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

  10. Implementation of an integrated safeguards approach for transfers of spent fuel to dry storage at multi-unit CANDU generating stations in Canada

    International Nuclear Information System (INIS)

    Full text: The Canadian Nuclear Safety Commission (CNSC) in 2002 undertook an IAEA Support Program task to develop and test an Integrated Safeguards (IS) approach for transfers of spent fuel to dry storage at multi-unit CANDU stations. The IS approach that Canada proposed in 2003 was successfully field tested in Canada by the IAEA in 2004. Since then the IAEA has drawn a positive conclusion for Canada and is preparing to implement IS in Canada based on the State-level Integrated Safeguards Approach (SLA) for Canada that was concluded in November 2005. Implementation of an IS approach for transfers to dry storage, similar to the one proposed by Canada, will be part of the initial phase of the implementation of the SLA. This paper briefly reviews the IS approach for transfers proposed by Canada and explains how the approach may be modified for implementation as part of the SLA to take into account lessons learned from the field test, operator and IAEA concerns, and developments in technology. (author)

  11. A two-dimensional discrete-ordinates analysis of the CANDU 9 end-shield penetrations

    International Nuclear Information System (INIS)

    The two-dimensional discrete-ordinates radiation transport code DORT has been used to calculate neutron and gamma fluxes and corresponding dose rates through the end-shield outlet penetration of a central fuel channel in a CANDU 9 480/NU reactor. The 67 group coupled neutron-gamma library cross sections used in the analysis are based on primary ENDF/B-IV data, but with significant updates from new ENDF/B-VI data for many materials common to CANDU applications. Calculations were made for normal steady -state full power operation and after reactor shutdown. The neutron and gamma dose rates were calculated throughout the end shield including the plane at the outer surface of the fuelling machine tube sheet and outside the end-fittings during reactor operation. (DM)

  12. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

  13. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

  14. Key thrusts in next generation CANDU. Annex 10

    International Nuclear Information System (INIS)

    Current electricity markets and the competitiveness of other generation options such as CCGT have influenced the directions of future nuclear generation. The next generation CANDU has used its key characteristics as the basis to leap frog into a new design featuring improved economics, enhanced passive safety, enhanced operability and demonstrated fuel cycle flexibility. Many enabling technologies spinning of current CANDU design features are used in the next generation design. Some of these technologies have been developed in support of existing plants and near term designs while others will need to be developed and tested. This paper will discuss the key principles driving the next generation CANDU design and the fuel cycle flexibility of the CANDU system which provide synergism with the PWR fuel cycle. (author)

  15. Analysis of IFR driver fuel hot channel factors

    International Nuclear Information System (INIS)

    Thermal-hydraulic uncertainty factors for Integral Fast Reactor (IFR) driver fuels have been determined based primarily on the database obtained from the predecessor fuels used in the IFR prototype, Experimental Breeder Reactor II. The uncertainty factors were applied to the channel factors (HCFs) analyses to obtain separate overall HCFs for fuel and cladding for steady-state analyses. A ''semistatistical horizontal method'' was used in the HCFs analyses. The uncertainty factor of the fuel thermal conductivity dominates the effects considered in the HCFs analysis; the uncertainty in fuel thermal conductivity will be reduced as more data are obtained to expand the currently limited database for the IFR ternary metal fuel (U-20Pu-10Zr). A set of uncertainty factors to be used for transient analyses has also been derived

  16. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  17. Maximum allowable power levels in HANARO fuel channels

    International Nuclear Information System (INIS)

    HANARO is a light-water-cooled and heavy-water-reflected research reactor designed and operated at a full power of 30 MWth. The compact core is composed of 23 hexagonal and 16 circular flow channels. Each hexagonal flow channel, formed by a hexagonal flow tube, is loaded with a hexagonal fuel bundle which has 36 fuel elements. The circular flow channel formed by a circular flow tube is loaded with a circular fuel bundle which has 18 fuel elements. In the thermalhydraulic design of the HANARO, the design limit values for the three design parameters were determined using the statistical thermal design method. The design basis is that for the normal operation and the operational occurrences, the physical phenomena like as CHF (Critical Heat Flux) or ONB(Onset of Nucleate Boiling) would not occur with the 95% of probability and the 95% confidence level as illustrated in Figure 1. In this paper, the maximum power levels satisfying the design limits were obtained and the safe operational ranges were made for the four different fuel channels

  18. A CANDU-6 versus ACR-1000 SDS1 performance comparison during some LOCA scenarios

    International Nuclear Information System (INIS)

    According to the Romanian Nuclear Strategy, the third and fourth units of the Cernavoda NPP will be commissioned by 2015. Improvements in operation and safety are expected to be applied for these CANDU-6 based units. On the other side, the need for innovation determined AECL to promote the ACR -1000 - an evolutionary Generation III+ power reactor design and a necessary step towards Generation IV inherently safe nuclear energy systems. CANDU-6 is recognized for having two independent fully capable shutdown systems. ACR-1000 also benefits for this strong safety feature. Two major achievements i.e. using of light water as coolant and using Low Enriched Uranium (LEU) as fuel in a compact heavy water moderated lattice allowed the obtaining of a slightly negative Coolant Void Reactivity (CVR) for the first time in a CANDU-type reactor. The main goal of the paper is to compare the response of SDS1 action during some LOCAs supposed to take place both in CANDU-6 and ACR-1000 reactors. In the considered scenarios, the initiation event was a Rupture of the Inlet Header (RIH) of 15, 25 or 35%. The analyses were performed using the point-kinetics approximation method implemented in the DIREN code - a 3D diffusion tool developed in INR Pitesti. The CANDU-6 core model is based on as-built data from Cernavoda Unit 1, while the ACR-1000 DIREN core model was recently developed during the PhD stage of the main author. The SOR reactivities, flux amplitude, maximum channel and bundle powers were the key parameters pursued in analyses. The results emphasized the net ACR-1000 safety improvement gained from its design innovations. (authors)

  19. Assessment of the integrity of KANUPP fuel channels

    International Nuclear Information System (INIS)

    The KANUPP reactor first produced power in 1972. In 1983 one of the fuel channels, G12 was difficult to refuel due to the South end fitting being retracted relative to the adjacent channels. In November 1993, eight fuel channels were inspected non-destructively for defects. The dimensions and shape of the pressure tubes were measured, and channel G12 was removed for examination and testing at the Chalk River Laboratories. Examination of channel G12 showed that the South end calandria tube rolled joint had been leaking moderator heavy water into the annulus. The inboard bearing was completely corroded away which resulted in the South end fitting seizing in the debris. The Inconel X750 garter springs were both completely broken into small rings due to Stress Corrosion Cracking (SCC) in acids produced by radiolysis of the moderator heavy water. The inspection of the eight fuel channels did not reveal any serious defects. The length and diameter changes in the pressure tubes were both small. The deuterium ingress had been low with the result that the hydrogen isotope concentration of the pressure tubes was below Terminal Solid Solubility (TSS) even close to the end fittings. Irradiation had increased the strength and reduced the ductility and fracture toughness of the tubes and the Delayed Hydride Cracking (DHC) properties were similar to irradiated cold worked Zr-2.5Nb pressure tubes. The examination and testing showed that the KANUPP pressure tubes are all probably in good condition and fit for several more years service. (author)

  20. A statistical assessment of CANDU fuel manufacturing parameters and their impact on in-reactor performance during transient conditions

    International Nuclear Information System (INIS)

    Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to generate input for ELOCA, an industry-standard fuel performance modeling code that predicts fuel element behaviour under transient conditions. The chosen accident for this study is a hypothesized 80% Reactor Outlet Header break LBLOCA. 105 simulations are conducted, and it is shown that the distributions of key output quantities are well below the limit values established by industrial acceptance criteria, implying the existence of margin in the current design. The results of this probabilistic study are then compared with those of a deterministic case, and the contrast between the two methods is quantified. (author)