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Sample records for candu reactors

  1. 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

  2. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    The CANDU 6 power reactor is visionary in its approach, renowned for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Limited (AECL), the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980s as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first set of CANDU 6 plants - Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea - have been in service for more than 22 years and are still producing electricity at peak performance; to the end of 2004, their average Lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customers' needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology, as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed the ''Enhanced CANDU 6'' (EC6), which incorporates several attractive but proven features that make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that are being incorporated into the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  3. 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.)

  4. 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

  5. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    The technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and AECL, has been safely,economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world

  6. 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

  7. Systems analysis of the CANDU 3 Reactor

    International Nuclear Information System (INIS)

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ''significant to safety,'' and identification of key operator actions for the analyzed events

  8. 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)

  9. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author)

  10. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

  11. 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

  12. ACR-1000TM - advanced Candu reactor design

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU ReactorTM- 1000 (ACR-1000TM) as an evolutionary advancement of the current CANDU 6TM reactor. This evolutionary advancement is based on AECL's in-depth knowledge of CANDU structures, systems, components and materials, gained during 50 years of continuous construction, engineering and commissioning, as well as on the experience and feedback received from 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. These innovations improve economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The Canadian nuclear reactor design evolution that has reached today's stage represented by the ACR-1000, has a long history dating back to the early 1950's. In this regard, Canada is in a unique situation, shared only by a very few other countries, where original nuclear power technology has been invented and further developed. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 and Phase 2 pre-project design reviews in December 2008 and August 2009, respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The final stage of the ACR-1000 design is currently underway and will be completed by fall of 2011, along with the final elements of the safety analyses and probabilistic safety analyses supporting the finalized design. The generic Preliminary Safety Analysis Report (PSAR) for the ACR-1000 was completed in September 2009. The PSAR demonstrates ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. (authors)

  13. 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)

  14. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Hydrogen atom has two isotopes: deuterium 1H2 and tritium 1H3. The deuterium oxide D2O is called heavy water due to its density of 1105.2 Kg/m3. Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D2O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D2O management is required to preserve it. (author)

  15. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    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 UO{sub 2} 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

  16. 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

  17. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  18. 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.

  19. Drift effects in CANDU reactors

    International Nuclear Information System (INIS)

    The diffusion equation is an approximation to the transport equation which relies on the validity of Fick's law. Since this law is not explicitly integrated in the transport equation it can only be derived approximately using homogenization theories. However, such homogenization theories state that when the cell is not symmetric Fick's law breaks down due to the presence of an additional term to the neutron current, called the drift term. In fact, this term can be interpreted as a transport correction to Fick's law which tends to increase the neutron current in a direction opposite to that specified by the flux gradient. In this paper, we investigate how the presence of asymmetric liquid zone controllers will modify the flux distribution inside a CANDU core. 5 refs., 2 figs., 1 tab

  20. 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

  1. 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.

  2. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  3. 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.

  4. Leak detection capability in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Azer, N.; Barber, D.H.; Boucher, P.J. [and others

    1997-04-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  5. 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)

  6. Leak detection capability in CANDU reactors

    International Nuclear Information System (INIS)

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis

  7. 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)

  8. Simulation-based reactor control design methodology for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Kattan, M.K.; MacBeth, M.J. [Atomic Energy of Canada Limited, Saskatoon, Saskatchewan (Canada); Chan, W.F.; Lam, K.Y. [Cassiopeia Technologies Inc., Toronto, Ontario (Canada)

    1996-07-01

    The next generation of CANDU nuclear power plant being designed by AECL is the 900 MWe CANDU 9 station. This design is based upon the Darlington CANDU nuclear power plant located in Ontario which is among the world leading nuclear power stations for highest capacity factor with the lowest operation, maintenance and administration costs in North America. Canadian-designed CANDU pressurized heavy water nuclear reactors have traditionally been world leaders in electrical power generation capacity performance. This paper introduces the CANDU 9 design initiative to use plant simulation during the design stage of the plant distributed control system (DCS), plant display system (PDS) and the control centre panels. This paper also introduces some details of the CANDU 9 DCS reactor regulating system (RRS) control application, a typical DCS partition configuration, and the interfacing of some of the software design processes that are being followed from conceptual design to final integrated design validation. A description is given of the reactor model developed specifically for use in the simulator. The CANDU 9 reactor model is a synthesis of 14 micro point-kinetic reactor models to facilitate 14 liquid zone controllers for bulk power error control, as well as zone flux tilt control. (author)

  9. 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.

  10. Distinctive safety aspects of the CANDU-PHW reactor design

    International Nuclear Information System (INIS)

    Two lectures are presented in this report. They were prepared in response to a request from IAEA to provide information on the 'Special characteristics of the safety analysis of heavy water reactors' to delegates from member states attending the Interregional Training Course on Safety Analysis Review, held at Karlsruhe, November 19 to December 20, 1979. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (auth)

  11. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    It has been demonstrated on a large scale that CANDU reactors can produce industrial process steam as well as electricity, reliably and economically. The advantages of cogeneration have led to the concept of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development in the province of Ontario. For steam demands between 300,000 and 500,00 lb/h (38-63 kg/s) and an annual load factor of 80%, the estimated cost of nuclear steam at the Bruce site boundary is $3.21/MBtu ($3.04GJ), which is at least 30% cheaper than oil-fired steam at the same site. The most promising near term application of nuclear heat is likely to be found within the energy-intensive chemical industry. Nuclear energy can substitute for imported oil and coal in the eastern provinces if the price remains competitive, but low cost coal and gas in the western provinces may induce energy-intensive industries to locate near those sources of energy. In the long term it may be feasible to use nuclear heat for the mining and extraction of oil from the Alberta tar sands. (auth)

  12. Development of Regulatory Requirements and Inspection Guides for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Kim, K.; Ryu, Y. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Ro, H. Y.; Jin, T. E. [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2009-05-15

    The first domestic CANDU power reactor, Wolsong unit 1, has been operated for about twenty years since commercial operation in 1983, and has been raised common aging issues of CANDU reactors in pressure tubes, calandria tubes, feeder pipes, etc. To solve these aging issues, utility is promoting the refurbishment activities for these major components. Therefore, confirmation and improvement for insufficient requirements considering the CNSC regulatory documents, regulatory principles between regulatory body and utilities related with refurbishment activities are required. These review contents are described herein, and representative review results are presented.

  13. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  14. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  15. 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)

  16. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    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 to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling 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 schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include 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. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  17. Safety benefits from CANDU reactor replacement. A case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  18. Safety benefits from CANDU reactor replacement - a case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  19. Impact of lattice geometry distortion due to ageing on selected physics parameters of a CANDU reactor

    International Nuclear Information System (INIS)

    In this paper, results related to a limited scope assessment of the geometry-distortion-induced effects on key reactor physics parameters of a CANDU reactor are discussed. These results were generated by simulations using refined analytical methods and detailed modeling of CANDU reactor core with aged lattice cell geometry. (authors)

  20. 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

  1. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  2. Assessment of LOCA with loss of class IV power for CANDU-6 reactors using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Recently, there is an effort to improve the accuracy and reality in the transient simulation of nuclear power plants. In the prediction of the system transient, the system code simulates the system transient using the power transient curve predicted from the reactor core physics code. However, the pre-calculated power curve could not adequately predict the behavior of power distribution during transient since the coolant density change has influence on the power shape due to the change of the void reactivity. Therefore, the consolidation between the reactor core physics code and the system thermal-hydraulic code takes into consideration to predict more accurate and realistic for the transient simulation. In this regard, there are two codes are developed to assess the safety of CANDU reactor. RELAP-CANDU is a thermal-hydraulic system code for CANDU reactors developed on the basis of RELAP5/MOD3 in such a way to modify inside model for simulating the thermal-hydraulic characteristics of horizontal type reactors. SCAN (SNU CANDU-PHWR Neutronics) is a three dimensional neutronics nodal code to simulate the core physics characteristics for CANDU reactors. To couple SCAN code with RELAP-CANDU code, SCAN code was improved as a spatial kinetics calculation module in such a way to generate a SCAN DLL (dynamic linked library version of SCAN). The coupled code system, RELAP-CANDU/SCAN, enables real-time feedback calculations between thermal-hydraulic variables of RELAP-CANDU and reactor powers of SCAN. To verify the reliability of RELAP-CANDU/SCAN coupled code system, an assessment of 40% reactor inlet header (RIH) break loss of coolant accident (LOCA) with loss of Class IV power (LOP) for Wolsong Unit 2 conducted using RELAP/CANDU-SCAN coupled system. The LOCA with LOP is one of GAI (Generic Action Items) for CANDU reactors issued by CNSC (Canadian Nuclear Safety Commission) and IAEA (International Atomic Energy Agency)

  3. Conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    A Waste Treatment Centre (WTC) is being constructed at the Chalk River Nuclear Laboratories to develop and demonstrate processes for converting reactor wastes to a form suitable for disposal. The WTC contains a starved air incinerator for reducing the volume of combustible solid wastes, a reverse osmosis section for reducing the volume of liquid wastes and an immobilization section for incorporating the conditioned wastes in bitumen. The incinerator is commissioned on inactive waste: approximately 16.5 Mg of waste packaged in polyethylene bags has been incinerated in 17 burns. Average weight and volume reductions of 8.4:1 and 32:1, respectively, have been achieved. Construction of the reverse osmosis section of WTC is complete and inactive commissioning will begin in 1982 January. The reverse osmosis section was designed to process 30,000 m3/a of dilute radioactive waste. The incinerator ash and concentrated aqueous waste will be immobiblized in bitumen using a horizontal mixer and wiped-film evaporator. Results obtained during inactive commissioning of the incinerator are described along with recent results of laboratory programs directed at demonstrating the reverse osmosis and bituminization processes

  4. 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

  5. CANDU in-reactor quantitative visual-based inspection techniques

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

    This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is

  6. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    International Nuclear Information System (INIS)

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  7. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  8. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: lxia4@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)

    2014-10-15

    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  9. Advanced CANDU reactor development: a customer-driven program

    International Nuclear Information System (INIS)

    The Advanced CANDU Reactor (ACR) product development program is well under way. The development approach for the ACR is to ensure that all activities supporting readiness for the first ACR project are carded out in parallel, as parts of an integrated whole. In this way design engineering, licensing, development and testing, supply chain planning, construct ability and module strategy, and planning for commissioning and operations, all work in synergy with one another. Careful schedule management :ensures that program focus stays on critical path priorities.'This paper provides an overview of the program, with an emphasis on integration to ensure maximum project readiness, This program management approach is important now that AECL is participating as the reactor vendor in Dominion Energy's DOE-sponsored Combined Construction/Operating License (COL) program. Dominion Energy selected the ACR-700 as their reference reactor technology for purposes of demonstrating the COL process. AECL's development of the ACR is unique in that pre-licensing activities are being carded out parallel in the USA and Canada, via independent, but well-communicated programs. In the short term, these programs are major drivers of ACR development. The ACR design approach has been to optimize to achieve major design objectives: capital cost reduction, robust design with ample margins, proveness by using evolutionary change from existing :reference plants, design for ease :of operability. The ACR development program maintains these design objectives for each of the program elements: Design: .Carefully selected design innovations based on the SEU fuel/light water coolant:/heavy water moderator approach. Emphasis on lessons-learned review from operating experience and customer feedback Licensing: .Safety case based on strengths of existing CANDU plus benefits of optimised design Development and Test: Choice of materials, conditions to enable incremental testing building on existing CANDU and LWR

  10. Survey of considerations involved in introducing CANDU reactors into the United States

    Energy Technology Data Exchange (ETDEWEB)

    Till, C E; Bohn, E M; Chang, Y I; van Erp, J B

    1977-01-01

    The important issues that must be considered in a decision to utilize CANDU reactors in the U.S. are identified in this report. Economic considerations, including both power costs and fuel utilization, are discussed for the near and longer term. Safety and licensing considerations are reviewed for CANDU-PHW reactors in general. The important issues, now and in the future, associated with power generation costs are the capital costs of CANDUs and the factors that impact capital cost comparisons. Fuel utilization advantages for the CANDU depend upon assumptions regarding fuel recycle at present, but the primary issue in the longer term is the utilization of the thorium cycle in the CANDU. Certain safety features of the CANDU are identified as intrinsic to the concept and these features must be examined more fully regarding licensability in the U.S.

  11. Ultrasonic measurement method of calandria tube sagging in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor (calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the calandria tube (made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, calandria tube and liquid inject ion tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here

  12. 3D simulation of CANDU reactor regulating system

    International Nuclear Information System (INIS)

    Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

  13. Ultrasonic crack-tip diffraction in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Currently there is no reliable method of measuring defect depths in CANDU reactor pressure tubes. The demonstrated success of crack-tip diffraction (or time-of-flight-testing) in round-robins on thick components has promoted an interest in this technique. In CANDU reactors, pressure tubes are effectively accessible only from the inside. Development work has concentrated on outside surface defects using 45 degree shear waves in contrast to the longitudinal waves usually used for testing thick components with this technique. Due to the small wall thickness of the pressure tubes (4.2 mm) and the typical sizes of defects of interest (0.15 mm or greater), frequencies of the order of 20 MHz are being used. A further complication comes from the orientation of the defects, which may be at any angle in pressure tubes. Initial studies have been performed on a series of outside surface notches and slots, plus a real fatigue crack. This crack was on the inside surface, so the technique required measuring this defect's depth from the outside. Initial results are encouraging. Even without signal processing, crack-tip diffracted signals were detectable from all but very large (2.5 mm) and very small (less than 0.076 mm) notches. Errors in estimates of defect depths were typically less than 0.1 mm for all the notches, and the results were consistent. Measurements on the fatigue crack showed similar random errors, though there appeared to be a deterministic error of about 0.1 mm as well

  14. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    The assembly known as a CANDU power reactor consists of a number of standardized fuel channels or 'power modules'. Each of these channels produces about 5 thermal megawatts on average. Within practical limitations on fuel enrichment and ultimately on economics, the number of these channels is variable between about 50 and approximately 700. Small reactors suffer from inevitable disadvantages in terms of specific cost of design/construction as well as operating cost. Their natural 'niche' for application is in remote off-grid locations. At the same time this niche application imposes new and strict requirements for staff complement, power system reliability, and so on. The distinct advantage of small reactors arises if the market requires installation of several units in a coordinated installation program - a feature well suited to power requirements in Canada's far North. This paper examines several of the performance requirements and constraints for installation of these plants and presents means for designers to overcome the consequent negative feasibility factors.

  15. 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.

  16. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahmed, Rizwan; Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2011-11-15

    The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 x 10{sup 16} Bq, 2.09 x 10{sup 3} W, 5.31 x 10{sup 14} m{sup 3}-water, 4.69 x 10{sup 5} kg, and 7.38 x 10{sup 1} m{sup 3}, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  2. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  3. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  4. The advanced CANDU reactor: The next step in safety and economics

    International Nuclear Information System (INIS)

    The Advanced CANDU Reactor (ACRTM) is the 'Next Generation' CANDUR reactor, aimed at safe, reliable power production at a capital cost significantly less than that of current reactors such as the very successful CANDU 6 reactors (e.g., Wolsong 1-4). The Wolsong 1-4 units are being joined by the Qinshan Phase 3 units in China as the current successful examples of CANDU technology. The ACR design builds on this knowledge base, adding a selected group of innovations to obtain substantial cost reduction while retaining a proven design basis. The ACR maximizes the use of components and equipment applications that are well proven through CANDU and other nuclear experience. Joint development of equipment designs and specifications with manufactures has been emphasized. Similarly, the ACR design emphasizes constructability, and takes advantage of inherent CANDU features to enable short project and construction schedules. Overall, the ACR design represents a balance of proven design basis and innovations to give step improvements in safety, reliability and economics. The ACR development program, now well into the detail design stage, includes parallel formal licensing in the USA and Canada

  5. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    International Nuclear Information System (INIS)

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

  6. Development of the safety regulatory guides on the refurbishment for the CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Chin, T. E.; Rho, H. Y.; Park, H. B.; Yeom, H. G.; Hwang, G. M.; Hwang, B. G.; Seo, Y. H.; Lee, J. W. [Korea Power Engineering Co. Inc., Yongin (Korea, Republic of)

    2007-02-15

    In this study, requirements and standards concerned with safety performance for CANDU type reactors and review guidelines for facilities and performance concerned with refurbishment of major facilities such as pressure tubes, calandria tubes, and feeder popes were developed. To develop review guidelines for facilities and performance review concerned with refurbishment of CANDU reactors, review activities related with refurbishment and performance were categorized into designing and planning of equipments, removal and refurbishment of equipment, and confirmation of installation and inspection. As a result, following detailed review guidelines concerned with refurbishment of pressure tubes, calandria tubes, and feeder pipes in directly or indirectly referring to FSAR, design manual, startup-test manual were developed.

  7. Application of principal component analysis for the diagnosis of neutron overpower system oscillations in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca

    2014-04-01

    Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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)

  13. 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)

  14. 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)

  15. The Impact of Power Coefficient of Reactivity on CANDU 6 Reactors

    International Nuclear Information System (INIS)

    The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant

  16. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  17. Once-through CANDU reactor models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  18. Analysis of flow induced density wave oscillations in the CANDU supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, Goutam, E-mail: gd@iiitdmj.ac.in [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario, Canada N6A 5B9 (Canada); Zhang, Chao, E-mail: czhang@eng.uwo.ca [Department of Mechanical and Materials Engineering, University of Western Ontario, London, Ontario, Canada N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario, Canada N6A 5B9 (Canada)

    2015-05-15

    In this paper, a 1-D thermal-hydraulic model, THRUST, is developed to simulate and analyze the CANDU supercritical water reactor (SCWR) from the thermodynamic point of view without considering the effect of neutronic coupling. THRUST, where a characteristic-based finite difference scheme is used, is validated against the available numerical results. The model is, then, used for the analysis of the CANDU SCWR with a primary focus to determine the conditions for potential density wave oscillations. Extensive numerical studies are performed to obtain the marginal stability boundary in the operating regime of the reactor. The effect of various parameters, such as mass flow rate, operating pressure, axial heat flux profile, local pressure drop coefficient, and friction factor, on the stability thresholds of the reactor have been investigated.

  19. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

  20. Kinetic parameter calculation as function of burn-up of candu reactor

    International Nuclear Information System (INIS)

    Kinetic parameter calculation as function of burn-up of candu reactor. Kinetic marameter calculation as function of burp-up of CANDU reactor with Canflex fuel type-CANDU has been done. This type of fuel is currently being develop, so kinetic parameter such as effective delay neutron fraction (.......), delay neutron decay constant ( .... ) and prompt neutron generation time ( ...... ) are very important for analysis of reactor operation safety. WIMS-CRNL code was used to generate macroscopic cross section and reaction rate based on transport theory. Fast and thermal neutron velocity and macroscopic cross section fission product of the unit cell were determined by KINETIC Code. The result of calculation showed that the value of effective delay neutron fraction was 7,785616 x 10-3 at the beginning of operation at burn-up of 0 MWD/T and after the reactor operated at burn-up of 7,2231 x 10-3 MWD/T was 4,962766 x 10-3, or reduced by 36%. The value of prompt generation time was 9,982703 x 10-4 s at the beginning of operation at burn-up of 0 MWD/T and 8,965416 x 10-4 s after the reactor operated at burn-up of 7,2231 x 103 MWD/T, or reduced by 10%. The result of calculation showed that the values of effective delay neutron fraction and prompt neutron generation time are still great enough

  1. 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

  2. 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)

  3. 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

  4. 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)

  5. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  6. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  7. Mobile robotics for CANDU reactor maintenance: case studies and near-term improvements

    International Nuclear Information System (INIS)

    Although robotics researchers have been promising that robotics would soon be performing tasks in hazardous environments, the reality has yet to live up to the hype. The presently available crop of robots suitable for deployment in industrial situations are remotely operated, requiring skilled users. This talk describes cases where mobile robots have been used successfully in CANDU stations, discusses the difficulties in using mobile robots for reactor maintenance, and provides near-term goals for achievable improvements in performance and usefulness. (author)

  8. 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

  9. Backup and Ultimate Heat Sinks in CANDU Reactors For Prolonged SBO Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Brown, M. J. [Atomic Energy of Canada Limited, Ontario (Canada)

    2013-10-15

    In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ∼2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

  10. Validation of WIMS-AECL reactivity device calculations for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Donnelly, J. V. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-06-01

    An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. The incremental cross section properties of the Wolsung 1 and Point Lepreau adjusters, Mechanical Control Absorbers(MCA) and liquid Zone Control Unit (ZCU) is based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods. (author). 13 tabs., 4 figs., 11 refs.

  11. 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

  12. 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.

  13. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

  14. 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)

  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. 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

  17. Reactor physics innovations in ACR-700 design for next CANDU generation

    International Nuclear Information System (INIS)

    ACR-700 is the 'Next Generation' CANDU reactor, aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs. A key element of cost reduction is the use of H20 as coolant and Slightly Enriched Uranium fuel in a tight D20- moderated lattice. The innovations in the ACR core physics result in substantial improvements in economics, as well as significant enhancements in reactor controllability and waste reduction. Fuel design is chosen to balance fuel performance, cost, and reactor-physics characteristics. Full-core coolant void reactivity in ACR-700 is about 3 mk. Power coefficient is substantially negative. Discharge fuel burnup is about three times the current natural-uranium discharge burn-up. The result is a core design which provides a high degree of inherent safety with attractive power-production efficiency and stability. (author)

  18. 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)

  19. 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).

  20. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

  1. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models

  2. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, S. H.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models.

  3. 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)

  4. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  5. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  6. 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)

  7. 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.)

  8. Advanced CANDU reactor: an optimized energy source of oil sands application

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) is developing the ACR-700TM (Advanced CANDU Reactor-700TM) to meet customer needs for reduced capital cost, shorter construction schedule, high capacity factor while retaining the benefits of the CANDU experience base. The ACR-700 is based on the concept of CANDU horizontal fuel channels surrounded by heavy water moderator. The major innovation of this design is the use of slightly enriched uranium fuel in a CANFLEX bundle that is cooled by light water. This ensures: higher main steam pressures and temperatures providing higher thermal efficiency; a compact and simpler reactor design with reduced capital costs and shorter construction schedules; and reduced heavy water inventory compared to existing CANDU reactors. ACR-700 is not only a technically advanced and cost effective solution for electricity generating utilities, but also a low-cost, long-life and sustainable steam source for increasing Alberta's Oil Sand production rates. Currently practiced commercial surface mining and extraction of Oil Sand resources has been well established over the last three decades. But a majority of the available resources are somewhat deeper underground require in-situ extraction. Economic removal of such underground resources is now possible through the Steam Assisted Gravity Drainage (SAGD) process developed and proto-type tested in-site. SAGD requires the injection of large quantities of high-pressure steam into horizontal wells to form reduced viscosity bitumen and condensate mixture that is then collected at the surface. This paper describes joint AECL studies with CERI (Canadian Energy Research Institute) for the ACR, supplying both electricity and medium-pressure steam to an oil sands facility. The extensive oil sands deposits in northern Alberta are a very large energy resource. Currently, 30% of Canda's oil production is from the oil sands and this is expected to expand greatly over the coming decade. The bitumen deposits in the

  9. The Advanced Candu reactor annunciation system - Compliance with IEC standard and US NRC guidelines

    International Nuclear Information System (INIS)

    Annunciation is a key plant information system that alerts Operations staff to important changes in plant processes and systems. Operational experience at nuclear stations worldwide has shown that many annunciation implementations do not provide the support needed by Operations staff in all plant situations. To address utility needs for annunciation improvement in Candu plants, AECL in partnership with Canadian Candu utilities, undertook an annunciation improvement program in the early nineties. The outcome of the research and engineering development program was the development and simulator validation of alarm processing, display, and information presentation techniques that provide practical and effective solutions to key operational deficiencies with earlier annunciation implementations. The improved annunciation capabilities consist of a series of detection, information processing and presentation functions called the Candu Annunciation Message List System (CAMLS). The CAMLS concepts embody alarm processing, presentation and interaction techniques, and strategies and methods for annunciation system configuration to ensure improved annunciation support for all plant situations, especially in upset situations where the alarm generation rate is high. The Advanced Candu Reactor (ACR) project will employ the CAMLS annunciation concepts as the basis for primary annunciation implementations. The primary annunciation systems will be implemented from CAMLS applications hosted on AECL Advanced Control Centre Information System (ACCIS) computing technology. The ACR project has also chosen to implement main control room annunciation aspects in conformance with the following international standard and regulatory review guide for control room annunciation practice: International Electrotechnical Commission (IEC) 62241 - Main Control Room, Alarm Function and Presentation (International standard) US NRC NUREG-0700 - Human-System Interface Design Review Guidelines, Section 4

  10. 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)

  11. Ultrasonic measurement of gap between calandria tube and liquid injection nozzle in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, ti possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site

  12. Sizing cracks in thin-walled CANDU reactor pressure tubes using crack-tip diffraction

    International Nuclear Information System (INIS)

    The most practical nondestructive means of measuring the depth of cracks approximately 0.4 mm deep in CANDU reactor pressure tubes is the ultrasonic crack-tip diffraction method. Initially, optimum ultrasonic parameters for wave mode, transducer frequency, main-bang pulse characteristics, incident and diffracted angles were obtained on three fatigue cracks, based on the criteria of maximum signal amplitude and accuracy in determination of crack depth. In addition, three signal processing techniques, auto and cross-correlation, rectification and smoothing and the magnitude of the analytic signal, were used to obtain time measurements. The results of these measurements are presented. Except for the first fatigue crack, the depth calculations were accurate to within the specified range of ± 0.1 mm

  13. SMART- IST: a computer program to calculate aerosol and radionuclide behaviour in CANDU reactor containments

    International Nuclear Information System (INIS)

    The SMART-IST computer code models radionuclide behaviour in CANDU reactor containments during postulated accidents. It calculates nuclide concentrations in various parts of containment and releases of nuclides from containment to the atmosphere. The intended application of SMART-IST is safety and licensing analyses of public dose resulting from the releases of nuclides. SMART-IST has been developed and validated meeting the CSA N286.7 quality assurance standard, under the sponsorship of the Industry Standard Toolset (IST) partners consisting of AECL and Canadian nuclear utilities; OPG, Bruce Power, NB Power and Hydro-Quebec. This paper presents an overview of the SMART-IST code including its theoretical framework and models, and also presents typical examples of code predictions. (author)

  14. CAE advanced reactor demonstrators for CANDU, PWR and BWR nuclear power plants

    International Nuclear Information System (INIS)

    CAE, a private Canadian company specializing in full scope flight, industrial, and nuclear plant simulators, will provide a license to IAEA for a suite of nuclear power plant demonstrators. This suite will consist of CANDU, PWR and BWR demonstrators, and will operate on a 486 or higher level PC. The suite of demonstrators will be provided to IAEA at no cost to IAEA. The IAEA has agreed to make the CAE suite of nuclear power plant demonstrators available to all member states at no charge under a sub-license agreement, and to sponsor training courses that will provide basic training on the reactor types covered, and on the operation of the demonstrator suite, to all those who obtain the demonstrator suite. The suite of demonstrators will be available to the IAEA by March 1997. (author)

  15. 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

  16. 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)

  17. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    The market for CANDU reactor sales, both international and domestic, is reviewed. It is reasonable to expect that between five and ten reactors can be sold outside Canada before the end of the centry, and new domestic orders should be forthcoming as well. AECL International has been created to market CANDU, and is working together with the Canadian nuclear industry to promote the reactor and to assemble an attractive package that can be offered abroad. (L.L.)

  18. 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.

  19. 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.

  20. A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

    International Nuclear Information System (INIS)

    A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

  1. Design and development of the high performance shutdown rods for the CANDU 3 reactor

    International Nuclear Information System (INIS)

    The high performance Shutdown Rod (SDR) unit has been successfully developed to improve insertion performance significantly. Because this unit is the reactor's primary emergency shutdown device, reliability is very important. It was therefore prudent to use an evolutionary approach, based on the well-established Shutoff Rod (SOR) unit, rather than attempt radical new concepts. Based on analyses of the key performance influences, the absorber rod's weight was reduced 55% by thinning its absorber and sheath tubes. The release clutch was uprated by 60% by adding extra friction plates and utilizing its full rated voltage. Its release time was shortened simply by working it harder. The proven drive mechanism was otherwise unchanged. The rod and clutch change permitted increasing the accelerating spring force and stroke both 250%. These changes yielded a reduction of insertion time from 0.56 sec (old type SOR) to 0.33 sec, and is 0.11 sec faster than the original CANDU 3 design target. Full scale rig testing has verified it can reliably perform partial drop tests and also endure 3000 full drops, 5 times its rated service demand

  2. 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

  3. 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)

  4. Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

  5. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes

  6. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.

  7. 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).

  8. 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

  9. 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

  10. Modelling material effects on flow-accelerated corrosion in primary CANDU coolant and secondary reactor feed-water

    International Nuclear Information System (INIS)

    The effects of chromium content on flow-accelerated corrosion (FAC) of carbon steel have been predicted very well by including a passivating layer, which is a chromium-dependent diffusion barrier at the metal-oxide interface. By adjusting the properties of the chromium-dependent layer, described with a Passivation Parameter (PP), we can predict the FAC of carbon steel of different chromium contents in typical reactor feed-water environments (140oC and neutral or ammoniated chemistry). The model and an appropriate PP are also applied to the environment typical of carbon-steel feeders in the primary coolant of a CANDU reactor (310oC and lithiated chemistry). The model predicts FAC rate very well (with a deviation of 10% or less) in both situations. (author)

  11. 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)

  12. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDU{sup R} 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    Energy Technology Data Exchange (ETDEWEB)

    Mostofian, Sara; Boss, Charles [AECL Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga Ontario L5K 1B2 (Canada)

    2008-07-01

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  13. Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

  14. 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

  15. 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

  16. 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

  17. Ludwig: A Training Simulator of the Safety Operation of a CANDU Reactor

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

    Full Text Available This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data from a real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students.

  18. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    Science.gov (United States)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

  19. CANDU at the crossroads

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-11-01

    ''Ready for the challenge of the 90s'' was the theme of this year's gathering of the Canadian Nuclear Association held in Toronto, 3-6 June. What that challenge really entails is whether the CANDU system will survive as the last remaining alternative to the light water reactor in the world reactor market, or whether it will decline into oblivion along with the Advanced Gas Cooled reactor and so many other technically excellent systems which have fallen along the way. The fate of the CANDU system will not be determined by its technical merits, nor by its impeccable safety record. It will be determined by public perceptions and by the deliberations of an Environmental Assessment Panel established by the Government of Ontario. The debate at the Association meeting is reported. (author).

  20. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDUR 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    International Nuclear Information System (INIS)

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  1. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  2. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2004-02-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

  3. ADORE-GA: Genetic algorithm variant of the ADORE algorithm for ROP detector layout optimization in CANDU reactors

    International Nuclear Information System (INIS)

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

  4. Kinetics of iodine and cesium reactions in the CANDU reactor primary heat transport system under accident conditions

    International Nuclear Information System (INIS)

    Gas-phase reaction kinetics have been modelled for the release of cesium and iodine into steam and steam/hydrogen atmospheres. The conditions are those anticipated in a CANDU reactor fuel channel following some postulated loss-of-coolant accidents. A total of seventeen chemical species were used in the model, including all important cesium and iodine species. Reaction rate constants were taken from the literature, or calculated where possible, or estimated. The composition evolution of the system was calculated, following a burst release of cesium and iodine, as a function of total iodine and cesium concentrations, cesium/iodine release ratio, iodine release form (atomic I or CsI), fuel channel atmosphere, and radiolysis effects. In general, the calculation demonstrates that CsI and CsOH rapidly (-2 s) become the most important species in the system for virtually all conditions. Atomic I is found to be significant only for very low release concentrations, or for Cs:I ratios less than unity. The main body of the modelling was performed at 1000 K. Some calculations were also performed for a three-node temperature system - 1500 K, 1000 K and 750 K - with the fission products being transported from high to low temperature. Thus, a qualitative picture is provided of the evolution of the chemistry in the fuel channel as the fission products are swept out by the residual steam flow

  5. 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

  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. Simulation of a power pulse during loss of coolant accident in a CANDU-6 reactor by coupling the neutronic code PUMA and the thermalhydraulic code CATHENA

    International Nuclear Information System (INIS)

    In the frame of the safety analysis for a joint feasibility study (between Nucleoelectrica Argentina and Atomic Energy of Canada) of using slightly enriched uranium fuel (0.9 w% U235), Loss of Coolant Accidents (LOCAs) simulations were performed for Embalse NPP, a CANDU-6 type reactor (648. MWe gross). Being a reactor with a positive void reactivity coefficient, the void generation during the first seconds of LOCAs leads to an initial power increase, which is larger in the half of the reactor affected by the break. In order to simulate the power transient, which has a strong spatial variation in the flux and power distributions due to CANDU reactor features, two computer codes were used: the 3 dimensional diffusion, spatial kinetics neutronic program PUMA (developed in Argentina) and the thermal-hydraulics program CATHENA (developed in Atomic Energy of Canada). The codes were coupled by an iterative methodology: the CATHENA thermal-hydraulic simulation results (mainly temperatures of fuel and temperatures and densities of coolant) were used as input of the PUMA neutronic calculation, then the time dependent power distribution calculated by PUMA was applied as input for a new CATHENA calculation. The process was repeated up to convergence, which was obtained in a short number of iterations due to the relative minor effect of the power pulse and the strong influence of the break on the thermal-hydraulics Plant behavior during the analyzed time period. The method was utilized to simulate different accidental scenarios (break size and location, and initial conditions). (author)

  8. Computational Aid of CANDU Reactor Severe Accident Management Strategies for Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Kim, D. H.; Kim, S. D.; Park, S. Y.; Jin, Y. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Six computational aids (CA) have developed for the severe accident management strategies, plant specific features and behaviors must be studied by detailed analysis works. These computational aids (CA) are one of the tools that can be used to assess the hydrogen generation and challenges that may occur. This paper shows hydrogen related CA. The purpose of this CA is to define whether the hydrogen in the reactor building atmosphere is flammable, and to estimate the hydrogen concentration in the reactor building atmosphere based on an estimated oxidation percentage

  9. 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.

  10. 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.

  11. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

  12. 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

  13. Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Fong, R.W.L.; Coleman, C.E

    1998-08-01

    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 CHIF 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 degrees C. (author)

  14. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    The prime focus of the CANDU concept (natural uranium fuelled-heavy water moderated reactor) from the beginning has economy, heavy water losses and radiation exposures also were strong incentives for ensuring good design and reliable equipment. It was necessary to depart from previously accepted commercial standards and to adopt those now accepted in industries providing quality products. Also, through feedback from operating experience and specific design and development programs to eliminate problems and improve performance, CANDU has evolved into today's successful product and one from which future products will readily evolve. Many lessons have been learned along the way. On the one hand, short cuts of failures to understand basic requirements have been costly. On the other hand, sound engineering and quality equipment have yielded impressive economic advantages through superior performance and the avoidance of failures and their consequential costs. The achievement of lifetime economical performance demands quality products, good operation and good maintenance. This paper describes some of the basic approaches leading to high CANDU station reliability and overall excellent performance, particularly where difficulties have had to be overcome. Specific improvements in CANDU design and in such CANDU equipment as heat transport pumps, steam generators, valves, the reactor, fuelling machines and station computers, are described. The need for close collaboration among designers, nuclear laboratories, constructors, operators and industry is discussed. This paper has reviewed some of the key components in the CANDU system as a means of indicating the overall effort that is required to provide good designs and highly reliable equipment. This has required a significant investment in people and funding which has handsomely paid off in the excellent performance of CANDU stations. The close collaboration between Atomic Energy of Canada Limited, Canadian industry and the

  15. ACR technology for CANDU enhancements

    International Nuclear Information System (INIS)

    The ACR-1000 design retains many essential features of the original CANDU plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. This program has extended the database of knowledge on the CANDU design. The ACR-1000 design has been reviewed by the Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) which concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada after completing three phases of the pre-project design review. The generic PSAR for the ACR-1000 design was completed in September 2009. The PSAR contains the ACR-1000 design details, the safety and design methodology, and the safety analysis that demonstrate the ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. The ACR technology developed during the ACR-1000 Engineering Program and the supporting development testing has had a major impact beyond the ACR program itself: Improved CANDU components and systems; Enhanced engineering processes and engineering tools, which lead to better product quality, and better project efficiency; and Improved operational performance. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the EC6, or can be applied for servicing operating CANDU reactors. (author)

  16. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Commercial versions of CANDU reactors were put into service starting more than 25 years ago. The first unit of Ontario Hydro's Pickering A station was put into service in 1971, and Bruce A in 1977. Most CANDU reactors, however, are only now approaching their mid-life of 15 to 20 years of operation. In particular, the first series of CANDU 6 plants which entered service in the early 1980's were designed for a 30 year life and are now approaching mid life. The current CANDU 6 design is based on a 40 year life as a result of advancement in design and materials through research and development. In order to assure safe and economic operation of these reactors, a comprehensive CANDU Plant Life Management (PLIM) program is being developed from the knowledge gained during the operation of Ontario Hydro's Pickering, Bruce, and Darlington stations, worldwide information from CANDU 6 stations, CANDU research and development programs, and other national and international sources. This integration began its first phase in 1994, with the identification of most of the critical systems structures and components in these stations, and a preliminary assessment of degradation and mechanisms that could affect their fitness for service for their planned life. Most of these preliminary assessments are now complete, together with the production of the first iteration of Life Management Plans for several of the systems and components. The Generic CANDU 6 PLIM program is now reaching its maturity, with formal processes to systematically identify and evaluate the major CSSCs in the station, and a plan to ensure that the plant surveillance, operation, and maintenance programs monitor and control component degradation well within the original design specifications essential for the plant life attainment. A Technology Watch program is being established to ensure that degradation mechanisms which could impact on plant life are promptly investigated and mitigating programs established. The

  17. Validation of WIMS-CANDU using Pin-Cell Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U{sup 238} absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data.

  18. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  19. 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

  20. Modelling and simulation of dynamic characteristics of CANDU-SCWR

    International Nuclear Information System (INIS)

    Owing to the thermal properties of supercritical water and features of heat transfer correlation under supercritical pressure, a detailed thermal-hydraulic model with movable boundary of is developed for CANDU-SCWR (Supercritical Water-Cooled Reactor). Steady-state results of the model agree well with the design data. The dynamic responses of CANDU-SCWR to different disturbances are simulated and characteristics are analyzed. A dynamic model for ACR is also developed using CATHENA. Differences between dynamic characteristics of CANDU-SCWR and those of ACR are highlighted and investigated. It is concluded that CANDU-SCWR has a larger time constant, but with a higher response amplitude. (author)

  1. Pressure tube life management in CANDU-6 nuclear plant

    International Nuclear Information System (INIS)

    Operating parameters of pressure tube in CANDU-6 reactor, the relation between pressure tube life and plant life improvement of pressure tube by AECL in past years were summarized, and the factors affecting pressure tube life, idea and main measures of pressure tube life management in QINSHAN CANDU-6 power plant introduced

  2. Technology spin-offs from a CANDU development program

    International Nuclear Information System (INIS)

    Both Enhanced CANDU 6 (EC6) and ACR-1000 design retain many essential features of the operating CANDU 6 plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. The ACR technology developed during the ACR-1000 Basic Engineering Program and the supporting development testing has extended the database of knowledge on the CANDU design. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the Enhanced CANDU 6 (EC6), or can be applied for servicing operating CANDU reactors. (author)

  3. Influence of the flux axial form on the conversion rate and duration of cycle between recharging for ThPu and U{sub nat} fuels in CANDU reactors; Influence de la forme axiale du flux sur le taux de conversion et la duree du cycle entre rechargements pour du combustible ThPu et U{sub nat} dans les reacteurs CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, Richard [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier / CNRS-IN2P3, 53 Avenue des Martyrs, F-38026 Grenoble (France)

    2007-01-15

    To face the increasing world power demand the world nuclear sector must be continuously updated and developed as well. Thus reactors of new types are introduced and advanced fuel cycles are proposed. The technological and economic feasibility and the transition of the present power park to a renewed park require thorough studies and scenarios, which are highly dependent on the reactor performances. The conversion rate and cycle span between recharging are important parameters in the scenarios studies. In this frame, we have studied the utilisation of thorium in the CANDU type reactors and particularly the influence of axial form of the flux, i.e. of the recharging mode, on the conversion rate and duration of the cycle between recharging. The results show that up to a first approximation the axial form of the flux resulting from the neutron transport calculations for assessing the conversion rate is not necessary to be taken into account. However the time span between recharging differs up to several percents if the axial form of the flux is taken into consideration in transport calculations. Thus if the burnup or the recharging frequency are parameters which influence significantly the deployment scenarios of a nuclear park an approach more refined than a simple transport evolution in a typical cell/assembly is recommended. Finally, the results of this study are not more general than for the assumed conditions but they give a thorough calculation method valid for any recharging/fuel combination in a CANDU type reactor.

  4. 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

  5. Proceedings of the Canadian Nuclear Society CANDU maintenance conference

    International Nuclear Information System (INIS)

    The conference proceedings comprise 51 papers on the following aspects of maintenance of CANDU reactors: Major maintenance projects, maintenance planning and preparation, maintenance effectiveness, future maintenance issues, safety and radiation protection. The individual papers have been abstracted separately

  6. Korea signs for 2nd CANDU at Wolsong

    International Nuclear Information System (INIS)

    The sale of a second CANDU 6 reactor to Korea for the Wolsong site is discussed in relation to nuclear power in Korea, the Korean economy generally, Canadian trade with Korea, and cooperation between AECL and KAERI

  7. Preliminary evaluation of licensing issues associated with U. S. -sited CANDU-PHW nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    van Erp, J B

    1977-12-01

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S.

  8. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 6 - PRESENTATION OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

  9. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 9 - CUTTING AND EXTRACTING DEVICE FUNCTIONING

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper presents a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components. It's a flexible and modular device, which is designed to work inside the fuel channel and has the following functions: moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and flexible elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing variable travel speed through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three knife rolls for pressure tube cutting, using a system for cutting place video surveillance and pyrometers for monitoring cutting place temperature. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  10. CANDU 9 safety enhancements and licensability

    International Nuclear Information System (INIS)

    The CANDU 9 design has followed the evolutionary product development approach that has characterized the CANDU family of nuclear power plants. In addition to utilizing proven equipment and systems from operating stations, the CANDU 9 design has looked ahead to incorporate design and safety enhancements necessary to meet evolving utility and regulatory requirements both in Canada and overseas. To demonstrate licensability in Canada, and to assure overseas customers that the design had independent regulatory review in the country of origin, the pre-project Basic Engineering Program included an extensive two year formal review by the Canadian regulatory authority, the Atomic Energy Control Board (AECB). Documentation submitted for this licensing review included the licensing basis, safety requirements and safety analysis necessary to demonstrate compliance with regulations as well as to assess system design and performance. The licensing review was successfully completed in 1997 January. In addition, to facilitate licensability in Korea, CANDU 9 incorporates feedback from the application of Korean licensing requirements to the CANDU 6 reactors at Wolsong site. (author)

  11. CANDU safety and licensing framework and process

    International Nuclear Information System (INIS)

    Nuclear Safety is a shared responsibility of the Industry, public and the Government. The International Atomic Energy Agency's (IAEA) safety fundamentals, basic objectives and safety guides lay down the principles from which requirements, recommendations and methodologies for safety design of Nuclear Power Plants (NPP) are derived. Within the framework of the international regulations and those of the Canadian Nuclear Safety Commission (CNSC), this paper will discuss the overall safety objectives, the defence in depth philosophy guiding CANDU safety, as well as the licensing process defined to meet all applicable CNSC regulations. The application of such philosophy to the ACR design and safety approach will also be discussed along with aspects of its implementation. The role of deterministic analysis, and Probabilistic Safety Analysis (PSA) in the design and licensing process of the Advanced CANDU Reactor will be discussed. Postulated initiating events and their combinations, acceptance criteria, CANDU margins and limits, supporting methodologies and computer codes used in safety analysis will be reviewed. The paper will also note intrinsic safety characteristics of CANDU, some of the ACR passive safety features built-in by design, CANDU distinctive features with respect to severe core damage, mechanisms of heat rejection in those extreme conditions, emergency coolant injection system features and other post accident mitigating systems. Update on the ACR Canadian and US licensing progress will also be provided. (authors)

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. 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

  17. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    This paper consolidates the actual cost experience gained by Atomic Energy of Canada Limited, Ontario Hydro, and other Canadian electric utlities in the planning, design and construction of CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) generating stations over the past 30 years. For each of the major CANDU-PHWR generating stations in operation and under construction in Canada, an analysis is made to trace the evolution of the capital cost estimates. Major technical, economic and other parameters that affect the cost trends of CANDU-PHWR generating stations are identified and their impacts assessed. An analysis of the real cost of CANDU generating stations is made by eliminating interest during construction and escalation, and the effects of planned deferment of in-service dates. An historical trend in the increase in the real cost of CANDU power plants is established. Based on the cost experience gained in the design and construction of CANDU-PHWR units in Canada, as well as on the assessment of parameters that influence the costs of such projects, the future costs of CANDU-PHWRs are presented

  18. Preliminary evaluation of licensing issues associated with U.S.-sited CANDU-PHW nuclear power plants

    International Nuclear Information System (INIS)

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S

  19. 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

  20. CANDU 9 design for hydrogen in containment

    International Nuclear Information System (INIS)

    CANDU 9 is a single unit plant whose design is based on proven CANDU technology with some design improvements in plant performance and safety. One improvement is in the area of post-accident hydrogen control. The reactor building layout and hydrogen control system are designed to enhance atmospheric mixing and prevent unacceptably high local and global hydrogen concentrations. Preliminary safety analysis shows that a maximum of 300 kg of hydrogen is produced during the metal-water reaction phase of severe accident: a large LOCA with emergency core cooling unavailable. Even though the hydrogen production rate is conservatively overestimated, preliminary containment thermalhydraulic analysis predicts that the maximum hydrogen concentration in the accident vault peaks at 6.8% by volume without crediting hydrogen igniters and recombiners. After about one hour, the concentration throughout the reactor building is about 2.4%. (author)

  1. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Rhee, B. W.; Park, J. H.; Kim, H. T.; Choi, H. B.; Shim, J. I.; Yoon, C.; Yang, M. K

    2002-03-01

    To develop CANDU safety analysis system, methodology, and assessment technology, GAIs from CNSC and GSIs drived by IAEA are summarized. Furthermore, the following safety items are investigated in the present study. - It is intended to secure credibility of the void reactivity in the stage of nuclear design and analysis. The measurement data concerned with the void reactivity were reviewed and used to assess the physics code such as POWDERPUFS-V/RFSP, and the lattice code such as WIMS-AECL and MCNP-4B. - Reviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc. were examined. - The development of 3D CFD transient analysis model has been performed to predict local subcooling of the moderator in the vicinity of Calandria tubes in a CANDU-6 reactor in the case of Large LOCA transient. - The trip coverage analysis methodology based on CATHENA code is developed. The simulation of real plant transient showed good agreement. The trip coverage map was generated successfully for two typical depressurization and pressurization event. - The multi-dimensional analysis methodology for hydrogen distribution and hydrogen burning phenomena in PHWR containment is developed using GOTHIC code. The multi-dimensional analysis predicts the local hydrogen behaviour compared to the lumped parameter model.

  2. Ontario Hydro 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 report 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 for the workers and the public

  3. Ontario Hydro 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 report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

  4. 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)

  5. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  6. Heavy water detritiation to support CANDU station maintenance activities

    International Nuclear Information System (INIS)

    Heavy water and tritium control are important aspects of CANDU operation and have a major impact on station maintenance activities. Station personnel are trained to understand the importance of heavy water management and the economics and environmental impact of tritiated heavy water losses. This paper discusses new GE technology that can now make a major improvement in CANDU maintenance activities through significant reductions in station tritium levels. Tritium is of particular concern in the CANDU industry given the nature of heavy water reactors to build up high levels of tritium over time. High tritium levels in the reactor vault significantly slow down maintenance activities in the reactor vault due to the requirement for personnel protective equipment, including breathing apparatus and cumbersome plastic air suits. The difficulties increase as reactors age and tritium levels increase. Building upon GE's extensive operational experience in tritium management in CANDU reactors and its own tritium handling facility, GE. has developed a new large-scale diffusion based isotope separation process as an alternative to conventional cryogenic distillation. Having a tritium inventory an order of magnitude lower than conventional cryogenic distillation, this process is very attractive for heavy water detritiation, and applicable to single and multi-unit CANDU stations. This new process can now provide a step change reduction in CANDU heavy water tritium levels resulting in reduced environmental emissions and lowering reactor vault tritium MPC(a) levels. Reactor vault tritium can be reduced sufficiently for maintenance activities to be done without plastic suits, leading to shorter outages, improved station capacity factors, and improved station economics. (author)

  7. The CANTEACH project: preserving CANDU technical knowledge

    International Nuclear Information System (INIS)

    Almost sixty years have passed since the nuclear energy venture began in Canada. Fifty years have passed since the founding of AECL. Tens of thousands of dedicated people have forged a new and successful primary energy supply. CANDU technology is well into its second century. This specialty within the world's fission technology community is quite unique, first because it was established as a separate effort very early in the history of world fission energy, and second because it grew in an isolated environment, with tight security requirements, in its early years. Commercial security rules later sustained a considerable degree of isolation. The pioneers of CANDU development have finished their work. Most of the second generation also has moved on. As yet, we cannot point to a consistent and complete record of this remarkable achievement. We, as a nuclear enterprise, have not captured the design legacy in a form that is readily accessible to the current and future generation of professionals involved with CANDU reactors, be they students, designers, operations staff, regulators, consultants or clients. This is a serious failure. Young people entering our field of study must make do with one or two textbooks and a huge collection of diverse technical papers augmented by limited-scope education and training materials. Those employed in the various parts of the nuclear industry rely mostly on a smaller set of CANDU- related documents available within their own organization; documents that sometimes are rather limited in scope. University professors often have even more limited access to in-depth and up to date information. In fact, they often depend on literature published in other countries when preparing lectures, enhanced by guest lecturers from various parts of the industry. Because CANDU was developed mostly inside Canada, few of these text materials contain useful data describing processes important to the CANDU system. For many years it has been recognized that

  8. 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

  9. 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

  10. 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)

  11. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability.

  12. 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)

  13. Configuration management for CANDU feeder refurbishment

    International Nuclear Information System (INIS)

    The Canada Deuterium Uranium Reactor CANDU was originally designed to last twenty-five years. In 2005, Atomic Energy of Canada (AECL) made the decision to extend the life of the reactor by thirty years. One of the most critical elements of the life extension project was determining how to refurbish the Primary Heat Transport System. It was determined that the refurbishment required replacing the entire length of inlet and outlet feeders, from the end fittings to the header. The use of a robust Configuration Management program would have added significant value to the life extension project. (author)

  14. 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

  15. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  16. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  17. Mitigating aging in CANDU plants

    International Nuclear Information System (INIS)

    Aging degradation is a phenomenon we all experience throughout life, both on a personal basis and in business. Many industries have been successful in postponing the inevitable impact on their related systems and components through programs to maintain long-term reliability, maintainability and safety. However, this has not always been the case for nuclear power. While all power plants are experiencing the world trend of increasing operating costs with age, few (if any) have been able to fully define the parameters that solve the aging equation, particularly in relation to major components. Inspection and preventive maintenance have not been effective in predicting life-limiting degradation and failure. In CANDU nuclear plants, utilities are taking a comprehensive approach in dealing with the aging problem. Programs have been established to identify the current condition and degradation mechanisms of critical components, the failure of which would impact negatively on station competitiveness and safety. These include subcomponents under the general headings of reactor components, civil structures, piping (nuclear and conventional), steam generators, turbines and cables. In support of these efforts, R and D projects have been defined under the CANDU Owners Group to deal with generic issues on aging common to its members (e.g., investigation of degradation mechanisms, development of tools and techniques to mitigate the effects of aging, etc.). This paper describes recent developments of this cost-shared program with specific reference to concrete aging and crack repairs, flow-assisted corrosion in piping, elastomer service life, cable aging, degradation mechanisms in steam generators and lubricant breakdown. (author)

  18. 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

  19. Development of advanced CANDU PHWR -Development of the advanced CANDU technology-

    International Nuclear Information System (INIS)

    The target of this project is to assess the feasibility of improving PHWR and to establish the parameter of the improved concept and requirements for developing it. To set up the requirements for the Improved Pressurized Heavy Water Reactor: (1) Design requirements of PHWR main systems and Safety Design Regulatory Requirements for Safety Related System i.e. Reactor Shutdown System, Emergency Core Cooling System and Containment System were prepared. (2) Licensing Basis Documents were summarized and Safety Analysis Regulatory. Requirements were reviewed and analyzed. To estimate the feasibility of improving PHWR and to establish the main parameters of the concept of new PHWR in future: (1) technical level/developing trend of PHWR in Korea through Wolsong 2, 3 and 4 design experience and Technical Transfer Program was investigated to analyze the state of basic technology and PHWR improvement potential. (2) CANDU 6 design improvement tendency, CANDU 3 design concept and CANDU 9 development state in other country was analyzed. (3) design improvement items to apply to the reactors after Wolsong 2, 3 and 4 were selected and Plant Design Requirements and Conceptual Design Description were prepared and the viability of improved HWR was estimated by analyzing economics, performance and safety. (4) PHWR technology improving research and development plan was established and international joint study initiated for main design improvement items

  20. Independence and diversity in CANDU shutdown systems

    International Nuclear Information System (INIS)

    Atomic Energy Control Board regulations state that Canadian CANDU reactors shall have two fully effective, independent and diverse shutdown systems. The Darlington Nuclear Generating Station is the first power plant operated by Ontario Hydro to make use of software-based computer control in its shutdown systems. By virtue of the reliance placed on these systems to prevent exposure of the public to harmful radioactivity in the event of an accident, the shutdown system software has been categorized as safety critical software. An important issue that was considered in the design of the Darlington shutdown systems was how the software should be designed and incorporated into the systems to comply with the independence and diversity requirement. This paper describes how the independence and diversity requirement was complied with in previous CANDU shutdown system designs utilizing hardware components. The difference between systems utilizing hardware alone, and those utilizing both hardware and software are discussed. The results of a literature search into the issue of software diversity, the behaviour of multi-version software systems, and experience in other industries utilizing safety critical software are referred to. This paper advocates a systems approach to designing independent shutdown systems utilizing software. Opportunities exist at the system level for design decisions that can enhance software diversity and can reduce the likelihood of common mode faults in the systems. In the light of recent experience in implementing diverse safety critical software, potential improvements to the design process for CANDU shutdown systems are identified. (Author) 19 refs

  1. 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

  2. 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

  3. 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.

  4. 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

  5. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  6. 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)

  7. Licensing evaluation of CANDU-PHW nuclear power plants relative to U.S. regulatory requirements

    International Nuclear Information System (INIS)

    Differences between the U.S. and Canadian approach to safety and licensing are discussed. U.S. regulatory requirements are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to current Regulatory Requirements and Guides. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S. These modifications are proposed solely for the purpose of maintaining consistency within the current U.S. regulatory system and not out of a need to improve the safety of current-design CANDU-PHW nuclear power plants. A number of issues are identified which still require resolution. Most of these issues are concerned with design areas not (yet) covered by the ASME code. (author)

  8. Proof testing of CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Prior to commissioning of a CANDU reactor, a proof pressure test is required to demonstrate the structural integrity of the containment envelope. The test pressure specified by AECB Regulatory Document R-7 (1991) was selected without a rigorous consideration of uncertainties associated with estimates of accident pressure and conatinment resistance. This study was undertaken to develop a reliability-based philosophy for defining proof testing requirements that are consistent with the current limit states design code for concrete containments (CSA N287.3).It was shown that the upodated probability of failure after a successful test is always less than the original estimate

  9. 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)

  10. 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

  11. Operational improvements in the CANDU 9 control centre

    International Nuclear Information System (INIS)

    AECL has adopted an evolutionary approach to the design of the CANDU 9 control centre. Several factors have contributed to this decision including the desire to build on the successes of the current generation of CANDU stations, the changing roles and responsibilities of operations staff, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and evolving customer and regulatory requirements. Underlying this approach is a refined engineering design process that cost-effectively integrates operational feedback and human factors engineering to define the operating staff information and information presentation requirements. Based on this approach, the CANDU 9 control centre will provide utility operating staff with a layout and information organization that is better matched to operational tasks, thereby leading to reduced operations, maintenance and administration (OM and A) costs. Significant design features that contribute to the improved operational capabilities of the CANDU 9 control centre include: a control centre layout with improved functionality; a new Plant Display System that is separated from the digital control computer system; and an enhanced computerized reactor shutdown system. The paper will present a summary of the design process, a detailed description of the CANDU 9 control centre layout and features, a description of the plant control and display systems design, including findings from a regulatory review, and other improvements to enhance operability. (author)

  12. A comparison of the void reactivity effect between the CANDU standard and CANDU-6 SEU-43 cells

    International Nuclear Information System (INIS)

    In a CANDU type reactor the void coefficient of the reactivity is positive. The experimental data are available only for fresh fuel in cold conditions. On the other hand, taking into account the reactivity effects induced by changes of the coolant properties is often difficult. The safety analyses require an estimation of the calculation error. A comparison between models is an usual approach to obtain detailed information. In our paper a heterogeneous multi-stratified coolant model is used both for the CANDU standard fuel assembly cell and CANDU SEU-43 cell concept. The coolant is treated as a two phase (liquid and vapors) medium gravitationally separated. The results are inter-compared for different burnups in the partial or total void cases. (authors)

  13. 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

  14. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

    The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (Kr and Lr). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables Kr and Lr, the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

  15. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 10 - PRESENTATION OF THE DECOMMISSIONING DEVICE OPERATING

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU,

    2015-05-01

    Full Text Available This paper presents a solution proposed by the authors in order to achieve of a cutting and extracting device operating panel for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components, moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. All operations can be monitored and controlled from a operating panel. The PLC fully command the device in automatic or manually mode, to control the internal sensors, transducers, electrical motors, video surveillance and pyrometers for monitoring cutting place temperature. The device controller has direct access to the measured values with these sensors, interprets and processes them, preparing the next actionafter confirming the action in progress. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  16. CANDU Safety R&D Status, Challenges, and Prospects in Canada

    Directory of Open Access Journals (Sweden)

    W. Shen

    2015-01-01

    Full Text Available In Canada, safe operation of CANDU (CANada Deuterium Uranium; it is a registered trademark of Atomic Energy of Canada Limited reactors is supported by a full-scope program of nuclear safety research and development (R&D in key technical areas. Key nuclear R&D programs, facilities, and expertise are maintained in order to address the unique features of the CANDU as well as generic technology areas common to CANDU and LWR (light water reactor. This paper presents an overview of the CANDU safety R&D which includes background, drivers, current status, challenges, and future directions. This overview of the Canadian nuclear safety R&D programs includes those currently conducted by the COG (CANDU Owners Group, AECL (Atomic Energy of Canada Limited, Candu Energy Inc., and the CNSC (Canadian Nuclear Safety Commission and by universities via UNENE (University Network of Excellence in Nuclear Engineering sponsorship. In particular, the nuclear safety R&D program related to the emerging CANDU ageing issues is discussed. The paper concludes by identifying directions for the future nuclear safety R&D.

  17. Generic validation of computer codes used in safety analyses of CANDU power plants

    International Nuclear Information System (INIS)

    Since the 1960s, the CANDU industry has been developing and using scientific computer codes, validated according to the quality-assurance practices of the day, for designing and analyzing CANDU power plants. To provide a systematic framework for the validation work done to date and planned for the future, the industry has decided to adopt the methodology of validation matrices, similar to that developed by the Nuclear Energy Agency of the Organization for Economic Co-operation and Development for Light Water Reactors (LWR). Specialists in six scientific disciplines are developing the matrices for CANDU plants, and their progress to date is presented. (author)

  18. Technology transfer in CANDU marketing

    International Nuclear Information System (INIS)

    The author discusses how the CANDU system lends itself to technology transfer, the scope of CANDU technology transfer, and the benefits and problems associated with technology transfer. The establishment of joint ventures between supplier and client nations offers benefits to both parties. Canada can offer varying technology transfer packages, each tailored to a client nation's needs and capabilities. Such a package could include all the hardware and software necessary to develop a self-sufficient nuclear infrastructure in the client nation

  19. CANDU steam generator life management: laboratory data and plant experience

    International Nuclear Information System (INIS)

    As CANDU reactors enter middle age, and the potential value of the plants in a deregulated market is realized, life management and life extension issues become increasingly important. An accurate assessment of critical components, such as the CANDU 6 steam generators (SGs), is crucial for successful life extension, and in this context, material issues are a key factor. For example, service experience with Alloy 900 tubing indicates very low levels of degradation within CANDU SGs; the same is also noted worldwide. With little field data for extrapolation, life management and life extension decisions for the tube bundles rely heavily on laboratory data. Similarly, other components of the SGs, in particular the secondary side internals, have only limited inspection data upon which to base a condition assessment. However, in this case there are also relatively little laboratory data. Decisions on life management and life extension are further complicated--not only is inspection access often restricted, but repair or replacement options for internal components are, by definition, also limited. The application of CANDU SG life management and life extension requires a judicious blend of in-service data, laboratory research and development (R and D) and materials and engineering judgment. For instance, the available laboratory corrosion and fretting wear data for Alloy 800 SG tubing have been compared with plant experience (with all types of tubing), and with crevice chemistry simulations, in order to provide an appropriate inspection guide for a 50-year SG life. A similar approach has been taken with other SG components, where the emphasis has been on known degradation mechanisms worldwide. This paper provides an outline of the CANDU SG life management program, including the results to date, a summary of the supporting R and D program showing the integration with condition assessment and life management activities, and the approach taken to life extension for a typical

  20. Investigation of techniques for the application of safeguards to the CANDU 600 MW(e) nuclear generating station

    International Nuclear Information System (INIS)

    A cooperative program with the Canadian Atomic Energy Control Board, Atomic Energy of Canada Limited and the IAEA was established in 1975: to determine the diversion possibilities at the CANDU type reactors using a diversion path analysis; to detect the diversion of nuclear materials using material accountancy and surveillance/containment. Specific techniques and instrumentation, some of which are unique to the CANDU reactor, were developed. 10 appendices bring together the relevant reports and memoranda of results for the Douglas Point Program

  1. 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.)

  2. 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

  3. 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)

  4. 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)

  5. The Candu suppliers: Growing beyond Canada

    International Nuclear Information System (INIS)

    Canada has demonstrated its ability to successfully manage Candu projects. It has been shown domestically by Ontario Hydro over a period of many years. It has been demonstrated abroad by private sector service companies, under Nuclear Construction Managers, on the 600 MW Wolsung project in South Korea, where NPM companies carried out virtually all project and construction management, the plant was in service only 60 months after the first concrete was poured for the reactor building. In the past, however, the project management team for Candu nuclear plants overseas, and for some at home, was often assembled ad hoc. Though this approach can be successful, it does not guarantee success in project and construction management. Because of the size, diversity and flexibility of the private owners of NPM, they have been able to keep their experienced nuclear power teams on staff even though the nuclear power industry has been going through slow times, assigning these people to large projects in other industries that call for the same high level of project, construction and commissioning management skills. Yet they are in a position to reassemble this team at very short notice, and have them ready to go to work on a major project within weeks of the green light

  6. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  7. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    International Nuclear Information System (INIS)

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAPS/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAPS/MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAPS source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version

  8. 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)

  9. 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)

  10. An investigation into the relationship between local and global power excursions in CANDU

    International Nuclear Information System (INIS)

    If a reactor exhibits large neutronic decoupling of one core region from another, it may be able to sustain localized power excursions. If such an excursion is too localized, there may be no in-core detector close enough to detect it promptly. To confirm that CANDU reactors are unlikely to support local power excursions, we selected a credible local reactivity-insertion mechanism in the CANDU 600 and calculated the resulting neutron flux transient with a three-dimensional model. A coupled neutronics-thermohydraulics simulation of the transient was performed. The transient was subdivided into appropriate time steps, and iteration between the neutronics and thermohydraulics calculations was carried out at each step, so that the proper distribution of thermohydraulic feedback reactivity was obtained. The calculation shows that, with credible reactivity insertions, the neutronic characteristics of CANDU reactors do not allow local power excursions to occur without a comparable global power increase. The latter would not escape detection by the protective systems

  11. Ballooning of CANDU pressure tubes. Model assessment

    International Nuclear Information System (INIS)

    The transient creep equations used to analyze the possible ballooning and failure of Zr-2.5% Nb pressure tubes during a loss-of-coolant accident (LOCA) were developed and verified using as-received Zr-2.5% Nb pressure tube material. But in a CANDU reactor, the pressure tubes absorb deuterium and are exposed to a continuous neutron fluence. Consequently, a literature survey was done to determine how irradiation damage and deuterium might affect the creep rate and ductility of Zr-2.5% Nb pressure tubes in the temperature range from 600 to 800 degrees C. It was found that irradiation damage, dissolved deuterium and deuteride blisters could possibly affect the creep rate and ductility of ZR-2.5% Nb pressure tubes in this temperature range, but deuteride platelets are expected to have little effect. Further tests are required to determine the effect of irradiation damage and deuterium on the creep rate and ductility of pressure tubes

  12. 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)

  13. 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)

  14. 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)

  15. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  16. Advancement of safeguards inspection technology for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  17. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 7 - FUNCTIONING OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The scope of this paper is to achieve the device functioning steps for the commissioning of the horizontal fuel channels of calandria vessel. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. For the decommissioning operation design shall be taken to ensure all aspects of security, environmental protection during decommissioning operation steps and creating and implementing work procedures resulting from developed decommissioning plan. The fuel channel decommissioning device is designed for dismantling and extraction of the fuel channel and its components. The decommissioning operation consists of following major steps: platform with device positioning to the fuel channel to be dismantled; coupling and locking the device at the fuel channel; unblock, extract and store the channel closure plug; unblock, extract and store the channel shield plug; block and cut the middle and the end of the pressure tube; block, extract and store the end fitting; block, extract and store the half of pressure tube; mounting of the extended closing plug. The operations steps are performed by the Cutting and Extraction Device and by the extraction actuator from the device handling elements assembly. After each step of dismantling is necessary the confirmation its finalization in order to perform the next operation step. The dismantling operation steps of the fuel channel components are repeated for all the 380 channels of the reactor, from the front of calandria side (plane R as well as the rear side (plane R'.

  18. Thermal fatigue screening criteria for identifying susceptible piping components in CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Schefski, C.; Chen, Q.; Pentecost, S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    In December 1987, a fatigue failure in a non-isolable section of a safety injection line at the Farley-2 plant prompted the U.S. Nuclear Regulatory Commission (NRC) to issue Bulletin 88-08 requiring that U.S. utilities review all non-isolable branch lines to determine if they are susceptible to thermal fatigue. The thermal fatigue incident at Farley-2 was caused by stresses in the pipe wall resulting from large-scale temperature fluctuations. Shortly after the Farley-2 event, several other incidents with through-wall cracks due to thermal fatigue had occurred in plant subsystems and piping configurations similar to the Farley-2 safety injection line. Thermal fatigue cracks have also occurred in piping configurations with different geometries, such as drain, residual heat removal, and shutdown cooling suction lines in various pressurized water reactors (PWR) and boiling water reactors (BWR). Thermal fatigue, caused by local thermal stratification phenomena, has received significant attention in the PWR and BWR communities in the past two decades. Although CANDU stations have experienced relatively few thermal fatigue failures; the impact of this known fatigue mechanism for CANDU designs has not been rigorously assessed. Screening and evaluation methodology, which has been developed by Electric Power Research Institute (EPRI) to identify locations susceptible to thermal cycling in PWR systems, has recently been modified under a CANDU Owners Group (COG) project for application in CANDU piping systems. This paper describes a new software tool for evaluating locations susceptible to thermal fatigue in CANDU piping systems in an effort to avoid failures that lead to costly plant shutdowns. The software, combined with engineering judgement, will assist CANDU station staff to focus their inspections on key components, therefore reducing dose, time and cost during outages. Computational Fluid Dynamics (CFD) was used to form the basis for expanding the range of validity in

  19. 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

  20. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

    1997-07-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  1. 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)

  2. 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

  3. Experimental validation of Pu-Sm evolution model for CANDU-6 power transients

    International Nuclear Information System (INIS)

    Development of a methodology to evaluate the reactivity produced by Pu-Sm transient, effect displayed after power transients. This methodology allows to predict the behavior of liquid zones with which the fine control of CANDU reactor power is made. With this information, it is easier to foresee the refueling demand after power movements. The comparison with experimental results showed good agreement. (author)

  4. Data base for a CANDU-PHW operating on a once-through natural uranium 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)

  5. 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

  6. 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)

  7. An experimental study of a flashing-driven CANDU moderator cooling system

    International Nuclear Information System (INIS)

    The results of an experimental study to investigate the feasibility of using a passive flashing-driven natural circulation loop for CANDU-reactor moderator heat rejection are presented. A scaled loop was constructed and tested at conditions approximating those of a CANDU calandria cooling system. The results showed that stable loop operation was possible at simulated powers approaching normal full power. At lower powers, flow oscillations occurred as the flow in the hot-leg periodically changed from two-phase to single-phase. The results from earlier numerical predictions using the CATHENA thermalhydraulics code showed good qualitative agreement with the experimental results. (author). 6 refs., 11 figs

  8. Improvement of RELAP5/MOD3.2.2 models for the development of CANDU auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Jung, B. D.; Lee, W. J.; Lim, H. S. [KAERI, Taejon (Korea, Republic of); Bang, Y. S.; Kim, M. W.; Lee, S. H. [KINS, Taejon (Korea, Republic of)

    1999-10-01

    Thermal-hydraulic models of the model improvements of NRC PWR auditing tool , i.e. RELAP5/MOD3, current auditing tool for LWR licensing, have been improved were attempted to develop a than auditingermal hydraulic auditing code for the CANDU. Major in order to identify the thermal hydraulic phenomena for the key CANDU events for key eventswere identified for reactor systems and components. Based on this, the modeling limitation of current RELAP5/MOD3 for CANDU applications were derived and the model improvement areas were identified. By improving these models in the code, RELAP5/MOD3/CANDU version has been developed. The new version was written in FORTRAN90 and its application capability has been verified through the simple verification calculations.

  9. Improvement of RELAP5/MOD3.2.2 models for the development of CANDU auditing code

    International Nuclear Information System (INIS)

    Thermal-hydraulic models of the model improvements of NRC PWR auditing tool , i.e. RELAP5/MOD3, current auditing tool for LWR licensing, have been improved were attempted to develop a than auditingermal hydraulic auditing code for the CANDU. Major in order to identify the thermal hydraulic phenomena for the key CANDU events for key eventswere identified for reactor systems and components. Based on this, the modeling limitation of current RELAP5/MOD3 for CANDU applications were derived and the model improvement areas were identified. By improving these models in the code, RELAP5/MOD3/CANDU version has been developed. The new version was written in FORTRAN90 and its application capability has been verified through the simple verification calculations

  10. OpenFOAM Analysis of CANDU-6 Moderator Flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, Se-Myong [Kunsan National University, Gunsan (Korea, Republic of)

    2015-10-15

    In this study OpenFOAM (Open Field Operation and Manipulation), an open source CFD solver, is used to simulate the three-dimensional moderator flow in calandria tank of CANDU-6 reactor improving the computational efficiency by parallel computing which does not need any proprietary license. A prototype of CANDU-6 reactor is numerically analyzed about three-dimensional moderator flow in calandrian tank with OpenFOAM, an open source CFD code. The horizontal fuel channels in a CANDU-6 reactor (a pressurized heavy water reactor) are submerged in the heavy water (D{sub 2}O) pool which is contained by a cylindrical tank, calandria. Each fuel channel consists of concentric tubes: a Pressure Tube (PT) and a Calandria Tube (CT). And the CO{sub 2} gas is filled between these tubes. Consequently, a heat flux is rapidly transferred to the outer CT so that a film boiling may occur in CT. As a result, it is important to keep the subcooling in the moderator. It is one of the major concerns in the CANDU safety analyses to estimate the local subcooling margin of the moderator inside the calandria tank. Previous experimental studies showed that the film boiling would be unlikely to occur if the local moderator subcooling is sufficient. Therefore, an accurate prediction of the moderator temperature distribution in the calandria tank is needed to confirm the channel integrity. There have been numerous computational efforts to estimate the thermal hydraulics in the calandria tank using CFD codes. Hadaller et al. obtained a tube bank pressure drop model for tube bundle region of the calandria tank and implemented it into the MODTURC{sub C}LAS code. Yoon et al. used the CFX code to develop a CFD model with a porous media approach for the core region. However, it is known that porous media modeling provide only average values of flow velocities and temperatures and do not give any information about local flow variables near tube solid walls, which are necessary to implement accurate heat

  11. 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.

  12. Qinshan CANDU 6 main heat transport system high accuracy performance tracking in support of regional overpower protection

    International Nuclear Information System (INIS)

    This paper deals with the Qinshan CANDU 6 main Heat Transport System (HTS) high accuracy performance tracking/adjustment up to about 7 years of operation in support of Regional Overpower Protection (ROP). Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Analytic predictions, in association with an optimized parameter tracking and adjustment methodology, confirm continued safe reactor operation. This paper demonstrates that Qinshan CANDU Unit 1, as compared to other CANDU 6 nuclear reactors of earlier design, continues to exhibit significantly improved performance with much reduced plant aging effects. This paper further demonstrates the high accuracy of the advanced performance tracking and adjustment methodology and applies it to Qinshan CANDU Unit 1, ensuring and demonstrating the continued excellent performance of the reference analytic models. The analytic methodology as well as the advanced performance tracking and analysis methodology can also beneficially be applied to both new and refurbished CANDU type nuclear reactors. (author)

  13. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  14. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwnag, M

    2001-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented.

  15. Hydrogen analysis using MELCOR at CANDU plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Chu; Park, Jae Hong; Kim, Han Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    As a part of a submitted Wolsong Severe Accident Management Plan(SAMP) at the end of 2009, KINS needs to the review of it. In this the focus is on the hydrogen behavior during the station blackout(SBO) because the hydrogen explosion during severe accident in CANDU plants such as Wolsong units has been safety issue. A SBO occurs with failure of the emergency power system when loss of class IV electric power happens because the loss of class IV power is the most dominant internal event causing core damage for Wolsong units. And following a SBO event, most of the engineered safety features(ESFs), including hydrogen igniters, are inoperable except the passive systems such as Dousing systems. Hydrogen is generated due to Zr-steam reaction in fuel channels and core debris oxidation in the suspended debris beds, jet breakup of molten debris in the water pool of the reactor vault and molten core-concrete interaction (MCCI). A combustible gas control system consisting of Passive Autocatalytic Recombiner(PAR) is currently installed at Wolsong unit 1. The results of MELCOR analysis are presented

  16. 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.

  17. Value added services to CANDU plants

    International Nuclear Information System (INIS)

    Over the last decade or so, nuclear power plants, just like other types of electricity generating plants, have been facing a number of challenges. Depending on the operating environment of the utility, these challenges are forcing plant owners to examine all facets of the operating costs. Privatization, deregulation and economics of alternative electricity generation methods are exerting enormous pressure on nuclear power plants to streamline costs and improve their operational performance. CANDU reactors are no exception to these forces and face similar pressures. In particular, operating plants that are contemplating plant life extensions are being required to clearly demonstrate the economics of continued operation over other forms of power generation available to the utility. Improvement of capacity factors has the effect of increasing the revenues from the plant and as these revenues increase, the fixed portion of the plant costs including OM and A costs become a smaller percentage of the total revenues. Similar results can be achieved by aiming to reduce the plant OM and A costs. In reality, most well-planned intervention schemes directed at reducing OM and A costs tend to also increase the plant availability. Following plant turnover after commissioning, AECL has been supporting the CANDU owners and utilities with an assortment of products and services dealing with plant operations and outage management issues. AECL has taken the lead in arranging specialized resources, products and services by teaming with other complementary organizations to provide a complete suite of services. Recent examples of such support to operating CANDU plants will be described in the paper. AECL is responding to this changing business environment in two important ways. First, AECL is changing from simply providing a service to its clients towards providing value, something much more important. To this end, AECL is looking to other organizations to form alliances, partnerships and

  18. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  19. 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).

  20. 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).

  1. The third generation CANDU control room

    International Nuclear Information System (INIS)

    In CANDU stations, as in most complex industrial plants, the man/machine interface design has progressed through three generations. First Generation control rooms consisted entirely on fixed, discrete components (handswitches, indicator lights, strip chart, recorder, annunciator windows, etc.). Human factors input was based on intuitive common sense factors which varied considerably from one designer to another. Second Generation control rooms incorporated video display units and keyboards in the control panels. Computer information processing and display are utilized. There is systematic application of human factors through ergonomic and anthropometric standards and cookbooks. The human factors are applied mainly to the physical layout of the control panels and the physical manipulation performed by the operators. Third Generation control rooms exploit the dramatic performance/cost improvements in computer, electronic display and communication technologies of the 1980's. Further applications of human factors address the cognitive aspects of operator performance. At AECL, second generation control rooms were installed on CANDU stations designed in the mid 70s and early 80s. Third generation features will be incorporated in the CANDU 3 station design and future CANDU stations. There have been significant improvements in the man/machine interface in CANDU stations over the past three decades. The continuing rapid technological developments in computers and electronics coupled with an increasing understanding and application of human factors principles is leading to further enhancements. This paper outlines progress achieved in earlier stations and highlights the features of the CANDU 3rd generation control room. (author). 13 refs, 5 figs

  2. A bibliography of AECL publications on reactor safety

    International Nuclear Information System (INIS)

    AECL Publications on Reactor Safety in CANDU Reactors are listed in this bibliography. The listing is chronological and the accompanying index is by subject. The bibliography will be brought up to date annually. (auth)

  3. Applications of ASTEC integral code on a generic CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Gabriela, E-mail: gabriela.radu@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Prisecaru, Ilie [Power Engineering Department, University “Politehnica” of Bucharest, 313 Splaiul Independentei, Bucharest (Romania)

    2015-05-15

    Highlights: • Short overview of the models included in the ASTEC MCCI module. • MEDICIS/CPA coupled calculations for a generic CANDU6 reactor. • Two cases taking into account different pool/concrete interface models. - Abstract: In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon. The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core–concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction–correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust.

  4. 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

  5. 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)

  6. CANDU 9-meeting evolving safety and licensing requirements

    International Nuclear Information System (INIS)

    To date the safety and licensing experience with CANDU has been good, with plants successfully licensed and operating safely both domestically and in other countries. Nevertheless it is now clearly recognized that the utilities need more formal evidence that the licensing risk is low, if they are to proceed with new nuclear power projects. The CANDU 9 is an evolutionary design generating over 900 MWe. It is based on already-operating CANDUs, with improvements in the areas of construct ability, economics, plant layout, operations and maintenance, and safety. Even so, both domestic and foreign potential users require evidence of safety and licensability in Canada as of 'today'. For CANDU 9, this involves two main activities : · incorporation by designers not just of new Canadian safety regulations, but also of world trends in areas such as severe accidents and human factors, and · a two-year formal intensive review by the Canadian regulatory agency, the AECB (Atomic Energy Control Board) which will end in January 1997, with a first report mid-1996. In the severe accident area, CANDU 9 has improved the ability of the water-filled shield tank to act as a 'core catcher' for accidents such as a combined failure of the reactor cooling system piping, plus loss of emergency core cooling, plus loss of the moderator as a backup heat sink. A reserve water tank supplies sufficient water to the shield tank to remove decay heat and prevent melt-through. The licensing review documents encompass all of the Safety Design Guides, Design Requirements and flowsheets of the major systems, a preliminary Safety Analysis Report, a preliminary Probabilistic Safety Assessment, a Human Factors Engineering Programme Plan, the procedures for producing safety-critical software, and others. The licensing review scope includes comparison of the design and safety assessment results with current Canadian licensing requirements. The review will be complete once there is a finding that there are 'no

  7. Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

  8. 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

  9. Heavy water: a distinctive and essential component of CANDU

    International Nuclear Information System (INIS)

    The exceptional properties of heavy water as a neutron moderator provide one of the distinctive features of CANDU reactors. Although most of the chemical and physical properties of deuterium and protium (mass 1 hydrogen) are appreciably different, the low terrestrial abundance of deuterium makes the separation of heavy water a relatively costly process, and so of considerable importance to the CANDU system. World heavy-water supplies are currently provided by the Girdler-Sulphide process or processes based on ammonia-hydrogen exchange. Due to cost and hazard considerations, new processes will be required for the production of heavy water in and beyond the next decade. Through AECL's development and refinement of wetproofed catalysts for the exchange of hydrogen isotopes between water and hydrogen, a family of new processes is expected to be deployed. Two monothermal processes, CECE (Combined Electrolysis and Catalytic Exchange, using water-to-hydrogen conversion by electrolysis) and CIRCE (Combined Industrially Reformed hydrogen and Catalytic Exchange, based on steam reforming of hydrocarbons), are furthest advanced. Besides its use for heavy-water production, the CECE process is a highly effective technology for heavy-water upgrading and for tritium separation from heavy (or light) water. (author). 10 refs., 1 tab., 7 figs

  10. CANDU bundle junction. Misalignment probability and pressure-drop correlation

    International Nuclear Information System (INIS)

    The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs

  11. 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)

  12. A model for coolant void reactivity evaluation in assemblies of CANDU cells

    International Nuclear Information System (INIS)

    Coolant void reactivity is a very important safety parameter in CANDU reactor analysis. Here we evaluate the coolant void reactivity in a 2 x 2 heterogeneous assembly of CANDU cells using the code DRAGON. Since the current version of DRAGON can only treat the coolant void reactivity for a single CANDU cell, an approximate model for the geometry must be considered to perform assembly calculations in a 2 x 2 pattern. The model we propose consists of replacing the annular fuel pins by equivalent square fuel pins. The equivalence between annular fuel pins and square fuel pins is brought about by homogenizing the fuel plus its sheath and subsequently conserving the reaction rates between the two geometries using a SPH equivalence procedure. The approximate CANDU cells constructed using square pins were used to perform the transport calculations in 2 x 2 assembly patterns. In addition, the model was used to evaluate coolant void reactivity in 2 x 2 checkerboard voiding patterns. These calculations reflect more accurately the actual voiding situation being studied. This helps in assessing the effects due to the coupling of neutrons born in one cell to those born in the neighbouring cells

  13. CANDU-OCR power station options and costs

    International Nuclear Information System (INIS)

    This report updates and in some cases expands the technical and economic parameters presented originally in AECL-4441. 'Summary report on the design of a prototypical 500 MWe CANDU-OCR power station.' Updating is desirable owing to the increasing number of inquiries that have been received by Atomic Energy of Canada Ltd. from government agencies and the private sector. Each is exploring the available options in their continuing endeavour to provide sufficient and economical energy. The organic-cooled reactor (OCR) concept is particularly interesting to the oil industry because the high steam pressures it can develop allow it to be used for heavy oil extraction can also be used economically for other large thermal and electrical energy production requirements such as those encountered in district heating schemes, heavy water production and electricity production. The report describes a reference OCR-500 MWe reactor. It includes an overview of organic reactor experience and areas for further development based on 14 years of operating experience with the WR-1 reactor. A discussion of several variations on the reference design is given including estimates of costs for various reactor sizes, enrichments and operating functions. Costs are presented in a form which allow easy comparison with those of competing energy options. (auth)

  14. 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)

  15. 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

  16. 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)

  17. 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

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-09-15

    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.

  20. 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.

  1. 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

  2. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  3. Characterisation of Laboratory-produced CANDU(r)-like workplace neutron fields

    International Nuclear Information System (INIS)

    Two neutron fields were produced in the Neutron Irradiation Facility (NIF) at the Chalk River Laboratories of the Atomic Energy of Canada Ltd. by direction (d,D) neutrons from a 150 kV neutron generator through a specially designed moderator assembly. Bonner sphere and proton recoil spectrometry systems were used to characterise these fields to determine whether they were CANDU(r)-like, i.e. whether they resembled neutron fields found in workplaces around pressurised heavy-water moderated power reactors such as CANDU(r) reactors. Similarities were found between the distributions in energy of neutron fluence and ambient dose equivalent of the neutron fields produced in the NIF and those measured previously in power plants. In addition, there was agreement between theoretical (Monte Carlo) data and measured data, thereby validating continued use of Monte Carlo modelling for field characterisations in the NIF. The CANDU(r)-like fields add to the repertoire of neutron fields available in the NIF and are expected to be useful for evaluating neutron dosemeters. (author)

  4. Experience in the application of NUSS and Canadian quality assurance standards for overseas CANDU projects

    International Nuclear Information System (INIS)

    The Canadian QA standards - the CSA Z299 series for manufacture, which first appeared in 1975, and the CSA N286 series for all other phases of plant life which appeared in 1979, have been based on experience with the CANDU reactor program. The CSA Technical Committee responsible for issue and for updating the two series have a direct liaison with the IAEA Technical Review Committee for Quality Assurance. Ontario Hydro, which has a substantial commitment to nuclear power using CANDU reactors, has played a large part in the Canadian QA standards program. Atomic Energy of Canada Limited has also taken a major part in the development of CSA QA standards. As a main contractor the Company has supplied CANDU plants in Canada and to Argentina, South Korea and Romania. Because of the compatibility of the Canadian QA standards used, the Embalse plant in Argentina and the Wolsung 1 plant in Korea, are essentially in compliance with NUSS QA standards. The plant under construction at Cernavoda in Romania similarly follows Canadian QA standards

  5. Measuring velocity profile in scaled CANDU6 moderator tank using particle image velocimetry

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Bang, In Cheol [UNIST, Ulsan, (Korea, Republic of); Kim, Hyoung Tae; Cha, Jae Eun [KAISt, Daejeon (Korea, Republic of)

    2012-10-15

    The Calandria vessel, which is called as CANDU6 moderator tank, is the actual reactor core including fuel channels and moderator in PHWR. The understanding of circulation patterns in Calandria vessel is important because the cooling capability is related to the moderator tank. Therefore, measuring velocity and temperature patterns in the Calandria vessel is the key factor for the safety of CANDU reactor. Various experimental and numerical efforts to predict and analyze the thermal hydraulic characteristic of Calandria vessel have been made. Yoon et al. developed a CFD model for the CANDU6 moderator analyzing the velocity profile and the temperature distribution. Khartabil et al. did an experiment to measure 3 dimensional velocity and temperature distribution in moderator circulation tests at a 1/4 scaled down facility. Laser Doppler Anememetry (LDA) was used to detect the velocity profile and thermocouples detect the temperature distribution. Previous study presented the velocity profile in a 1/40 scaled-down Calandria vessel and compared with CFD analysis. In the present work, Particle Image Velocimetry (PIV) technique is used to obtain velocity profiles in a 1/8 scaled down Calandria vessel.

  6. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    International Nuclear Information System (INIS)

    To provide information on two-phase natural circulation in a CANDU-type coolant circuit a series of tests has been performed in the RD-12 loop at the Whiteshell Nuclear Research Establishment. RD-12 is a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behaviour is discussed briefly with reference to a simple stability model

  7. Uncertainty Analysis of the Potential Hazard of MCCI during Severe Accidents for the CANDU6 Plant

    Directory of Open Access Journals (Sweden)

    Sooyong Park

    2015-01-01

    Full Text Available This paper illustrates the application of a severe accident analysis computer program to the uncertainty analysis of molten corium-concrete interaction (MCCI phenomena in cases of severe accidents in CANDU6 type plant. The potential hazard of MCCI is a failure of the reactor building owing to the possibility of a calandria vault floor melt-through even though the containment filtered vent system is operated. Meanwhile, the MCCI still has large uncertainties in several phenomena such as a melt spreading area and the extent of water ingression into a continuous debris layer. The purpose of this study is to evaluate the MCCI in the calandria vault floor via an uncertainty analysis using the ISAAC program for the CANDU6.

  8. Assessment of databases and modeling capabilities for the CANDU 3 design

    International Nuclear Information System (INIS)

    The NRC staff has been conducting a preliminary review of the CANDU 3 (Canadian Deuterium Uranium Model 3) reactor design, a new heavy-water design developed by Atomic Energy of Canada Limited through its US affiliate, AECL Technologies. The review has been aimed at identifying key technical areas and policy issues that will have to be addressed for standard design certification. As part of the research program associated with the preliminary review, the NRC Office of Nuclear Regulatory Research (RES) has completed an assessment of databases and modeling capabilities that might be needed to support the CANDU 3 design. To ensure full coverage of the design, a detailed assessment methodology was developed by the RES staff and was implemented with help from research projects at three national laboratories. This report integrates and summarizes the database and modeling assessments, including major contributions from these laboratories. 395 refs

  9. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  10. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    1998-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)

  11. 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)

  12. LOCA power pulse analysis for CANDU-6 CANFLEX-RU core

    International Nuclear Information System (INIS)

    The power pulses following a large LOCA are analyzed for CANDU-6 reactor core fuelled with CANFLEX-RU fuel. The coupled simulations for reactor physics and channel thermal-hydraulic phenomena are done using RFSP and CATHENA codes. The 55% pump suction, 35% reactor inlet header and 100% reactor outlet header breaks are selected. The highest power pulse is predicted for 100% reactor outlet header break and it is higher than that for the standard 37-element natural fuel. However, the summation of initial stored energy and transient pulse energy of hottest pin has the minimum 17% margin to the fuel break up. Therefore, it is expected that there is no fuel breakup during the LOCA for CANFLEX-RU core

  13. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    Science.gov (United States)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  14. A review of CANDU plant lifetime management

    International Nuclear Information System (INIS)

    In recent years, plant lifetime management(PLIM) including life extension has become the focus of the nuclear industry worldwide due to a number of factors which have arisen over the past decade : new siting difficulties, imbalance of power supply and demand, and high construction costs. In order to solve the problems, the PLIM program is being developed for the purpose of life extension and improvement of plant availability and safety. This paper describes the current activities and prospects of AECL and CANDU utilities, the conceptional evaluation results for the degradation mechanisms, and PLIM regulatory aspects. In addition, this paper provides the applicability of CANDU PLIM to Wolsong Unit 1 which has been operated for 17 years

  15. Application of a Zircaloy/Steam Oxidation Model to a CFX-10 Code for its Validation against a CANDU Fuel Channel Experiment: CS28-2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2007-11-15

    Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production.

  16. Detector response in a CANDU low void reactivity core

    International Nuclear Information System (INIS)

    The response of the in-core flux detectors to the CANFLEX Low-Void-Reactivity Fuel (LVRF) [1] bundles for use in the CANDU reactor at Bruce nuclear generation station has been studied. The study was based on 2 detector types - platinum (Pt)-clad Inconel and pure Inconel detectors, and 2 fuel types - LVRF bundles and natural-uranium (NU) bundles. Both detectors show a decrease of thermal-neutron-flux to total-photon-flux ratio when NU fuel bundles are replaced by LVRF bundles in the reactor core (7% for Inconel and 9% for Pt-clad detectors). The ratio of the prompt component of the net electron current to the total net electron current (PFe) of the detectors however shows a different response. The use of LVRF bundles in place of NU fuel bundles in the reactor core did not change the PFe of the Pt-clad Inconel detector but increased the PFe of the pure Inconel detector by less than 2%. The study shows that the Inconel detector has a larger prompt-detector response than that of the platinum-clad detector; it reacts to the change of fluxes in the reactor core more readily. On the other hand, the Pt-clad detector is less sensitive to perturbations of the neutron-to-gamma ratio. Nevertheless the changes in an absolute sense are minimal; one does not anticipate a change of the flux-monitoring system if the NU fuel bundles are replaced with the CANFLEX LVRF bundles in the core of the Bruce nuclear generating station. (authors)

  17. Validation of computer codes used in safety analyses of CANDU power plants

    International Nuclear Information System (INIS)

    Since the 1960s, the CANDU industry has been developing and using scientific computer codes for designing and analysing CANDU power plants. In this endeavour, the industry has been following nuclear quality-assurance practices of the day, including verification and validation of design and analysis methodologies. These practices have resulted in a large body of experience and expertise in the development and application of computer codes and their associated documentation. Major computer codes used in safety analyses of operating plants and those under development have been, and continue to be subjected to rigorous processes of development and application. To provide a systematic framework for the validation work done to date and planned for the future, the industry has decided to adopt the methodology of validation matrices for computer-code validation, similar to that developed by the Nuclear Energy Agency of the Organization for Economic Co-operation and Development and focused on thermalhydraulic phenomena in Light Water Reactors (LWR). To manage the development of validation matrices for CANDU power plants and to engage experts who can work in parallel on several topics, the CANDU task has been divided into six scientific disciplines. Teams of specialists in each discipline are developing the matrices. A review of each matrix will show if there are gaps or insufficient data for validation purposes and will thus help to focus future research and development, if needed. Also, the industry is examining its suite of computer codes, and their specific, additional validation needs, if any, will follow from the work on the validation matrices. The team in System Thermalhydraulics is the furthest advanced, since it had the earliest start and the international precedent on LWRs, and has developed its validation matrix. The other teams are at various stages in this multiphase, multi-year program, and their progress to date is presented. (author)

  18. 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)

  19. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  20. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    International Nuclear Information System (INIS)

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  1. Opportunities for CANDU for the Alberta oil sands

    International Nuclear Information System (INIS)

    The Alberta oil sands bitumen deposits comprise of one of the largest sources hydrocarbon in the world, and have emerged as the fastest growing, soon to be dominant, source of crude oil in Canada. The oil industry has made great strides in improving the effectiveness of gathering this resource. In particular, alternatives to open-pit mining have been developed which enable in-site recovery of underground deposits with a minimum of environmental disruption. The main challenge that remains is the large quantity of energy needed in the process of extracting the oil and upgrading it to commercial levels. For a typical in-situ extraction project, about 18% of the energy content of the oil produced is used up in the extraction process, while a further 5% is used in generating hydrogen to upgrade the bitumen to synthetic crude oil. Looking ahead, even as improvements in energy use efficiency, (and hydrocarbon use efficiency) counterbalance the increases in hydrocarbon demand from economic growth (particularly in the developing world), Canada and Alberta recognize that the oil sands resource will be needed, and both support the development of this resource in an environmentally responsible way. The large energy requirement for the oil sands extraction process represents a challenge with regard to both environmental impact and security of supply. The use of natural gas, the current energy supply, has impacts in terms of air quality (via NOX and other emissions) and also represents a large greenhouse gas emissions component. As the oil sands industry expands, the availability of natural gas also becomes a concern, as does price and price stability. With this background, the opportunity for nuclear reactors to provide an economical, reliable, virtually zero-emission source of energy for the oil sands becomes very important. Over the last few years, developments in oil sands extraction technology, and developments in CANDU technology through the Advanced CANDU Reactor, (ACR

  2. The CANDU contribution to environmentally friendly energy production

    International Nuclear Information System (INIS)

    National prosperity is based on the availability of affordable, energy supply. However, this need is tempered by a complementary desire that the energy production and utilization will not have a major impact on the environment. The CANDU energy system, including a next generation of CANDU designs, is a major primary energy supply option that can be an important part of an energy mix to meet Canadian needs. CANDU nuclear power plants produce energy in the form of medium pressure steam. The advanced version of the CANDU design can be delivered in unit modules ranging from 400 to 1200 MWe. This Next Generation of CANDU designs features lower cost, coupled with robust safety margins. Normally this steam is used to drive a turbine and produce electricity. However, a fraction of this steam (large or small) may alternatively be used as process steam for industrial consumption. Options for such steam utilization include seawater desalination, oil sands extraction and heating. The electricity may be delivered to an electrical grid or alternatively used to produce quantities of hydrogen. Hydrogen is an ideal clean transportation fuel because its use only produces water. Thus, a combination of CANDU generated electricity and hydrogen distribution for vehicles is an available, cost-effective route to dramatically reduce emissions from the transportation sector. The CANDU energy system contributes to environmental protection and the prevention of climate change because of its very low emission. The CANDU energy system does not produce any NOx, SOx or greenhouse gas (notably CO2) emissions during operation. In addition, the CANDU system operates on a fully closed cycle with all wastes and emissions fully monitored, controlled and managed throughout the entire life cycle of the plant. The CANDU energy system is an environmentally friendly and flexible energy source. It can be an effective component of a total energy supply package, consistent with Canadian and global climate

  3. 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

  4. A criterion for limiting design mitigation features for CANDU 3U

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) sponsored a peer review group composed of recognized national and international experts in reactor safety to review NUREG-1150. The conclusions and recommendations were published in NUREG-1420, dated August 1990. The Kouts Peer review Committee recommended that a residual boundary or event families and plant damage states with accident scenario frequencies below about 10minus7/yr should be neglected in probabilistic risk assessment. This concept should have an impact on the NRC review of the CANDU 3U for a final design approval and standard design certification

  5. Multi-step approach to Code-coupling for progression induced severe accidents in CANDU NPPs (MACPISA-CANDU)

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, D.J.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada); Giannotti, W.; D' Auria, F. [Univ. of Pisa, Dept. of Mechanical, Nuclear and Production Engineering, Pisa (Italy)

    2009-07-01

    This paper reviews the progression of severe accidents, describes computer codes currently employed for analysis of severe accidents and outlines a new methodology to modelling the progression of severe accidents in CANDU nuclear power plants (NPPs) called the Multi-step Approach to Code-coupling for Progression Induced Severe Accidents in CANDU NPPs (MACPISA-CANDU). The MACPISA-CANDU methodology was used to couple the U.S. NRC codes SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4) and MELCOR (1.8.5) in order to model a small break loss of coolant accident with loss of emergency coolant injection (SBLOCA-LOECI) under natural circulation in a CANDU 6 NPP. Using this model it was shown that the sheath temperature did not exceed the zirconium melting temperature of 2098 K and hence the progression of the severe accident was terminated as expected. (author)

  6. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R. [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  7. The AECL reactor development programme

    International Nuclear Information System (INIS)

    The modem CANDU-PHWR power reactor is the result of more than 50 years of evolutionary design development in Canada. It is one of only three commercially successful designs in the world to this date. The basis for future development is the CANDU 6 and CANDU 9 models. Four of the first type are operating and four more will go an line before the end of this decade. The CANDU 9 is a modernized single-unit version of the twelve large multi-unit plants operated by Ontario Hydro. All of these plants use proven technology which resulted from research, development, design construction, and operating experience over the past 25 years. Looking forward another 25 years, AECL plans to retain all of the essential features that distinguish today's CANDU reactors (heavy water moderation, on-power fuelling simple bundle design, horizontal fuel channels, etc.). The end product of the planned 25-year development program is more than a specific design - it is a concept which embodies advanced features expected from ongoing R and D programs. To carry out the evolutionary work we have selected seven main areas for development: Safety Technology, Fuel and Fuel Cycles, Fuel Channels, Systems and Components, Heavy Water and Tritium Information Technology, and Construction. There are three strategic measures of success for each of these work areas: improved economics, advanced fuel cycle utilization, and enhanced safety/plant robustness. The paper describes these work programs and the overall goals of each of them. (author)

  8. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    OpenAIRE

    Kim, Hyoung Tae; Chang, Se-Myong; Shin, Jong-Hyeon; Kim, Yong Gwon

    2016-01-01

    The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchm...

  9. 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)

  10. A model for analyzing CANDU-6 SDS No.2 poison injection system

    International Nuclear Information System (INIS)

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called. ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For the validation, an attempt was made to validate this model against a poison injection experiment performed at Bhabha Atomic Research Center (BARC) of India. The interim progress will be presented and the validation analysis result is discussed

  11. Estimation of Aging Effects on LOHS for CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

  12. Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.J.; Kotzer, T.G.; Chant, L.A

    2001-06-01

    In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

  13. Experience teaching CD-ROM-based course on CANDU nuclear-power-plant systems and operation

    International Nuclear Information System (INIS)

    This paper presents personal experience garnered from teaching a CD-ROM-based course on CANDU Power-Plant Systems and Operation. This course was originally developed by Prof. G.T. Bereznai as research in distance-learning techniques when he was directing the Thai-Canadian Human Resources Development Project at Chulalongkorn University in Bangkok. The course has been offered in a number of universities, including McMaster University and the University of Ontario Institute of Technology. All the course material, including lectures, assignments, and a simulator, is provided on a CD-ROM. Lectures include a spoken soundtrack covering the material. The class often includes both undergraduate and graduate students. I found that most students appreciate having the material on electronic format, which they can view and review at will and on their own time. Students find this course quite intensive - it covers all major systems in the CANDU reactor and power plant in detail. A very important component of the course is the simulator, which teaches students how systems operate in normal operation, in power manoeuvres, and during process-system malfunctions. Effort in absorbing the material and performing assignments can often exceed 10 hours per week. Some of the simulator assignments involve tricky manoeuvres, requiring several tries to achieve the expected result. Some assignments may take several hours, especially if the manoeuvres requiring repetition take 30 minutes or more in real time. I found that some instruction in the basic theory of reactor physics and systems is appreciated by students. A few possible enhancements to the simulator model were identified. Graduate students taking the course are required to do an additional project; I assigned an investigation of the effects of xenon-concentration changes during 1 week of load cycling. In summary, this course provides to students the opportunity to learn a great deal about the workings of CANDU-plant systems. (author)

  14. Application of microprocessor based technology in CANDU Stations

    International Nuclear Information System (INIS)

    In early 80s, absence of more efficient, cost effective, reliable control hardware was experienced by the nuclear industry. Usage of microprocessor based technology was not prevalent in the nuclear industry due to lack of experience and infancy of the new technology at that time. In 1985, microprocessor based technology, 'Chameleon micro DCI Controllers' manufactured by Fischer and Porter was the 1st application of such technology in any safety related applications in a CANDU Station. The application was for In-Core LOCA logic in Emergency Coolant Injection System (ECIS) modification in Pickering A Nuclear Station. This was followed by use of the same hardware for modification of 'Dump-Arrest' logic for shut-down system in Pickering A in 1986. Other applications were 'Heat Transport Pump Trip Logic' in Bruce A in 1989. The hardware performed very well and far exceeded its predicted availability during the last 22 years of operation. In early 90's Power House Emergency Venting system used 22 micro DCI controllers for Pickering A and B Stations. The use of microprocessor based technology helped meeting the timing requirements of initiating Powerhouse Emergency Venting within 3 seconds of a steam break accident in reactor auxiliary bay. Development of digital trip meter was undertaken in 1991 to solve human-factors related issue of interpretation of HT system saturation trip margin as the analog indicator lacked accuracy and resolution. This was a significant milestone in use of digital technology in safety related applications in any CANDU Station. The process offered an opportunity of trial use of software standards developed by OPG and AECL for category 1 safety critical software systems. This application generated significant learning experience for OPG (then Ontario Hydro) and the industry. OPG was ahead of other nuclear utilities in implementation of microprocessor based technology in early 80s and 90s. OPG initiatives generated lot of design, maintenance and

  15. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  16. Development and validation of a model for CANDU-6 SDS2 poison injection analysis

    International Nuclear Information System (INIS)

    In CANDU-6 reactor there are two independent reactor shutdown systems. The shutdown system no. 2(SDS2) injects the liquid poison into the moderator tank by high pressure via small holes on the 6 nozzle pipes and stops the nuclear chain reaction. To ensure the safe shutdown of a reactor loaded with either DUPIC or SEU fuels it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. The motivation for this work arose from the fact that the computer code package for performing this task is not transfered to Korea yet. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a commercial CFD code developed by AEA technology, to simulate the formation of the poison jet curtain inside the moderator tank. For the validation, a simulation for a generic CANDU-6 SDS2 design poison jet growth experiment was made to evaluate this model's capability against experiment. As no concentration field was measured and only the growth of the poison jet height was obtained by high speed camera, the validation was limited as such. The result showed that if one assume the jet front corresponds to 200 ppm of the poison the model succeed to

  17. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2016-01-01

    Full Text Available The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor, has been modeled in multidimension for the computation based on CFD (computational fluid dynamics technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

  18. The status of improved pressurized heavy water reactor development - A new option for PHWR -

    International Nuclear Information System (INIS)

    Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

  19. 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

  20. 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)

  1. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  2. CANDU energy for steam assisted gravity drainage

    International Nuclear Information System (INIS)

    Traditional open-pit mining has been used by industry for many years to remove oil sands from shallow deposits. To increase production capacity, the industry is looking for new technology to exploit bitumen from deep deposits. Among them, SAGD (Steam-Assisted Gravity Drainage) appears to be the most promising approach. It uses steam to remove bitumen from underground reservoirs. Recently, the SAGD recovery process has been put into commercial operation by major oil companies.Atomic Energy Canada Limited has assessed the use of the ACR-1000 as a source of heat and electricity for oil sand extraction and processing. The ACR-1000 design is an evolutionary development of the familiar CANDU technology, adding innovations to enhance economics, operations, and safety margins. The net electrical output from a standard ACR-1000 will be close to 1100 MWe, depending on local cooling water temperature

  3. 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)

  4. GOTHIC analysis of post-accident hydrogen mixing behaviour in CANDU fuelling machine vault

    International Nuclear Information System (INIS)

    The GOTHIC code was used to assess the post-accident hydrogen gas mixing patterns in a CANDU reactor containment and demonstrate the acceptability of Ontario Hydro Nuclear's hydrogen mitigation methods. The fuelling machine vault, being a small volume room containing major reactor piping, is the room of most concern with respect to hydrogen concentrations. Detailed three dimensional modelling of the gas mixing patterns in the fuelling machine vault was completed. Results showed that, even without forced air circulation, there is enough dispersion of hydrogen to other parts of containment to preclude the build-up of sensitive mixtures in the vault. For a brief time during the peak period of hydrogen release, hydrogen concentrations rise to close to the lower ignition limit in a small portion of the vault, but these hydrogen-steam-air mixtures are considered acceptable. Natural mixing alone is sufficient to preclude damaging hydrogen burns. (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. 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)

  7. Subchannel analysis for thorium-based advanced CANDU reactor

    International Nuclear Information System (INIS)

    This paper presents the subchannel thermal hydraulic analysis in different thorium loading modes, different power and different lifetime using ASSERT-PVV3R1, a subchannel computer code developed by ACEL, to determine the best loading mode from the point of view of subchannel thermal hydraulics. According to the results for subchannel the mass flux, void fraction and outset of dry power, a best loading mode is established from the point of view of being calm and security to power, i.e., the loading mode of ThO2 and SEU mixing in pellets is obviously better than the loading mode in which the only ThO2 is loaded in the internal eight rods. (authors)

  8. Studies on acoustic modelling techniques for CANDU reactors

    International Nuclear Information System (INIS)

    This paper reviews the current technology being used to predict acoustic resonance in fluid-filled piping systems. The paper also reports on the analysis of a simple benchmark experiment that yielded some valuable insights and understanding into acoustic damping. A volumetric drag formula for the ABAQUS code is presented. Its application in experiments has yielded better results than previously obtained using a constant volumetric drag. (author). 24 refs., 1 tab., 12 figs

  9. Evolution of the CANDU ICS-90+ control room design

    International Nuclear Information System (INIS)

    The design of the CANDU Control Room and the associated design process has evolved considerably over several generations of plants, from the first commercial scale demonstration CANDU at Douglas Point through to the large scale CANDUs at Darlington, and beyond, for the next generation of CANDU plant, ICS-90+, represented by new designs like CANDU 3. In the early plants, the control room configuration was based on designers' projections of control interface requirements. With succeeding generations, of designs, there has been an evolution towards: increasing attention to formal requirements definition, incorporation into the Human Machine Interface (HMI) of a larger base of operational experience, more systematic consideration of Human Factors (HF) aspects of the design and the application of a more powerful computer based HMI. For the newest plant, the CANDU 3, a Human Factors Engineering Program Plan (HFEPP) defines the overall HF engineering process, the associated requirements and HF engineering standards to be followed in each stage, and for all HMI aspects of the control room and plant design. The CANDU 3 control room also incorporates several new design innovations that will facilitate operating crew performance improvements. These are based on past experience with operating CANDU plants, incorporated with the use of formal design and validation methods plus results from Canadian research program to support control centre design and operation. For example, there are design improvements to facilitate: operator tracking of plant state, problem solving, alarm filtering, annunciation system interrogation, special safety system testing features, etc. The present paper will expand and elaborate on each of the above topics. (author). 7 refs, 2 figs, 1 tab

  10. 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

  11. COMET solutions to whole core CANDU-6 benchmark problems

    International Nuclear Information System (INIS)

    In this paper, the coarse mesh transport code COMET is used to solve CANDU-6 benchmark problems in two and three dimensional geometry. These problems are representative of a simplified quarter core reactor model. The COMET solutions, the core eigenvalue and the fuel pin fission density distribution, are compared to those from the Monte Carlo code MCNP using two-group cross sections. COMET decomposes the core volume into a set of non-overlapping sub-volumes (coarse meshes) and uses pre-computed heterogeneous response functions that are constructed using Legendre polynomials as boundary conditions to generate a user selected whole core solution (e.g., the core eigenvalue and fuel pin fission density distribution). These response functions are pre-computed by performing fixed source calculations with a modified version of MCNP in only the unique coarse meshes in the core. Reference solutions are calculated by MCNP5 with a two-group energy library generated with the HELIOS lattice code. In the 2-D problem, the angular current on the coarse mesh interfaces in COMET is expanded to 2. order in both spatial and angular variables. The COMET eigenvalue error is 0.09%. The corresponding average error in the fission density over all 3515 fuel pins is 0.5%. The maximum error observed is 2.0%. For the 3-D case, with 4. order expansion in space and azimuthal angle and 2. order expansion in the cosine of the polar angle, the eigenvalue differs from the reference solution by 0.05%. The average fission density error over the 42180 fuel pins is 0.7% with a maximum error of 3.3%. (authors)

  12. 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

  13. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  14. A New Fuel Design for Two Different HW Type Reactors

    Directory of Open Access Journals (Sweden)

    Daniel O. Brasnarof

    2011-01-01

    Full Text Available A new fuel element (called CARA designed for two different heavy water reactors (HWRs is presented. CARA could match fuel requirements of both (one CANDU and one unique Siemens's design Argentine HW reactors. It keeps the heavier fuel mass density and hydraulic flow restriction in both reactors together with improving both thermomechanic and thermalhydraulic, safety margins of present fuels. In addition, the CARA design could be considered as another design line for the next generation of CANDU fuels intended for higher burnup.

  15. 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

  16. Verification of revised EOP using the CATHENA code for Wolsong NPP- Candu 6

    International Nuclear Information System (INIS)

    EOP (Emergency Operating Procedure) is the operating procedure document which describes the operator's action needed for the maintenance of nuclear power plant (NPP) safety when a certain accident takes place. EOPs of Wolsong NPP Units 2, 3, 4 (Candu 6) have been revised aiming at the optimization of procedure to recover the NPP from an accident and the minimization of human errors. The revised overall EOPs are composed of 2 EOPs describing the accident diagnosis and recovery of critical safety parameters and 13 EOPs describing the operator actions needed for the individual accidents which include LOCA and MSLB and so on. The purpose of the present study is to verify the revised EOPs if the procedures and operator actions described in each EOP are fully appropriate and effective for the safe shutdown of NPP leading to temperature low enough to start the long term cooling using the shutdown cooling system and for the maintenance of fuel and pressure boundary integrity. The verification was carried out with the deterministic safety analysis methodology using the CATHENA computer program which is a Thermalhydraulic numerical code used in the safety analysis of Candu 6 reactors. CATHENA can simulate the various thermalhydraulic behaviors which are expected to happen during each accident and by various operator actions. From the verification results it can be assured that the revised individual EOPs have the enough appropriateness with respect to the engineering aspects and effectiveness in view of fuel and pressure boundary integrity. (authors)

  17. 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)

  18. Evolution of the CANDU control centre retrofit and new stations

    International Nuclear Information System (INIS)

    Significant event data from operating nuclear plants in many countries consistently indicates human errors are the root cause for 40-60% of operating station significant events. Because so much information is already in digital form, opportunities exist to improve the CANDU control centre with retrofits that exploit this information. These opportunities are enhanced because of rapid technological development in computers and electronics, coupled with significant progress in the behavioural sciences that greatly increases our knowledge of the cognitive strengths and weaknesses of human beings. CANDU control rooms are undergoing retrofits and for future CANDU stations, a new concept of the control centre is emerging. The objective is to significantly reduce the incidence of human error, reduce operations and maintenance costs and improve both reliability and safety

  19. Advancing CANDU experience to the world steam generator market

    International Nuclear Information System (INIS)

    Tube degradation in certain recirculating nuclear steam generators has provided a market for steam generator replacement. Prior to this need, B and W supplied over 200 steam generators for CANDU nuclear plants. With this experience, and implementing extensive research and development improvements in material selection, design enhancements, and new manufacturing and analytical methods, B and W has supplied or secured orders for the replacement of 26 steam generators. Along with plans for new replacement orders, B and W will continue to supply steam generators for future CANDU plants. This paper will review the progression of B and W's CANDU experience to meet the replacement steam generator market, and examine the continuous improvements required for today's increasingly demanding nuclear specifications. (author). 1 tab., 4 figs

  20. 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

  1. Replacement steam generators for pressurized water reactors

    International Nuclear Information System (INIS)

    Babcock and Wilcox Canada has developed an Advanced Series steam generator for PWR Systems. This design incorporates all of the features that have contributed to the successful CANDU steam generator performance. This paper presents an overview of the design features and how the overall design relates to the requirements of a PWR reactor system

  2. Evolution of the CANDU control centre design process

    International Nuclear Information System (INIS)

    The design of the CANDU NPP control centre and the associated control centre design process has evolved considerably over several generations of plants, from Douglas Point through Darlington, and beyond, to new designs like CANDU 3. In the early plants, the control centre configuration had to be based on designers' projections of control interface requirements. With succeeding generations of designs, along with the introduction of advancing computer control technology, a larger based of operational experience has been factored into the control interface design, and increasing attention has been given to more formal requirements definition, and more systematic consideration of human factors aspects of the design

  3. 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)

  4. The Chernobyl-4 Reactor and the possible causes of the accident

    International Nuclear Information System (INIS)

    A description and information about the Chernobyl nuclear reactor is given. Some comparison elements between the RBMK reactor type and GCR, CANDU, SGHWR and Hanford N reactor types are presented. A scenario of the possible causes of the accident is discussed. (A.F.)

  5. 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)

  6. 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

  7. 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.

  8. 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)

  9. Candu plant life management - safeguarding the investment

    International Nuclear Information System (INIS)

    With a large number of the CANDU NPPs getting to the mid-point of their design life, a number of programs have been initiated by the Utilities and AECL to proactively manage plant aging, hence ensuring a safe and reliable operation for the remaining design life. An integrated aging management program is being formulated in response to the regulatory authority the Atomic Energy Control Board (AECB). This integrated program will take into account safety, performance and economic requirements. Elements of this program cover a variety of areas; Phase 1 covers life assessment studies of the critical systems, structures and components (SSCs) including a methodology for defining the critical SSCs. Other programs within Phase 2 address the preparation of a unified set of industry guidelines covering maintenance and inspection requirements during early, middle and late years of plant operation. A 'technology watch' program has also been initiated. The objective of this program is to identify far in advance potential aging phenomena and failure modes that could affect plant performance along with the inspection and maintenance required to monitor and mitigate such aging effects. The importance of maintaining the plant within a well-established 'licensing basis' envelope is also discussed. Plant licensing is carried out for an initial set of conditions and equipment status which may vary during plant life. Canadian utilities are required by the licensing authority, the AECB, to assess the impact of aging on the licensing case for each plant in operation. So far, the main focus of these assessments has been the impact of aging on key parameters such as fuel thermal margin and containment leak rate. Ongoing activities covering characterization of aging, maintenance, monitoring, and appropriate refurbishments are examined. Emphasis of R and D performed to characterize degradation mechanisms, support inspection and fitness for service guidelines are also briefly outlined. (author)

  10. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  11. The Candu system - The way for nuclear autonomy

    International Nuclear Information System (INIS)

    The experience acquired by Canada during the development of Candu System is presented. Some basic foundations of technology transfer are defined and, the conditions of canadian nuclear industry to provide developing countries, technical assistence for acquisition of nuclear energy autonomy, are analysed. (M.C.K.)

  12. A short history of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    This paper provides a short historical summary of the evolution of the CANDU nuclear power system with emphasis on the roles played by Ontario Hydro and private sector companies in Ontario in collaboration with Atomic Energy of Canada Limited (AECL). (author). 1 fig., 61 refs

  13. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    International Nuclear Information System (INIS)

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  14. A proposed structural, risk-informed approach to the periodicity of CANDU-6 nuclear containment integrated leak rate testing

    Energy Technology Data Exchange (ETDEWEB)

    Saliba, N. [McGill Univ., Dept. of Civil Engineering and Applied Mechanics, Montreal, Quebec (Canada); Komljenovic, D. [Hydro-Quebec, Gentilly-2 Nuclear Power Plant, Becancour, Quebec (Canada); Chouinard, L. [McGill Univ., Dept. of Civil Engineering and Applied Mechanics, Montreal, Quebec (Canada); Vaillancourt, R.; Chretien, G. [Hydro-Quebec, Gentilly-2 Nuclear Power Plant, Becancour, Quebec (Canada); Gocevski, V. [Hydro-Quebec Equipements, Montreal, Quebec (Canada)

    2010-07-01

    As ultimate lines of defense against leakage of large amounts of radioactive material to the environment in case of major reactor accidents, containments have been monitored through well designed periodic tests to ensure their proper performance. Regulatory organizations have imposed types and frequencies of containment tests based on highly-conservative deterministic approaches, and judgments of knowledgeable experts. Recent developments in the perception and methods of risk evaluation have been applied to rationalize the leakage-rate testing frequencies while maintaining risks within acceptable levels, preserving the integrity of containments, and respecting the defense-in-depth philosophy. The objective of this paper is to introduce a proposed risk-informed decision making framework on the periodicity of nuclear containment ILRTs for CANDU-6 nuclear power plants based on five main decision criteria, namely: 1) the containment structural integrity; 2) inputs from PSA Level-2; 3) the requirements of deterministic safety analyses and defense-in-depth concepts; 4- the obligations under regulatory and standard requirements; and 5) the return of experience from nuclear containments historic performance. The concepts of dormant reliability and structural fragility will guide the assessment of the containment structural integrity, within the general context of a global containment life cycle management program. This study is oriented towards the requirements of CANDU-6 reactors, in general, and Hydro-Quebec's Gentilly-2 nuclear power plant, in particular. The present article is the first part in a series of papers that will comprehensively detail the proposed research. (author)

  15. Numerical Analysis of CANDU-6 Moderator System Using OpenFOAM

    International Nuclear Information System (INIS)

    On the moderator of CANDU-6 reactor, thanks to the rapid development of CFD (Computational Fluid Dynamics), the 1-D model code can be substituted to the 3-D simulation codes. The three-dimensional computation becomes not so expensive that now we can enjoy the benefit of innovation about CFD technology. In this study, we have modeled the Calandria tank system as simplified models preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors. The use of OpenFOAM is a very important point for the present study. The OpenFOAM is based on the object-oriented programming using C++ language. The solvers and libraries of physical properties, for example, are declared as classes to produce a new code with the reproduction from the existing classes. As this code is fully open to the public, the development of CFD code with OpenFOAM should be very prospective to the future design of system codes, not just restricted in the area of hydro-thermal system concerning atomic reactors

  16. Improved operability of the CANDU 9 control centre

    Energy Technology Data Exchange (ETDEWEB)

    Macbeth, M. J.; Webster, A. [Atomic Energy of Canada Limited, Saskatoon (Canada)

    1996-04-15

    The next generation CANDU nuclear power plant being designed by AECL is the 900 MWe class CANDU 9 station. It is based upon the Darlington CANDU station design which is among the world leaders in capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This Control Centre design includes the proven functionality of existing CANDU control centres (including the Wolsong 2,3, and 4 control centre improvements, such as the Emergency Core Cooling panels), the characteristics identified by systematic design with human factors analysis of operations requirements and the advanced features needed to improve station operability which is made possible by the application of new technology. The CANDU 9 Control Centre provides plant staff with an improved operability capability due to the combination of proveness, systematic design with human factors engineering and enhanced operating features. Significant features which contribute to this improved operability include: {center_dot} Standard NSP, BOP and F/H panels with controls and indicators integrated by a standard display/presentation philosophy. {center_dot} Common plant parameter signal database for extensive monitoring, checking, display and annunciation. {center_dot} Powerful annunciation system allowing alarm filtering, prioritizing and interrogation to enhance staff recognition of events, plant state and required corrective procedural actions. {center_dot} The use of an overview display to present immediate and uncomplicated plant status information to facilitate operator awareness of unit status in a highly readable and recognizable format. {center_dot} Extensive cross checking of similar process parameters amongst themselves, with the counterpart safety system parameters and as well as with 'signature' values obtained from known steady state conditions. {center_dot} Powerful calculation capabilities, using the plant wide database, providing immediate recognizable and readable and

  17. Development of ANC-type empirical two-phase pump model for full size CANDU primary heat transport pump

    International Nuclear Information System (INIS)

    The development of an ANC-type empirical two-phase pump model for CANDU (Canadian Deuterium) reactor primary heat transport pumps is described in the present paper. The model was developed based on Ontario Hydro Technologies' full scale Darlington pump first quadrant test data. The functional form of the ANC model which is widely used was chosen to facilitate the implementation of the model into existing computer codes. The work is part of a bigger test program with the aims: (1) to produce high quality pump performance data under off-normal operating conditions using both full-size and model scale pumps; (2) to advance our basic understanding of the dominant mechanisms affecting pump performance based on more detailed local measurements; and (3) to develop a 'best-estimate' or improved pump model for use in reactor licensing and safety analyses. (author)

  18. Compatibility analysis of DUPIC fuel (Part II) - Reactor physics design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Choi, Hang Bok; Rhee, Bo Wook; Roh, Gyu Hong; Kim, Do Hun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The compatibility analysis of the DUPIC fuel in a CANDU reactor has been assessed. This study includes the fuel composition adjustment, comparison of lattice properties, performance analysis of reactivity devices, determination of regional over-power (ROP) trip setpoint, and uncertainty estimation of core performance parameters. For the DUPIC fuel composition adjustment, three options have been proposed, which can produce uniform neutronic characteristics of the DUPIC fuel. The lattice analysis has shown that the characteristics of the DUPIC fuel is compatible with those of natural uranium fuel. The reactivity devices of the CANDU-6 reactor maintain their functional requirements even for the DUPIC fuel system. The ROP analysis has shown that the trip setpoint is not sacrificed for the DUPIC fuel system owing to the power shape that enhances more thermal margin. The uncertainty analysis of the core performance parameter has shown that the uncertainty associated with the fuel composition variation is reduced appreciably, which is primarily due to the fuel composition adjustment and secondly the on-power refueling feature and spatial control function of the CANDU reactor. The reactor physics calculation has also shown that it is feasible to use spent PWR fuel directly in CANDU reactors without deteriorating the CANDU-6 core physics design requirements. 29 refs., 67 figs., 60 tabs. (Author)

  19. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  20. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  1. 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

  2. Current safety issues of CANDU licensing

    International Nuclear Information System (INIS)

    As requested by Korea Institute of Nuclear Safety(KINS), the status of five generic licensing issues has been examined and their potential impact on a new plant that would be constructed in Canada has been evaluated. The results and conclusions of this evaluation are summarized as follows: steam explosion in calandria, hydrogen explosion in containment, use of PSA in reactor licensing, human factors, safety critical software

  3. 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

  4. Alternatives to proposed replacement production reactors

    International Nuclear Information System (INIS)

    To insure adequate supplies of plutonium and tritium for defense purposes, an independent evaluation was made by Los Alamos National Laboratory of the numerous alternatives to the proposed replacement production reactors (RPR). This effort concentrated on the defense fuel cycle operation and its technical implications in identifying the principal alternatives for the 1990s. The primary options were identified as (1) existing commercial reactors, (2) existing and planned government-owned facilities (not now used for defense materials production), and (3) other RPRs (not yet proposed) such as CANDU or CANDU-type heavy-water reactors (HWR) for both plutonium and tritium production. The evaluation considered features and differences of various options that could influence choice of RPR alternatives. Barring a change in the US approach to civilian and defense fuel cycles and precluding existing commercial reactors at government-owned sites, the most significant alternatives were identified as a CANDU-type HWR at Savannah River Plant (SRP) site or the Three Mile Island commercial reactor with reprocessing capability at Barnwell Nuclear Fuel Plant and at SRP

  5. Streamlined Reliability Centred Maintenance (RCM) application to CANDU 6 stations

    International Nuclear Information System (INIS)

    Over the past five years, Atomic Energy of Canada Ltd. (AECL) has been working with CANDU utilities on Plant Life Management (PLiM) programs that will see existing CANDU plants through their design life and beyond. As part of this initiative, AECL and New Brunswick Power have partnered to develop a Systematic Approach to Maintenance program applied to selected critical plant systems. This paper will describe how streamlined Reliability Centred Maintenance (RCM) techniques have been applied on systems at the Point Lepreau Generating Station to provide a sound documented basis for maintenance strategies. These strategies have emphasised a hierarchy of condition based maintenance, time based maintenance and, where appropriate, corrective maintenance. The major steps in the process are described. The clear benefits of focusing maintenance in areas where it is needed and effective from the context of impact on system function requirements are described. The basis of the maintenance program is fully documented at the individual task level. The results of the program are also used to define maintenance strategies for future CANDU power plants. (author)

  6. Analyse du transfert de chaleur et de la perte de pression pour des ecoulements supercritiques dans le reacteur CANDU-SCWR

    Science.gov (United States)

    Zoghlami, Sarra

    The supercritical water reactor is one of the six concepts of generation IV nuclear reactors that has been selected by the International Generation IV Forum (GIF). Canada has chosen to conduct advanced research on this type of reactor. For the design and safety analysis of the reactor concept, the development of numerical simulation codes is needed. The ARTHUR code is a thermal-hydraulic computer code developed by Fassi-Fehri (2008), at the Ecole Polytechnique de Montreal, to analyse the CANDU-6 reactor. The purpose of this project is to modify this numerical code so that it can be used to treat the CANDU-SCWR. To calculate the coolant thermal-hydraulics properties in the fuel channel of a CANDU-SCWR, it was assumed that the water flows under supercritical conditions is a one-phase flow. Thus within this code, we developed the conservation equations for one-phase flow. Hydraulic resistance and heat transfer at supercritical pressure are two important aspects to be considered in the modeling of a fuel channel in a nuclear reactor. To choose the accurate correlation to predict the pressure friction factor, we compared numerical calculations, using different correlations found in literature, to experimental data. We concluded that the Garimella (2008) correlation is the most consistent, to be incorporated in the ARTHUR &barbelow;SCWR code. We proved that the choice of the friction factor correlation affects slightly the distribution of thermal-hydraulic properties in the fuel channel. Under supercritical conditions, water thermal-physical properties are characterized by significant variations in the pseudo-critical region. This behavior influences the forced convection heat transfer phenomena. To choose the adequate correlation to calculate the forced convection heat transfer coefficient, we compared numerical results to experimental data, and we found that the standard deviation given by Mokry et al. (2010) correlation is the lowest. In order to model the fuel

  7. 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.

  8. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  9. Automatic control of the D20 vapor recovery dryers at Cernavoda NPP U2 (CANDU Project 82)

    International Nuclear Information System (INIS)

    The 'D20 vapor Recovery System' is built in the frame of CANDU NPP Cernavoda, Romania, in order to preserve the atmosphere dry inside the Reactor Building. The system includes ten dryers, four tanks and three pumps grouped in thirteen equipment units with cyclic operation. When desiccant bed of the dryers is full of humidity the operating regime is switched automatically from absorption to regeneration. During regeneration, the air from blower discharge is passed through electric heaters and desiccant bed to condenser where the D20 vapor is recovered, and again to fan inlet. The dryer regeneration is completed when its desiccant bed is hot. Also, transferring the contents of the tanks is a batch processing operation. Considering the Cernavoda U1 operating experience, the CANDU project 82, for Cernavoda NPP U2, is now under an improving process, which implies an Automatic Control for the 'D20 vapor Recovery System'. This Automatic Control, performed by Industrial Computers and Data Acquisition System, is able automatically to detect the completion of the equipment unit phases, to switch the equipment unit to appropriate operation mode, to check the proper operating mode and to trip the equipment unit if necessary. This paper intends to present the problems surmounted during the technical specification issuing for designing process. (author)

  10. Design requirements, criteria and methods for seismic qualification of CANDU power plants

    International Nuclear Information System (INIS)

    This report describes the requirements and criteria for the seismic design and qualification of systems and equipment in CANDU nuclear power plants. Acceptable methods and techniques for seismic qualification of CANDU nuclear power plants to mitigate the effects or the consequences of earthquakes are also described. (auth)

  11. Power mapping in a Canada deuterium uranium reactor using Kalman filtering technique

    International Nuclear Information System (INIS)

    For a Canada Deuterium Uranium 600 MWe (CANDU 6) reactor, a new power mapping method has been developed by using detector readings as boundary conditions. In this study, the measured detector readings are combined with the diffusion theory with the Kalman filtering (DIKAL) method. The measured detector readings are transformed into the measured mesh flux through appropriate approximation. And, the difference between calculated and measured mesh flux is filtered out by Kalman filtering technique. Then, the measured mesh fluxes are used as an internal boundary condition in the diffusion equation. The performance of the DIKAL method has been assessed for the various core states, and has been also applied to the calculation of power and flux distribution calculation in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL is quite stable to the detector random noise and systematic errors. Also, it is shown that the DIKAL approach is more accurate than the currently used flux synthesis approach in CANDU 6 reactors. (author)

  12. Thermal Analysis of CANDU6-Moderator System for Loss of Cooling%CANDU6慢化剂系统丧失冷却情况下的温度分析

    Institute of Scientific and Technical Information of China (English)

    徐珍

    2012-01-01

    The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, was compared with the experiment data and evaluated in this paper. the thermal analysis of moderator system for loss of cooling the system failure caused by the temperature raising wasevaluated in this paper.%压力管卧式重水反应堆(CANDU6)具有相互独立的冷却剂系统和慢化剂系统。慢化剂系统将堆芯高能裂变中子慢化到能维持持续裂变所需的热中子水平,并将慢化中子过程中产生的热量带出。在反应堆大修期间,需要对再循环冷却水系统(RCW)进行检修,则需要并投入其备用系统,但是RCW备用系统仅对反应堆冷却剂系统进行冷却,不提供慢化剂系统热交换器冷却水。所以在RCW备用系统投入的情况下,慢化剂系统丧失冷却。为判断在此情况下慢化剂的温度变化情况,本文对CANDU6大修期间慢化剂系统丧失冷却情况下的温度变化进行分析并与试验结果进行比较,评估是否会由于温度过高而导致系统失效。

  13. Safety of CANDU nuclear power stations

    International Nuclear Information System (INIS)

    A nuclear plant contains a large amount of radioactive material which could be a potential threat to public health. The plant is therefore designed, built and operated so that the risk to the public is low. Careful design of the normal reactor systems is the first line of defense. These systems are highly resistant to an accident happening in the first place, and can also be effective in stopping it if it does happen. Independent and redundant safety sytems minimize the effects of an accident, or stop it completely. They include shutdown systems, emergency core cooling systems, and containment systems. Massive impairment of any one safety system together with an accident can be tolerated. This 'defence in depth' approach recognizes that men and machines are imperfect and that the unexpected happens. The nuclear power plant need not be perfect to be safe. To allow meaningful judgements we must know how safe the plant is. The Atomic Energy Control Board guidelines give one such measure, but they may overestimate the true risk. We interpret these guidelines as an upper limit to the total risk, and trace their evolution. (author)

  14. Role of accident analysis in development of severe accident management guidance for multi-unit CANDU nuclear power plants

    International Nuclear Information System (INIS)

    This paper discusses the role of accident analysis in support of the development of Severe Accident Management Guidance for domestic CANDU reactors. In general, analysis can identify what types of challenges can be expected during accident progression but it cannot specify when and to what degree accident phenomena will occur. SAMG overcomes these limitations by monitoring the actual values of key plant indicators that can be used directly or indirectly to infer the condition of the plant and by establishing setpoints beyond which corrective action is required. Analysis can provide a means to correlate observed post-accident plant behavior against predicted behaviour to improve the confidence in and quality of accident mitigation decisions. (author)

  15. 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

  16. Measurements of elastic modulus in Zr alloys for CANDU applications

    International Nuclear Information System (INIS)

    Measurements of elastic modulus as a function of temperature from 20 to 400°C were carried out on specimens of Zr-2.5Nb, Zircaloy-4, Zircaloy-2 and Excel Zr alloy using an ultrasonic resonance technique. The specimens were machined from CANDU pressure tubes, a calandria tube and commercial sheet material. Effects of crystallographic texture, neutron irradiation and hydrogen on elastic modulus were investigated. The results show that elastic modulus of the Zr alloys (1) decreases with increasing temperature, (2) depends strongly on crystallographic texture, and (3) increases slightly with neutron irradiation. (author)

  17. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  18. An improved ultrasonic downcomer flow-measurement system for CANDU steam generators

    International Nuclear Information System (INIS)

    Ultrasonic measurements of downcomer flow velocity have been successfully used in the past to determine re-circulation ratios and water inventory in CANDU steam generators. Knowledge of these process conditions allows operators to assess the effectiveness of maintenance programs, monitor the effects of tube fouling, observe flow conditions following component modifications, and provides designers with a means to validate or improve code predictions. Non-intrusive ultrasonic measurement systems were recently installed on four steam generators at the Bruce B Nuclear Generating Station as part of an investigation into the possible effects of long-term degradation due to internal Flow-Accelerated Corrosion (FAC). The most recent version of AECL's downcomer-flow measurement technology was used in which buffered ultrasonic transducers are magnetically attached and then welded to the steam-generator outer shell. This method of attachment eliminates the complications of precision surface preparation and high-temperature couplants. The paper outlines the new attachment method and summarizes flow velocities measured during start-up, shut-down and normal operation. It also briefly describes how the information may be used to assess thermalhydraulic conditions, verify design calculations, and support the case for reactor uprating. (author)

  19. Aging of elastomers in CANDU pressure boundary service

    International Nuclear Information System (INIS)

    This report describes the properties and aging of elastomers, and examines the performance of major elastomeric components in CANDU pressure boundary service. The components examined are vacuum building roof seals, pressure relief duct seals, airlock door seals, fuelling machine hoses, and cable penetrations. For each of these components, the design requirements, technical specifications and component testing procedures are compared with applicable standards. Information on actual and recommended monitoring and maintenance methods is presented. Operational and environmental stressors are identified. Component failure modes, causes and frequencies are described, as well as the remedial action taken. Many different elastomers are used in CANDU plants, for many different applications. Standards and manufacturers' recommendations are not consistent and may vary from one component to another. Accordingly, the monitoring, maintenance and replacement practices tend to vary from one application to another, and may also be different at different stations. Recommendations are given in this report for improved monitoring and maintenance, in an attempt to provide more consistency in approach. A summary of some experiences with elastomers from non-Canadian sources is contained in the last section. 125 refs

  20. 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)

  1. Future CANDU nuclear power plant design requirements document executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Usmani, S.A. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new.

  2. 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

  3. 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)

  4. 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)

  5. Project planning and scheduling techniques for the CANDU programme - an overview

    International Nuclear Information System (INIS)

    The energy crisis and higher costs have imposed the need for tighter control of completion time for the construction of CANDU nuclear power plants. System procedures and techniques to meet this challenge are described

  6. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  7. 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

  8. 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.

  9. 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

  10. 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 %.

  11. 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

  12. Integrated control centre concepts for CANDU power plants

    International Nuclear Information System (INIS)

    The size and complexity of nuclear power plants has increased significantly in the last 20 years. There is general agreement that plant safety and power production can be enhanced if more operational support systems that are significantly different from the ones based on the more conventional technologies used in plant control rooms. In particular, artificial intelligence and related technologies will play a major role in the development of innovative methods for information processing and presentation. These technologies must be integrated into the overall management and control philosophy of the plant and not be treated as vehicles to implement point solutions. The underlying philosophy behind our approach is discussed in this paper. Operator support systems will integrate into the overall control philosophy by complementing the operator. Four support systems are described; each is a prototype of a system being considered for the CANDU 3 control centre

  13. An analysis to determine industry's preferred option for an initial generic reliability database for CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Komljenovic, D. [Hydro Quebec, Montreal, Quebec (Canada)], E-mail: komljenovic.dragan@hydro.qc.ca; Chan, E. [Bruce Power, Tiverton, Ontario (Canada); Ganguli, S. [Ontario Power Generation, Ontario (Canada); Wu, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Parmar, R. [Nuclear Safety Solutions Ltd., Toronto, Ontario (Canada)

    2008-07-01

    The paper presents a multi-attribute analysis approach that was used to choose the industry's preferred option in developing a generic reliability database for CANDU. The Risk and Reliability Working Group (sponsored by CANDU Owners Group - Nuclear Safety Committee - COG NSC) was faced with the decision to assess in depth the following four options: a) Use External Data Base Only; b) Full CANDU generic database (All Utilities); c) CANDU Specific plus External Generic Databases; d) Ontario Power Generation (OPG)/ Bruce Power (BP) Experience Only. Nine decision criteria were utilized to rank the proposed options (alternatives). The Analytic Hierarchy Process (AHP) was used in carrying out the ranking process. (author)

  14. 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

  15. Critical heat flux in CANDU moderator following a pressure tube to calandria tube contact - part I

    International Nuclear Information System (INIS)

    Heavy water moderator surrounding each fuel channel is one of the important features in CANDU reactors that act as a heat sink for the fuel in the situations where other means of heat removal fail. In the critical break LOCA scenario, fuel cooling becomes severely degraded due to rapid flow reduction in the affected flow pass of the heat transport system. This can result in pressure tubes experiencing significant heat-up while coolant pressure is still high, thereby causing uniform thermal creep strain (ballooning) of the pressure tube (PT) into contact with its calandria tube (CT). The contact of the hot PT with the CT causes rapid redistribution of stored heat from the PT to CT and a large spike in heat flux from the CT to the moderator fluid. For lower subcooling conditions of the moderator, this heat flux spike can cause dryout of the CT. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The focus of this research is to develop a mechanistic model to predict Critical Heat Flux (CHF) on the CT surface following a contact with its pressure tube. A COMSOL multi-physics model using a two-dimensional transient fluid-thermal analysis of the CT surface undergoing heat up is used to predict flow and temperature profile on the CT surface. A mechanistic CHF model is to be proposed based on a concept of wall dry patch formation, prevention of rewetting and subsequent dry patch spreading. (author)

  16. CANDU 6 steam generator thermalhydraulic modeling and simulation

    International Nuclear Information System (INIS)

    The main objective of this paper is to describe the process of accurately modeling the dynamic behavior of CANada Deuterium Uranium 6 (CANDU 6) steam generator and its control algorithm. The mathematical model of the steam generator was developed mainly through differential equations calculated from the physical properties of the components. In addition, empirical modeling techniques were utilized in order to incorporate more complex properties of steam generators. The controller design and resulting controller performance on the actual plant are both strongly dependent on the accuracy of the mathematical model used to describe the plant. During power manoeuvrings, the level control in a steam generator is complicated by the thermal reverse effects known as 'shrink and swell'. This non-linear non-minimum phase characteristic of the steam generators is most challenging to model. Based on various recent publications, the root cause of the deficiencies in model-based controller designs is the fact that the 'shrink and swell' phenomenon is not captured by the plant model. The model developed in this project captures this phenomenon and clearly presents the adverse effect of such characteristic in the performance of conventional controllers. The resulting mathematical model developed in this project is implemented in Simulink, a graphical block diagramming tool offered by MATLAB. In addition to the plant model, the complete control algorithm of the CANDU 6 steam generator is modeled in this project. This model duplicates the actual control strategy applied in the existing steam generators in order to create a realistic interaction between the plant and its control algorithm. The control model is developed in discrete-time in order to accurately simulate the output of the digital controller. The analog model of the plant is dynamically integrated with the digital control model in the Simulink environment creating a realistic presentation of the actual communication

  17. R and D in support of CANDU plant life management

    International Nuclear Information System (INIS)

    One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

  18. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    International Nuclear Information System (INIS)

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

  19. Detecting, locating and identifying failed fuel in Canadian power reactors

    International Nuclear Information System (INIS)

    This document summarizes how defected fuel elements are detected, located and identified in Canadian CANDU power reactors. Fuel defects are detected by monitoring the primary coolant for gaseous fission products and radioiodines, while location in core is usually performed on-power by delayed neutron monitoring of coolant samples from individual fuel channels or off-power by gamma-ray monitoring of the channel feeder pipes. The systems and techniques used to detect and locate defected fuel in both Ontario Hydro and CANDU 6 power stations are described, along with examples provided by station experience. The ability to detect and locate defected fuel in power stations was greatly enhanced by a fundamental R and D program, which provided an understanding and models of fission-product release and transport, and the post-defect deterioration of failed fuel. Techniques and equipment used to identify and store defected fuel after it has been discharged from the reactor are briefly reviewed

  20. Etude de l'influence du champ magnetique dans une section d'essais thermohydraulique d'un canal de reacteur nucleaire CANDU 6

    Science.gov (United States)

    Landry-Lavoie, Renaud

    This memoir deals with the effects of the magnetic fields present in a thermal hydraulic test section of the Canadian nuclear industry. This test section is used to determine the thermal hydraulic conditions that can lead to critical heat flux in a channel of a CANDU 6 nuclear reactor. To perform their series of experiments the STERN Company used strong electric currents to heat the simulation bundles with a thermal power similar to the one found in a channel of a CANDU reactor. The materials constituting the simulation channel and its supports are of ferromagnetic nature. The strong magnetic field generated by the bundles implies that they are subjected to a magnetostatic force due to the magnetization of the ferromagnetic materials. The nuclear industry wants to know if these efforts, combined with the force of gravity, are sufficient to maintain the bundles in place in the simulation channel. The question also arises whether or not the magnetic field present in the channel can affect the parameters of boiling heat transfer. To determine the magnetic field distribution in the simulation channel, we had recourse to the magnetostatic image method and the integral method of calculation of magnetization. The results of the calculations show that the magnetostatic forces exerted by the ferromagnetic elements of the test section are inferior in magnitude to the one estimated by the STERN laboratorie. We used the mechanistic model of Sullivan et al. (1964) to evaluate the possible influence of the magnetic fields on the departure diameter of the vapor bubbles. The deviation in the frequency of bubble emission was evaluated by using the correlations of Zuber et al. (1959) and Cole (1960). By introducing a magnetostatic force in the boiling model and in the correlations, we demonstrated that the magnetic field present in the STERN test section has a negligible effect on the bubble departure diameter and their emission frequency. We conclude that the conditions in the test

  1. A framework for operator support systems for CANDU

    International Nuclear Information System (INIS)

    The size and complexity of nuclear power plants has increased significantly in the last 20 years. There is a general agreement that both plant safety and availability can be enhanced by providing the operator with more operational support if than can be done without overloading him/her with unnecessary information. Recent advances in computer technology provide opportunities for implementing operator support systems that are significantly different from the ones based on the more conventional technologies used in plant control rooms. In particular, artificial intelligence and the related technologies will play a major role in the development of innovative methods for information processing and presentation. Our approach to these technologies is that they must be integrated into the overall management and control philosophy of the plant and are not to be treated as vehicles to implement point solutions. The underlying philosophy behind this approach and the design objectives and goals for the CANDU 3 control centre are discussed in this paper. Operator support systems will integrate into the overall control philosophy by complementing the operator. There is also a role for such systems in assisting the operator to be a situation manager, organizing, managing and planning the running of the plant. Four support systems that incorporate the underlying philosophy are described

  2. A design basis for the development of advanced CANDU control centres

    International Nuclear Information System (INIS)

    The basic design for current CANDU control centres was established in the early 1970's. Plants constructed since then have, for the most part, retained the same basic design. Several factors have led to the need to re-examine CANDU control centre design for plants to be built beyond the year 2000. These factors include the changing roles and responsibilities for the operations staff, an improved understanding of operational issues associated with supervisory control, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and marketing pressures. This paper describes the proposed design bases for the development of advanced control centres to be implemented in CANDU plants beyond the year 2000. Four areas have been defined covering design goals, design principles, operational bases, and plant functional bases. (author)

  3. Evaluation the coolant void reactivity of assemblies of CANDU-6 lattices

    International Nuclear Information System (INIS)

    It is known that the coolant void reactivity (CVR) is positive in an isolated CANDU-6 fuel cell. Here we will attempt to evaluate the CVR for a heterogeneous assembly of CANDU-6 cells using the code DRAGON. Because the existing version of DRAGON has inherent geometry limitations, CANDU-6 assembly calculations cannot be performed in the exact geometry. Accordingly, an approximate model for the geometry must be considered. Here, the model we propose consists in replacing the annular fuel pins by equivalent square fuel pins. This model can be used to calculate the effect of homogeneous or heterogeneous draining of the coolant in a single cell. It will also be used to evaluate the effect of fine mesh discretization at the corner of the lattices on cell eigenvalue as well as the CVR for the checkerboard voiding of a 2 x 2 assembly of cells. (author)

  4. Application of the SASSI soil structure interaction method to CANDU 6 NPP seismic analysis

    International Nuclear Information System (INIS)

    The standard CANDU 6 NPP has been conservatively qualified for a Design Basis Earthquake (DBE) peak horizontal ground acceleration of 0.2 g. Currently there are potential opportunities for siting the CANDU 6 at higher seismicity sites. In order to be able to extend the use of a standardized design for sites with higher seismicity than the standard plant, various design options, including the use of the SASSI Soil Structure Interaction (SSI) analysis method, are being evaluated. This paper presents the results of a study to assess the potential benefits from utilization of the SASSI computer program and the use of more realistic damping ratios for the structures

  5. Application of the SASSI soil structure interaction method to CANDU 6 NPP seismic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ricciuti, R.A.; Elgohary, M.; Usmani, S.A. [Atomic Energy of Canada Ltd., Mississauga, Ontario (Canada)

    1996-12-01

    The standard CANDU 6 NPP has been conservatively qualified for a Design Basis Earthquake (DBE) peak horizontal ground acceleration of 0.2 g. Currently there are potential opportunities for siting the CANDU 6 at higher seismicity sites. In order to be able to extend the use of a standardized design for sites with higher seismicity than the standard plant, various design options, including the use of the SASSI Soil Structure Interaction (SSI) analysis method, are being evaluated. This paper presents the results of a study to assess the potential benefits from utilization of the SASSI computer program and the use of more realistic damping ratios for the structures.

  6. A study for good regulation of the CANDU's in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Joe, S. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2002-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the second phase of this 2 years study, following research activities were done. Review the technical basis and framework of the new Canadian Regulation System and IAEA. Analysis on the interview of Wolsung operation staffs to identify important safety issues and regulation problems experienced at operation. Providing a plan of CANDU regulation system enhancement program.

  7. Analysis of F/M duty cycle and O/M cost for four-bundle shift refuelling scheme in CANDU6 NPP

    International Nuclear Information System (INIS)

    A four-bundle shift refuelling method, a refuelling scheme that can reduces local flux peak compared to the current eight-bundle shift refuelling method used in CANDU6 NPP, is analyzed to see how much Fuel Handling System load and management cost increase are required due to the change. The current eight-bundle shift refuelling method requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The four-bundle shift refuelling method increases Fuelling Machine duty cycle and operator load. The study showed that the refuelling scheme change from the eight-to four-bundle shift increases the operation and maintenance cost about 35% from the current figure by conservative estimate and that the Fuel Handling System has enough flexibility to meet the demand of a more frequent refuelling scheme

  8. Safety design of next generation SUI of CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada); Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Review of current SUI technologies and challenges. Black-Right-Pointing-Pointer Propose a new type of SUI detectors. Black-Right-Pointing-Pointer Propose a new SUI system architecture and layout. Black-Right-Pointing-Pointer Propose implementation procedure for SUI with reduced risks. - Abstract: Due to the age and operating experience of Nuclear Power Plants, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. This paper summarizes the research into a methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this research is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today's digital technology, the objectives of this study do not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Nuclear Power Plants (NPPs) will be identified and a structured approach for developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU (Canadian Deuterium Uranium) stations is proposed.

  9. Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

  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. 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

  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. Super critical water reactor for use in steam generation for recovery of bitumen resources

    International Nuclear Information System (INIS)

    The process of recovering the bitumen (oil sand) resources in Alberta requires steam at high pressures. To help reduce the carbon footprint of exploiting these fuel resources, an innovative new design of a CANDU super critical water reactor (CANDU-SCWR) is being considered to provide the high pressure steam required for the steam assisted gravity drainage (SAGD) process. The high temperature and pressure associated with the CANDU-SCWR allow for the high pressure, temperature steam to be produced without supplementary energy. The Petroleum Technology Alliance of Canada (PTAC) has specified the SAGD process requires steam at 11 MPa and near 100% steam quality, and net electrical power of 106 MWe. This paper examines steam cycle and design options to meet the steam and power requirements defined by PTAC. Steam cycle options are examined focusing on the optimization of steam and power conversion. Additionally passive safety and cooling for both the heat transport and moderation systems are considered and their impact on performance are examined. As the CANDU-SCWR is at a preliminary stage of design, basic design parameters have been defined based on preliminary assessments. This paper is focused on a reactor with the following basic design assumptions: Vertical fuel channel; Re-entrant fuel channels; Pu-Th fuel; and Batch refuelling. (author)

  14. Control and instrumentation systems for the 600 MWe CANDU PHW nuclear power plants

    International Nuclear Information System (INIS)

    The control and instrumentation systems for CANDU power plants are designed for high reliability and availabilty to meet stringent safety and operational requirements. To achieve these goals a 'defence-in-depth' design philosophy is employed. The diversely functioning systems designed to satisfy these requirements are described, along with the extensive use of computers for important plant control and man-machine functions

  15. Monte Carlo Few-Group Constant Generation for CANDU 6 Core Analysis

    Directory of Open Access Journals (Sweden)

    Seung Yeol Yoo

    2015-01-01

    Full Text Available The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analysis, this paper proposes utilizing a B1 theory augmented Monte Carlo (MC few-group constant generation method (B1 MC method which has been devised for the PWR fuel assembly analysis method. To examine the applicability of the B1 MC method for the CANDU 6 core analysis, the fuel bundle cell and supercell calculations are performed using it to obtain the two-group constants. By showing that the two-group constants from the B1 MC method agree well with those from WIMS-AECL and that core neutronics calculations for hypothetical CANDU 6 cores by a deterministic diffusion theory code SCAN with B1 MC method generated two-group constants also agree well with whole core MC analyses, it is concluded that the B1 MC method is well qualified for both fuel bundle cell and supercell analyses.

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    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.

  18. The influence of lattice structure and composition on the coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

    The neutronic mechanisms responsible for the production of positive void reactivity in CANDU lattices are presented. Based on an understanding of these, a method of void reactivity reduction is developed. Calculations show that a zero void reactivity lattice, which requires neutron poison in the fuel and fuel enrichment, is possible. (Author) 1 ref., 8 tabs

  19. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    International Nuclear Information System (INIS)

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  20. Ajustement du rechargement et des mecanismes de reactivite des reacteurs CANDU pour les cycles de combustible avances

    Science.gov (United States)

    St-Aubin, Emmanuel

    This research project main objectives are to set up and apply a methodology that can determine the potential of advanced thorium-based fuel cycles in CANDU reactors and that is able to adjust reactivity devices, in such a way as to maintain their reference efficiency for these new fuels. In order to select these fuel cycles, a large alternative fuel envelope is submitted to several discriminating criteria. A coarse parametric core modeling, that takes into account standard reactivity devices, is first used to highlight candidates presenting the best economical performances and to eliminate non viable options. Then, for the best candidates, the neutronic modeling is optimized before considering reactivity devices adjustment. For every reactivity device managed by the reactor regulating system, innovative generic optimization methods are used to achieve specific objectives for every fuel cycle, all of them being based on the reference natural uranium cycle behavior. Specific optimization objectives are assessed by simulating advanced fuel cycle for specific operating conditions, including : normal on-power refueling period, spurious reactor trip and fueling machine unavailibility. Unlike the generalized perturbative approach proposed in the OPTEX code, we have successfully implemented a multi-step method able to maximize both the energy extracted from the fuel using an equilibrium refueling optimization, and the reactivity devices adequacy. We also propose new reactivity device supercell models that provides accurate reactor databases for a fraction of the computing cost usually needed using a full model with a similar spatial discretization. Our approach is verified by comparing our simulation results with results published in the literature for the reference fuel cycle. The methodology developed identified advanced fuel cycles, containing up to 60%v. thorium, thereby increasing resources utilization by more than 50% and multiplying the fuel average exit burn-up by

  1. Applicability of a track-based multiprocess portable robot to some maintenance tasks in CANDU nuclear plants

    International Nuclear Information System (INIS)

    Hydro-Quebec has developed a six-axis, track-based, multiprocess robot. This lightweight (30 kg) compact robot travels on a bent track with a radius of curvature ranging from 1 m to infinity (straight track). Standard and tandem wires GMAW, FCAW and Narrow gap TIG welding as well as plasma gouging and cutting, electrical and pneumatic rough and precision grinding, and profile measurement functionalities have been incorporated. A description of this technology an its newly developed functionalities is given in this paper. Since 1995, a number of industrial and R and D projects have been performed using this technology now called the Scompi technology. The main field of application is the in situ repair of hydraulic turbine runners. However some applications have been developed in the nuclear field. One particular development was funded by the International Thermonuclear Experimental Reactor (ITER) project. Scompi was selected by the ITER US Home Team for a demonstration of remote techniques for welding, cutting and rewelding the 30 m diameter, 17 m high, vacuum vessel. The demonstration involved all position robotic plasma cutting and NG-TIG welding of a 316L, 40 mm thick, double wall. In 1998, two Scompi robots working in tandem performed in York, Pa, the joint welding and cutting of a full scale portion of the vacuum vessel. In 1995, the applicability of the Scompi technology to the repair of the divider plates in the four steam generators at Gentilly-2 was evaluated based on a joint proposal by Ontario Hydro Technologies (now Ontario Power Technologies-OPT) and Hydro-Quebec. A MIG welding procedure was proposed for the horizontal and vertical divider plates welds. A complete simulation of the robot and primary head demonstrated the feasibility of the concept. However, based on cost and scheduling, it was decided to proceed with a manual repair. Nevertheless it is anticipated that this technology will find its niche in the maintenance of Candu reactors. (author)

  2. Applicability of a track-based multiprocess portable robot to some maintenance tasks in CANDU nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Hazel, B.; Fihey, J.-L.; Laroche, Y. [Hydro-Quebec, Varennes, Quebec (Canada)

    2000-07-01

    Hydro-Quebec has developed a six-axis, track-based, multiprocess robot. This lightweight (30 kg) compact robot travels on a bent track with a radius of curvature ranging from 1 m to infinity (straight track). Standard and tandem wires GMAW, FCAW and Narrow gap TIG welding as well as plasma gouging and cutting, electrical and pneumatic rough and precision grinding, and profile measurement functionalities have been incorporated. A description of this technology an its newly developed functionalities is given in this paper. Since 1995, a number of industrial and R and D projects have been performed using this technology now called the Scompi technology. The main field of application is the in situ repair of hydraulic turbine runners. However some applications have been developed in the nuclear field. One particular development was funded by the International Thermonuclear Experimental Reactor (ITER) project. Scompi was selected by the ITER US Home Team for a demonstration of remote techniques for welding, cutting and rewelding the 30 m diameter, 17 m high, vacuum vessel. The demonstration involved all position robotic plasma cutting and NG-TIG welding of a 316L, 40 mm thick, double wall. In 1998, two Scompi robots working in tandem performed in York, Pa, the joint welding and cutting of a full scale portion of the vacuum vessel. In 1995, the applicability of the Scompi technology to the repair of the divider plates in the four steam generators at Gentilly-2 was evaluated based on a joint proposal by Ontario Hydro Technologies (now Ontario Power Technologies-OPT) and Hydro-Quebec. A MIG welding procedure was proposed for the horizontal and vertical divider plates welds. A complete simulation of the robot and primary head demonstrated the feasibility of the concept. However, based on cost and scheduling, it was decided to proceed with a manual repair. Nevertheless it is anticipated that this technology will find its niche in the maintenance of Candu reactors. (author)

  3. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    Science.gov (United States)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  4. Geoneutrinos and reactor antineutrinos at SNO+

    CERN Document Server

    Baldoncini, M; Wipperfurth, S A; Fiorentini, G; Mantovani, F; McDonough, W F; Ricci, B

    2016-01-01

    In the heart of the Creighton Mine near Sudbury (Canada), the SNO+ detector is foreseen to observe almost in equal proportion electron antineutrinos produced by U and Th in the Earth and by nuclear reactors. SNO+ will be the first long baseline experiment to measure a reactor signal dominated by CANDU cores ($\\sim$55\\% of the total reactor signal), which generally burn natural uranium. Approximately 18\\% of the total geoneutrino signal is generated by the U and Th present in the rocks of the Huronian Supergroup-Sudbury Basin: the 60\\% uncertainty on the signal produced by this lithologic unit plays a crucial role on the discrimination power on the mantle signal as well as on the geoneutrino spectral shape reconstruction, which can in principle provide a direct measurement of the Th/U ratio in the Earth.

  5. Characterization of excel alloy pressure tube material for CANDU SCW reactors

    International Nuclear Information System (INIS)

    The phase transformation temperatures, aging response, and creep rupture strength of Zr alloy Excel (Zr- 3.5%Sn- 0.8%Nb- 0.8%Mo) pressure tube material were investigated. The α → α+β and α+β → β transus temperatures were found to be in the range of 600-690 °C and 962-975 °C respectively. Precipitation hardening was observed in the microstructures water-quenched from high in the α+β or β regions followed by aging at 400-500 °C for 1 hr. The results of creep-rupture experiments at 400 °C suggest that a fully martensitic and aged microstructure has better creep properties at high stress levels (>700 MPa) and a microstructure obtained by air-cooling from high in the α+β region shows good creep properties at lower stresses (<560 MPa). (author)

  6. Timing of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    This paper documents the analytical methodology used to evaluate severe accident sequences. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression are described. An example of a specific accident scenario is provided in order to illustrate the application of the severe accident progression methodology. The postulated sever accidents analyzed mainly differ in the timing to reach and progress through each defined 'core damage state'. (author)

  7. Failure mode and effect analysis experimental reliability determination for the CANDU reactor equipment

    International Nuclear Information System (INIS)

    This paper describes the experimental tests performed in order to prove the reliability parameters for certain equipment manufactured in INR Pitesti, for NPP Cernavoda. The tests were provided by Technical Specifications and test procedures. A comparison, referring to the reliability parameters, between Canadian equipment and INR manufactured equipment ones is also given. The results of tests and conclusions are shown. (author)

  8. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  9. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  10. 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. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    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-VT (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 keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), 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}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)

  11. 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)

  12. Assessment of In-Core Damage for Feeder Stagnation Break in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    A feeder break is a single channel accident while the other channels remain intact in the CANDU core. For some ranges of feeder break size, a flow in the channel can become stagnate due to a force balance between the upstream and the downstream ends. In the extreme, this can lead to a rapid fuel heat up and fuel damage, and the failure of a fuel channel. This break scenario is called a feeder stagnation break. Following the feeder stagnation break, the fuel and pressure tube in the affected channel heat up quickly. The channel fails due to overheating and the channel contents begin to discharge into the moderator. The discharge is composed of steam, some hydrogen produced by possible metal-water reaction, and solid fuel elements of fuel fragments with molten material. The severity of the transient is primarily determined by the amount of molten material discharged into the moderator, and by the interaction between the molten material and the moderator, which determines the rate of energy release. After a channel rupture (pressure tube and calandria tube) some SOR (Shut-Off Rod) guide tubes, which are located in the vicinity of the break in the core, may be damaged. If the damage to the guide tube is substantial, some SORs may not be able to descend into the moderator, and therefore, not contribute to the shut down of the reactor. The increase in system reactivity, due to factors such as poison dilution from discharging coolant and void formation, may challenge the reactivity worth of the available undamaged SORs. Therefore, an analysis of the reactivity worth of the partially impaired SDS 1 (Shut-Down System 1) is required to determine that it can compensate for the increase in reactivity and shut down the reactor. In this study, the hydrodynamic transient, due to the dispersed molten material and the discharged steam, was calculated following the feeder stagnation break. The timing of the channel failure and the mass of the molten material were provided from the

  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. IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)

    International Nuclear Information System (INIS)

    Description of program or function: Romanian irradiation tests concerned with Candu type fuel elements behavior and with the limits of the design parameters. A particular feature of the Candu fuel project is the small plenum (void volume) added for relaxation of the fission gases, which are inherently released during the fuel irradiation. Two irradiation tests in the C2 device from the TRIGA 14 MW reactor were performed between the years 1985-1987. The tests were done to evaluate the effect of the fuel density on the time-evolution of the fission gas pressure. Experimental fuel elements were adequately instrumented with pressure transducers to follow the fission gas pressure changes during fuel irradiation. The first irradiation test was conducted on the fuel element coded No.89 whose main characteristics were the nominal values of the main fuel design parameters. The second one was conducted on the fuel element coded No.51. Because of the axial flux asymmetry inside the TRIGA reactor core, the experimental elements are shorter in length than the Candu fuel design. The irradiation tests consisted in evaluation of the time-evolution of the internal pressure from two experimental fuel elements having the main design characteristics as the Romanian Candu type fuel element design and to follow the dependence of the internal pressure of the fission gas on the fuel density

  15. Burnup-dependent effect of lattice-level homogenization and group condensation on calculated kinetics parameters for CANDU-type lattices

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.nichita@uoit.ca [University of Ontario Institute of Technology, 2000 Simcoe Str. North, Oshawa, ON, Canada L1H 7K4 (Canada); Serghiuta, D.; Podobed, S. [Canadian Nuclear Safety Commission, 280 Slater Street, P.O. Box 1046 Station B, Ottawa, ON, Canada K1P 5S9 (Canada)

    2015-05-15

    Highlights: • CANDU-type-lattice kinetics parameters are calculated using different adjoint-weighting approximations at different burnups. • Fine-group space-dependent adjoint weighting is the most accurate method of calculating the kinetics parameters. • Two-group lattice-homogenized adjoint weighting overestimates the effective delayed-neutron fraction by approximately 5%. • Fine-group lattice-homogenized adjoint weighting overestimates the effective delayed neutron fraction only by approximately 2%. - Abstract: Modern analysis of nuclear reactor transients uses space-time reactor kinetics methods. In the Canadian nuclear industry, safety analysis calculations use almost exclusively the Improved Quasistatic (IQS) flux factorization method. The IQS method, like all methods based on flux factorization, relies on calculating effective point kinetics parameters, which dominate the time behavior of the flux, using adjoint-weighted integrals. The accuracy of the adjoint representation influences the accuracy of the effective kinetics parameters. Routine full core calculations are not performed using detailed models and transport theory, but rather using a cell-homogenized model and two-group diffusion theory. This work evaluates the effect of homogenization and group condensation at different burnups, for three fuel types: natural-uranium (NU) fuel, low-void reactivity (LVR) fuel and Advanced CANDU Reactor (ACR) fuel. Results show that the use of a two-group lattice-homogenized adjoint consistently overestimates the effective delayed neutron fraction by approximately 5% for all three fuel types and over a wide burnup range. The use of a two-group lattice-homogenized adjoint also introduces errors in the effective neutron generation time, but these are at most 1.3% (and their sign changes with burnup). Errors tend to vary with burnup by approximately 1% (of the individual parameter value). If a 69-group lattice-homogenized adjoint is used, the errors drop to

  16. Management of a 600 MW CANDU project to facilitate electricity export

    International Nuclear Information System (INIS)

    The export of electricity from 600-MW CANDU nuclear power plants built in Canada remains feasible providing certain requirements continue to be met. The principal objective in developing nuclear power resources for export is that they must produce economically attractive electricity. A review of the experience of construction and operation of Point Lepreau Unit 1 suggests an inherent ability to reduce construction costs and shorten construction schedules so as to make electrical power output from these stations even more attractive to export customers

  17. Database and modeling assessments of the CANDU 3, PIUS, ALMR, and MHTGR designs

    International Nuclear Information System (INIS)

    As part of the research program to support the preapplication reviews of the CANDU 3, PIUS, ALMR, and MHTGR designs, the NRC has completed preliminary assessments of databases and modeling capabilities. To ensure full coverage of all four designs, a detailed assessment methodology was developed that follows the broad logic of the NRC's Code Scaling, Applicability, and Uncertainty (CSAU) methodology. This paper describes the methodology of the database assessments and presents examples of the assessment process using preliminary results for the ALMR design

  18. Review of the Safety Concern Related to CANDU Moderator Temperature Distribution and Status of KAERI Moderator Circulation Test (MCT) Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Bo W.; Kim, Hyoung T. [Severe Accident and PHWR Safety Research Division, Daejeon (Korea, Republic of); Kim, Tongbeum [University of the Witwatersrand, Johannesburg (South Africa); Im, Sunghyuk [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    threshold temperature and no further deformation is expected. Consequently, a sufficient condition to ensure fuel channel integrity following a large LOCA, is the avoidance of sustained calandria tubes dryout. If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The temperature oscillations observed in reactor and in test measurements such as MTF need to be characterized and quantified to show that it does not jeopardize the currently available safety margins. Because of the importance of an accurate prediction of moderator temperature distributions and the related moderator subcooling, a 1/4 scaled-down moderator tank of a CANDU-6 reactor, called Moderator Circulation Test (MCT), was erected at KAERI and the current status of MCT experiment progress is described and further experiments are expected to be carried out to generate the experimental data necessary to validate the computer codes that will be used to analyze the accident analysis of operating CANDU-6 plants.

  19. COG CANDU outage optimization project at Wolsong, Qinshan, Cernavoda, Point Lepreau, Darlington and Pickering

    Energy Technology Data Exchange (ETDEWEB)

    McWilliams, L. [CANDU Owners Group, Toronto, Ontario (Canada)

    2011-07-01

    This COG initiated project objective is to reduce the timeline CANDU power plants are shutdown for planned maintenance outages through knowledge sharing, benchmarking and completion of a Gap Analysis. The following CANDU Nuclear Power Stations/Facilities formed a partnership to achieve the objective: Korean Hydro and Nuclear Power Company (Wolsong), Qinshan, Societatea Nationala 'Nuclearelectrica' (Cernavoda), New Brunswick Power Nuclear (Point Lepreau), Candu Energy Inc. (formerly AECL) and Ontario Power Generation Inc. (Darlington and Pickering). Project participants selected ten focus areas to evaluate and optimize. Benchmarking studies were conducted at each utility. A Gap Analysis was performed between the stations and site specific recommendations have been made considering: Critical path improvement opportunities (Unit Shutdown, RBLRT and Unit Start up); Major work program improvement opportunities (Turbine/Generator, Electrical Maintenance, MOT, SST and the Valve Program); Recommended modifications to reduce outage durations; and, Process Improvements (standardized clearance process). A final report has been issued to each station identifying: Gap Analysis Comparison results; Best Practices for each area studied; Site specific improvement opportunities; Most Effective Process for Outage Preparation; Most Effective Outage Execution Practices; and, Contingency Plan Preparation. Results were discussed during the presentation.

  20. Long-term passive CANDU containment response after a design-basis accident

    International Nuclear Information System (INIS)

    A passive CANDU reg-sign containment system, currently being developed, is aimed at limiting the consequences of a postulated accident, by ensuring the structural integrity of the containment building and limiting fission-product release to within siting dose limits, without operator action or reliance on ac power for up to 3 d. All main functions of the containment system, i.e. energy removal, hydrogen mitigation, and fission-product retention, are to be accomplished passively. The passive CANDU containment relies on the passive emergency water system (PEWS) for energy removal after an accident and on passive autocatalytic recombiners (PAR) for hydrogen removal. The key feature of this concept, is a recirculating, buoyancy-driven flow through the recombiners and the tube banks of the PEWS. This paper presents preliminary design calculations for the PEWS tank and tube banks and a simulation of the long-term passive containment response, based on the current CANDU-6 containment, to a large loss-of-coolant/loss-of- emergency coolant injection (LOCA/LOECI) using the GOTHIC code. It is shown that a 1500-M3 PEWS tank, connected to tube banks with a total surface area of 1800 m2, can limit the second pressure peak to about 300 kPa(a) if a recirculating flow is established in the containment building. The PEWS tank water is boiling in the long term, and the peak containment temperature is 114 degrees C. 6 refs., 4 figs

  1. Standard compliance - NDE performance demonstration/inspection in the CANDU industry

    International Nuclear Information System (INIS)

    CANDU nuclear power plants are operated in 3 provinces in Canada for electric power generation. A table in the paper will show the built and operating plants in Ontario, Quebec, New Brunswick and overseas. The regulator for nuclear power in Canada is the Canadian Nuclear Safety Commission (CNSC). The CNSC holds the plant licensees accountable for compliance to CSA N285.4 for periodic inspections. The Standard basically specifies the 'what, when, where, how, how much and how frequently' NDE is to be done on pressure retaining systems and components in CANDU nuclear power plants. In inspection methods, the Standard specifies they must be non-destructive. The NDE methods were grouped into visual, dimensional, surface, volumetric and integrative. The Standard also specifies that the licensees are responsible for the performance demonstration (PD) of the adequacy of the procedures and the proficiency of the personnel. This paper describes the Standard's requirement in NDE qualification and presents a joint project participated by Canadian and overseas CANDU owners. The sub-project for NDE included providing evidence and technical justification on the adequacy of the procedures and the proficiency of the personnel. The paper describes the qualification methodology followed by the participants. This will be followed by how the participants produced Inspection Specification, tools and procedures, personnel training and qualification programs, test and qualification samples, independent peer reviews and Technical Justification. (author)

  2. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Science.gov (United States)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  3. The U.S.-Russian joint studies on using power reactors to disposition surplus weapons plutonium as spent fuel

    International Nuclear Information System (INIS)

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years

  4. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chebeskov, A.; Kalashnikov, A. [State Scientific Center, Obninsk (Russian Federation). Inst. of Physics and Power Engineering; Bevard, B.; Moses, D. [Oak Ridge National Lab., TN (United States); Pavlovichev, A. [State Scientific Center, Moscow (Russian Federation). Kurchatov Inst.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  5. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  6. A magnetic flux leakage NDE system for CANDU feeder pipes

    Science.gov (United States)

    Mak, Thomas Don

    This work examines the application of different magnetic flux leakage (MFL) inspection concepts to the non destructive evaluation (NDE) of residual (elastic) stresses in CANDURTM reactor feeder pipes. The stress sensitivity of three MFL inspection techniques was examined with flat plate samples, with stress-induced magnetic anisotropy (SMA) demonstrating the greatest stress sensitivity. A prototype SMA testing system was developed to apply magnetic NDE to feeders. The system consists of a flux controller that incorporates feedback from a wire coil and a Hall sensor (FCV2), and a magnetic anisotropy prototype (MAP) probe. The combination of FCV2 and the MAP probe was shown to provide SMA measurements on feeder pipe samples and predict stresses from SMA measurements with a mean accuracy of +/-38MPa.

  7. BDDR, a new CEA technological and operating reactor database

    Energy Technology Data Exchange (ETDEWEB)

    Soldevilla, M.; Salmons, S.; Espinosa, B. [CEA-Saclay, CEA/DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette (France); Clanet, M.; Boudin, X. [CEA-Bruyeres-le-Chatel, 91297 Arpajon (France)

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  8. Study of the developmental status and operational features of heavy water reactors. Final report

    International Nuclear Information System (INIS)

    The Canadian Heavy Water Reactor System, CANDU, and future plans for it as reported in a number of recent Canadian papers, are briefly reviewed. This has been done recognizing several leading observers of the nuclear industry suggest CANDU may represent a promising longer term route to nuclear fuel self sufficiency. It now appears that the concept is well established with a demonstrated performance comparable to the best LWR installations, but it presently has fuel utilization efficiency which is of the same order as that of the current LWRs. It should also be noted that CANDU is designed to a different set of safety and regulatory criteria; a point having potential for at least a second order economic significance to many non-Canadian markets. The feature of particular interest, however, is that the system is indicated by the Canadian Government Corporation-Atomic Energy of Canada Ltd.- to be capable of upgrading to a thermal near breeder. As such, this is a point of substantial import, given that system hardware is already proven, of demonstrated safety, reliability and economy, and one for which some supply industry capability is now in place. Remaining areas of uncertainty in the planned CANDU program are those related to the near breeder fuel cycle with its implied fuel processing and recycling requirements, the economics of future heavy water supply, and the actual demonstration of the physics of a unity converter/near breeder. It is noted that recently a variant of this LWBR using heavy water coolant, has been proposed by U.S. workers. Thus LWBR related efforts may be a more appropriate direction for U.S. investigations than a direct focus on CANDU itself. It is recommended that further investigation of such improved conversion developments for existing LWRs be conducted; they could represent a potential addition to the present U.S. program for a 100 percent LMFBR future

  9. 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)

  10. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  11. The safety of Ontario's nuclear power reactors. A scientific and technical review. A submission to the Ontario Nuclear Safety Review by Atomic Energy Canada Limited

    International Nuclear Information System (INIS)

    This submission comments on the evolution of the Canadian nuclear program, the management of safety, and the reactor design, analysis, operation and research programs that contribute to the safety of the CANDU reactor and provide assurance of safety to the regulatory agency and to the public. The CANDU reactor system has been designed and developed with close cooperation between Atomic Energy of Canada Ltd. (AECL), utilities, manufacturers, and the Atomic Energy Control Board (AECB). The AECB has the responsibility, on behalf of the public, for establishing acceptable standards with respect to public risk and for establishing through independent review that these standards are satisfied. The plant designer has responsibility for defining how those standards will be met. The plant operator has responsibility for operating within the framework of those standards. The Canadian approach to safety design is based on the philosophy of defence in depth. Defence in depth is achieved through a high level of equipment quality, system redundancy and fail-safe design; regulating and process systems designed to maintain all process systems within acceptable operating parameters; and, independent safety systems to shut down the reactor, provide long-term cooling, and contain potential release of radioactivity in the event of an accident. The resulting design meets regulatory requirements not only in Canada but also in other countries. Probabilistic safety and risk evaluations show that the CANDU design offers a level of safety and least as good as other commercially available reactor designs

  12. 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

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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)

  18. 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

  19. Requirements for the support power systems of CANDU nuclear power plants

    International Nuclear Information System (INIS)

    This Standard covers principal criteria and requirements for design, fabrication, installation, qualification, inspection, and documentation for assurance that support power will be available as required. The minimum requirements for support power are determined by the special safety systems and other safety-related systems that must function to ensure that the public health risk is acceptably low. Support power systems of a CANDU nuclear power plant include those parts of the electrical systems and instrument air systems that are necessary for the operation of safety-related systems

  20. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    Directory of Open Access Journals (Sweden)

    Urva Pancholi

    2016-07-01

    Full Text Available The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR, generally of a Canadian Deuterium Uranium (CANDU type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniques, instruments and specific data for such components mentioned here are synthesized results from primary research and knowledge available in this area, in order to produce coherent argument focused on quality control of end fittings.

  1. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART II: CUTTING AND EXTRACTING PRESSURE TUBE PROCESS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The cutting and extracting pressure tube processes present few particularities due to special adopted technical solutions: a special module with three cutting rollers (system driven by an actuator, a guiding-extracting and connecting module (three fixing claws which are piloted by an actuator and block the device in the connecting position with extracting plugs. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons.

  2. A foundation for allocating control functions to humans and machines in future CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Since the control room for the Atomic Energy of Canada Limited CANDU 6 plant was designed in the 1970s, requirements for control rooms have changed dramatically as a result of new licensing requirements, evolution of major new standards for control centre design and technological advances. The role of the human operator has become prominent in the design and operation of industrial and, in particular, nuclear plants. Major industrial accidents in the last decade have highlighted the need for paying significantly more attention to the requirements of the human as an integral part of the plant control system. A Functional Design Methodology has been defined that addresses the issues related to maximizing the strengths of the human and the machine in the next generation of CANDU plants. This method is based, in part, on the recently issued international standard IEC 964. The application of this method will lead to the definition of the requirements for detailed design of the control room, including man-machine interfaces, preliminary operating procedures, staffing and training. Further, it provides a basis for the verification and validation of the allocation of functions to the operator and the machine

  3. CFX analysis of the CANDU moderator thermal-hydraulics in the Stern Lab. Test Facility

    International Nuclear Information System (INIS)

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data. It is shown that the present CFD prediction without the empirical correlation based on the pressure drop test is in good agreement with the test results. The prediction becomes more accurate, as the flow conditions become more turbulent with a higher Reynolds number. However, the temperature fluctuation is observed during iteration steps for a steady-state simulation of the thermal-hydraulic test. This result shows that the flow and temperature distribution inside the moderator tank may not be stable in the actual test

  4. 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

  5. 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

  6. 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)

  7. Joint submission of the Canadian Nuclear Association and the Organization of CANDU Industries to the Ontario Nuclear Safety Review

    International Nuclear Information System (INIS)

    The manufacturing company members of the Canadian Nuclear Association and the Organization of CANDU Industries are proud to have played their part in the development of the peaceful application of nuclear technology in Ontario, and the achievement of the very real benefits discussed in this paper, which greatly outweigh the hypothetical risks

  8. 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)

  9. Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)

    2014-05-15

    The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.

  10. Operational transparency: an advanced safeguards strategy for future on-load refuelled reactors

    International Nuclear Information System (INIS)

    The IAEA's system for tracking fuel movement in an on-load refuelled heavy-water reactor is robust, but an opportunity remains to exploit the wealth of data streaming from the reactor vault during operation and provide real-time, third-party monitoring of reactor status and history. This concept of Operational Transparency would require that large amounts of operational data be reduced in near-real time to a small subset of high-level information. Operational Transparency would enhance the IAEA's ability to monitor the state of the core to an unprecedented level. This paper provides an overview of the novel concept of Operational Transparency in heavy water reactors, using potential application to CANDU reactors as an example, and explores some of the technical challenges that will need to be solved for efficient implementation. (author)

  11. 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

  12. Development of fault diagnostic technique using reactor noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  13. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  14. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  15. 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

  16. The development of CANDU technology and training at the institute for advanced engineering in Korea

    Energy Technology Data Exchange (ETDEWEB)

    MacBeth, M.J.; Cho, U.Y. [Electrical Power Systems, Institute for Advanced Engineering (Korea, Republic of); Muzumdar, A.P. [Atomic Energy of Canada Limited (Canada)

    1998-07-01

    This paper presents an overview of the cooperative agreement between the Institute for Advanced Engineering (IAE) and Atomic Energy of Canada Limited (AECL) to facilitate the transfer of CANDU technology in South Korea. This paper will present those AECL technology program activities worked on by IAE staff with AECL support along with the associated issues which these activities addressed and the expertise nature of this work. The training methods utilized and an assessment of their success will be discussed to show the potential applicability of these methods to the nuclear power industry staff of other countries. The spin-off cooperative work initiated with other Korean organizations as part of this initiative will also be considered. (author)

  17. Design principles for CANDU control centres in response to evolving utility business needs

    International Nuclear Information System (INIS)

    Nuclear generation operators are facing a challenging business environment at the beginning of the new millennium. Evolving changes in business context, competitive commercial pressures, and changes in technology have dictated recurring evaluation of operational practices and the adequacy of supporting tools, and the pursuit of opportunities for operational improvement. A key area of utility operations that has been impacted by these changes is the nuclear plant control centre. Changes to workspace layout, equipment provisions, staffing, and work organization are examples of some of the adjustments being introduced to improve operational and safety effectiveness. This paper discusses some of the key factors influencing these changes and identifies additional design principles for CANDU control centres that will enable new control centre designs and retrofits of existing control centres to remain relevant and responsive to utility needs. (author)

  18. 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)

  19. Estimate of man-rem expenditures for a mature CANDU 600 MW(e) station

    International Nuclear Information System (INIS)

    In recent years, man-rem expenditures at operating stations have come under close scrutiny in order to reduce operating personnel dosage. This increased awareness has led to concerted efforts to improve station design and to improve operating procedures to achieve lower man-rem expenditures. This paper is intended to highlight design improvements that have been made in the CANDU 600 MW(e) design and to show how these improvements will reduce man-rem expenditures. Other considerations, such as station decontaminations of the primary heat transport system and the fuelling machines and stricter chemistry control are presently available to help reduce man-rem consumption. Also, station management operating policy should emphasize man-rem awareness. (author)

  20. 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

  1. Thermal stability of chloroform in the steam condensate cycle of CANDU-PHW nuclear power plant

    International Nuclear Information System (INIS)

    Analysis of samples taken at the Gentilly 2 (Quebec) CANDU-PHW (CANadian Deuterium Uranium - Pressurized Heavy Water) plant after chlorination and demineralization revealed the presence of all four trihalomethanes (THMs) (CHCl3, CHBrCl2, CHBr2Cl and CHBr3) and other unidentified halogenated volatile compounds. Among the THMs, chloroform was the major contaminant. A study of its thermal stability in water at different temperatures confirmed the degradation of the CHCl3 molecule according to the equation CHCl3 + H2O → CO + 3 HCl. The reaction follows first order kinetics and has an activation energy of 100 kJ/mol. The estimated half-life is six seconds at 260 deg C, the maximum temperature of the steam condensate cycle

  2. 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)

  3. Development of linear sensitivity matrix method for fast evaluation of CANDU refuelling schemes%用于CANDU堆换料方案快速评价的线性敏感矩阵方法

    Institute of Scientific and Technical Information of China (English)

    王文聪; 刘宇轩; 冯进军; 王军; 樊申; 张少泓

    2009-01-01

    为了实现CANDU堆换料方案的快速评价,本文提出了线性敏感矩阵方法(LSM).该方法基于各种扰动引起的堆芯状态变化互不干涉且堆芯对扰动的响应与扰动量成正比的假设,借助于事先针对参考堆芯形成的敏感矩阵,主要的堆芯参数经简单代数运算就可得到,无需耗时的三维扩散方程求解.秦山三期1号机组部分运行历史检验结果表明,该方法不但可大幅提高换料方案评价的速度,而且还具有较高的精度,能满足现场工程应用的要求.%In order to develop a numerical tool for the fast evaluation of CANDU refuelling schemes, a Linear Sensitivity Matrix method (LSM) is developed. It assumes that all the effects caused by various perturbations to the core state are independent to each other and the core response to a perturbation is proportional to its magnitude. In this way, the main core parameters of a refuelling scheme can be derived by simple algebraic operations with the use of pre-calculated sensitivity matrices, without resorting to the complicated and time-consuming 3D core calculation. Verification against the QinShan CANDU reactor operation history demonstrates that LSM is capable of generating accurate results and running very fast for evaluating a refuelling scheme.

  4. 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.

  5. 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)

  6. Candu Energy's Aging and Obsolescence Program and its Application to Operating Facilities and New Plant Design

    International Nuclear Information System (INIS)

    While plant aging is inevitable, predictable and 'graceful 'aging' behavior can be achieved through the implementation of a comprehensive and integrated Plant Life Management (PLiM) program. Despite organizations like the IAEA and INPO placing more emphasis on equipment reliability, there is still a lack of completely integrated programs in the industry as evidenced by: - Piece-meal, often crisis-driven, implementation comprising many different, partial solutions; - Duplication of effort often seen when different groups work in 'silos'. A strategy which fits with existing plant processes and programs, and which coordinates a broad range of equipment reliability activities is key to achieving the desired results. An example of such a program is the Aging and Obsolescence Program (AOP). AOP follows application of INPO AP-913 guidance for equipment reliability. The program is augmented to include single point vulnerability identification, unified approach to short and long lived components, risk management, spare parts management, and the identification and resolution of obsolescence issues. The systematic nature of the program provides the needed foundation to old and new stations alike. For existing operating stations some of the key uses include outage interval extension, reduced forced outages, and/or outage time reduction, any of which can translate into improving plant performance, competitiveness, and significant dollars saved. Program elements applied to new plant design are commensurate with the industry direction to 'design for reliability', and has allowed Candu Energy to learn and to improve upon what it can offer to operating stations. This paper intends to describe the basic elements of Candu Energy's Aging and Obsolescence Program and will share some of the experience having applied it to existing operating stations, consider applications to support expanding regulatory requirements, and describe the integration into the design of new plants, promoting

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

    Energy Technology Data Exchange (ETDEWEB)

    Paul, U.K.; Xu, Z.; Xu, S.; Wang, X.; Chakraborty, K. [Atomic Energy of Canada Limited, Mississauga (Canada)

    2010-07-01

    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)

  8. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    International Nuclear Information System (INIS)

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

  9. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  10. Characteristics of outage radiation fields around various reactor components

    International Nuclear Information System (INIS)

    Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

  11. Two-phase control absorber development program: out-reactor measurements with hoorizontal absorber elements

    International Nuclear Information System (INIS)

    The two-phase control absorber works on the principle that the neutron flux in a nuclear reactor can be regulated by changing the density of a two-phase fluid flowing through U-tubes in the reactor core. The concept is considered to be a strong candidate for use in future CANDU nuclear reactors with either vertical or horizontal pressure tubes. In addition to the experiments carried out previously on vertically oriented U-tubes and reported separately, a series of tests with horizontal U-tubes was performed. The results confirmed that U-tube orientation has no measurable effect on the performance of the two-phase control absorber concept. In particular, the measured pressure drops, mixture densities, fluid velocities and void propagation velocities, at given operating conditions, were identical in the two orientations, within experimental error. The results of the experiments and analyses were incorporated in a steady-state design code that was used in the conceptual design of a Two-Phase Absorber Control System for a CANDU-PHW-1250 power reactor. The experimental data are available separately as AECL-6532 Supplement. (auth)

  12. 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)

  13. Measurement of Velocity and Temperature Profiles in the 1/40 Scaled-Down CANDU-6 Moderator Tank

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2015-01-01

    Full Text Available In order to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions, a scaled-down moderator test facility has been constructed at Korea Atomic Energy Institute (KAERI. In the present work an experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a particle image velocimetry (PIV technique under an isothermal state, and the temperature field is measured using a laser induced fluorescence (LIF technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment.

  14. 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

  15. 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

  16. Report of the COG/IAEA international workshop on managing nuclear safety at CANDU (PHWR) plants. Working material

    International Nuclear Information System (INIS)

    The workshop, hosted by COG and co-sponsored by the International Atomic Energy Agency (IAEA, Vienna) was held in Toronto, April 28 - May 1st, 1997. The 40 participants included senior managers from IAEA member countries operating or constructing CANDU (PHWR) stations. All the offshore utilities with PHWR stations in Korea, Romania, India, Argentina, Pakistan, and China were present with their domestic counterparts from Ontario Hydro Nuclear, Hydro Quebec, New Brunswick Power, and AECL. The objectives of the workshop were to: provide a forum for exchange of ideas among nuclear safety managers operating CANDU (PHWR) stations and to learn from each other's experiences; to foster sharing of information on different operating approaches to managing safety and, in particular, to highlight the strategies for controlling the overall plant risk to a low level; to identify and discuss issues of mutual interest pertinent to PHWR stations and to define future follow-up activities. Refs, figs

  17. Nuclear Reactor Simulations for Unveiling Diversion Scenarios: capabilities of the antineutrino probe

    Energy Technology Data Exchange (ETDEWEB)

    Bui, V.M.; Fallot, M.; Giot, L.; Guillon, B.; Martino, J.; Yermia, F. [SUBATECH - CNRS-IN2P3 - Univ. of Nantes - EMN, Nantes (France); Nuttin, A. [LPSC - CNRS-IN2P3/UJF/INPG, Grenoble (France)

    2009-06-15

    After many years of fundamental research, physicists have a good understanding of the neutrinos detection techniques. It is now possible to apply neutrino physics as a new tool to monitor nuclear power plants. We already know that modest size detectors are achievable to fulfill that task such as the SONGS 1 and the future Nucifer detectors. In parallel, sophisticated simulations of reactors and their associated antineutrino flux and energy spectrum have been developed to predict the neutrino signature of the fuel burnup and of a diversion. Taking advantage of the tremendous quantity of information available nowadays in nuclear databases, the total {beta} spectrum of a reactor is built by adding the contributions of all the {beta} branches involved in the decay of all fission products (FP). A package called MCNP Utility for Reactor Evolution (MURE) computes the fuel and FP inventories by simulating the neutronics and time evolution of a reactor core. MURE, initially developed by CNRS/IN2P3/LPSC Grenoble and IPN Orsay to study Generation IV reactors, is a precision code written in C++ which automates the preparation and computation of successive MCNP calculations either for precision burnup or thermal-hydraulics purpose. MURE will be soon available at NEA. The only user-defined inputs driving the time evolution of the isotopic composition of the core are the initial fuel composition, the refueling scheme, and the thermal power. The evolution of the antineutrino flux and energy spectrum with the fuel burnup, as well as the effect of neutron capture on various nuclei are taken into account. Nonproliferation scenarios and burnup monitoring with antineutrinos have been studied using these tools for PWR and Candu reactors. A full core simulation of an N4-PWR will be presented in a first part. Gross unveiling diversion scenarios using a PWR have been simulated in order to test the ability of the antineutrino probe. A channel of a Heavy Water Reactor (Candu 600) loaded with

  18. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  19. 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

  20. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor