WorldWideScience

Sample records for advanced candu reactor

  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. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

  3. Advanced fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. 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)

  6. Advanced CANDU reactor, evolution and innovation

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the ACRTM (Advanced CANDU(1) ReactorTM) to meet today's market challenges. It is a light water tube type pressurized water reactor and is the latest evolution of CANDU technology. The design was launched to be cost-competitive with other generating sources, while building on the unique safety and operational advantages of the CANDU design. The ACR is an evolutionary design that retains the proven CANDU features delivered at Qinshan Phase III, while incorporating a set of innovative features and proven state-of-the-art technologies that have emerged from AECL's ongoing Research, Development and Demonstration programs. This approach ensures that key design parameters are well supported by existing reactor experience and R and D. The result is a design that delivers a new threshold in safety, performance and economics while retaining ample design margin. AECL has developed the enabling technologies and components for the ACR design, and has applied them to two plant sizes, ACR-700 and ACR-1000. The ACR integrates hallmark characteristics of traditional CANDU plants (e.g. horizontal pressure tubes, on power fuelling, automated reactor control systems, and dual independent shutdown systems), new innovations (e.g. state-of-the- art control room, extensive use of modular construction techniques, smaller reactor core, enriched uranium fuel), and certain PWR features (e.g. light water coolant, negative void reactivity). The ACR is designed for a high capacity factor and low operation and maintenance costs. It fully exploits the construction techniques that contributed to the impressive schedule accomplishments at Qinshan Phase III and therefore features a very short construction schedule, 40 months construction schedule (First Concrete to Fuel Loading ) for the first unit with improvements to 36 months for later units. The ACR is a true Gen-III plus product with a broad application. It has been proven to be an ideal

  7. 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

  8. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

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

  9. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

  10. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

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

  11. Pre-licensing of the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

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

  12. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

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

  15. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)

  16. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  17. 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

  18. Development of the Advanced CANDU Reactor control centre

    International Nuclear Information System (INIS)

    The next generation CANDU control centre is being designed for the Advanced CANDU Reactor (ACR) station. The design is based upon the recent Qinshan control room with further upgrades to meet customer needs with respect to high capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This evolutionary design includes the long proven functionality at several existing CANDU control centres such as the 4-unit station at Darlington, with advanced features made possible by new control and display technology. Additionally, ACR control centres address characteristics resulting from Human Factors Engineering (HFE) analysis of control centre operations in order to further enhance personnel awareness of system and plant status. Statistics show that up to 70% of plant significant events, which have caused plant outages, have a root cause attributable to the human from such sources as complex interfaces, procedures, maintenance and management practices. Consequently, special attention is made for the application of HFE throughout the ACR design process. The design process follows a systematic analytical approach to define operations staff information and information presentation requirements. The resultant human-system interfaces (HSI) such as those for monitoring, annunciation and control information are then verified and validated against the system design requirements to provide a high confidence level that adequate and correct information is being provided in a timely manner to support the necessary operational tasks. The ACR control centre provides plant staff with an improved operability capability due to the combination of systematic design and enhanced operating features. Significant design processes (i.e. development) or design features which contribute to this improved operability, include: Design Process: Project HFE Program Plan - intent, scope, timeliness and interfacing; HFE aspects of design process - procedures and instructions

  19. Development of the advanced CANDU reactor control centre

    International Nuclear Information System (INIS)

    The next generation CANDU control centre is being designed for the Advanced CANDU Reactor (ACR) station. The design is based upon the recent Qinshan control room with further upgrades to meet customer needs with respect to high capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This evolutionary design includes the long proven functionality at several existing CANDU control centres such as the 4-unit station at Darlington, with advanced features made possible by new control and display technology. Additionally, ACR control centres address characteristics resulting from Human Factors Engineering (HFE) analysis of control centre operations in order to further enhance personnel awareness of system and plant status. Statistics show that up to 70% of plant significant events, which have caused plant outages, have a root cause attributable to the human from such sources as complex interfaces, procedures, maintenance and management practices. Consequently, special attention is made for the application of HFE throughout the ACR design process. The design process follows a systematic analytical approach to define operations staff information and information presentation requirements. The resultant human-system interfaces (HSI) such as those for monitoring, annunciation and control information are then verified and validated against the system design requirements to provide a high confidence level that adequate and correct information is being provided in a timely manner to support the necessary operational tasks. The ACR control centre provides plant staff with an improved operability capability due to the combination of systematic design and enhanced operating features. Significant design processes (i.e. development) or design features which contribute to this improved operability, include: Design Process: Project HFE Program Plan - intent, scope, timeliness and interfacing; HFE aspects of design process - procedures and instructions

  20. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

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

  1. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Full text: The CANDU 6 power reactor is visionary in its approach, remarkable for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Ltd, the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980's as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first 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 21 years and are still producing electricity at peak performance and 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 customer's 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 as the 'Enhanced CANDU 6' (EC6)- which incorporates several attractive but proven features that will make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that will be incorporated in 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

  2. Recent advances in thorium fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

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

  3. Qualification of the reactor physics toolset for the design and analysis of the advanced CANDU reactor

    International Nuclear Information System (INIS)

    The qualification of reactor physics toolset for Advanced CANDU Reactor (ACR) applications is described in this paper. The qualification process follows AECL standard code validation methodology. The ACR nuclear design incorporates certain features that challenge the physics code-suite capabilities. The physics codes were first assessed, and development work required to meet these challenges was undertaken. A Validation Matrix Document was prepared to identify the physics phenomena that could arise during postulated accident events, and specify the experimental data required for code validation. Key issues related to physics modelling and code validation are also discussed. (author)

  4. Advancing the technologies of CANDU

    International Nuclear Information System (INIS)

    CANDU standard product design will continue to evolve, building upon the success of current operating units. Progressive improvements and enhancements will continue to be made to the CANDU system with heavy water moderated, pressure tube reactor technology of high neutron efficiency, based on the results of advanced technology R and D and operational experience from operating CANDU stations. The directions of development will respond to customer's requirements for economical, reliable and safe generating stations

  5. 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

  6. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  9. 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

  10. 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

  11. 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

  12. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  17. An advanced CANDU reactor with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    AECL is studying an advanced CANDU reactor concept, with supercritical steam as coolant. The coolant, being a high density gas, at a pressure above 22 MPa and temperatures above 370 deg C, does not encounter the two-phase region with its associated fuel-dryout and flow-instability problems. Increased coolant temperature leads directly to increased plant thermodynamic efficiency, thereby reducing unit energy cost through reduced specific capital cost and reduced fueling cost. The reduced coolant in-core density leads to sufficiently reduced void reactivity, so that light water becomes a coolant option. The use of supercritical water coolant also opens up the possibility of enhanced safety with a natural circulation primary flow, taking advantage of the gas expansion coefficient. To preserve neutron economy, especially at high coolant temperatures, a fuel channel that is currently being developed has a pressure tube that is thermally insulated from high-temperature coolant and is in contact with the cold heavy-water moderator. Two stages of development of a supercritical-cooled CANDU reactor were identified. The first uses conventional or near-conventional zirconium-alloy fuel cladding with coolant core-mean temperatures near 400 deg C, and the second uses advanced high-temperature fuel cladding at coolant core-mean temperatures near 500 deg C. A first-stage cost reduction of 20% from the CANDU 6 design is estimated as a result of improved thermodynamic efficiency. A large change in coolant density across the core leads to a factor 3 or 4 reduction in heavy-water inventory and a corresponding reduction in coolant void reactivity. The latter leads to improved fuel burnup and reduced demands on the safety shutdown systems. (author)

  18. 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

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. Advanced CANDU control centre

    International Nuclear Information System (INIS)

    The CANDU 9 design is based upon the 900 MWe class Darlington station in Canada, which is among the world leading nuclear power stations for capacity factor with low operation, maintenance and administration costs. The CANDU 9 design provides an advanced control centre with enhanced operations features. The advanced AECL control centre design includes the proven functionality of existing CANDU control centres, those implementable characteristics identified by systematic design combined with a human factors analysis of operations requirements and features needed to improve station operability which are made possible by the application of current technology. The design strategy is to preserve the general main control room operations staff work area as unchanged as possible to facilitate the inclusion of past features and operational experience while incorporating operability improvements. The author will present those features of the advanced CANDU control centre which facilitates improved operability capabilities. As well, aspects of the design process utilized, application of simulation technology and conclusions regarding this design approach will be reviewed

  2. Maintenance based design and equipment reliability for AECL's advanced CANDU reactor

    International Nuclear Information System (INIS)

    This paper will describe how the elements of AECL's Maintenance Based Design will enable the Advanced CANDU Reactor to sustain high equipment reliability and capacity factors over the 60-year design life of the plant. The elements of Maintenance Based Design are; 1-Design Reliable Systems,Structures and Components (SSCs); 2-Select and Procure Reliable Components; 3-Incorporate Monitoring Capabilities and Facilities for SSCs; 4-Develop Maintenance Strategies and Programs for SSCs; 5-Apply Lessons Learned From Previous Plants; 6-Incorporate Maintainability and Event Free Features in the Design; 7-Provide Enhanced Maintenance Management Information and Tools to the Customer; 8-Optimize Chemistry and Materials in the Design. All these elements will be discussed with a detailed focus on the following; Design Reliable SSCs Using the techniques outlined in INPO AP-913, Equipment Reliability Process Description, each CANDU system that has caused any station past unavailability is analyzed as part of the ACR design in order to identify the critical components and any Single Points of Vulnerability (SPVs). All SPVs are then analyzed further in order to determine if they can be practically designed out or otherwise mitigated by the design. Developing Maintenance Strategies and Programs for SSCs Equipment degradation begins as soon as a component is manufactured and accelerates during initial commissioning and eventual operation. In order to sustain high levels of equipment reliability a maintenance strategy must be developed during the design phase and be ready for implementation before the start of commissioning. This maintenance strategy is developed for all critical components using the techniques of INPO AP-913 and other best industry practices. The strategy can be expanded and customized in conjunction with a future owner. Specific examples from the current ACR-1000 design will be used to show how these elements are being implemented.

  3. CANDU: The fuel conserving reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Thorium fuel studies for CANDU reactors

    International Nuclear Information System (INIS)

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

  6. CANFLEX - an advanced fuel bundle for CANDU

    International Nuclear Information System (INIS)

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

  7. 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

  8. The advanced carrier bundle - comprehensive irradiation of materials in CANDU power reactors

    International Nuclear Information System (INIS)

    Improved methods of measuring element profiles on new CANDU fuel bundles were developed at the Sheridan Park Engineering Laboratory, and have now been applied in the hot cells at Whiteshell Laboratories. For the first time, the outer element profiles have been compared between new, out-reactor tested, and irradiated fuel elements. The comparison shows that irradiated element deformation is similar to that observed on elements in out-reactor tested bundles. In addition to the restraints applied to the element via appendages, the element profile appears to be strongly influenced by gravity and the end loads applied by local deformation of the endplate. Irradiation creep in the direction of gravity also tends to be a dominant factor. (author)

  9. Future trends in the design of CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. Engineering work is well underway to refine the design of the CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule for CANDUs of the 1990s. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components, but with an innovative new plant layout that makes it cost competitive with coal-fired plants. For the long term, work on advanced fuel cycles and major system improvements are underway ensuring that CANDU plants will stay competitive well into the next century

  10. 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.)

  11. 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)

  12. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

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

  13. 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

  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. Fuel for CANDU pressurized heavy water reactors

    International Nuclear Information System (INIS)

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

  16. 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

  17. CANDU: Shortest path to advanced fuel cycles

    International Nuclear Information System (INIS)

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

  18. 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

  19. 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)

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

    International Nuclear Information System (INIS)

    power and water systems; Other improvements to meet higher safety goals consistent with Canadian and International standards based on PSA studies; - Additional reactor trip coverage, based on refurbishment projects experience, to meet current Canadian Regulations. - Both CANDU 6 and EC6 offer flexible fuel cycle options including use of slightly enriched uranium from reprocessed LWR spent fuel, high burnup MOX fuel, thorium etc in a more efficient 43 element fuel bundle carrier called CANFLEX. - A target life up to 60 years with one mid-life refurbishment of critical equipment such as fuel channels and feeders. - Project elements have been optimized through feedback from past construction projects to arrive at an EC6 'in-service' schedule of 57 months from first concrete. Open-top construction method using a very-heavy-lift crane, concurrent construction, modularization and prefabrication and use of advanced computer technologies to minimize interferences are the key contributing elements for achieving this schedule. - State-of-the-art electronic tools for engineering, safety, licensing, procurement, drawing and project management are integrated to provide complete document control during all phases of the project, including construction and commissioning. This information, in electronic format, will be turned over to the Owner for operational and configuration management needs during plant life. - Advanced MACSTOR design for efficient dry spent fuel storage with optimized space usage. Summary. Capitalizing on the proven features of CANDU technology, AECL has designed the EC6 to be competitive with all forms of energy, including nuclear, while achieving high safety and performance standards. (author)

  1. Economic and system aspects of CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

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

  2. ASTEC code adaptability to CANDU type reactors

    International Nuclear Information System (INIS)

    ASTEC integral code is dedicated for severe accident (SA) analysis, mainly for PWR type reactors. In the last years, in the FP-6 NoE SARNET project framework, important efforts were focused on the extension of the ASTEC use to other reactors: WWER, RBMK, BWR and CANDU. The use of ASTEC at CANDU type reactors introduces many difficulties especially for the core degradation phenomena. The paper shows some results obtained in exploratory calculation with the modules SOPHAEROS, CPA, IODE, CESAR and DIVA in order to investigate the possibility to use or to adapt the models at CANDU type reactors. An important part of the paper is focused on the models for CANDU core degradation to be implemented in DIVAC module. (authors)

  3. The next generation of CANDU reactor: evolutionary economics

    International Nuclear Information System (INIS)

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

  4. Thermalhydraulic safety analysis of the Candu reactor

    International Nuclear Information System (INIS)

    The thermalhydraulic analysis requirements for the safety and licensing of the CANDU reactor are outlined. The unique features of the CANDU design are first described, and the specialized analysis requirements for the reactor are identified. Thermalhydraulic codes used to perform the analysis are presented and the experimental test programs used to validate the codes are described. The paper concludes with future plans for the experimental test programs, code development, and code validation. (authors). 11 figs., 1 tab., 19 refs

  5. 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

  6. 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

  7. 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)

  8. fuel management in candu reactors: RFSP code

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. 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)

  11. 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.

  12. 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

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

    International Nuclear Information System (INIS)

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

  14. ROP margin improvement options for CANDU reactors

    International Nuclear Information System (INIS)

    Highlights: • ROP aging challenge related to power derating is addressed. • ROP detector layout optimization and HSP re-classification are the options explored to mitigate power derating challenge. • Significant improvement in the ROP TSP value for an aged CANDU reactor can be realized from implementing these options. - Abstract: Over the past few years, Candu Energy Inc. (a wholly owned subsidiary of SNC-Lavalin Inc., which acquired the assets of Atomic Energy of Canada Limited’s Commercial Reactor Division) has been continuously developing and evaluating various options to improve the regional overpower protection (ROP) margin in aged CANDU 600 MW (CANDU 6®) reactors. This paper presents results from applying a couple of margin improvement options to a generic aged CANDU 6 reactor, namely ROP detector layout optimization and application of a revised handswitch position designation. Application of these options requires no change to the ROP analysis methodology, statistical approach or acceptance criterion. As such, any increase in ROP margin associated with these options carries little or no licensing risk and are not expected to require more than one standard outage to implement

  15. The CANDU Reactor System: An Appropriate Technology.

    Science.gov (United States)

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  16. Operator companion for CANDU reactors

    International Nuclear Information System (INIS)

    As CANDU nuclear power plants become more complex, and are operated under tighter constraints and for longer periods between outages, plant operations staff will have to absorb more information to correctly and rapidly respond to upsets. A development program is underway at Atomic Energy of Canada Limited to use expert systems and interactive media tools to assist operations staff of existing and future CANDU plants. The complete system for plant information access and display, on-line advice and diagnosis, and interactive operating procedures is called the Operator Companion. A prototype, consisting of operator consoles, expert systems and simulation modules in a distributed architecture, is currently being developed to demonstrate the concepts of the Operator Companion. (author). 5 refs, 2 figs

  17. The pressure tubes in the CANDU power reactor

    International Nuclear Information System (INIS)

    Nuclear power reactors using zirconium alloy pressure tubes generate electricity in several countries. In Ontario CANDU reactors generate about 30 percent of the electricity produced in the province. The pressure tubes of the first five CANDU reactors were made of cold-worked Zircaloy-2, an alloy of zirconium and tin developed by the US Navy. In 1958 the USSR published information on a Zr-2.5 wt percent Nb alloy, in which the Nb promotes stabilization of the β phase, thus presenting opportunities of exploiting metallurgically strong pressure tubes analogous to the heat-treatable α-β titanium alloys. After two reactors using Zr-2.5 wt percent Nb in a quenched and aged condition were constructed, an extensive development program on cold-worked Zr-2.5 wt percent Nb pressure tubes resulted in their becoming the reference tubes for all future CANDU reactors. Pressure tubes of Zr-3.3 wt percent Sn-0.8 wt percent Nb-0.8 wt percent Mo (Excel) are in an advanced state of development. These tubes will be used in an annealed condition; projections show that they will have improved dimensional stability over the lifetime of the reactors. These improvements result from experimental programs leading to an understanding of the relationship between microstructures and fabrication variables and effects of the environment during service in nuclear reactors. (author)

  18. 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

  19. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

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

  20. 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)

  1. 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

  2. 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

  3. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

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

  4. 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.

  5. New flux detectors for CANDU 6 reactors

    International Nuclear Information System (INIS)

    CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

  6. 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.

  7. 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

  8. 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)

  9. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments. The major application is in the health care industry where irradiators are used to sterilize single use medical products. These irradiators are designed and built by MDS Nordion and are used by manufacturers of surgical kits, gloves, gowns, drapes and other medical products. The irradiator is a large shielded room with a storage pool for the cobalt-60 sources. The medical products are circulated through the shielded room and exposed to the cobalt-60 sources. This treatment sterilizes the medical products which can then be shipped to hospitals for immediate use. Other applications for this irradiation technology include sanitisation of cosmetics, microbial reduction of pharmaceutical raw materials and food irradiation. The cobalt-60 sources are manufactured by MDS Nordion in their Cobalt Operations Facility in Kanata. More than 75,000 cobalt-60 sources for use in irradiators have been manufactured by MDS Nordion. The cobalt-60 sources are double encapsulated in stainless steel capsules, seal welded and helium leak tested. Each source may contain up to 14,000 curies. These sources are shipped to over 170 industrial irradiators around the world. This paper will focus on the MDS Nordion proprietary technology used to produce the cobalt-60 isotope in CANDU reactors. Almost 55 years ago MDS Nordion and Atomic Energy of Canada developed the process for manufacturing cobalt-60 at the Chalk River Labs, in Ontario, Canada. A cobalt-59 target was introduced into a research reactor where the cobalt-59 atom absorbed one neutron to become cobalt-60. Once the cobalt-60 material was removed from the research reactor it was encapsulated in stainless steel and seal welded using a Tungsten Inert Gas weld. The first cobalt-60 sources manufactured using material from the Chalk River Labs were used in cancer

  10. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

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

  11. CANDU improvement

    International Nuclear Information System (INIS)

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

  12. Role of water lubricated bearings in Candu reactors

    International Nuclear Information System (INIS)

    During the twentieth century a great emphasis was placed in understanding and defining the operating regime of oil and grease lubricated components. Major advances have been made through elastohydrodynamic lubrication theory in the quantifying the design life of heavily loaded components such as rolling element bearings and gears. Detailed guidelines for the design of oil and grease lubricated components are widely available and are being applied to the successful design of these components. However similar guidelines for water lubricated components are either not available or not well documented. It is often forgotten that the water was used as a lubricant in several components as far back as 1884 B.C. During the twentieth century the water lubricated components continued to play a major role in some high technology industries such as in the power generation plants. In CANDU nuclear reactors water lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication. This paper presents some examples of the major water lubricated components in a CANDU reactors. Major part of the paper is focused on presenting an example of successful operating history of water lubricated bearings used in the HT pumps are presented. Both types of bearings have been qualified by tests for operation under normal as well as under more severe postulated condition of loss-of-coolant-accident (LOCA). These bearings have been designed to operate for the 30 years in the existing CANDU 6 (600 MW) reactors. However for the next generation of CANDU 6 reactors which go into service in the year 2003, the HT pump bearing life has been extended to 40 years. (author)

  13. R and R programmes to advance CANDU technology

    International Nuclear Information System (INIS)

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

  14. Operational support analysis for CANDU reactors

    International Nuclear Information System (INIS)

    Enormous supporting analyses are required for each reactor power plant. Most supporting analyses are routine but some are urgent when dealing with unanticipated events. The operational support analysis involves assessment of trip parameters, defining the safety operating envelop, providing information for updating operational manuals and procedures, or investigating the causes of abnormal transients when plant data are not available to provide the information. The modules in the TUF code dealing with reactor controllers, thermal-hydraulics, reactor physics, heat conduction and heat transfer, system components and other auxiliary models for CANDU reactors are discussed. To qualify TUF as an analytical tool for operational support analysis, plant data have been used to validate TUF. To illustrate that, simulation of a Class IV power failure event at Darlington NGS is presented

  15. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

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

  16. 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.

  17. An approach to neutronics analysis of candu reactors

    International Nuclear Information System (INIS)

    An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)

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

    International Nuclear Information System (INIS)

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

  19. 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)

  20. Nuclear Archeology for CANDU Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, Bryan L [ORNL

    2011-01-01

    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-07-01

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

  4. 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

  5. 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.

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

    International Nuclear Information System (INIS)

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

  7. 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).

  8. Next generation CANDU plants

    International Nuclear Information System (INIS)

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

  9. 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)

  10. 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)

  11. 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)

  12. 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)

  13. 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)

  14. Tritium source identification in CANDU reactors

    International Nuclear Information System (INIS)

    Very small amounts of tritiated heavy water may escape from the moderator and heat-transport systems of CANDU reactors during maintenance and normal operation. Through comprehensive tritium management, the impact of this leaked heavy water on operating personnel and the environment can be controlled. One useful management technique is source identification, a set of methods for locating very small heavy-water leaks. This technique permits an operator to optimize plant performance, adapting to changes in plant conditions. Various identification methods are available, including local-hazard monitoring, tritium mapping and the correlation of measured hazards with emissions and hazards in other areas. The suitability of each method depends on the management objectives. In this report, each of these methods is reviewed and applications discussed. (author)

  15. Thermalhydraulic safety analysis of the Candu reactor

    International Nuclear Information System (INIS)

    The thermalhydraulic analysis requirements for the safety and licensing of the Candu reactor are outlined. The unique features of the Candu design are first described, and the specialized analysis requirements for the reactor are identified. Thermalhydraulic codes used to perform the analysis are presented and the experimental test programs used to validate the codes are described. The paper concludes with future plans for the experimental test programs, code development, and code validation. Future experimental work will largely focus on improving our understanding of the interaction of multiple parallel heated channels under upset conditions. This is, of course, related to the blowdown and refill thermal-hydraulics of full-size flow headers connecting the feeders. An increasing emphasis is currently being placed on refining the instrumentation for our test facilities. An Instrument Development Program has been recently implemented to provide instrumentation not currently available commercially. For example, conductivity probes are being developed to accurately measure the level in the RD-14M headers. As well, neutron scattering tomography is being evaluated to measure the void and void distribution in the heated channels of RD-14M and the CWIT facility. To facilitate easy access to experimental data, a program has been initiated to develop a fully relational data base of thermalhydraulic data obtained from our experimental programs. Software is being developed to display the information in a number of formats, including an 'animated' replay of the experiment, to aid the analyst in interpreting the experimental data. Future code development will also focus on accurately predicting the behaviour of the Candu header/feeder system under loss-of-coolant accident conditions. A multi-dimensional representation of the header will be required. At the same time, the computational efficiency of the codes will have to increase to handle the large number of parallel channels

  16. 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.

  17. 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)

  18. Thorium fuel-cycle studies for CANDU reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 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. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Seok, Ho Cheon; Na, Yeong Hwan; Seok, Soo Dong; Lee, Bo Uk; Kwak, Ho Sang; Kim, Bong Ki; Kim, Seok Nam; Min, Byeong Joo; Park, Jong Ryunl; Shin, Jeong Cheol; Lee, Kyeong Ho; Lee, Dae Hee; Lee, Deuk Soo; Lee, Yeong Uk; Lee, Jeong Yang; Jwon, Jong Seon; Jwon, Chang Joon; Ji, Yong Kwan; Han, Ki Nam; Kim, Kang Soo; Kim, Dae Jin; Kim, Seon Cheol; Kim, Seong Hak; Kim, Yeon Seung; Kim, Yun Jae; Kim, Jeong Kyu; Kim, Jeong Taek; Kim, Hang Bae; Na, Bok Kyun; Namgung, In; Moon, Ki Hwan; Park, Keun Ok; Shon, Ki Chang; Song, In Ho; Shin, Ji Tae; Yeo, Ji Won; Oh, In Seok; Jang, Ik Ho; Jeong, Dae Won; Jeong, Yong Hwan; Ha, Jae Hong; Ha, Jeong Koo; Hong, Hyeong Pyo; Hwang, Jeong Ki [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    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.

  3. Romanian progress in the advanced CANDU fuel manufacturing

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  4. 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)

  5. Power coefficient of reactivity in CANDU 6 Reactors

    International Nuclear Information System (INIS)

    The Power Coefficient of Reactivity (PCR) measures the change in reactor core reactivity per unit change in reactor power and 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 the inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity in all operating conditions are maintained within a safe range. The CANDU reactor design takes advantage of the inherent nuclear characteristics of small reactivity coefficient, minimal excess reactivity and very long prompt neutron lifetime to mitigate the magnitude of the demand on the engineered systems for controlling reactivity. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with its design characteristics, such that the overall design of the reactor does not depend on the sign of the PCR. This is a contrast to other reactor design concepts which are dependent on a PCR which is both large and negative in the design of their engineered systems for controlling reactivity. It will be demonstrated that during a Loss of Regulation Control (LORC) event, the impact of having a positive power coefficient, or of hypothesizing a PCR larger than that estimated for CANDU, has no significant impact on the reactor safety. Since the CANDU 6 PCR is small, its role in the operation or safety of the reactor is not significant

  6. CANDU 6 reactor core physics and site physicist role

    International Nuclear Information System (INIS)

    The CANDU reactor is fuelled on-line. There is thus an on-going need for fuel and core management which is supported by an on-site Reactor Physics group. The author outlines the role of the on-site Physics group at the Point Lepreau Generating Station. This role covers Production, Technical as well as Safety and Compliance aspects

  7. Investigation of CANDU reactors as a thorium burner

    International Nuclear Information System (INIS)

    Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in the form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates the prospects of exploiting the rich world thorium reserves in CANDU reactors. Two different fuel compositions have been selected for investigations: (1) 96% thoria (ThO2) + 4% PuO2 and (2) 91% ThO2 + 5% UO2 + 4% PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the reactor criticality k ∞ and the burn-up values of the reactor have been pursued by full power operation for >∼8 years. The reactor starts with k ∞ = ∼1.39 and decreases asymptotically to values of k ∞ > 1.06, which is still tolerable and useable in a CANDU reactor. The reactor criticality k ∞ remains nearly constant between the 4th year and the 7th year of plant operation, and then, a slight increase is observed thereafter, along with a continuous depletion of the thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn-up can be achieved with the same fuel (>160,000 MW D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burned up, provided that the fuel rods could be fabricated to withstand such high burn-up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically

  8. 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).

  9. 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.

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

    International Nuclear Information System (INIS)

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

  11. R and D directions for the development of CANDU reactors

    International Nuclear Information System (INIS)

    Full text: AECL is carrying out a comprehensive R and D programme to advance all aspects of CANDU reactor technology. These programs are focusing on three main strategic directions: improved economics, enhanced safety, and fuel cycle flexibility. R and D areas include fuel cycle development, heavy water technology, fuel channel development, safety technology, control and instrumentation, reactor chemistry, systems and components, and health and environment. In each case, the R and D programs have short, medium, and long-term goals to achieve the overall strategic directions. Most of the programs seek to further develop and exploit some of the unique characteristics of pressurized heavy water reactors. Examples of this include high neutron economy and on-power fueling which allow several different fuel cycles, the presence of large water heat sinks for enhanced safety, and modular components that can be easily replaced for plant life extension. This presentation reviews AECL's product development directions and the R and D programs that have been begun for their development

  12. 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

  13. Mathematical models and computer code ELESIM used for CANDU reactors

    International Nuclear Information System (INIS)

    Candu reactors are used in many countries all over the world for power generation. This is because the reactors use natural uranium fuel, with simple design, which permits local manufacturing of the reactor components, in addition to safety in operation. The operation of Candu reactors is accompanied by highly sensitive automatic control loops, which in turn are accompanied by using a set of computer codes to simulate the components of the reactor. One of those codes is ELESIM, which is a computer program for simulating the behaviour of fuel element under the normal operating conditions. In this report, the most important phenomena modelled in ELESIM are manipulated in accordance to their dependence on each other. When necessary the mathematical model used in each item is given, while the equations used in the code is represented in appendix. 6 FIG

  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. 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.

  16. 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)

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

    International Nuclear Information System (INIS)

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

  18. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

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

  19. Design research for accident prevention in CANDU reactor

    International Nuclear Information System (INIS)

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

  20. 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

  1. Controlling radiation fields in CANDU reactors using chemical decontamination technologies

    International Nuclear Information System (INIS)

    Radiation dose to personnel during major maintenance and reactor refurbishment of CANDU reactors can be controlled using chemical decontamination technologies. Technologies that have, and can be applied in CANDU reactors include; sub- and full-system decontaminations of the heat transport system using the CAN-DECON, CAN-DEREM and CAN-DEREM Plus processes, and removal of Sb-122 and Sb-124 from the reactor core using hydrogen peroxide. CAN-DECON is a dilute chemical decontamination process that employs ion-exchange technology to continuously remove dissolved metals and radionuclides and regenerate the components of the CAN-DECON formulation. Qualification of the CAN-DECON process, equipment requirements, process effectiveness, recent process improvements and future directions are discussed. Radioantimony deposited on in-core surfaces can be released into the HTS coolant by air ingress during maintenance. At Gentilly-2, where large amounts of in-core antimony are present, these releases have resulted in increased radiation fields around the reactor, making outage dose planning difficult and contributing significantly to the radiation exposure of maintenance personnel. An antimony removal process developed by KWU for PWR's and adapted to meet CANDU specific conditions, has been successfully applied at Gentilly-2. Optimization of process conditions, and improvements in the in-core antimony removal process are described. (author)

  2. Controlling radiation fields in CANDU reactors using chemical decontamination technologies

    International Nuclear Information System (INIS)

    Radiation dose to personnel during major maintenance and reactor refurbishment of CANDU reactors can be controlled using chemical decontamination technologies. Technologies that have, and can be applied in CANDU reactors include; sub- and full-system decontamination of the heat transport system using the CAN-DECON CAN-DEREM and CAN-DEREM Plus processes; and removal of Sb-122 and Sb-124 from the reactor core using hydrogen peroxide. CAN-DECON is a dilute chemical decontamination process that employs ion-exchange technology to continuously remove dissolved metals and radionuclides and regenerate the components of the CAN-DECON formulation. Qualification of the CAN-DECON process, equipment requirements, process effectiveness, recent process improvements and future directions are discussed. Radioantimony deposited on in-core surfaces can be released into the HTS coolant by air ingress during maintenance. At Gentilly-2, where large amounts of in-core antimony are present, these releases have resulted in increased radiation fields around the reactor, making outage dose planning difficult and contributing significantly to the radiation exposure of maintenance personnel. An antimony removal process developed by KWU for PWR's and adapted to meet CANDU specific conditions, has been successfully applied at Gentilly-2. Optimization of process conditions, and improvements in the in-core antimony removal process are described. (author)

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

    International Nuclear Information System (INIS)

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

  4. 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)

  5. 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)

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

    International Nuclear Information System (INIS)

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

  7. Review of Regulations on Continued Operation for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Sang Kyu; Yoo, Kun Joong; Kim, Hyun Koon; Ryu, Yong Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Roh, Heui Young; Jin, Tae Eun [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2007-10-15

    The first CANDU type reactor, Wolsong Unit 1, has been operating for twenty four years since the commencement of its commercial operation in 1983 and its lifetime will be completed until end of 2012. Hence the licensee, KHNP, is considered a continued operation for Wolsong Unit 1 in economic point of view. Regarding to the license of the continued operating of nuclear power plants including CANDU reactors, a regulatory body is developing the regulatory requirements on continued operation for reviewing the technical requirements of safety assessment and management of aging for structures, systems and components (SSC) in the nuclear power plants. Regarding to this, in this paper the review contents are described and general review results are presented.

  8. Review of Regulations on Continued Operation for CANDU Reactors

    International Nuclear Information System (INIS)

    The first CANDU type reactor, Wolsong Unit 1, has been operating for twenty four years since the commencement of its commercial operation in 1983 and its lifetime will be completed until end of 2012. Hence the licensee, KHNP, is considered a continued operation for Wolsong Unit 1 in economic point of view. Regarding to the license of the continued operating of nuclear power plants including CANDU reactors, a regulatory body is developing the regulatory requirements on continued operation for reviewing the technical requirements of safety assessment and management of aging for structures, systems and components (SSC) in the nuclear power plants. Regarding to this, in this paper the review contents are described and general review results are presented

  9. 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)

  10. 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

  11. CANDU load following test in ICN research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbas, S.; Palleck, S. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Hohoianu, G.; Parvan, M. [Inst. for Nuclear Research, ICN (Romania)

    2008-07-01

    To study the performance of CANDU fuel under Load Following (LF) conditions, a CANDU 6 type fuel element with 8.0 wt% {sup 235}U enrichment was irradiated in the 14 MW TRIGA materials testing reactor at the Institute for Nuclear Research (ICN) in Romania. This experiment was developed under the INR-AECL Memorandum for Co-operation in research and development of nuclear energy and technology. The fuel element underwent a successful demonstration of LF capability, where the fuel element withstood 200 daily cycles from 27 to 54 kW/m (average element linear power), as well as additional ramps due to reactor trips and restarts during the test period. The fuel element underwent a series of post-irradiation destructive and nondestructive examinations after the LF test irradiation. No performance or integrity issues were observed. This paper presents a description of the test facility, details of the test irradiation conditions and the post-irradiation examination results with discussion on their relation to CANDU fuel performance and integrity. (author)

  12. 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)

  13. 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)

  14. Explaining the absence of Co-58 radiation fields around CANDU reactor primary circuit

    International Nuclear Information System (INIS)

    Radiation fields from Co-58 are rarely detected in CANDU plants. For example, Ge(Li) surveys of the Inconel 600 steam generators at some CANDU plants may show radiation attributed to Co-58 only early in plant life, and most artefacts removed from the primary circuit later in plant operation show no Co-58 present. However, Pressurized Water Reactor plants experience relatively large fields from Co-58 on their isothermal piping, e.g., steam generator channel head, and steam generators tube sampling programs do show deposits in the tubes with significant Co-58 compared to other radionuclides such as Co-60. CANDU reactors have high concentrations of dissolved iron due to the extensive use of carbon steel for the isothermal piping, e.g., feeders, headers, and steam generator channel heads. A dissolved iron transport diagram that was proposed recently for the primary circuit of CANDU plants has been validated by comparison of predicted deposit weights with plant deposit data from various components. One feature of the diagram is dissolved iron precipitation inside the steam generators tubes. An hypothesis is advanced here in which precipitating dissolved iron is proposed to occlude dissolved nickel. This removal mechanism may prevent the solubility of dissolved nickel from being exceeded anywhere around the primary circuit. In particular, this mechanism could avoid NiO precipitation in the core and the generation of large quantities of Co-58. Using this mechanism along with the known solubility behaviour of NiO with temperature, a dissolved nickel transport diagram has been proposed for CANDU plants. (authors)

  15. Evaluation of the applicable reactivity range of a reactivity computer for a CANDU-6 reactor

    International Nuclear Information System (INIS)

    Recently, a CANDU digital reactivity computer system (CDRCS) to measure the worth of the liquid zone controller in a CANDU-6 was developed and successfully applied to a physics test of refurbished Wolsong Unit 1. In advance of using the CDRCS, its measurable reactivity range should be investigated and confirmed. There are two reasons for this investigation. First, the CANDU-6 has a larger reactor and smaller excore detectors than a general PWR and consequently the measured reactivity is likely to reflect the peripheral power variation only, not the whole core. The second reason is photo neutrons generated from the interaction of the moderator and gamma-rays, which are never considered in a PWR. To evaluate the limitations of the CDRCS, several tens of three-dimensional steady and transient simulations were performed. The simulated detector signals were used to obtain the dynamic reactivity. The difference between the dynamic reactivity and the static worth increases in line with the water level changes. The maximum allowable reactivity was determined to be 1.4 mk in the case of CANDU-6 by confining the difference to less than 1%.

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

    International Nuclear Information System (INIS)

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

  17. EC6TM - Enhanced Candu 6TM reactor safety characteristics

    International Nuclear Information System (INIS)

    The EC6 is a 740 MWe-class natural-uranium-fuelled, heavy-water-cooled and -moderated pressure-tube reactor, which has evolved from the eleven (11) CANDUR 6 plants operating in five countries (on four continents). CANDU 6 has over 150 reactor-years of safe operation. The most recent CANDU 6 - at Qinshan, in China - is the Reference Design for EC6. The EC6 shares many inherent, passive and engineered safety characteristics with the Reference Design. However EC6 has been designed to meet modern regulatory requirements and safety expectations. The resulting design changes have improved these safety characteristics, and this paper provides a convenient summary. The paper addresses the safety functions of reactivity control, heat removal, and containment of radioactive material. For each safety function, the EC6 characteristics are categorized as inherent, passive, or engineered. The paper focuses mostly on the first two. The Enhanced CANDU 6 uses an appropriate mix of passive, inherent, and engineered safety functions. Reactivity transients are generally slow, mild and inherently limited due to the natural uranium core and use of on-power refuelling. Only the coolant void coefficient can cause a large reactivity insertion, particularly in a large LOCA. This is mitigated by the long prompt neutron lifetime and the large delayed neutron fraction, and terminated by either of the two shutdown systems. For EC6, the large LOCA power transient has been reduced significantly by speeding up the slower of the two shutdown systems. Redundant shutdown and the LOCA power pulse improvements mitigate the limiting large positive reactivity insertion. Decay heat removal shows a very high component of passive safety, from thermo-siphoning in the Reactor Coolant System to passive heat removal in severe accidents via the moderator or reactor vault. The latter two can maintain the fuel in a more predictable and favourable geometry than 'core on the floor'. The containment structure is

  18. Optimization of decontamination strategy for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Theoretical models of the decontamination process are developed and combined with an existing model of 60Co production in CANDU PHW reactors to predict the effects of decontamination on long term 60Co build-up in reactor primary heat transport systems. The effects of decontamination interval, decontamination factor, and post-decontamination corrosion release are calculated. An optimum decontamination strategy for a Pickering G.S. type reactor is developed on the basis of a cost-benefit analysis. This study indicates that the optimum decontamination interval is approximately six years. This optimum interval is relatively insensitive to variations in the costs of personnel exposure, the cost of a decontamination, the decontamination factor, and the post-decontamination corrosion model used. (author)

  19. CANDU 9 - Overview

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-15

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

  1. 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

  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. Time-Average Calculation using FEM in a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Eun Hyun; Park, Joo Hwan; Song, Yong Man; Lee Chung Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

    2012-05-15

    To get a much accurate result and to be sure about the calculated reactor physics value, new code system which is appropriate to the CANDU reactor and has high fidelity is required. This study here is to understand and analyze the existing code system, WIMS-RFSP. Because the FEM codes used here can calculate multiplication factor, group flux, channel power easily with cross section data from WIMS and geometrical data from GMSH, the results of FEM are good examples to compare with RFSP results. With the comparison process itself and numerical experiments, it is expected that the basis of new code system become abundant. Time-average module is mainly discussed with regular process in RFSP

  4. Time-Average Calculation using FEM in a CANDU Reactor

    International Nuclear Information System (INIS)

    To get a much accurate result and to be sure about the calculated reactor physics value, new code system which is appropriate to the CANDU reactor and has high fidelity is required. This study here is to understand and analyze the existing code system, WIMS-RFSP. Because the FEM codes used here can calculate multiplication factor, group flux, channel power easily with cross section data from WIMS and geometrical data from GMSH, the results of FEM are good examples to compare with RFSP results. With the comparison process itself and numerical experiments, it is expected that the basis of new code system become abundant. Time-average module is mainly discussed with regular process in RFSP

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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

  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. 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

  10. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. 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.

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

    International Nuclear Information System (INIS)

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

  15. 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)

  16. 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

  17. 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

  18. Radiological Characteristics of decommissioning waste from a CANDU reactor

    International Nuclear Information System (INIS)

    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 1016 Bq, 2.09 x 103 W, 5.31 x 1014 m3-water, 4.69 x 105 kg, and 7.38 x 101 m3, 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-15

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

  20. CANDU 9 safety improvements

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Economic potential of advanced fuel cycles in CANDU

    International Nuclear Information System (INIS)

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

  3. 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.

  4. Advanced reactor design and safety objectives. The heavy water reactor perspective

    International Nuclear Information System (INIS)

    The development objectives of advanced heavy water reactors (AHWRs) should be guided by the requirements of the operating utilities. The paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for AHWR systems are based on systematic reviews of current operating CANDU stations. Hence these requirements represent those for the evolutionary development of AHWR systems, factoring in the considerable operating experiences of the CANDU stations. The requirements for the new HWR designs can be summarized under economic objectives, safety objectives, operational objectives and other utility requirements. 2 refs

  5. U.S. utilities may seek CANDU reactors

    International Nuclear Information System (INIS)

    The economic advantage of the CANDU fuel cycle, using natural uranium, over the LWR cycle, using enriched uranium, is outlined. It is argued that President Carter's decision to shelve the fast breeder and plutonium recycle should make CANDU more attractive to American utilities. Possible effects of American licensing on the Canadian industry are discussed. Another possibility would be American assistance in developing the CANDU OCR thorium near-breeder. (N.D.H.)

  6. 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)

  7. Advanced reactor design and safety objectives - The heavy water reactor perspective

    International Nuclear Information System (INIS)

    This paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for Advanced Heavy Water reactor Systems are based on systematic reviews of current operating CANDU stations to identify opportunities for generic improvements in reliability, operability, maintainability and to address emerging licensing or safety issues. Hence these requirements represent those for the evolutionary development of the Advanced Heavy Water Reactor systems factoring in the considerable operating experience of the CANDU stations. This evolutionary approach to the development of advanced heavy water reactors will be consistent with a philosophy of minimizing the risk to future reactor owners whose requirements are for a reliable, low cost unit

  8. 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)

  9. 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.

  10. Tritium in heat transport and moderator systems of CANDU reactors

    International Nuclear Information System (INIS)

    The production rates of tritium in the heavy-water moderator and heat transport systems of CANDU reactors are calculated from the neutron fluxes generated in reactor-physics analyses of the lattice cell and radiation-physics analyses of the primary radial shields. The concentrations of tritium activity in the heavy water can then be calculated assuming a simple build-up of a decaying radioactive species. This simple treatment has been compared with tritium concentrations measured in the domestic CANDU 6 stations. These comparisons show that the predicted concentrations need to take account of the heavy water management practices of the stations as well as the operating history of the plant. The success of the individual station at segregating the heat transport and moderator heavy water systems also has some impact on the tritium concentrations in the heat transport systems. There is some evidence to show that there are different levels of success at achieving separation. The current build-up of tritium in the heavy water systems shows no evidence for significant tritium production from helium-3, the decay product of tritium, via the 3He(n,p)3H-reaction. The paper introduces explanations for the absence of this effect. The agreement between the predicted concentrations and the measured concentrations is an indirect validation of the thermal fluxes calculated in the lattice and radiation physics codes. The agreement suggests that a comparison between predicted and measured tritium levels in the heavy water systems would allow an operator to monitor the success of the plant at maintaining segregation between the heavy water systems. The paper presents details of the flux levels used to predict tritium production, typical rates of heavy water loss from the plant and the operating histories over a 15-year period. The paper indicates how heavy water recoveries having tritium concentrations between those of the heat transport and moderator systems affect the tritium

  11. 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

  12. CANDU advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

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

  13. 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)

  14. Fuel Management in Candu Reactors Using Tabu Search

    International Nuclear Information System (INIS)

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

  15. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACRTM-1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  16. Computer based core monitoring system for an operating CANDU reactor

    International Nuclear Information System (INIS)

    The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same(0.008%), which showed that the CCMS could monitor core behaviors well

  17. Hydride blister formation simulation in Candu type reactors

    International Nuclear Information System (INIS)

    We have developed a computer code for the probability study of hydride blister formation in pressure tubes named BLIFO. The basic hypothesis of the model are: the pressure tube is divided into five areas according to the existence of four garter springs. For each area the probability of blister formation is the probability of the hydrogen content exceeding a critical threshold when contact tube is present; the probability of a blister in a tube is the OR combination of the probabilities of a blister in each area; the tube contact is a function of the garter springs location, and the time; the critical hydrogen threshold is sorted over the areas within the pressure tube; hydrogen pick-up rate was sorted with a Gaussian distribution; the initial hydrogen content values for each tube were measured before the ensamble and they are used in the code. For Embalse evaluation, we build up a subroutine that simulate Gaussian distribution using the parameters of a typical nuclear power Candu reactor garter spring distribution. (author)

  18. An optimization strategy for refueling simulation of a Candu reactor

    International Nuclear Information System (INIS)

    The AUTOREFUEL program can perform a large amount of refueling simulations within a short period, which is a strong advantage especially when a series of sensitivity calculations is needed. It also has the capability to keep the maximum channel and bundle powers less than the license limits. However, there is a chance that zone controller unit (ZCU) level exceeds the typical operating range during the refueling simulation because of incomplete modeling of the relationship between zone power and ZCU levels. In order to reserve a large enough operating margin of the reactor, the ZCU level should be kept within the typical operating range. Therefore, a deterministic method has been needed to accurately estimate the ZCU level during the refueling operation, which enables the optimum refueling channel selection. In this study, a fuel management method is proposed for the selection of refueling channels using the constraint on the ZCU level. The estimated ZCU level is used as a primary index for optimum channel selection. In this study, a generalized perturbation theory (GPT) program GENOVA, which was developed to perform the deterministic estimation of the ZCU level change due to a perturbation, is briefly described. Then, the refueling channel selection strategy proposed in this study is explained and the result of application to natural uranium CANDU-6 core refueling simulation is presented. (authors)

  19. 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.

  20. 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

  1. Assessment of the WIMS-D5 applicability to CANDU reactors

    International Nuclear Information System (INIS)

    The purpose of this study is to develop a WIMS/CANDU code for a lattice calculation on the basis of WIMS-D5 code for the safety analysis of CANDU reactors. To assess the WIMS-D5 applicability to a CANDU reactor, a lattice model was developed For CANDU-6 reactors at the Wolsong site. As for the benchmark of the code validation, the code-to-code comparison was performed between the WIMS-D5 code with both the 69- and 172-energy groups of ENDF/B-VI nuclear data library and the WIMS-AECL code with the 89-energy group. The comparison studies of the reactor physics parameters such as void reactivity', coolant/fuel/moderator temperature coefficients were conducted with the change of the internal isotopic composition due to the fuel burning-up using both WIMS-AECL and POWDERPUFS-V (PPV) codes. The results show that the present results between the WIMS-D5 code and WIMS-AECL code agreed well with those of the PPV at the beginning of the fuel horn-up phase. As burning-up progresses, the results of WIMS-D5 show a large deviation from those of PPV for CANDU 6 reactors. (author)

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

    International Nuclear Information System (INIS)

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

  3. 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.

  4. 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

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

    International Nuclear Information System (INIS)

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

  6. The next generation CANDU 6

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. 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)

  9. Design features of Candu 9

    International Nuclear Information System (INIS)

    Thirty-two nuclear generating units with an aggregate installed capacity of 19,119 MWe worldwide are equipped with heavy water moderated and cooled pressure tube reactors of the Canadian Candu line. The list includes nine reactors of the 700 MWe category, and twelve reactors of the 900 MWe category in the Candu 6 series. On the basis of the 900 MWe units, Atomic Energy of Canada Ltd. (AECL) developed the advanced Candu 9 series by evolution. This series has been designed for a service life of sixty years. The use of modular, simplified units and systems in the Candu 9 design is to shorten the planning and construction phase, increase safety, and improve plant operation. AECL will offer this reactor on the world market, first to its customers in (South) Korea, which is one of the reasons why the safety parameters have been chosen especially under the aspect of seismic characteristics. (orig.)

  10. Next generation CANDU plants

    International Nuclear Information System (INIS)

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

  11. CANDU-3: Features of next generation CANDU

    International Nuclear Information System (INIS)

    CANDU 3, with a net electrical output of 450 MW, is the latest and smallest version of the CANDU power system. Significant innovation built on proven CANDU reactor technology is the basis of the CANDU 3 design. This, coupled with the commitment to reduce plant cost, increase performance capacity factor, enhance safety features and incorporate technological improvements, makes CANDU 3 an advanced, world class product. This paper describes the following CANDU 3 features: A station layout to provide a flexible construction sequence, good system separation and ease of maintenance and operation. An up-front engineering and licensing process prior to beginning construction. Enhancement and simplification of safety features with extended time scales for response, thus limiting reliance on operator action for accident mitigation. Enhanced design capabilities through the use of the latest Computer Aided Design and Drafting (CADD) technology. A 38 month construction schedule achieved by using modularization and open-top construction and installation techniques. A more passive containment system incorporating a steel liner and eliminating the need for active spray. A grouping and separation philosophy for maximum protection of redundant safety systems. Ease of equipment qualification and maximum protection of critical components. Replacement of centralized control and monitoring computers with a redundant distributed control system and modern plant display system. A consistent, logical approach to control room design founded on human factors, automation and event management. (author). 6 refs, 3 figs

  12. 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.

  13. 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

  14. CANDU 300

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. 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)

  17. Control of reactor inlet header temperature (RIHT) rise in CANDU

    International Nuclear Information System (INIS)

    Fouling of tube surfaces in a CANDU steam generator is analyzed using a mathematical model and is shown to account for a major portion of the observed Reactor Inlet Header Temperature (RIHT) rise. First, a detailed heat transfer model is made to account for tube wall temperature at every point along a tube, then the solubility of magnetite is calculated at that wall temperature to check for primary side fouling. Once fouling can occur by magnetite crystal growth on the wall, the rate of fouling is determined by the mass transfer of dissolved iron from the water to the tube surface. The fouling deposit increases heat transfer resistance and thus the heavy water outlet temperature (RIHT) rises. This rise is followed through time and a detailed prediction of deposit weight profiles expected in the tubes is made. The inlet dissolved iron concentration to the boiler is calculated by using a simple flow-assisted corrosion model of outlet feeders. Primary side fouling of the boiler tubes is predicted to be a major contributor to RIHT rise. Coolant pH has a strong effect on flow-assisted corrosion of the outlet feeders and thus on the amount of iron entering the boilers. Deposit weights are predicted accurately for both Gentilly-2 (G-2) and Pickering-1 steam generators using solubility data from Sweeton and Baes at pH 10.3 or 1.4 mg/kg Li. Operation at the low end of the specified pH range seems desirable, e.g. 0.35 mg Li/kg, to reduce the fouling rate of the boiler tubes. Secondary side fouling is predicted to be equally important, but deposit data specific to each plant are needed to assess the precise contribution of this to RIHT rise. A gradual rise in primary side steam quality at the boiler inlet due to RIHT rise will itself generate an RIHT rise simply due to a reduction in boiler area available for sensible heat transfer. Finally, mechanical effects such as divider plate leakage and changes in primary side flow due to pressure tube creep and to increased surface

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

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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)

  1. 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)

  2. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    International Nuclear Information System (INIS)

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

  5. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    International Nuclear Information System (INIS)

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D2O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  6. Material control and accounting at a CANDU reactor: the instrumented safeguards scheme

    International Nuclear Information System (INIS)

    While CANDU reactors differ from LWRs quite markedly in the way they operate, the principles of materials accounting and safeguards are equally applicable. Indeed, since CANDU fuel is not reprocessed, the relatively simple procedure of item accounting is sufficient for CANDUs. However, on-power refueling means that automatic item counting is needed to independently confirm operator records. Surveillance and sealing techniques for spent fuel are needed for a practical system. The equipment developed has allowed the IAEA to apply safeguards at reasonable cost and with minimal interference to the utility operating the station

  7. 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

  8. The accident at Chernobyl and its implications for the safety of CANDU reactors

    International Nuclear Information System (INIS)

    In August 1986, a delegation of Canadians, including two members of the staff of the AECB (Atomic Energy Control Board), attended a post-accident review meeting in Vienna, at which Soviet representatives described the accident and its causes and consequences. On the basis of the information presented at that meeting, AECB staff conducted a study of the accident to ascertain its implications for the safety of CANDU nuclear reactors and for the regulatory process in Canada. The conclusion of this review is that the accident at Chernobyl has not revealed any important new information which would have an effect on the safety requirements for CANDU reactors as presently applied by the AECB. All important aspects of the accident and its causes have been considered by the AECB in the licensing process for currently licensed reactors. However a number of recommendations are made with respect to aspects of reactor safety which should be re-examined in order to reinforce this conclusion

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

    International Nuclear Information System (INIS)

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

  10. 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

  11. The CANDU 9 distributed control system design process

    International Nuclear Information System (INIS)

    Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. Plant control for the CANDU 9 station design is performed by a distributed control system (DCS) as compared to centralized control computers, analog control devices and relay logic used in previous CANDU designs. The selection of a DCS as the platform to perform the process control functions and most of the data acquisition of the plant, is consistent with the evolutionary nature of the CANDU technology. The control strategies for the DCS control programs are based on previous CANDU designs but are implemented on a new hardware platform taking advantage of advances in computer technology. This paper describes the design process for developing the CANDU 9 DCS. Various design activities, prototyping and analyses have been undertaken in order to ensure a safe, functional, and cost-effective design. (author)

  12. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    International Nuclear Information System (INIS)

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

  13. 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

  14. Application of the 3-D Nodal Equivalence Theory to the CANDU Reactor

    International Nuclear Information System (INIS)

    The RFSP code is found to be subject to inconsistency issue mainly due to the lack of nodal equivalence. In Ref. 2, it has been shown that nodal equivalence theory can be effective for the 2-D CANDU core analysis. In this work, the 3-D nodal equivalence theory was applied to see its effectiveness in a 3-dimensional CANDU reactor analysis. The 3-D nodal equivalence is applied to the whole core analysis of a clean CANDU6 core. Both the radial and the axial DFs are quite different for different reactivity devices inside the fuel lattice. It has been demonstrated that the application of the conventional 2-D nodal equivalence theory gives better accuracy in terms of the k-eff and power profiles, while the 3-D equivalence theory only results in marginal improvements. The relative ineffectiveness of the axial discontinuity factor may be ascribed to simplifications of the very complicated core geometry and some assumptions in modeling both radial and axial reflectors of the CANDU reactor. For a more accurate evaluation of the 3-D equivalence theory, more realistic reflector models are currently under development

  15. Designing large motor bearings for a reliable operating life of 40 years in Candu reactors

    International Nuclear Information System (INIS)

    CANDU nuclear reactors have an excellent operating history in various utilities both in Canada and abroad. A large measure of this success is attributed to special attention paid in designing all critical equipment components. One such component is the Heat Transport (HT) pump motor set, which removes heat from the reactor core by circulating pressurized heavy water through the reactor core. Because of the critical role of the HT pump motor set in successful operation of the CANDU reactors, a high degree of reliability is essential at all times over its operating life. A CANDU 6(600 MW unit) reactor has four HT pumps each driven by a 9000 horse power (6.7 MW) induction bearings are one of the few wearable components in the motors, a decision was made at the very early stage of motor design to ensure that these bearings are designed to last the life time of the plant. To date the longest running bearings are still in excellent condition. An extrapolation of the observed data shows that the bearings will easily achieve a designed life of 30 years or longer. Based on this evidence, it has been concluded that the next generation of CANDU 6 reactors which go into operation in 2003 A.D. will have no problems in meeting a 40 year life. This paper presents the multidisciplinary approach involving equipment design, tribology, condition monitoring and maintenance which enabled these motor bearings to be designed for a reliable and long life estimated to be at least 40 years. (author)

  16. 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.)

  17. A Preliminary Assessment of the Adjuster Rod Depletion Effect in the CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    Lifetime of the Wolsong-1 CANDU reactor, which will be shutdown in April, 2009. Major reactor components such as the pressure tube are to be replaced and it is expected that the CANDU reactor can be operated for additional 25-30 years. Meanwhile, all the reactivity devices including the adjuster rods (ADJ) are supposed to be continuously used without any change. In the CANDU reactor, 21 stainless steel (SS) ADJs are used to control the core power distribution and compensate for some reactivity loss during several abnormal cases. The ADJs are normally fully inserted and the SS absorber should undergo a slow depletion through neutron irradiation for a long time. In April, 2009, the accumulated FPY (Full Power Day) of Wolsong-1 is about 23 years. Depletion of ADJs should result in a smaller ADJ worth and a higher fuel burnup and the core power distribution should also be affected by the ADJ depletion. In this work, the effects of the ADJ depletion have been assessed in terms of ADJ worth, time-average core characteristics.

  18. 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)

  19. Synergistic CANDU-LWR fuel cycles

    International Nuclear Information System (INIS)

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

  20. 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

  1. 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.

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

    International Nuclear Information System (INIS)

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

  3. 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)

  4. Thermal hydraulics safety analysis of Candu reactor using the RELAP5 system code

    International Nuclear Information System (INIS)

    In this study, the response of CANDU-6 nuclear reactor to several transients are investigated. The simulation of the system is performed by using RELAP5 thermalhydraulic system code. AECL performes the transient simulations of CANDU reactor by using the FIREBIRD code, developed by AECL for thermal hydraulic analysis of CANDU. All analysis for LOCA and ECCS effectiveness were done by using the FIREBIRD code. The investigations concerning the RELAP5 analysis of CANDU system are too few. Better normal operating conditions are achieved, the effect of pipe interconnecting the outlet headers in a loop is observed. It is found that, with the reactor outlet headers interconnected, the system is stable to perturbations but would exhibit divergent pressure, quality and flow oscillations if the interconnection is removed and if the quality at the reactor outlet header region is greater than 1-2% but less than 8%, specific large (100% of flow area) and small (10% of flow area) breaks in both inlet and outlet headers and in the pump suction are analysed. Results indicate that, l00% break in the inlet header has more probability of fuel failure than the same size break in the outlet header. The worst break location is found to be the pump suction with a break size of 100%. Higher void fractions, higher outlet header quality and heat temperatures are observed in the large break transients than that of small break transients. For small break transients, the break location in the inlet header results higher void fraction, outlet header quality and sheath temperatures than that of outlet header break transients. Emergency core cooling system (ECCS) is found to be effective for the cases analysed. Initiating trip parameters and time for scram and ECCS injection is also investigated

  5. 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)

  6. An integrated control system and operator interface design for retubing of CANDU reactors

    International Nuclear Information System (INIS)

    Experience gained during the successful retubing of four CANDU reactors at Pickering, Ontario, has led to the development of an integrated control system and operator interface which utilizes video and real time graphical animations for feedback. The system provides a uniform interface for many different tools, while the supervisory computer is able to provide prompts of potential problems and monitor the status of subsystems. (author)

  7. Numerical simulation of moderator flow and temperature distributions in a CANDU reactor vessel

    International Nuclear Information System (INIS)

    This paper describes numerical predictions of the two-dimensional flow and temperature fields of an internally-heated liquid in a typical CANDU reactor vessel. Turbulence momentum and energy transport are simulated using the k-epsilon model. Both steady-state and transient results are discussed. The finite control volume analogues of the conservation equations are solved using a modified version of the TEACH code

  8. A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  9. A passive emergency heat sink for water cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. Their effective heat transport combines with the large heat-transfer coefficients of tube banks, thereby reducing containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  10. 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

  11. A Development of Preliminary Evaluating Model of Crept Pressure Tube Diameter for CANDU Reactor (2)

    International Nuclear Information System (INIS)

    Pressure tube of CANDU reactor can be expanded toward both radial and axial directions due to irradiation under the high pressure and temperature condition. As the irradiation period increases, the radial expansion due to creep of the pressure tube increases. The radial expansion of the pressure tube comes out the reduction of the coolability and it results in the power deration. Although the radial expansion of the pressure tube directly affect the safety and economy of the currently operated CANDU reactor, there is no domestic evaluation model to predict the pressure tube diameter. Accordingly, it is necessary to develop the prediction model of the pressure tube diameter and the is the motivation of this study. The objectives of the current work is to develop the basic evaluation model of the pressure tube diameter for CANDU reactor especially for Wolsong NPP (Nuclear Power Plant). In order to develop the diameter evaluation model, measured data for total 86 channels were collected from Wolsong NPP 1, 2, 3 and 4 and analyzed. Based on the provided data, the operational conditions such as a temperature, pressure and neutron flux along the axial direction were derived. All data were analysed to derive the correlation between the pressure tube diameter and the other operation parameters. The evaluation model of pressure tube diameter was modeled by using the neural network algorithm. Neural network algorithm has been widely used to derive the non-linear relation between the input and output data. The developed neural network model was learned based on the data from Wolsong NPP 2, 3, and 4 and was tested by using data from Wolsong NPP 1. The current project will be carried out by IAEA CRP in which all CANDU nations are going to participate

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

    International Nuclear Information System (INIS)

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

  13. 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)

  14. 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.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  17. 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

  18. A compact, low cost, tritium removal plant for CANDU-6 reactors

    International Nuclear Information System (INIS)

    Tritium concentrations in CANDU-6 reactors are currently around 40 Ci/kg in moderator systems and around 1.5 Ci/kg in primary heat transport (PHT) systems. It is expected that tritium concentrations in moderator systems will continue to rise and will reach about 80 Ci/kg at maturity. A more detailed description of the increase in tritium concentrations in the moderator and PHT systems of CANDU-6 reactors is given in the next section of this paper. While moderator systems currently contribute more than 50% to tritium emissions, the impact of acute releases of moderator water is more severe at higher tritium concentrations. This impact can be substantially reduced by the addition of an isotope separation system for lowering the tritium level in the moderator system. In addition, lower tritium levels in CANDU systems will inevitably result in reduced occupational exposures, or will provide economic benefits due to ease of maintenance because less protective measures are required and maintenance activities can be more efficient

  19. 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)

  20. Spectral effects on stress relaxation of Inconel X-750 springs in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M.; Butcher, F.J.; Ariani, I.; Douglas, S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Garner, F.A.; Greenwood, L. [Pacific Northwest National Laboratory, Richland, Washington (United States)

    2008-07-01

    CANDU reactors have been operating for periods up to about 25 years. During this time there are changes to the nuclear reactor core components that are a function of operating environment and time. It is important to know how the properties of critical core components are likely to change over the life of a reactor and therefore their behaviours are characterised long before the end of the reactor design life. Tests are typically conducted in materials test reactors. The behaviour of a material is often characterised as a function of fast neutron fluence and the expected effect of operating time is established by simply extrapolating as a function of fluence. This may be appropriate when the neutron energy spectrum for the materials test reactor matches closely the neutron spectrum where the component resides in the power reactor. However, in cases where the spectrum is very different one has to convert the accumulated dose into a unit that is common in its effect on the material properties. For many property changes in nuclear reactor cores this unit is displacements per atom (dpa). There are different processes that cause atomic displacements and the main ones have to be included in any dpa calculation in order to accurately predict how a given component will perform. One property that is significantly affected by irradiation is stress. Irradiation-induced stress relaxation is a phenomenon that has been used as a method for studying in-reactor creep. Stress relaxation also results in a loss of tension in springs if these springs are in a reactor core environment. This paper describes the stress relaxation of Inconel X-750 in the National Research Universal (NRU) materials test reactor and relates this to the expected relaxation of springs that are installed in the periphery of CANDU reactors. The results show that spectral effects are particularly significant for certain components at the edge of the CANDU reactor core where the neutron spectrum is changing

  1. 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).

  2. Modularized construction, structural design and analysis of CANDU 3 plant

    International Nuclear Information System (INIS)

    CANDU 3 is rated at 450 MW electric, and is a smaller and advanced version of CANDU reactors successfully operating in Canada and abroad. The design uses modularization to minimize the construction schedule and thereby reduce cost. The paper (which is published only as a long summary), deals with the concept of modularization, and with stress analysis of the various civil structures

  3. 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

  4. Detailed Monte Carlo Evaluation of the Power Coefficient of Reactivity of CANDU Reactor

    International Nuclear Information System (INIS)

    The value of FTC can even be positive due to the 239Pu buildup during the fuel depletion and also the neutron up-scattering by the oxygen atoms in the fuel. Unlike the pressurized light water reactor, CANDU-6 has a positive coolant void reactivity (CVR) and coolant temperature coefficient (CTC). As a result the power coefficient of reactivity (PCR) is known to be slightly positive during full power operation. To improve the inherent stability and the generic safety features, a negative PCR is essential. Due to the small value of the FTC and PCR of CANDU, high-fidelity physics approaches are necessary for the precise estimation of the safety parameters. During the reactor analysis, the asymptotic scattering kernel has been used and neglects the thermal motion of nuclides such as U-238. However, it is well accepted that in a scattering reaction, the thermal movement of the target can affect the scattering reaction in the vicinity of scattering resonance and enhance neutron capture by the capture resonance. Some recent works have revealed that the thermal motion of U-238 affects the scattering reaction and that the resulting Doppler broadening of the scattering resonance enhances the FTC of the thermal reactor including PWRs by 10-15%. In order to observe the impacts of the Doppler broadening of the scattering resonances on the criticality and FTC, a recent investigation was done for a clean and fresh CANDU fuel lattice using Monte Carlo code MCNPX for analysis. In ref. 3 the so-called DBRC (Doppler Broadened Rejection Correction) method was adopted to consider the thermal movement of U-238. In this study, the safety parameter of CANDU-6 is re-evaluated by using the continuous energy Monte Carlo code SERPENT 2 which uses the DBRC method to simulate the thermal motion of U-238. The analysis is performed for a full 3-D CANDU-6 core and the PCR is evaluated near equilibrium burnup. For a high-fidelity Monte Carlo calculation, an extremely large number of neutron

  5. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  6. 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)

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. 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).

  10. Remote metallurgical investigations on pressure tubes removed from CANDU power reactors

    International Nuclear Information System (INIS)

    As part of the periodic in-service inspection program for CANDU reactors, pressure tubes are periodically removed for destructive examination. The procedures, equipment, and facilities used to perform metallurgical examinations on these highly irradiated components are described. The initial examinations of the tubes from the generating station are performed underwater in inspection bays. Detailed visual examination and metallography are subsequently performed in shielded hot-cell facilities; a description of the remote metallographic equipment and preparation techniques used is given. Examinations of two recently removed Zr-2.5%Nb pressure tubes containing fretting-wear flaws and a lamination flaw are used to highlight the techniques employed

  11. 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)

  12. CANDU fuel performance

    International Nuclear Information System (INIS)

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

  13. Modelling the activity of 129I in the primary coolant of a CANDU reactor

    Science.gov (United States)

    Lewis, Brent J.; Husain, Aamir

    2003-01-01

    A mathematical treatment has been developed to describe the activity levels of 129I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10 -9 has been determined for the coolant activity ratio of 129I/ 131I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ˜2×10 -8 for 129I/ 137Cs. This ratio can be used to estimate the levels of 129I in reactor waste based on a measurement of the activity of 137Cs.

  14. 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)

  15. Candu fuel and fuel cycles

    International Nuclear Information System (INIS)

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

  16. 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.

  17. 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

  18. 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

  19. Maintenance of ageing CANDU reactors. A regulatory perspective

    International Nuclear Information System (INIS)

    The subject of this paper is, 'requirements for maintenance of ageing reactors from the perspective of a regulator', with a focus on the particular theme of; 'continuing safety assurance'. A major role of maintenance is to ensure the continuing reliability and effectiveness of safety related systems and equipment. Continuing safety assurance is an issue the Atomic Energy Control Board has been wrestling with for some time. From my perspective, much remains to be done before the AECB can be confident that Canadian nuclear plants have the necessary programs in place to achieve continuing safety assurance. To introduce the topic, it would be appropriate to say a few words about the AECB's position with respect to the situation at the Pickering NGS. Why did we blow the whistle last August and, what are we doing about it? (author)

  20. Inteligent control system for a CANDU 600 type reactor process

    International Nuclear Information System (INIS)

    The present paper is set on presenting a highly intelligent configuration, capable of controlling, without the need of the human factor, a complete nuclear power plant type of system, giving it the status of an autonomous system. The urge for such a controlling system is justified by the amount of drawbacks that appear in real life as disadvantages, loses and sometimes even inefficiency in the current controlling and comanding systems of the nuclear reactors. The application stands in the comand sent from the auxiliary feedwater flow control valves to the steam generators. As an environment fit for development I chose Matlab Simulink to simulate the behaviour of the process and the adjusted system. Comparing the results obtained after the fuzzy regulation with those obtained after the classical regulation, we can demonstrate the necessity of implementing artificial intelligence techniques in nuclear power plants and we can agree to the advantages of being able to control everything automatically. (authors)

  1. Seismic analysis of fuelling machine support structure for CANDU6 reactor

    International Nuclear Information System (INIS)

    The fueling machine in the CANDU nuclear power plants is used to perform on-line refueling of the reactor. Canadian safety philosophy requires that the fueling machine survive the design basis earthquake. In the CANDU6 nuclear power plant there are two fueling machines, one on each side of the reactor and located in the reactor building. During reactor operation both machines can either be attached to the reactor (fueling mode) or unattached (stand-by mode). Both cases are considered for seismic qualification. The fueling machine can travel horizontally and vertically and assume any of the 380 fuel channel positions. A number of dynamic models for the fueling machine support structure are prepared using beam elements and lumped masses. Special attention is given to realistically model the linkage points between various components of the system. Spring mechanisms are represented by nonlinear spring elements in the model. The spring characteristics are determined using pull back testing of parts of the machine. These models are analyzed using multiple-level acceleration time-histories at the support points. The analysis is done using the time-history, direct integration method. PC micro computers are used to perform most of the computation work. Different routines of the STARDYNE computer program are used for that purpose. The seismic responses obtained from the analysis are used for stress analysis and verification of load ratings of components. The nonlinear time-history analysis is found to be a practical way of analyzing such a machine. The methodology, modeling techniques and results of this analysis are described in this paper

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  3. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    International Nuclear Information System (INIS)

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  4. The thermalhydraulic behavior of CANDU 600 reactor core fuelled with SEU 43

    International Nuclear Information System (INIS)

    CANDU 600 nuclear reactors are usually fuelled with Standard 37 rods fuel bundles, and natural uranium (NU) dioxide (UO2) is used as fuel composition. A new fuel bundle geometry is proposed with 43 rods and slightly enriched uranium fuel (SEU 43 with 0.96% enrichment of 235U). In this paper a comparative analysis of the behavior of the primary circuit during a LOCA 35% RIH accident is performed for two core types (normal core 37 pin/bundles and a 43 pin/bundles proposed core). This kind of accident is considered to be a severe accident for CANDU type fuel elements. This analysis uses FIREBIRD code coupled with bipoint kinetics module. The bipoint kinetics module includes the models for the neutronic measurement instrumentation (the platinum detectors and the ion chambers) and the RRS (Reactor Regulating System) module. For the large LOCA, the RRS is of no effect, therefore the RRS option was not used for this analysis. The main conclusions of the analysis are the following: in this case of 35% RIH, LOCA, the thermalhydraulic behavior of the 43 pin/bundles core is better than the normal core. while the fuel and the sheath temperature reached not the melting point. (authors)

  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. Research on the separation of hydrogen isotopes from liquid wastes from CANDU nuclear reactors

    International Nuclear Information System (INIS)

    The separation of hydrogen isotopes is very important in operation of CANDU nuclear reactors fueled with natural uranium. This paper refers to separation of tritium from liquid wastes from CANDU nuclear reactors. The tritium recovery from wastes is of importance for the following reasons: - the process has a high nuclear yield; - it contributes to the radioprotection of operation personnel and environment. The separation has been carried out through isotope exchange process between hydrogen and liquid water using metal/support catalysts. Pt/SDB/PS were used as catalysts. The experiments were performed under the following conditions: - radioactive concentration of tritiated heavy water, 4.34 mCi/ml; - pressure, 1 atm; - temperature of exchange column, 26-33 deg. C; - deuterium concentration, 10 % D/(D+H); - migration speed through catalytic bed, 0.02-0.34 m/s; - contact time: 0.22-7.2 s. The experiments have showed the catalytic efficiency of Pt/SDB/PS for both deuterium exchange and tritium exchange. The results showed that the deuterium exchange is faster than that of tritium and that, due to the high catalytic efficiency of the catalyst used, it is particularly adequate for tritium separation from liquid wastes. (authors)

  7. Elasto-plastic behaviour of the pressure tube (Zr-2.5Nb%) of CANDU reactor

    International Nuclear Information System (INIS)

    To ensure the structural integrity and meet the safety conditions in the pressure tubes of Cernavoda CANDU type reactor, subjected to a severe thermodynamic operational environment (operation temperature 560 K - 585 K, tangential stress coefficient σe = 110-130 MPa), to the corroding ambient and to the radiation field (a fast neutron flux of 1017 n/m2s), the knowledge of mechanical parameter evolution, along the lifetime span (about 30 - 40 years), is needed. The present work reports the materials data of samples extracted from a pressure tube of Cernavoda reactor (alloy-Zr97.5Nb2.5), axially submitted, to ambient and 300 deg. C temperatures. The thermodynamic stress monitoring and the experimental data acquisition and processing were carried out by an analog-to-digital converter. The experimental data obtained by this procedure were correlated rather well by means of the Hsu-Bertels law, adequate to describe the mechanical fatigue of elasto-plastic materials. The parameters determined are used in computing and predicting the behaviour of CANDU pressure tube in normal operation conditions

  8. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

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

  9. 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)

  10. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

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

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

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

  16. Decrease of the CANDU spent nuclear waste inventories in fusion-fission (hybrid) reactors

    International Nuclear Information System (INIS)

    The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D,T) and catalyzed (D,D) modes. The analysis is conducted for a CANDU spent nuclear fuel which was used up to a total enrichment grade of 0.418%. The behavior of the spent fuel is observed during 48 months for discrete time intervals of Δt = 6 months. The cooling of the fissile fuel zone is considered with three different coolants, notably gas (preferably He), Flibe (Li2BeF4) and natural lithium. A rejuvenation period of 8 months is evaluated for a final fissile fuel enrichment grade of 1% for all coolant types in the fissile zone under a first-wall fusion neutron current load of 1,014 (2.45-MeV n/cm2.s) and 1,014 (14.1-MeV n/cm2.x), corresponding to 2.64 MW/m2 by a plant factor of 75% for the catalyzed (D,D) fusion reactor mode. The rejuvenation period increases to 12 months for the same fissile fuel enrichment grade using the (D,T) fusion reactor mode under a first-wall fusion neutron current load of 1,014 (2.45-MeV n/cm2.s), corresponding to 2.25 MW/m2 by a plant factor of 75%. This enrichment would be sufficient for a re-utilization in a CANDU reactor

  17. Development of new MMS modules for CANDU reactors, using CompGen program

    International Nuclear Information System (INIS)

    Modelling of thermodynamic processes in stationary and dynamic regimes are very important for nuclear power plants from both design and operation point of view. Since safety requirements for NPPs are higher than the ones for classical power plants simulation analyses are necessary in this field. In the last years there where developed and improved many computing codes which can be used in such analyses. One of these codes is MMS. MMS package was developed by Framatome Technologies, Bacok and Wilcox and nHance Technologies Inc. It uses a modular architecture based on ACSL program. MMS is provided with a rich library of modules including control, electromechanical, gas, water, boundary, fossil and nuclear modules. Unfortunately, the water and nuclear modules were developed initially only for light LWR NPPs. The main problem of using MMS for CANDU NPPs was the lack of heavy water properties. Recently nHance Technologies improved some of existing modules allowing to work also with heavy water systems. However, development of new modules appropriate to CANDU reactors is necessary to simulate various systems in CANDU NPPs. New MMS modules developed by the authors, characteristic for CANDU NPPs concern: Calandria, heat exchanger, delay tank, end shield., fuel storage pool, calandria vault. These modules were built with means of CompGen program, which is included in the MMS package. They were used at Center of Technology and Engineering for Nuclear Projects - CITON, at Bucharest to simulate operational regimes of the primary circuit in Cernavoda NPP. In this present paper the capabilities of MMS package regarding thermal-hydraulic simulation of nuclear systems are shown . Also, presented are the results of a joint research of Power Plants Simulation Laboratory of University 'Politehnica' at Bucharest and CITON on several MMS modules, developed by using CompGen. These results serve for safety analyses for Cernavoda 2 NPP. As an illustration, a 'real' safety analysis for the

  18. 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)

  19. The blister phenomena in relation to pressure tube integrity in CANDU reactor

    International Nuclear Information System (INIS)

    Zirconium alloy pressure tubes in CANDU type PHWR reactors are exposed to aqueous conditions embracing high temperature, fast neutron flux and high pressure. Two properties, dimension and hydrogen concentration, represent the main properties where changes are important to the life of a pressure tube. Rupture of a cold worked Zircaloy-2 pressure tube in Pickering Unit 2 in 1983 occurred when a crack developed from an array of hydride blisters. These have been observed on the outside surface of the pressure tube where it contacted the surrounding calandria tube. The contact of the pressure tube with the calandria tube can occur during the operation time and produces in the pressure tube a localized cooling. The consequence of the local heating of the calandria tube is some localized hydride precipitation. Under certain conditions, hydrogen will migrate down the temperature gradient and accumulate in the coldest region. Such precipitation, when it occurs under operating conditions, is considered to be the start of blister formation. When the blister cracking threshold is reached, the blister cracking can initiate a crack on the tube body. The failure mechanisms in zirconium alloy pressure tubes involve the presence of hydrogen in the initiation process and then a propagation process. If crack, originating from a hydride blister on the outside of the pressure tube, is developed then crack growth is possible in the axial direction to a partial thickness unstable length. Unstable pressure tube rupture is an event in a CANDU reactor that has potentially serious economic and safety consequences and reactor operation under conditions which entail the risk of such failure should be avoided. (authors)

  20. Thermalhydraulic analysis of Candu 6 100% reactor outlet header break using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I.; Ghitescu, P. [Power Plant Engineering Faculty, Politehnica University, Bucharest (Romania); Negut, G. [Institute for Nuclear Research, Pitesti (Romania)

    2007-07-01

    One of the postulated large break losses of coolant accident (LBLOCA) in a Candu reactor is the Reactor Outlet Header (ROH) break. After such an event, the normal coolant flow in the channels downstream the break is disrupted and late stagnation occurs after the pump head is degraded and before emergency core cooling (ECC) injection takes place. Given that the fuel decay and stored energy decreased significantly by that time, the heatup of the fuel sheath is much lower than in the case of 35% Reactor Inlet Header (RIH) break. However, the coolant pressure is much lower than the corresponding one at the 35% RIH break.The combination of a high fuel clad temperature and coolant low-pressure lead to more fuel failure events. Thus, the 100% ROH break has the highest potential for radioactivity release. The paper presents the thermal hydraulic analyses of a 100% reactor outlet header break. The study is done with RELAP5/SCDAP mod 3.4 and the results were compared with those of CATHENA. RELAP5 predicts a slightly faster inventory loss, an extended flow stagnation period and a higher clad temperature.

  1. 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

  2. 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

  3. 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)

  4. Suitability of CR-39 dosimeters for personal dosimetry around CANDU reactors

    International Nuclear Information System (INIS)

    The capabilities and limitations of CR-39 damage track detectors have been evaluated for their use as personal neutron dosimeters around CANDU reactors. Since the energy response is a critical characteristic, the neutron energy spectra expected within CANDU containments were studied. In the boiler rooms, around the moderator cooling systems, and in most of the fueling machine vaults, the spectra vary considerably, but the majority of the dose is expected to be delivered by neutrons above 80 keV, the approximate threshold for electrochemically-etched CR-39 detectors. In the Pickering A fueling machine vault, and in areas in other stations to which neutrons from reactors have been multiply scattered, lower energy neutrons may be important. In nearly all areas where people work, it appears that working times will be limited by gamma rays rather than by neutrons. The characteristics of other neutron dosimeters - bubble and superheated drop detectors, albedo detectors, and Si real-time detectors - were also reviewed. For workers who typically receive neutron doses that are small compared with regulatory limits, CR-39 is the most suitable available dosimeter for demonstrating compliance. All single dosimeters have poor angular response over the range 0 to 180 degrees because of the shielding of the body. Albedo and Si detectors have particularly poor energy responses over the energy range of importance. Bubble and superheated drop detectors have the advantages of immediate readout and high sensitivity, but the disadvantages of inability to integrate doses over a long period, temperature dependence, very limited range and higher cost. (Author) (110 refs., 45 figs.)

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. Verification of source term analysis system for decommissioning wastes from a CANDU reactor

    International Nuclear Information System (INIS)

    There are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear capacity becomes higher and installations age, the Korean government and industry have launched R and D to estimate appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsong unit 1 was used as a neutron source to activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with measured data within 20% difference. Therefore, it can be concluded that MCNP/ORIGEN system could be a reliable tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant flux irradiation and snapshot equilibrium core model as a reference core. (author)

  11. 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 PWRs. Canadian Deuterium Uranium (CANDU trademark) 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 resulted in a decrease in steam generator-related station unavailability of Canadian CANDU reactors. 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 (R and D) work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for speciality tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service (FFS) guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. This paper will also show how recent advances in cleaning technology are integrated into a life management strategy. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New steam generator 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 strategic plans to ensure that good future operation is ensured. (orig.)

  12. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  13. Assessment of aging of Zr-2.5Nb pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    In modern CANDU nuclear generating stations, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D2O) coolant. The pressure tubes operate at an internal pressure of about 10 MPa and temperatures ranging from about 250oC at the inlet to about 310oC at the outlet. Over the expected 30 year lifetime of these tubes they will be subjected to a total fluence of approximately 3 x 1026 n m-2. In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen-deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. A fitness-for-service methodology has been developed which assures that this will not happen. A key element in this methodology is the acquisition of data and understanding-from surveillance and accelerated aging testing-to assess and predict changes in the DHC initiation threshold, the DHC velocity and the fracture toughness (critical crack length) as a function of service time. The most recent results of the DHC and fracture toughness properties of CANDU pressure tubes as a function of time in service are presented and used to suggest procedures for mitigation and life extension of the pressure tubes. (author)

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

    International Nuclear Information System (INIS)

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

  15. 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.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Benefit of chromium in reducing the rates of flow accelerated corrosion of carbon steel outlet feeders in CANDU reactors

    International Nuclear Information System (INIS)

    In the mid 1990's, wall thinning of outlet feeders due to flow accelerated corrosion (FAC) was recognized as an active mechanism in the outlet feeders of CANDU reactors. To address wall thinning of outlet feeders in new reactor construction and refurbishment projects, AECL introduced a minimum Cr concentration in its specification for the SA-106 carbon steel feeder pipe. The effectiveness of Cr in reducing FAC was subsequently demonstrated in in-reactor and out-reactor loops at AECL's Chalk River Laboratories. More recently, wall-thinning rates have been determined from wall thickness data collected from outlet feeders, containing a specified minimum Cr concentration, installed in the Point Lepreau Generating Station in 2001. This paper presents the FAC rates determined from in-service outlet feeders and compares the rates with data from previous in-reactor and out-reactor test loops, highlighting the consistency observed in results from the three sources. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Recovery of tritium from CANDU reactors, its storage and monitoring of its migration in the environment

    International Nuclear Information System (INIS)

    Tritium is produced in CANDU heavy water reactors mainly by neutron activation of deuterium. The typical production rate is 2.4 kCi per megawatt-year (89 TBq. per megawatt-year. In Pickering Generating Station the average concentration of tritium in the moderators has reached 16 Ci.kg-1 (0.6 TBq.kg-1) and in coolants, 0.5 Ci.kg-1 (0.02 TBq.kg-1). Concentrations will continue to increase towards an equilibrium determined by the production rate, the tritium decay rate and heavy water replacement. Tritium removal methods that are being considered for a pilot plant design are catalytic exchange of DTO with D2 and electrolysis of D2O/DTO to provide feed for cryogenic distillation of D2/DT/T2. Storage methods for the removed tritium - as elemental gas, as metal hydrides and in cements - are also being investigated. Transport of tritiated wastes should not be a particularly difficult problem in light of extensive experience in transporting tritiated heavy water. Methods for determining the presence of tritium in the environment of any tritium handling facility are well established and have the capability of measuring concentrations of tritium down to current ambient values. (author)

  3. PCI-OGRAMS: application of CANDU fuelogram methodology to PCI data from light water reactors

    International Nuclear Information System (INIS)

    The FUELOGRAM model was derived to predict PCI defect probablilities for CANDU fuel bundles that had experienced power increases after being irradiated to burnups mostly in the range 100 +- 60 MW.h/kg U. It is inappropriate to extrapolate the FUELOGRAM model to predict the performance of differently designed fuels at burnups up to 600 MW.h/kg U Therefore data obtained from the operaton of a Boiling Water Reactor were analyzed using the FUELOGRAM methodology to assess fuel performance criteria at high burnups. The resultant PCI-OGRAMS evaluate defect probabilities in terms of power increase (ΔP), ramped power (P), and the burnup (ω) of the most highly rated rod in a fuel assembly. Defect probability also depends on the dwell time (t), of fuel at the ramped power. The predictions of the PCI-OGRAM, FUELOGRAM and other models are compared in three-dimensional sketches of P, ΔP, and ω with the dwell time t held constant. (author)

  4. 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

  5. Numerical analysis of zirconium hydride blisters in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    CANDU nuclear reactors use zirconium alloy pressure tubes for primary containment of fuel and coolant. The 1983 failure of a pressure tube in Unit 2 of the Pickering Nuclear Generating Station was attributed to the formation of large precipitates of zirconium hydride, referred to as blisters. These blisters formed at localized cold spots on the pressure tube surface where it had come into contact with the colder calandria tube. The high hydrogen concentrations in the Zircaloy-2 pressure tubes used only in the first two Pickering Units were a major contributing factor to blister formation and the ultimate failure. In an effort to better understand the mechanism of crack initiation at a blister, a program was undertaken to use finite element methods to model the stresses generated by the formation of a blister in a tube. The preliminary results in this work have been published elsewhere. This paper summarizes the recent refinements to the model and our present understanding of the development of stresses in and around hydride blisters. (orig./GL)

  6. 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)

  7. 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

  8. Research and development initiatives in support of the conceptual design for the CANDU Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

    Canada's Generation-IV National Program has been established to provide research and development (R and D) supports for the concept of the CANDU SuperCritical Water-cooled Reactor (SCWR). It focuses primarily on the key technology areas, such as material, chemistry, thermal-hydraulics, safety, physics, and hydrogen production. Design challenges surrounding out-of-core components are also examined. R and D results of these areas will be applied in finalizing the design options currently being considered in developing Canada's conceptual design. In this paper, a brief description of the conceptual design options is presented and the linkage of the R and D projects to support the decision in selecting the reference design option for the CANDU SCWR is described. Various participants contributing the R and D information to the Gen-IV National Program are introduced. (author)

  9. 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)

  10. 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

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

    International Nuclear Information System (INIS)

    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

  12. The CANDUR Reactor - The Practical Path to RU and TH use in Nuclear Reactors

    International Nuclear Information System (INIS)

    The CANDU heavy water reactor has unrivalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and Thorium (Th). Recently, this unique CANDU reactor feature attracted considerable attention due to favourable commercial, environmental and strategic needs. This paper summarizes the solid progress over the last three years and outlines CANDU Energy Incorporated's (CEI) multi-stage vision of utilizing various fuels in currently operational and new build CANDU reactors. In CEI's fuel-cycle vision, CANDU reactors will operate in conjunction with other reactor types and use advanced fuels to produce more energy and ensure the most efficient and least costly method of utilizing Light Water Reactor (LWR) used fuel. With this vision and the tandem goal of systematic adoption of Thorium based fuels, CANDU reactors will be a strong technology partner in ensuring the availability of long-term stable resources for nuclear power plants

  13. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

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

  14. CANDU fuel cycle flexibility

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Gustavo Boroni; Alejandro Clausse

    2011-01-01

    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...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. 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)

  19. 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.

  20. 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

  1. 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).

  2. Acquisition, processing and evaluation of the corrosion data for primary system of CANDU 6 reactor

    International Nuclear Information System (INIS)

    Full text: The safety in operation of structural components from Primary Heat Transport System (PHTS) depends principally on chemical characteristics of heavy water used in CANDU reactor coolant and moderator, temperature, as well as on structural materials properties. To achieve or exceed the design life of the plant and to ensure a minimal degradation of PHTS structural materials, besides a rigorous chemical control of coolant, it is necessary to know and to understand the degradation phenomena by structural materials corrosion. It is also necessary to minimize and control the generation of radioactive products and waste in the plant, as well as to keep deposits, which reduce heat transfer in the reactor core or steam generators, to a minimum. To this goal, a chemical control of the water chemistry and structural materials corrosion testing program, as well as a system for data acquisition and processing, were developed and are presented. Structural materials corrosion testing programme has the following objectives: - elaboration of the evaluation methodology of the structural materials corrosion behavior; - development of the database regarding corrosion behavior and modeling of some oxidation processes; - finding the causes of failure in some components and possible remedies. To this aim, several of corrosion experiments were performed, namely: - out-of-pile corrosion experiments in different conditions of water chemistry and temperature; - corrosion experiments in autoclaves assembled in by-pass of CANDU - 6 PHTS; - corrosion analyses performed on some corroded components. Accumulated data were stored in databases and a data processing system allowed us: - to keep under observation the plant water chemistry; - to evidence the appearance and the evolution of some accelerated processes in primary circuit; - to determine the corrosion, deposition and releasing of the corrosion products, as well as the characteristics of the corrosion films formed on different

  3. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  4. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  5. 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

  6. Investigation of the Power Coefficient of Reactivity of 3D CANDU Reactor through Detailed Monte Carlo Analysis

    International Nuclear Information System (INIS)

    The heat is removed by the heavy water coolant completely separated from stationary moderator. Due to the good neutron economy of the CANDU reactor, natural uranium fuel is used without enrichment. Because of the unique core configuration characteristic, there is less resonance absorption of neutron in fuel which leads to a relatively small fuel temperature coefficient (FTC). The value of FTC can even be positive due to the 239Pu buildup during the fuel depletion and also the neutron up-scattering by the oxygen atoms in the fuel. Unlike the pressurized light water reactor, it is well known that CANDU-6 has a positive coolant void reactivity (CVR) and coolant temperature coefficient (CTC). In a traditional reactor analysis, the asymptotic scattering kernel has been used and neglects the thermal motion of nuclides such as U-238. However, it is well accepted that in a scattering reaction, the thermal movement of the target can affect the scattering reaction in the vicinity of scattering resonance and enhance neutron capture by the capture resonance. Some recent works have revealed that the thermal motion of U-238 affects the scattering reaction and that the resulting Doppler broadening of the scattering resonance enhances the FTC of the thermal reactor including PWRs by 10- 15%. In order to observe the impacts of the Doppler broadening of the scattering resonances on the criticality and FTC, a recent investigation was done for a clean and fresh CANDU fuel lattice using Monte Carlo code MCNPX for analysis.. In ref. 3 the so-called DBRC (Doppler Broadened Rejection Correction) method was adopted to consider the thermal movement of U-238. In this study, the safety parameter of CANDU-6 is re-evaluated by using the continuous energy Monte Carlo code SERPENT 2 which uses the DBRC method to simulate the thermal motion of U-238. The analysis is performed for a full 3-D CANDU-6 core and the PCR is evaluated near equilibrium burnup. For a high-fidelity Monte Carlo calculation

  7. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  8. 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.

  9. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  10. 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

  11. Analysis of transient dry patch behavior on CANDU reactor calandria tubes in a LOCA with late stagnation and impaired ECI

    International Nuclear Information System (INIS)

    An analytical method to describe the behavior of transient dry patches on CANDU reactor calandria tubes has been developed. Dry patches may form following the sagging of a pressure tube onto a calandria tube in certain low-probability scenarios in which a loss-of-coolant accident occurs with subsequent failure or impairment of the emergency cooling injection function. Results of the analysis show that the dry patches will not grow beyond a few degrees on each side of the bottom of the calandria tube and will rewet within a few tens of seconds, with the values depending on the specific CANDU reactor design and the mechanism of dry patch formation and rewetting. Maximum local calandria tube temperatures reached during the transient will be about 5500C to 7000C. There will be no significant effects (0C) on fuel, sheath and maximum pressure tube temperatures. The analytical results provide confidence that pressure tube and calandria tube integrity will not be threatened by dry-patch formation in the LOCA scenarios studied

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  17. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    The Three Mile Island accident spurred a world-wide interest in severe accidents. The initial reaction was to increase the preventative measures in existing designs, followed by development of predictive capabilities to improve the management of severe accidents. Recently, emphasis has been placed in new designs on mitigative measures which slow down or contain the progression of a severe accidents. U.S. requirements for Advanced Light Water Reactor designs must now: provide reactor cavity floor space to enhance debris spreading; provide a means to flood the reactor cavity to assist in the cooling process. The paper describes how CANDU Pressurized Heavy Water Reactors (PHWRs) have severe accident prevention and mitigation inherent in the design; in particular, the U.S. severe accident requirements can be met without significant change to the design of current CANDUs. (author). 32 refs, 7 figs, 1 tab

  18. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  19. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

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

  20. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

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

  1. 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

  2. ICONE-4: Proceedings. Volume 2: Advanced reactors

    International Nuclear Information System (INIS)

    The proceedings for this conference are contained in 5 volumes. This volume is divided into the following areas: advanced reactor requirements; advanced reactor design and analysis; arrangement and construction; specific reactor designs; demonstration testing; safety systems and analysis; component demonstration testing; advanced reactor containment design; licensing topics and updates; accelerator applications and spallation sources; and advanced reactor development. Separate abstracts were prepared for most papers in this volume

  3. 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.)

  4. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

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

  5. 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)

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

    International Nuclear Information System (INIS)

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

  7. MARATHON - a computer code for the probabilistic estimation of leak-before-break time in CANDU reactors

    International Nuclear Information System (INIS)

    The presence of high levels of moisture in the annulus gas system of a CANDU reactor indicates that a leaking crack may be present in a pressure tube. This will initiate the shutdown of the reactor to prevent the possibility of fuel channel damage. It is also desirable, however, to keep the reactor partially pressurized at hot shutdown for as long as it is necessary to unambiguously identify the leaking pressure tube. A premature full depressurization may cause an extended shutdown while the leaking tube is being located. However, fast fracture could occur during an excessively long hot shutdown period. A probabilistic methodology, together with an associated computer code (called MARATHON), has been developed to calculate the time from first leakage to unstable fracture in a probabilistic format. The methodology explicitly uses distributions of material properties and allows the risk associated with leak-before-break to be estimated. A model of the leak detection system is integrated into the methodology to calculate the time from leak detection to unstable fracture. The sensitivity of the risk to changing reactor conditions allows the optimization of reactor management after leak detection. In this report we describe the probabilistic model and give details of the quality assurance and verification of the MARATHON code. Examples of the use of MARATHON are given using preliminary material property distributions. These preliminary material property distributions indicate that the probability of unstable fracture is very low, and that ample time is available to locate the leaking tube

  8. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  9. 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)

  10. Experimental and computational thermalhydraulics research related to CANDU reactor operation and safety

    International Nuclear Information System (INIS)

    This paper describes recent, ongoing and planned research projects at the University of Ottawa, whose objective is to enhance our knowledge of flow and heat transfer in CANDU rod bundles and header/feeder systems and to assist the Canadian nuclear industry in the analysis of operation and safety of CANDU components as well as in designing improved ones. Several experimental facilities are being developed, including a refractive-index matching flow loop for detailed measurements of flows in eccentric annuli and rod bundles, a large-scale, heated rod-bundle facility with air as medium, matching the Reynolds number of the CANDU core and suitable for the study of the effects of geometrical distortions (e.g., pressure tube creep, spacers and fuel element bow) and transients, and an air-water loop for the testing of the operation of wire-mesh sensors and the study of two-phase flows in simple header/feeder vessels. Extensive CFD work on similar topics is also been conducted in parallel with the experiments using the experimental results for its validation. (author)

  11. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  12. 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

  13. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  14. Calibration method of liquid zone controller using the ex-core detector signal of CANDU 6 reactor

    International Nuclear Information System (INIS)

    Highlights: ► We developed a new LZC calibration method and measurement system. ► Photo-neutron effect, reactor core size, and detector position were evaluated and tested. ► We applied the new method and system to Wolsong NPP Unit 1. ► The LZC calibration test was well completed, and the requirement of the test was satisfied. - Abstract: The Phase-B test (low-power reactor physics test) is one of the commissioning tests for Canada Deuterium Uranium (CANDU) reactors that ensures the safe and reliable operation of the core during the design lifetime. The Phase-B test, which includes the approach to the first criticality at low reactor powers, is performed to verify the feasibility of the reactor’s physics design and to ensure the integrity of the control and protection facilities. The commissioning testing of pressurized heavy water moderated reactors (PHWRs) is usually performed only once (at the initial commissioning after construction). The large-scale facilities of the Wolsong nuclear power plant (NPP) Unit 1 have been gradually improved since May 2009 to extend its lifetime. The refurbishment was completed in April 2011 – then this NPP has been in operation again. We discusses the new methodology and measurement system that uses an ex-core detector signal for liquid zone controller (LZC) calibration of the Phase-B test instead of conventional methods. The inverse kinetic equation in the reactivity calculator is modified to treat the 17 delayed neutron groups including 11 photo-neutron fractions. The signal acquisition resolution of the reactivity calculator was enhanced and installed reactivity calculating module by each channel. The ex-core detector was confirmed to be applicable to a large reactor core, such as the CANDU 6 by comparison with the in-core flux detector signal. A preliminary test was performed in Wolsong NPP Unit 2 to verify the robustness of the reactivity calculator. This test convincingly demonstrated that the reactivity calculator

  15. Aerosol deposition at high-temperature gradients in geometries relevant to a CANDU nuclear reactor primary cooling system

    International Nuclear Information System (INIS)

    During some postulated severe loss-of-coolant/loss-of-emergency coolant injection accidents, temperatures in a CANDU reactor fuel channel may rise high enough to cause release of vapours of core materials (UO2 fuel, Zircaloy-4 cladding and fission products such as Cs and I). The released vapour mass is expected to be carried by the coolant steam into the cooler parts of the primary cooling system in the form of aerosol particles. A fraction of the aerosol mass containing the fission products will be deposited in various component geometries, such as flow channels containing fuel rodbundles and pipes with changing cross-sectional area and bends. As part of an aerosol program, we have conducted experiments on aerosol transport in two geometries, focussing on thermophoresis and diffusion. The results obtained in these experiments and their analysis are presented in this paper. (author)

  16. Overpressure protection requirements for primary heat transport systems in CANDU power reactors fitted with two shutdown systems

    International Nuclear Information System (INIS)

    The overpressure protection requirements of Article NB 7000 of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) are incorporated in the National Standard of Canada N285.1. These requirements do not, however, refer to a particular nuclear system design. This is recognized in paragraphs NCA-2141 and NB-7120 of the ASME Code which make reference to the requirements of the appropriate regulatory authority for guidance. For CANDU power reactors fitted with two shutdown systems, some guidance is given in the Atomic Energy Control Board (AECB) Regulatory Document R-10, but this does not address overpressure protection as a specific topic and further clarification is required. This document seeks to provide such clarification

  17. 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

  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. The CANDU 3 containment structure

    International Nuclear Information System (INIS)

    The design of the CANDU 3 nuclear power plant is being developed by AECL CANDU's Saskatchewan office. There are 24 CANDU nuclear power units operating in Canada and abroad and eight units are under construction is Romania and South Korea. The design of the CANDU 3 plant has evolved on the basis of the proven CANDU design. The experiences gained during construction, commissioning and operation of the existing CANDU plants are considered in the design. Many technological enhancements have been implemented in the design processes in all areas. The object has been to develop an improved reactor design that is suitable for the current and the future markets worldwide. Throughout the design phase of CANDU 3, emphasis has been placed in reducing the cost and construction schedule of the plant. This has been achieved by implementing design improvements and using new construction techniques. Appropriate changes and improvements to the design to suit new requirements are also adopted. In CANDU plants, the containment structure acts as an ultimate barrier against the leakage of radioactive substances during normal operations and postulated accident conditions. The concept of the structural design of the containment structure has been examined in considerable detail. This has resulted in development of a new conceptual design for the containment structure for CANDU 3. This paper deals with this new design of the containment structure

  20. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

  1. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring, and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs) including the Soviet designed water moderated and water cooled energy reactors (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

  2. Using the CompGen code for developing new MMS modules specific to CANDU type reactors

    International Nuclear Information System (INIS)

    The Modular Modeling System, MMS, is an advanced tool for thermal hydraulic analyses. The MMS code is built upon a modular philosophy allowing users to construct their own schemes from the existing MMS modules. Also, it is possible to develop new MMS modules by using the CompGen sub-code. Developing new modules in the frame of MMS program was necessary because the standard modules of the program libraries are customized to the American reactors, particularly to PWR type reactors. As known the Cernavoda NPP is of PHWR type having equipment and components different from the structure point of view, hence differ from the point of view of the equation describing the transient process. An example is the module of low pressure part of the turbine in the secondary circuit of the plant, where the moisture extraction coefficients measured from equipment operation are different from the coefficient introduced in the module of the MMS library. The new modules developed in this work were compared with data from the thermal balance supplied by General Electric, with functional data from operation, and with the data from the Commissioning Report of the Cernavoda NPP Unit 1 commissioning. In this work new modules which or developed are presented as well as two examples of using the steam generator modules, high pressure turbine, low pressure turbine, and superheater separator

  3. 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.

  4. On-line power monitoring in CANDU using PMFD

    International Nuclear Information System (INIS)

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

  5. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Applying operating experience to design the CANDU 3 process

    International Nuclear Information System (INIS)

    The CANDU 3 is an advanced, smaller (450 MWe), standardized version of the CANDU now being designed for service later in the decade and beyond. The design of this evolutionary nuclear power plant has been carefully planned and organized to gain maximum benefits from new technologies and from world experience to date in designing, building, commissioning and operating nuclear power stations. The good performance record of existing CANDU reactors makes consideration of operating experience from these plants a particularly vital component of the design process. Since the completion of the first four CANDU 6 stations in the early 1980s, and with the continuing evolution of the multi-unit CANDU station designs since then, AECL CANDU has devised several processes to ensure that such feedback is made available to designers. An important step was made in 1986 when a task force was set up to review and process ideas arising from the commissioning and early operation of the CANDU 6 reactors which were, by that time, operating successfully in Argentina and Korea, as well as the Canadian provinces of Quebec and New Brunswick. The task force issued a comprehensive report which, although aimed at the design of an improved CANDU 6 station, was made available to the CANDU 3 team. By that time also, the Institute of Power Operations (INPO) in the U.S., of which AECL is a Supplier Participant member, was starting to publish Good Practices and Guidelines related to the review and the use of operating experiences. In addition, details of significant events were being made available via the INPO SEE-IN (Significant Event Evaluation and Information Network) Program, and subsequently the CANNET network of the CANDU Owners' Group (COG). Systematic review was thus possible by designers of operations reports, significant event reports, and related documents in a continuing program of design improvement. Another method of incorporating operations feedback is to involve experienced utility

  9. Research for thorium cycle high conversion in the CANDU and PWR type reactors. Development of simulation and study of symbiotic scenarios

    International Nuclear Information System (INIS)

    In the frame of a sustainable nuclear energy, this study assesses Thorium-fueled CANDU and PWR competitiveness to reduce access difficulties to cheap uranium resources and Gen. IV cost and availability problems. It focuses on neutronic analysis of two thorium fuel management options: 233U production from Th/Pu fuels and 233U conversion in these reactors. In particular, breeding in multi-recycled Th/U CANDU has been established. Before this work, simulation methods and nuclear data have been validated by cross-checking two different types of codes (probabilistic and deterministic). Symbiotic scenarios, with various reactor and fuel combinations have been evaluated and compared. Resources savings have been quantified through core slight modifications. Deeper modifications towards breeding in PWR have been proposed and preliminarily studied. (author)

  10. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  11. An integrated CANDU system

    International Nuclear Information System (INIS)

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

  12. The hierarchy of essential CANDU reactor control functions in a distributed system

    International Nuclear Information System (INIS)

    Control functions in CANDU nuclear generating stations are programmed within two centralized and redundant minicomputers while safety functions are covered by conventional analog systems. This set-up is a product of standards, economic and technical considerations which are now being modified by the maturing of microprocessors, the progress in digital communications and the development of mathematical process models. Starting from the control and safety systems installed in Gentilly-2, this paper analyses trends that will affect the implementation of essential control functions within a distributed system. In particular, it emphasizes the characteristics of future software systems that must be built-in in order to comply with important operational requirements of nuclear generating stations. (auth)

  13. The Romanian experience on introduction of CANDU-600 reactor at the Cernavoda NPP

    International Nuclear Information System (INIS)

    The Cernavoda Nuclear Power Plant (NPP) Project is a key component of the Romanian nuclear development program. Selection of the CANDU design represents a major contribution to this development, due to the technological feasibility for manufacturing of parts, components and the nuclear fuel based on the uranium resources in Romania. The Romanian nuclear development program also involves a nuclear fuel manufacturing plant, a heavy water production plant and organizations specialized in research, engineering, manufacturing and completion for systems and components. The agreement on technological transfer between Canada and Romania is supporting the Romanian involvement to the achievement of the Project, with a degree of participation that is gradually increasing from the first to the last NPP Unit. (author)

  14. The advanced technique of on-line D2O leakage monitoring using FT-IR in CANDU NPP

    International Nuclear Information System (INIS)

    In a CANDU reactor, it is necessary to monitor the leakage of D2O that serves as a moderator and a coolant. The present monitoring systems at Wolsong-1 is dispersive infrared spectrometer. However, owing to the problems, such as low sensitivity and degradation of the equipment, it is needed to develop a new instrumental method for the monitoring of D2O leakage. This demand made us to carry out a series of experiments to test the possibility of FT-IR as a new monitoring method for the D2O leakage. Our experimental results show that the D2O on-line monitoring using FT-IR gives the precision of ±5.1 ppm that is an improved value of dispersive infrared spectrometer by about 50 %. Therefore, it is worthwhile to develop the FT-IR method as a future technique for the D2O leakage monitoring system

  15. The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

    International Nuclear Information System (INIS)

    A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

  16. Development of the advanced CANDU technology -Development of basic technology for HWR design

    International Nuclear Information System (INIS)

    It is believed that it is easier for Korea to become self-reliant in PHWR technology than in PWR technology, mainly because of the lower design pressure and temperature and because of the simplicity, economy, flexibility of the fuel cycle in comparison with PWR systems. Even though one has no doubt about the safety and the economics of the PHWR's that are now being operated or constructed in Korea. It is necessary to develop the advanced design technology for even safer and more economical PHWR systems to overcome the ever growing international resistance to sharing of nuclear technology and to meet the even more stringent requirements for the future public acceptance of nuclear power. This study is to develop the more advance design technology compared to the existing one, especially in the field of reactor physics, safety systems and safety evaluation to realize the above requirements. 71 tabs., 147 figs., 143 refs. (Author)

  17. Development of the advanced CANDU technology -Development of basic technology for HWR design-

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Hoh Chun; Lee, Sang Yong; Suk, Soo Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    It is believed that it is easier for Korea to become self-reliant in PHWR technology than in PWR technology, mainly because of the lower design pressure and temperature and because of the simplicity, economy, flexibility of the fuel cycle in comparison with PWR systems. Even though one has no doubt about the safety and the economics of the PHWR`s that are now being operated or constructed in Korea, it is necessary to develop the advanced design technology for even safer and more economical PHWR systems to overcome the ever growing international resistance to sharing of nuclear technology and to meet the even more stringent requirements for the future public acceptance of nuclear power. This study is to develop the more advance design technology compared to the existing one, by performing in-depth studies especially in the field of reactor physics, safety systems and safety evaluation to realize the above requirements. 90 figs, 50 tabs, 38 refs. (Author).

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

    International Nuclear Information System (INIS)

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

  19. 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

  20. CANDU passive shutdown systems

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. Investigation of acoustic and vibratory aspects with CIRCUS code. Application to the coolant pumps of the future Candu-9 type reactor

    International Nuclear Information System (INIS)

    In nuclear reactors, coolant pumps are the main sources of noise. The acoustic waves propagate along pipes and by resonance effects can lead to important and damaging vibrations. AVA and CIRCUS are 2 codes dealing with acoustic and vibration phenomena. They have been used, compared and applied to the different configurations of pumps in the future Candu-9 reactor. It is highlighted that a coupling between mechanical and acoustic aspects should be taken into account since resonance peaks could be amplified in the discharge line. (A.C.)

  4. 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.

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

    International Nuclear Information System (INIS)

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 deg. C. (author)

  6. 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)

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. Current issues in the management of low- and intermediate-level radioactive wastes from Ontario Hydro's CANDU reactors

    International Nuclear Information System (INIS)

    Nuclear generating stations (NGSs) in Canada are operated by utilities in Ontario, Quebec, and New Brunswick. Ontario Hydro, with a committed nuclear program of 13,600 MW(electric) is the major producer of CANDU pressurized heavy-water reactor (PHWR) low- and intermediate-level radioactive wastes. All radioactive wastes with the exception of irradiated fuel are processed and retrievably stored at a centralized facility at the Bruce Nuclear Power Development site. Solid-waste classifications and annual production levels are given. Solid-waste management practices at the site as well as the physical, chemical, and radiochemical characteristics of the wastes are well documented. The paper summarizes types, current inventory, and estimated annual production rate of liquid waste. Operation of the tritium recovery facility at Darlington NGS, which removes tritium from heavy water and produces tritium gas in the process, gives rise to secondary streams of tritiated solid and liquid wastes, which will receive special treatment and packaging. In addition to the treatment of radioactive liquid wastes, there are a number of other important issues in low- and intermediate-level radioactive waste management that Ontario Hydro will be addressing over the next few years. The most pressing of these is the reduction of radioactive wastes through in-station material control, employee awareness, and improved waste characterization and segregation programs. Since Ontario Hydro intends to store retrievable wastes for > 50 yr, it is necessary to determine the behavior of wastes under long-term storage conditions

  10. Analysis of the source term formation in a severe accident initiated by end fitting failure in CANDU type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, Marin; Constantin, Alina; Apostol, Minodora [Institute for Nuclear Research, Pitesti (Romania)

    2014-04-15

    CANDU type reactors have some peculiarities in initiation and progression of severe accidents. In the present paper one of the specificities - the End Fitting Failure accident - is analysed from the point of view of source term formation. The accident is initiated by a failure of the re-fueling machine. Fuel bundles are ejected in the re-fuelling machine room and fuel elements suffer a significant fragmentation by mechanical impact and by the rapid increase of the temperature. A direct transfer of the fission products occurs directly to the containment. The source term in the containment and also the source term to the environment is calculated supposing an open venting communication to the external atmosphere. The simulation is performed by using the ASTEC code in coupled calculation CPA-IODE-ISODOPE-DOSE option. The evolution of the distributions for the most important released fission products is presented for different regions and for different hosts. The most important factors of influence on the source term formation are identified and discussed. (orig.)

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

    International Nuclear Information System (INIS)

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

  12. 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.

  13. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

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

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    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.

  15. 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

  16. Modelling the activity of 99Tc the primary heat transport system of a CANDU reactor

    International Nuclear Information System (INIS)

    A physical model has been developed to describe the coolant activity behaviour of 99Tc, during constant and reactor shutdown operations. This analysis accounts for the fission production of technetium and molybdenum, in which their chemical form and volatility is determined by a thermodynamic treatment using a Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix and vaporization from the fuel-grain surface. Based on several in-reactor tests with defective fuel elements, and as supported by the thermodynamic analysis, the model accounts for the washout of molybdenum from the defective fuel on reactor shutdown. The model also considers the recoil release of both 99Mo and 99Tc from uranium contamination, as well as a corrosion source due to activation of 98Mo. The model has provided an estimate of the activity ratio 99Tc/137Cs in the ion-exchange columns of the Darlington Nuclear Generating Station, i.e., 6 x 10-6 (following ∼200 d of steady reactor operation) and 4 x 10-6 (with reactor shutdown). These results are consistent with that measured by the Battelle Pacific Northwest Laboratories with a mixed-bed resin-sampling device installed in a number of Pressurized Water Reactor and Boiling Water Reactor plants. (author)

  17. Advanced gas-cooled reactors (AGR)

    International Nuclear Information System (INIS)

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given

  18. Assessing CANDU requirements for irradiation - Research facilities

    International Nuclear Information System (INIS)

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

  19. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  20. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    An integrated approach to plant life management has been developed for CANDU reactors. Strategies, methods, and procedures have been developed for assessment of critical systems structures and components and for implementing a reliability centred maintenance program. A Technology Watch program is being implemented to eliminate 'surprises'. Specific work has been identified for 1998. AECL is working on the integrated program with CANDU owners and seeks participation from other CANDU owners

  1. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  2. Next Generation CANDU Performance Assurance

    International Nuclear Information System (INIS)

    AECL is developing a next generation CANDU design to meet market requirements for low cost, reliable energy supplies. The primary product development objective is to achieve a capital cost substantially lower than the current nuclear plant costs, such that the next generation plant will be competitive with alternative options for large-scale base-load electricity supply. However, other customer requirements, including safety, low-operating costs and reliable performance, are being addressed as equally important design requirements. The main focus of this paper is to address the development directions that will provide performance assurance. The next generation CANDU is an evolutionary extension of the proven CANDU 6 design. There are eight CANDU 6 units in operation in four countries around the world and further three units are under construction. These units provide a sound basis for projecting highly reliable performance for the next generation CANDU. In addition, the next generation CANDU program includes development and qualification activities that will address the new features and design extensions in the advanced plant. To limit product development risk and to enhance performance assurance, the next generation CANDU design features and performance parameters have been carefully reviewed during the concept development phase and have been deliberately selected so as to be well founded on the existing CANDU knowledge base. Planned research and development activities are required only to provide confirmation of the projected performance within a modest extension of the established database. Necessary qualification tests will be carried out within the time frame of the development program, to establish a proven design prior to the start of a construction project. This development support work coupled with ongoing AECL programs to support and enhance the performance and reliability of the existing CANDU plants will provide sound assurance that the next generation

  3. 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)

  4. 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

  5. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  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. 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.

  8. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

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

  9. 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)

  10. Update of operating experience with cold-worked Zr-2.5%Nb pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Zr-2.5 Nb pressure tubes are now used in all CANDU reactors. To ensure they perform reliably, their performance is carefully monitored. Both in situ inspection and sampling and testing techniques for tubes periodically removed from reactors have been developed. The data from these inspections and tests, together with models developed from research programs give confidence that pressure tubes will function effectively and safely for their design life. This presentation will describe how service life affects changes in the major material parameters in pressure tubes and the resulting maintenance activities resulting from those changes. Thermal creep, irradiation creep and irradiation growth change the dimensions during service, and axial elongation due to growth and sag due to creep in the older reactors have resulted in major maintenance programs. However, the dimensional changes continue to follow the behaviour predicted by the design equations and in the newer reactors should not limit service life. Extensive in situ sampling and the analysis of the tubes recently removed from Pickering Unit 3 indicate that hydrogen ingress into the pressure tubes from corrosion on the inside surface is very low and tests on irradiated material indicate that it should continue to remain low. The ingress rate from the annulus gas side can be significant if the integrity of the oxide on the outside surface is not maintained as a barrier. To maintain te integrity of the autoclave oxide, the recommended annulus gas is carbon dioxide, with oxygen addition, and adequate flow must be ensured. An explanation of the cause of relatively high hydrogen concentrations in a few Pickering A Zr-2.5% Nb pressure tubes has been developed defining the role of annulus side ingress. The model developed to predict the time and conditions to initiate blisters in pressure tubes that are in contact with their calandria tubes has been validated by the inspection, removal and examination of tubes and gives

  11. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  12. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    -plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  13. Advances in reactor safety research

    International Nuclear Information System (INIS)

    The Nuclear Safety Project is an important part of the German reactor safety research programme. It works on problems concerning safety and environemental risks of LWR reactors and reprocessing plants and investigates accident consequences. At the 1978 annual meeting, the core behaviour on cooling and reactivity disturbances was discussed, as well as release, retention, and possible radiological effects of radioactive pollutants. Among other subjects, fission product retention in LWR reactors and reprocessing plants were reported on as well as hypothetic core meltdown. (orig.)

  14. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  15. 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)

  16. 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)

  17. CANDU 9 nuclear power plant design description

    International Nuclear Information System (INIS)

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

  18. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    International Nuclear Information System (INIS)

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  19. Analytical chemistry requirements for advanced reactors

    International Nuclear Information System (INIS)

    The nuclear power industry has been developing and improving reactor technology for more than five decades. Newer advanced reactors now being built have simpler designs which reduce capital cost. The greatest departure from most designs now in operation is that many incorporate passive or inherent safety features which require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. India is developing the Advanced Heavy Water Reactor (AHWR) in its plan to utilise thorium in nuclear power program

  20. 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)

  1. 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)

  2. 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)

  3. Isotope distributions in primary heat transport and containment systems during a severe accident in CANDU type reactor

    International Nuclear Information System (INIS)

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code. The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivation to begin with a simplified model. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU loss of coolant accident sequence (CATHENA code results). The source term of FPs introduced into the PHT was estimated by ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species were obtained by using SOPHAEROS module. The distributions within the containment are obtained by the CPA module (thermalhydraulic calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts of the containment and different hosts. (authors)

  4. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  5. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  6. 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

  7. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

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

  8. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  9. Advanced Reactor Development in the United States

    International Nuclear Information System (INIS)

    In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE

  10. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  11. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  12. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  13. A high-speed data acquisition system to measure low-level current from self-powered flux detectors in CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Self-powered flux detectors are used in CANDU nuclear power reactors to determine the spatial neutron flux distribution in the reactor core for use by both the reactor control and safety systems. To establish the dynamic response of different types of flux detectors, the Chalk River Nuclear Laboratories have an ongoing experimental irradiation program in the NRU research reactor for which a data acquistion system has been developed. The system described in this paper is used to measure the currents from the detectors both at a slow, regular logging interval, and at a rapid, adaptive rate following a reactor shutdown. Currents that range from 100 pA to 1 mA full scale can be measured from up to 38 detectors and stored at sampling rates of up to 20 samples per second. The dynamic characteristics of the detectors can be computed from the stored records. The data acquisition system comprises a DEC LSI-11/23 microcomputer, dual cartridge disks, floppy disks, a hard copy and a video display terminal. The RT-11 operating system is used and all application programs are written in FORTRAN

  14. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

    The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

  15. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  16. Relative safeguards risks of advanced reactor concepts

    International Nuclear Information System (INIS)

    The purpose of this report is to develop a procedure to quantitatively assess the risk to diversion of the nuclear material in the fuel cycle for seven advanced reactor design concepts (LWR-Pu, LMFBR, HTGR, GCFR, MSBR, LWBR, HWR). Each stage in each of the seven reactor fuel cycles is evaluated and the result of the evaluation is a comparison among the various fuel cycles. This method of evaluation is used to determine the stages in a nuclear reactor fuel cycle that are most susceptible to diversion; it also gives an indication of what factors contribute to that susceptibility

  17. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  18. Advancement of light water reactor technology

    International Nuclear Information System (INIS)

    The Japanese technology of light water reactors is based on the technology imported from abroad around 1970, and the experience has been accumulated by the construction, operation and repair of light water reactors as well as the countermeasures to various troubles, moreover, the improvement and standardization of light water reactors have been promoted. As the result, recently the high capacity ratio has been attained, and the LWR technology has firmly taken root in Japan. The Subcommittee for the Advancement of Light Water Reactor Technology of the Advisory Committee for Energy has examined the subjects of technical development and the way the development should be in order to decide the strategy to advance LWR technology, and drawn up the interim report. The change of situation around the LWRs in Japan and the necessity to advance the technology, the target of advancing LWR technology and the subjects of the technical development, the system for the technical development and the securement of fund, and international cooperation are reported. The subjects of development are the pursuit of higher reliability and economic efficiency, the extension of plant life, the improvement of repairability and the reduction of radiation exposure, the improvement of operational capability, the reduction of wastes, the techniques for reactor decommissioning and the diversified location. (Kako, I.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. CANDU-BLW-250

    International Nuclear Information System (INIS)

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

  3. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

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

  4. 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)

  5. Evolutionary CANDU 9 plant improvements

    International Nuclear Information System (INIS)

    The CANDU 9 is a 935 MW(e) nuclear power plant (NPP) based on the multi-unit Darlington and Bruce B designs with additional enhancements from our ongoing engineering and research programs. Added to the advantages of using proven systems and components, CANDU 9 offers improvement features with enhanced safety, improved operability and maintenance including a control centre with advanced man-machine interface, and improved project delivery in both engineering and construction. The CANDU 9 NPP design incorporated safety enhancements through careful attention to emerging licensing and safety issues. The designers assessed, revised and evolved such systems as the moderator, end shield, containment and emergency core cooling (ECC) systems while providing an integrated final design that is more passive and severe-accident-immune. AECL uses a feedback process to incorporate lessons learned from operating plants, from current projects experiences and from the implementation or construction phase of previous projects. Most of the requirements for design improvements are based on a systematic review of current operating CANDU stations in the areas of design and reliability, operability, and maintainability. The CANDU 9 Control Centre provides plant staff with improved operability and maintainability capabilities due to the combination of systematic design with human factors engineering and enhanced operating and diagnostics features. The use of advanced engineering tools and modem construction methods will reduce project implementation risk on project costs and schedules. (author)

  6. Series lecture on advanced fusion reactors

    International Nuclear Information System (INIS)

    The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

  7. Methods of evaluation of accuracy with multiple essential parameters for eddy current measurement of pressure tube to calandria tube gap in CANDU reactors

    International Nuclear Information System (INIS)

    The purpose of inspection qualification of a particular inspection system is to show that it meets applicable inspection specification requirements. Often a requirement of the inspection system is that it meets a particular accuracy. In the case of a system with multiple inputs accompanied by additional influential parameters, calculation of the system's output accuracy can be formidable. Measurement of pressure-tube to calandria tube gap in CANDU reactors using an eddy current based technique is presented as a particular example of a system where multiple essential parameters combine to generate a final uncertainty for the inspection system. This paper outlines two possible methods of calculating such a system's accuracy, and discusses the advantages and disadvantages of each. (author)

  8. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  9. Advances in PHWR design

    International Nuclear Information System (INIS)

    Recent advances by AECL in improved performance, cost reduction and safety improvement of CANDU reactors are described. Topics include: computer-aided design tools, up-front licensing, site utilization, plant life management, construction techniques, plant control, safety-critical software, advanced fuels, human-machine interface, heat sinks, radiation protection, feedback to design, emergency core cooling and probabilistic safety assessment

  10. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  11. Development of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Future nuclear power plants should not only have the features of improved safety and economic competitiveness but also provide a means to resolve spent fuel storage problems by minimizing volume of high level wastes. It is widely believed that liquid metal reactors (LMRs) have the highest potential of meeting these requirements. In this context, the LMR development program was launched as a national long-term R and D program in 1992, with a target to introduce a commercial LMR around 2030. Korea Advanced Liquid Metal Reactor (KALIMER), a 150 MWe pool-type sodium cooled prototype reactor, is currently under the conceptual design study with the target schedule to complete its construction by the mid-2010s. This paper summarizes the KALIMER development program and major technical features of the reactor system. (author)

  12. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  13. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  14. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  15. Electrical, control and information systems in the Enhanced CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    De Grosbois, J.; Raiskums, G.; Soulard, M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2011-07-01

    This paper describes the electrical, control, and information system (EC and I) design feature improvements of the Enhanced CANDU 6 (EC6). These additional features are carefully integrated into the EC6 design platform, and are engineered with consideration of operational feedback, human factors, and leveraging the advantages of digital instrumentation and control (I and C) technology to create a coherent I and C architecture in support of safe and high performance operation. The design drivers for the selection of advanced features are also discussed. The EC6 nuclear power plant is a mid-sized Pressurized Heavy Water Reactor design, based on the highly successful CANDU 6 family of power plants, and upgraded to meet today's Canadian and international safety requirements and to satisfy Generation 3 design expectations. (author)

  16. 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

  17. System-integrated modular advanced reactor (SMART)

    International Nuclear Information System (INIS)

    Since the Kori nuclear power plant unit 1, the first nuclear power plant unit ever dedicated in Korea, began commercial operations with a generating capacity of 587 MW in 1978, much research and development has been conducted in the nuclear industry. In the middle 1980s, the Korean standard nuclear power plant (KSNP) was first developed under the 'nuclear power promotion plan' promulgated by the government with reference to system 80 of ABBCE of the USA. Applying indigenously accumulated technologies and up-to-date design standards from both home and abroad, the initial KSNP project began with the construction of the Younggwang NPP units No. 3 and 4. In addition, the Korea Atomic Energy Research Institute (KAERI) designed and constructed a high performance multipurpose research reactor based on experience in the operation of previous reactors and accumulated nuclear technology. Timed with completion of construction in April 1995, the reactor was named HANARO (high-flux advanced neutron application reactor), which, in Korean means, 'uniqueness'. In the middle of the 1990s, research and development was launched related to small and medium sized reactors (SMRs) to promote the utilization of nuclear energy. SMRs are under development worldwide for various purposes such as district heating, seawater desalination, nuclear ship propulsion, as well as electricity production. Generally, modern SMRs for power generation are expected to have greater simplicity of design, economy of mass production, and reduced capital costs. Many SMRs also have advantages of reactor safety and economics by implementing advanced design concepts and technology. Since 1997, KAERI has been developing the system-integrated modular advanced reactor (SMART), an advanced integral pressurized water reactor (PWR). The SMART is a promising, advanced SMR and has an integral type reactor with a rated thermal power of 330 MW. All major primary components, such as reactor core, steam generator (SG), main

  18. Engineering design of advanced marine reactor MRX

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    JAERI has studied the design of an advanced marine reactor (named as MRX), which meets requirements of the enhancement of economy and reliability, by reflecting results and knowledge obtained from the development of N.S. Mutsu. The MRX with a power of 100 MWt is intended to be used for ship propulsion such as an ice-breaker, container cargo ship and so on. After completion of the conceptual design, the engineering design was performed in four year plan from FY 1993 to 1996. (1) Compactness, light-weightiness and simplicity of the reactor system are realized by adopting an integral-type PWR, i.e. by installing the steam generator, the pressurizer, and the control rod drive mechanism (CRDM) inside the pressure vessel. Because of elimination of the primary coolant circulation pipes in the MRX, possibility of large-scale pipe break accidents can be eliminated. This contributes to improve the safety of the reactor system and to simplify the engineered safety systems. (2) The in-vessel type CRDM contributes not only to eliminate possibilities of rod ejection accidents, but also to make the reactor system compact. (3) The concept of water-filled containment where the reactor pressure vessel is immersed in the water is adopted. It can be of use for emergency core cooling system which maintains core flooding passively in case of a loss-of-coolant accident. The water-filled containment system also contributes essentially light-weightness of the reactor system since the water inside containment acts as a radiation shield and in consequence the secondary radiation shield can be eliminated. (4) Adoption of passive decay heat removal systems has contributed in a greater deal to simplification of the engineered safety systems and to enhancement of reliability of the systems. (5) Operability has been improved by simplification of the whole reactor system, by adoption of the passive safety systems, advanced automatic operation systems, and so on. (J.P.N.)

  19. Local AREA networks in advanced nuclear reactors

    International Nuclear Information System (INIS)

    The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

  20. Advanced simulation for fast reactor design

    International Nuclear Information System (INIS)

    Full text: This talk broadly reviews recent research aimed at applying advanced simulation techniques specifically to fast neutron reactors. By advanced simulation we generally refer to attempts to do more science-based simulation - that is, to numerically solve the three-dimensional governing physical equations on fine scales and observe and study the holistic phenomena that emerge. In this way simulation is treated more akin to a traditional physical experiment, and can can be used both separately and in conjunction with physical experiments to develop more accurate predictive theories on reactor behavior. Many existing fast reactor modeling tools were developed for last generation's computational resources. They were built by engineers and physicists with deep physical insight - insight that both shaped and was informed by existing theory, and was underpinned by a vast repository of experimental data. Their general approach was to develop models that were tailored to varying degrees to the details of the reactor design, using free model parameters that were subsequently calibrated to match existing experimental data. The resulting codes were thus extremely useful for their specific purpose but highly limited in their predictive capability (neutronics to a lesser degree). They tended to represent more the state-of-the-art in our understanding rather than tools of exploration and innovation. Recently, a number of researchers have attempted to study the feasibility of solving more fundamental governing equations on realistic, three-dimensional geometries for different fast reactor sub-domains. This includes solving the Navier-Stokes equations for single-phase sodium flow (Direct Numerical Simulation, Large Eddie Simulation, and Reynolds Averaged Navier Stokes Equations) in the core, upper plenum, primary and intermediate loop, etc.; the non-homogenized transport equations at very fine group, angle, and energy discretization, and thermo-mechanical feedback based on

  1. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  2. Advanced Fast Reactor - 100 - Design Overview

    International Nuclear Information System (INIS)

    The Advanced Fast Reactor-100 (AFR-100) is a small modular sodium-cooled fast reactor with an electrical power output of 100MWe. The AFR-100 has a long-lived core that does not require refueling for 30 years. The concept contains various innovations such as a small compact modular core (both vented and non-vented fuel pins), advanced core shielding materials, a compact fuel handling system, advanced electromagnetic pumps, compact intermediate heat exchangers, and a direct reactor auxiliary cooling system. These advanced systems and components were adopted in order to reduce the overall size of the primary heat transport system (and therefore the overall commodities and cost), enhance safety, and to improve overall plant performance. This paper presents the summary results of a year-long study that culminated in the design of two primary heat transport configurations for the AFR-100. The paper describes those innovations and shows how they are integrated into the overall AFR-100 primary heat transport system design. (author)

  3. Role of advanced reactors for sustainable development

    International Nuclear Information System (INIS)

    Availability of energy is an important prerequisite for socio-economic development in all parts of the world. Nuclear energy is an essentially unlimited energy resource with the potential to provide energy in the form of electricity, district heat and process heat under environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of national safety requirements, economic competitiveness and public acceptance. Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power plants. This experience forms a sound basis for further improvements. Nuclear programmes in Member States are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety. The IAEA, as a global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes. The paper gives an overview about trends in advanced reactor development programmes in Member States and the possible application of advanced reactors for electricity generation and heat production. (author)

  4. Development of CP{sub -}2D, a two dimensional transport FFCP computer code for CANDU reactor neutronic analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

    The code CP{sub -}2D was developed in INR Pitesti, between 1997-2001. It is a transport code in the first flight collision probability formalism, able to treat exactly a lot of complicated geometry (such as CANDU clusters, TRIGA and PWR fuel assemblies). The first version CP{sub -}2D1.0 was released in 1998. The second, CP{sub -}2D2.0, was released in 1999 and uses a multistratified coolant model for CANDU loss of coolant accident analysis. The third version, CP{sub -}2D3.0 (2000), have incorporated a generalized burning scheme. The exact treatment of the geometry is based on the power of the FFCP method and on the decomposition of the target geometry into factorial geometries. An user friendly interface was developed in 2001. It helps the user to prepare the input, to view the introduced geometry and to collect the interest information directly into tables and figure. (author)

  5. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  6. Mirror Advanced Reactor Study interim design report

    International Nuclear Information System (INIS)

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design

  7. Advanced methods in teaching reactor physics

    International Nuclear Information System (INIS)

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

  8. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

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

    International Nuclear Information System (INIS)

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

  10. Development of advanced nuclear reactors in Russia

    International Nuclear Information System (INIS)

    Several advanced reactor designs have been so far developed in Russia. The AES-91 and AES-92 plants with the VVER-1000 reactors have been developed at the beginning of 1990. However, the former design has been built in China and the latest which is certified meeting European Utility Requirements is being built in India. Moreover, the model VVER-1500 reactor with 50-60 MWd/t burn-up and an enhanced safety was being developed by Gidropress about 2005, excepting to be completed in 2007. But, this schedule has slipped in favor of development of the AES-2006 power plant incorporating a third-generation standardized VVER-1200 reactor of 1170 MWe. This is an evolutionary development of the well-proven VVER-1000 reactor in the AES-92 plant, with longer life, greater power and efficiency and its lead units are being built at Novovoronezh II, to start operation in 2012-13. Based on Atomenergoproekt declaration, the AES-2006 conforms to both Russian standards and European Utility Requirements. The most important features of the AES-2006 design are mentioned as: a design based on the passive safety systems, double containment, longer plant service life of 50 years with a capacity factor of 92%, longer irreplaceable components service life of 60 years, a 28.6% lower amount of concrete and metal, shorter construction time of 54 months, a Core Damage Frequency of 1x10-7/ year and lower liquid and solid wastes by 70% and 80% respectively. The presented paper includes a comparative analysis of technological and safety features, economic parameters and environmental impact of the AES-2006 design versus the other western advanced reactors. Since the Bushehr phase II NPP and several other NPPs are planning in Iran, such analysis would be of a great importance

  11. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  12. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  13. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

  14. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  15. CANDU 9 - the CANDU product to meet customer and regulator requirements now and in the future

    International Nuclear Information System (INIS)

    CANDU reactors developed under Canadian licensing regulations that placed the primary responsibility for safety on the licensee. The Atomic Energy Control Board (AECB), Canada's nuclear regulatory agency, state in their regulations what is expected in terms of safety performance so that designers are free to propose the best means of meeting this performance. This goal-oriented approach, besides encouraging innovation, allowed CANDU to be licensed in other jurisdictions. The latest design - the large, single unit, CANDU 9 - explicitly incorporates licensability in Canada through a formal AECB review of the design; lessons learned from licensing CANDU 6 in Asian countries, particularly with Wolsong 2, 3 and 4 in Korea, and more recently with Qinshan in China; utility requirements for modem evolutionary plants; and emerging international standards for safety, sponsored or issued by the IAEA. By combining the assurance of acceptability in Canada with compliance with foreign and international requirements, CANDU 9 becomes an internationally licensable product. (author)

  16. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  17. Final report on study of advanced fission power reactor development for the United States. Volume III

    International Nuclear Information System (INIS)

    This three-volume set details a multistage research study on the comparison of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. Volume III presents the basic data and other input information utilized in the process of the study. Detailed reactor and fuel-cycle information is contained on the HTGR, GCFR, MSBR, LWBR, and CANDU-HWR, as obtained by BCL staff from numerous sources. These data have been assembled, critically reviewed, and modified where necessary to assure consistency, both internally for each reactor system, and between reactor types. Similar, comparative information was required for the LWR mass balances, fuel cycle, etc., but this was generally accepted from ERDA sources without the critical evaluation to which the other data were subjected. Other detailed information that relates to the objective of the study, e.g., projections of nuclear power growth and uranium availability, generic safeguards information, description of computerized models, etc., is incorporated for ready reference

  18. Design and performance of CANDU steam generators

    International Nuclear Information System (INIS)

    The recirculating U-tube steam generators at Ontario Hydro's Pickering and Bruce CANDU Nuclear Generating Stations have demonstrated excellent reliability over many years of operation. Tube failures have been rare, contributing to high plant capacity factors. Of the approximately 390,000 steam generator U-tubes at the Pickering and Bruce plants, one tube leak has occurred at Pickering to date and 12 tubes have leaked at Bruce in a total of 89 reactor years of operation. The success of these units is attributed to an age old respect for steam generation equipment, an ongoing pursuit of research and design advancements, and extensive cooperative efforts on the part of the utility, the system designer, and the supplier. The supplier's involvement began with the steam generator design and manufacture for the very first CANDU plant. The utility's involvement began with their direct participation in the earliest stages of nuclear plant design. The utilities' contribution to the success of these units relates to the rigorous approach used in definition of requirements, in understanding the supplier's design in detail and in making every effort to operate, monitor, and service the equipment with appropriate care. This paper presents the supplier's and utility's approach to achieving these remarkable results. In order to present these viewpoints separately and in some detail, this paper is divided into two parts: Part 1 - The Equipment Supplier Perspective, and Part 2 - The Utility Perspective

  19. 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)

  20. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including