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Sample records for candu reactor building

  1. Prediction of hydrogen distribution in the reactor building in CANDU6 plant

    International Nuclear Information System (INIS)

    Jin, Y.; Song, Y.

    2008-01-01

    The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

  2. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  3. Build your own Candu reactor

    International Nuclear Information System (INIS)

    Carruthers, J.

    1979-01-01

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

  4. Advanced CANDU reactors

    International Nuclear Information System (INIS)

    Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

    1988-12-01

    AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

  5. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  6. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    Lipsett, J.J.

    1989-01-01

    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

  7. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    McKay, Paul.

    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

  8. CANDU reactor - supporting the nuclear renaissance

    International Nuclear Information System (INIS)

    Oberth, R.

    2010-01-01

    'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

  9. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Azeez, S.; Alizadeh, A.; Girouard, P.

    2005-01-01

    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

  10. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  11. Homogeneous Thorium Fuel Cycles in Candu Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

    2009-06-15

    The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

  12. AECL's advanced CANDU reactor - the ACR

    International Nuclear Information System (INIS)

    Alizadeh, Ala; Allsop, Peter; Hedges, Ken; Hopwood, Jerry; Yu, Stephen

    2003-01-01

    The ACR, the next generation CANDU design, represents the next step in development of the CANDU family of designs. AECL has achieved significant incremental improvements to the mid-size CANDU 6 nuclear power plant through successive projects, both in design and in project delivery. Building on this knowledge base, AECL is continuing to adapt the CANDU design to develop the ACR. This paper summarizes the ACR design features, which include major improvements in economics, inherent safety characteristics, performance and construction methods. Aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs, the ACR is an evolutionary design based on the very successful CANDU 6 reactor. The new ACR product is specifically designed to produce power at a cost competitive with other forms of power generation while achieving short construction times, improved safety, international licensability, high investor returns, and low investor risk. It achieves these targets by taking advantage of the latest advances in both pressure-tube and pressure-vessel reactor technologies and experience. The flexibility and development potential of the fuel channel approach also enables designs to be developed that address priorities identified in international long-term specification programs such as the US Department of Energy (DOE) sponsored Generation IV program and IAEA hosted INPRO program. ACR-700 can be built in 36 months with a 48 month project duration, and deliver a lifetime capacity factor in excess of 90%. 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. Based on the status of the ACR design and AECL's on-going experience delivering reactor projects on-time and on-budget, the first ACR could be in service by

  13. CANDU technology for generation III + AND IV reactors

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    2005-01-01

    Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU?reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU ReactorTM (ACRTM), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R and D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating

  14. Thermosyphoning in the CANDU reactor

    International Nuclear Information System (INIS)

    Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

    1984-01-01

    Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

  15. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    Azeez, S.; Girouard, P.

    2006-01-01

    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

  16. Reactors based on CANDU technology

    International Nuclear Information System (INIS)

    Bjegun, S.V.; Shirokov, S.V.

    2012-01-01

    The paper analyzes the use CANDU technology in world nuclear energy. Advantages and disadvantages in implementation of this technology are considered in terms of economic and technical aspects. Technological issues related to the use of CANDU reactors and nuclear safety issues are outlined. Risks from implementation of this reactor technology in nuclear energy of Ukraine are determined

  17. Pre-service proof pressure and leak rate tests for the Qinshan CANDU project reactor buildings

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The Qinshan CANDU Project Reactor Buildings (Units 1 and 2) have been successfully tested for the Pre-Service Proof Pressure and Integrated Leak Rate Tests. The Unit 1 tests took place from May 3 to May 9, 2002 and from May 22 to May 25, 2002, and the Unit 2 tests took place from January 21 to January 27, 2003. This paper discusses the significant steps taken at minimum cost on the Qinshan CANDU Project, which has resulted in a) very good leak rate (0.21%) for Unit 1 and excellent leak rate (0.130%) for Unit 2; b) continuous monitoring of the structural behaviour during the Proof Pressure Test, thus eliminating any repeat of the structural test due to lack of data; and c) significant schedule reduction achieved for these tests in Unit 2. (author)

  18. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  19. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

    1987-01-01

    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. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Hopwood, Jerry; Duffey, Romney B.; Yu, Steven; Torgerson, Dave F.

    2006-01-01

    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

  1. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    Kakaria, B. K.

    1994-01-01

    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

  2. Corrosion control in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Lesurf, J.E.

    1974-01-01

    Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

  3. MATLAB/SIMULINK model of CANDU reactor for control studies

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2006-01-01

    In this paper a MATLAB/SIMULINK model is developed for a CANDU type reactor. The data for the reactor are taken from an Indian PHWR, which is very similar to CANDU in its design. Among the different feedback mechanisms in the core of the reactor, only xenon has been considered which plays an important role in spatial oscillations. The model is verified under closed loop scenarios with simple PI controller. The results of the simulation show that this model can be used for controller design and simulation of the reactor systems. Adding models of the other components of a CANDU reactor would ultimately result in a complete model of CANDU plant in MATLAB/SIMULINK. (author)

  4. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

    2006-01-01

    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 550 0 C. 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

  5. MATLAB/SIMULINK platform for simulation of CANDU reactor control system

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2007-01-01

    In this paper a simulation platform for CANDU reactors' control system is presented. The platform is built on MATLAB/SIMULINK interactive graphical interface. Since MATLAB/SIMULINK are powerful tools to describe systems mathematically, all the subsystems in a CANDU reactor are represented in MATLAB's language and are implemented in SIMULINK graphical representation. The focus of the paper is on the flux control loop of CANDU reactors. However, the ideas can be extended to include other parts in CANDU power plants and the same technique can be applied to other types of nuclear reactors and their control systems. The CANDU reactor model and xenon feedback model are also discussed in this paper. (author)

  6. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Slack, J.; Norton, J.L.; Malkoske, G.R.

    2003-01-01

    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

  7. 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.

  8. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    Liu, W.S.; Gold, M.; Sills, H.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Mottram, R.; Millard, J.W.F.; Purdy, P.

    2011-01-01

    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)

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

    International Nuclear Information System (INIS)

    Mottram, R.; Millard, J.W.F.; Purdy, P.

    2011-01-01

    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)

  11. Conceptual designs for very high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    2000-07-01

    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 310{sup o}C, and exits at {approx}410{sup o}C. 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

  12. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

    2000-01-01

    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 310 o C, and exits at ∼410 o C. 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

  13. Localization of CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ala

    2010-09-15

    The CANDU pressurized heavy water reactor's principal design features suit it particularly well for technology transfer and localization. When the first commercial CANDU reactors of 540 MWe entered service in 1971, Canada's population of less than 24 million supported a 'medium' level of industrial development, lacking the heavy industrial capabilities of larger countries like the USA, Japan and Europe. A key motivation for Canada in developing the CANDU design was to ensure that Canada would have the autonomous capacity to build and operate nuclear power reactors without depending on foreign sources for key components or enriched fuel.

  14. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Zamonsky, G.

    1991-01-01

    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) [es

  15. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    Bain, A.S.

    1997-01-01

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada 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

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

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    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)

  17. Fuel deposits, chemistry and CANDU reactor operation

    International Nuclear Information System (INIS)

    Roberts, J.G.

    2013-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  18. An approach to neutronics analysis of candu reactors

    International Nuclear Information System (INIS)

    Gul, S.; Arshad, M.

    1982-12-01

    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)

  19. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    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.

  20. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Norton, J.L.; Slack, J.

    2002-01-01

    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, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

  1. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

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

    International Nuclear Information System (INIS)

    Lane, A.D.; McDonnell, F.N.; Griffiths, J.

    1988-11-01

    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

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

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Rizoiu, A.C.

    2009-01-01

    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 ThO 2 -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

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

    International Nuclear Information System (INIS)

    Hopwood, J.; Soulard, M.; Hastings, I.J.

    2011-01-01

    Atomic Energy of Canada (AECL) is finalizing development of the Enhanced CANDU 6 (EC6), which incorporates the CANDU 6's well-proven features, and adds 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)

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

    International Nuclear Information System (INIS)

    Hopwood, J.; Soulard, M.; Hastings, I.J.

    2011-01-01

    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)

  6. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  7. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  8. Thermo-Economic Assessment of Advanced,High-Temperature CANDU Reactors

    International Nuclear Information System (INIS)

    Spinks, Norman J.; Pontikakis, Nikos; Duffey, Romney B.

    2002-01-01

    Research underway on the advanced CANDU examines new, innovative, reactor concepts with the aim of significant cost reduction and resource sustainability through improved thermodynamic efficiency and plant simplification. The so-called CANDU-X concept retains the key elements of the current CANDU designs, including heavy-water moderator that provides a passive heat sink and horizontal pressure tubes. Improvement in thermodynamic efficiency is sought via substantial increases in both pressure and temperature of the reactor coolant. Following on from the new Next Generation (NG) CANDU, which is ready for markets in 2005 and beyond, the reactor coolant is chosen to be light water but at supercritical operating conditions. Two different temperature regimes are being studied, Mark 1 and Mark 2, based respectively on continued use of zirconium or on stainless-steel-based fuel cladding. Three distinct cycle options have been proposed for Mark 1: the High-Pressure Steam Generator (HPSG) cycle, the Dual cycle, and the Direct cycle. For Mark 2, the focus is on simplification via a Direct cycle. This paper presents comparative thermo-economic assessments of the CANDU-X cycle options, with the ultimate goal of ascertaining which particular cycle option is the best overall in terms of thermodynamics and economics. A similar assessment was already performed for the NG CANDU. The economic analyses entail obtaining cost estimates of major plant components, such as heat exchangers, turbines and pumps. (authors)

  9. Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

    1999-01-01

    High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

  10. Application of fuel management calculation codes for CANDU reactor

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun

    2003-01-01

    Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

  11. 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.

  12. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

    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

  13. New flux detectors for CANDU 6 reactors

    International Nuclear Information System (INIS)

    Cuttler, J.M.; Medak, N.

    1992-06-01

    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

  14. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  15. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

    Hedges, K. R.; Allen, P. J.; Hopwood, J. M.

    2000-01-01

    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)

  16. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Nanis, R.

    2000-01-01

    Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . 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 D 2 O 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, D 2 O management is required to preserve it. (author)

  17. ROP design for Enhanced CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J.; Scherbakova, D; Kastanya, D.; Ovanes, M. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    The Enhanced CANDU 6 (EC6) nuclear power plant is a mid-sized pressurized heavy water reactor design, based on the highly successful CANDU 6 (C6) family of power plants, upgraded to meet today's Canadian and international safety requirements and to satisfy Generation III expectations. The EC6 reactor is equipped with two independent Regional Overpower Protection (ROP) systems to prevent overpowers in the reactor fuel. The ROP system design, retaining the traditional C6 methodology, is determined to cover the End-of-Life (EOL) reactor core condition since the reactor operating/thermal margin gradually decreases as plant equipment ages. Several design changes have been incorporated into the reference C6 plant to mitigate the ageing effect on the ROP trip margin. This paper outlines the basis for the EC6 ROP physics design and presents the ROP related improvements made in the EC6 design to ensure that full power operation is not limited by the ROP throughout the entire life of the reactor. (author)

  18. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-01-01

    CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

  19. The Canadian R and D program targeted at CANDU reactors

    International Nuclear Information System (INIS)

    Moeck, E.O.

    1988-01-01

    CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

  20. Comparison of neutron parameters between a CANDU and ACR reactors; Comparacao de parametros neutronicos entre um reator CANDU e um ACR

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Gabriel H.P.; Silva, Clarysson A.M. da; Pereira, Claubia, E-mail: gabrielhpd@yahoo.com.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    CANDU (Canadian Deuterium Uranium) is a type of reactor that uses heavy water (D{sub 2}O) as a moderator and as a refrigerant. Despite having chemical characteristics similar to light water (H{sub 2}O), heavy water has a high moderation ratio for neutrons. This feature enables CANDU to use natural uranium as fuel. However, research has evaluated the possibility of using H{sub 2}O as a refrigerant and D{sub 2}O as a moderator aiming at reducing the volume of heavy water. Such changes would imply the use of lightly enriched uranium due to the presence of H{sub 2}O. In this context, the concept of ACR (Advanced CANDU Reactor) has been developed. This reactor has an innovative design which combines of the current CANDU with the characteristics of PWR (Pressurized Water Reactor) type reactors. Studies by AECL (Atomic Energy Canada Limited) show that compared to CANDU, the ACR presents a cost reduction in construction, improved firing performance, improved operation safety and longer life. The present work aims to evaluate, in steady state, some of the main neutron parameters of CANDU-6 and ACR-1000. The MCNPX 2.6.0 code (Monte Carlo N-Particle eXtended -version 2.6.0) was used to simulate such types of reactors. The results show that the models configured in the MCNPX adequately reproduce the neutron behavior of the studied reactors. These models may be used in future work for analysis of fuel burn and evolution.

  1. Radioactive effluents from CANDU 6 reactors during normal operation

    International Nuclear Information System (INIS)

    Boss, C.R.; Allsop, P.J.

    1995-12-01

    During routine operation of a CANDU 6 reactor, various gaseous, liquid, and solid radioactive wastes are generated. The layout of the CANDU 6 reactor and the design of its systems ensure that these are minimized, but small quantities of gaseous and liquid wastes are continually discharged at very low concentrations. This report discusses the make-up of these chronically generated gaseous and liquid effluents. From a safety perspective, the doses to individual members of the public resulting from radioactive wastes chronically discharged from CANDU 6 reactors have been negligible. Similarly, doses to the regional and global populations have been negligible, generally less than 0.001% of background. While far below regulatory limits, releases of tritium, noble gases and gross β - -γ have been the most radiologically significant emissions, while radioiodine and particulates have had the greatest potential to deliver public dose. (author). 8 refs., 16 tabs., 3 figs

  2. Conceptual designs for advanced, high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

    2000-07-01

    AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

  3. Development Directions For CANDU and Slowpoke Reactors

    International Nuclear Information System (INIS)

    Brooks, Gordon L.

    1990-01-01

    This paper provides a broader-based discussion of overall development directions foreseen for CANDU reactors, particularly those which have further evolved sine the earlier paper. The paper then discusses development directions for the Slowpokes Energy System which is a small nuclear heat source intended to meet local heating needs for building complexes and municipal heating systems. In evolving a sound development direction, it is necessary, firstly, to address the question of requirements, viz., what are the requirements which future nuclear power plants must satisfy if they are to be successful? Today, some in the nuclear industry believe that the most important of such requirements relates to the need for 'safer' reactors. Indeed, some proponents of this view would seem to suggest that if only we could develop such 'safer' reactors, suddenly all of our problem s with public acceptance would disappear and utilities would form long lines waiting to purchase such marvellous machines. I do not share such a simplistic view nor, indeed, do many of my colleagues in the international nuclear power industry

  4. Assessment studies on plutonium recycle in CANDU reactors

    International Nuclear Information System (INIS)

    1978-11-01

    This paper describes the CANDU reactor system in detail and goes on to explore the potential for using the system with plutonium recycle fuelling to improve fuel utilisation and to meet the long-term challenge of economic supplies of nuclear fuel. The paper includes comments on costs and non-proliferation aspects. It concludes that: recycle fuelling is feasible with little modification to the reactor design and no degradation of safety, and could offer over 50% savings in uranium requirements. However, recycle fuelling costs do not appear competitive with natural uranium in the CANDU system under current economic conditions

  5. A fast-running fuel management program for a CANDU reactor

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2000-01-01

    A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

  6. CANDU - a versatile reactor for plutonium disposition or actinide burning

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Gagnon, M.J.N.; Boczar, P.G.; Ellis, R.J.; Verrall, R.A.

    1997-10-01

    High neutron economy, on-line refuelling, and a simple fuel-bundle design result in a high degree of versatility in the use of the CANDU reactor for the disposition of weapons-derived plutonium and for the annihilation of long-lived radioactive actinides, such as plutonium, neptunium, and americium isotopes, created in civilian nuclear power reactors. Inherent safety features are incorporated into the design of the bundles carrying the plutonium and actinide fuels. This approach enables existing CANDU reactors to operate with various plutonium-based fuel cycles without requiring major changes to the current reactor design. (author)

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

    International Nuclear Information System (INIS)

    Prodea, I.; Mărgeanu, C. A.; Rizoiu, A.; Olteanu, G.

    2014-01-01

    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)

  8. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  9. A New In-core Production Method of Co-60 in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

    2016-05-15

    This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

  10. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    International Nuclear Information System (INIS)

    Allen, Francis; Bonin, Hugues

    2008-01-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  11. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

    2008-07-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  12. Fuel Management Study for a CANDU reactor Using New Physics Codes Suite

    International Nuclear Information System (INIS)

    Kim, Won Young; Kim, Bong Ghi; Park, Joo Hwan

    2008-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. The primary reactivity control in a CANDU reactor is the on-power refueling on a daily basis and an additional reactivity control is provided through an individual reactivity device movement, which includes 21 adjusters, 6 liquid zone controllers, 4 mechanical control absorbers and 2 shutdown systems. The refueling in CANDU is carried out on power and this makes the in-core fuel management different from that in a reactor refueled during shutdowns. The objective of a fuel management is to determine a fuel loading and fuel replacement procedure which will result in a minimum total unit energy cost in a safe and reliable operation. In this article, the in-core fuel management for the CANDU reactor was studied by using the new physics code suite of WIMS-IST/DRAGON-IST/RFSP-IST with the model of Wolsong-1 NPP

  13. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  14. The water chemistry of CANDU PHW reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-01-01

    This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

  15. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  16. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  17. Decommissioning of the CANDU-PHW reactor

    International Nuclear Information System (INIS)

    Unsworth, G.N.

    1977-04-01

    This report contains the results of a study of various aspects of decommissioning of reactors. The study places in perspective the size of the job, the hazards involved, the cost and the environmental impact. The three internationally agreed ''stages'' of decommissioning, namely, mothballing, entombment, and dismantling are defined and discussed. The single unit 600 MW(e) CANDU is chosen as the type of reactor on which the discussion is focussed but the conclusions reached will provide a basis for judgement of the costs and problems associated with decommissioning reactors of other sizes and types. (author)

  18. Wet steam turbines for CANDU-Reactors

    International Nuclear Information System (INIS)

    Westmacott, C.H.L.

    1977-01-01

    The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

  19. Role of water lubricated bearings in Candu reactors

    International Nuclear Information System (INIS)

    Kumar, Ashok N.

    1999-01-01

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    2005-01-01

    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

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

    International Nuclear Information System (INIS)

    Charak, I.; Kier, P.H.

    1995-04-01

    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)

  3. Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

    2003-01-01

    The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

    2013-07-01

    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)

  5. Fuel channel design improvements for large CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

  6. A JAVA applet to simulate a CANDU reactor

    International Nuclear Information System (INIS)

    Varin, E.; Desarmenien, J.

    2004-01-01

    Here we present a CANDU nuclear power plant simulator, directly available on a web page. The developed applet has two mains objectives: to expose the CANDU technology to a large public on the internet; and to construct a realistic simulator to be used as a pedagogical tool for nuclear introduction to high school or under-graduate students. The neutronic behavior and control algorithms of the reactor are simulated. Java programming language enables a very flexible environment for public information and user interaction with the plant. Examples of shutdown and power maneuver are explained. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

    2007-06-01

    The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

  9. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  10. Role of operator response guidelines in CANDU 9 design program

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    2000-01-01

    The CANDU 9 is a large version of the CANDU Pressurized Heavy Water Reactor (PHWR) system developed in Canada. With an electrical output of approximately 935 MWe, the CANDU 9 complements the established mid-size CANDU 6 (700 MWe) and makes use of proven technology updated with state of the art features resulting from ongoing development. The CANDU 9 builds on the reactor and process system designs of the operating Darlington and Bruce B plants, and incorporates a modified CANDU 6 station layout, as well as improved construction methods and operational features. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9; all key components (reactor core, steam generators, coolant pumps, pressure tubes, etc.) are of the same design as those proven in service in the operating CANDU power stations. Including Probabilistic Safety Assessment (PSA) as part of the CANDU 9 design process from the outset of the program was seen as key to ensuring completeness of safety related requirements. The PSA work provided an in-depth understanding of the plant response to various postulated accidents. As well, the time frame for recovery and the related operator actions were identified. This information together with AECL's experience in supporting the development of Emergency Operating Procedures (EOPs) for the operating CANDU reactors are the basis for preparation of CANDU 9 Operator Response Guidelines (ORGs). Technical content, format and human factors considerations adopted for the ORGs are such that these can be readily converted to EOPs. The scope of ORGs includes generic as well as event specific ORGs. This dual approach is required to provide defense-in-depth. This paper describes the process used to prepare ORGs for the CANDU 9 reactor and discusses important benefits gained from the application of ORGs as input to the control center design and future preparation of the EOPs. (author)

  11. Space-time neutronic analysis of postulated LOCA's in CANDU reactors

    International Nuclear Information System (INIS)

    Luxat, J.C.; Frescura, G.M.

    1978-01-01

    Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

  12. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.

    1997-01-01

    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 D 2 O 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)

  13. Analysis of ASTEC code adaptability to severe accident simulation for CANDU type reactors

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2008-01-01

    In order to prepare the adaptation of the ASTEC code to CANDU NPP severe accident analysis two kinds of activities were performed: - analyses of the ASTEC modules from the point of view of models and options, followed by CANDU exploratory calculation for the appropriate modules/models; - preparing the specifications for ASTEC adaptation for CANDU NPP. The paper is structured in three parts: - a comparison of PWR and CANDU concepts (from the point of view of severe accident phenomena); - exploratory calculations with some ASTEC modules- SOPHAEROS, CPA, IODE, CESAR, DIVA - for CANDU type reactors specific problems; - development needs analysis - algorithms, methods, modules. (authors)

  14. THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

    Directory of Open Access Journals (Sweden)

    D. KASTANYA

    2013-10-01

    Full Text Available 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.

  15. Once-through uranium thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Ozdemir, S.; Cubukcu, E.

    2000-01-01

    In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)

  16. Economics of CANDU

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.

    1981-02-01

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

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

    International Nuclear Information System (INIS)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-01-01

    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

  18. 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.

  19. Dimensional measurement of fresh fuel bundle for CANDU reactor

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

    2005-01-01

    This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

  20. Effect of elevated temperatures on heavy concrete structural strength in Qinshan phase 3 CANDU 6 reactor buildings

    International Nuclear Information System (INIS)

    Alikhan, S.; Khan, A.F.; Chen, S.

    2005-01-01

    Heavy concrete is commonly used inside the Qinshan Phase 3 CANDU 6 reactor buildings for radiation shielding functions in order to provide access to key areas during reactor operation. In some cases, the heavy concrete elements are also structural elements. Concerns have been raised about the functional performance of the heavy concrete structural elements, specifically the primary heat transport pump (PHTS) supporting slabs, surrounding the feeder cabinets when subjected to elevated temperatures between 42 degree C and 121 degree C and their corresponding temperature gradients on a long-term basis during the normal operation of the plant. This paper presents the results of a test investigation on the strength of heavy concrete under elevated temperature conditions being experienced by the heavy concrete structural elements around the feeder cabinet to confirm that these structural elements meet their functional requirements. The loading conditions consist subjecting the specimens to the elevated temperatures and temperature gradient noted during commissioning, including the effect of epoxy coating. The heavy concrete mix proportion and materials of the test samples (ilmenite aggregate and Portland cement) are identical to those used for heavy concrete structural elements surrounding the feeder cabinet. Subsequent to the confirmation of the functional requirements of the heavy concrete structural elements, alarm limits are recommended for these structural elements. (authors)

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

    International Nuclear Information System (INIS)

    Chung, Jang Hwan; Suk, Ho Cheon; Jeong, Moon Ki; Park, Joo Hwan; Jeong, Heung Joon; Jeon, Ji Soo; Kim, Bok Deuk

    1994-07-01

    This project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year Out-of-pile hydraulic tests for the prototype of CANFLEX bundle was conducted in the CANDU-hot test loop at KAERI. Thermalhydraulic analysis with the assumption of CANFLEX-NU fuel loaded in Wolsong-1 was performed by using thermalhydraulic code, and the thermal margin and T/H compatibility of CANFLEX bundle with existing fuel for CANDU-6 reactor have been evaluated. (Author)

  2. Audit of ECCS Availability for CANDU Reactors with an extended O/H interval

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong Soo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    KINS conducts regulatory periodic inspections of the safety and performance of each nuclear installation during the planned outage every 20 months, pursuant to the Atomic Energy Act. For CANDU reactors, planned outage or overhaul (O/H) have been performed every 15 months. KHNP has been making efforts to extend the O/H intervals of CANDU reactors into 20 months since 2001. Low ECCS availability is one of the regulatory pending issues in the related licensing

  3. Life extension of CANDU reactor cores

    International Nuclear Information System (INIS)

    Millard, J.; Kerker, J.; Albert, M.

    2011-01-01

    Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

  4. Leak-before-break experience in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Moan, G.D.; Coleman, C.E.

    1988-01-01

    In the Canada deuterium uranium (CANDU) reactor, each of the ∼ 400 hot pressure tubes containing the fuel bundles and the pressurized heat transport water is surrounded and insulated from the cold moderator by a calandria tube. The pressure tubes are made from cold-worked Zr-2.5 Nb with a minimum wall thickness of 4.19 mm, and the calandria tubes are made from annealed Zircaloy-2 with a minimum wall thickness of 1.37 mm. The annulus between these two tubes contains an inert gas. Leak-before-break has developed into an operational tool in CANDU reactors to prevent unstable failure of pressure tubes. A procedure for leak detection and reactor response has been developed from the use of the annulus gas, whose dew point is measured to ascertain if leaks have crept into the annulus. The characteristics of the crack are used to establish the response time for leak detection. The reactor is required to be shut down before the length of the slowly growing crack has reached the critical stage. This critical crack length, determined using slit burst tests on tubes, is the crack length at which the crack growth becomes unstable. The most likely crack growth mechanism is delayed hydride cracking. This mechanism requires three conditions to occur simultaneously: the material must be sensitive to delayed hydride cracking; zirconium hydrides must be present in the material; and the tensile stress must be sufficiently great

  5. Modernization of the NESTLE-CANDU reactor simulator and coupling to scale-processed cross sections

    International Nuclear Information System (INIS)

    Hart, S.; Maldonado, G.I.

    2012-01-01

    The original version of the NESTLE computer code for CANDU applications, herein referred as the NESTLE-CANDU or NESTLE-C program, was developed under sponsorship by the CNSC as a “stand-alone” program. In fact, NESTLE-C emerged from the original version of NESTLE, applicable to light water reactors, which was written in FORTRAN 77 to solve the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). Accordingly, NESTLE-C can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source or eigenvalue initiated transient problems for CANDU reactor fuel arrangements and geometries. This article reports a recent conversion of the NESTLE-C code to the Fortran 90 standard, in addition, we highlight other code updates carried out to modularize and modernize NESTLE-C in a manner consistent with the latest updates performed with the parent NESTLE code for light water reactor (LWR) applications. Also reported herein, is a simulation of a CANDU reactor employing 37-element fuel bundles, which was carried out to highlight the SCALE to NESTLE-C coupling developed for two-group collapsed and bundle homogenized cross-section generation. The results presented are consistent with corresponding simulations that employed HELIOS generated cross-sections. (author)

  6. Modernization of the NESTLE-CANDU reactor simulator and coupling to scale-processed cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Hart, S.; Maldonado, G.I. [Univ. of Tennessee, Knoxville, Tennessee (United States)

    2012-07-01

    The original version of the NESTLE computer code for CANDU applications, herein referred as the NESTLE-CANDU or NESTLE-C program, was developed under sponsorship by the CNSC as a “stand-alone” program. In fact, NESTLE-C emerged from the original version of NESTLE, applicable to light water reactors, which was written in FORTRAN 77 to solve the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). Accordingly, NESTLE-C can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source or eigenvalue initiated transient problems for CANDU reactor fuel arrangements and geometries. This article reports a recent conversion of the NESTLE-C code to the Fortran 90 standard, in addition, we highlight other code updates carried out to modularize and modernize NESTLE-C in a manner consistent with the latest updates performed with the parent NESTLE code for light water reactor (LWR) applications. Also reported herein, is a simulation of a CANDU reactor employing 37-element fuel bundles, which was carried out to highlight the SCALE to NESTLE-C coupling developed for two-group collapsed and bundle homogenized cross-section generation. The results presented are consistent with corresponding simulations that employed HELIOS generated cross-sections. (author)

  7. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

    Jagannath, D.V.; Oldaker, I.E.

    1976-01-01

    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)

  8. 1200 FPD refuelling simulation of RUFIC fuel in a CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soon Young; Jeong, Chang Joon; Min, Byung Joo; Suk, Ho Chun

    2001-07-01

    The refuelling strategy of RUFIC (Recovered Uranium Fuel in CANDU) fuel as a high-burnup fuel for a CANDU 6 reactor is studied to determine the achievable operation characteristics of the fuel and reactor. In this study, three refuelling schemes of 4-, 2-, and 3-bundle shift for 0.92 w/o RUFIC fuel in an CANDU 6 reactor were individually evaluated through 1200 FPD(Full Power Day)refuelling simulaltions where the 0.92 w/o RUFIC is equivalent to CANFLEX 0.9 w/o SEU(Slightly Enriched Uranium) in reactivity and burnup respects. The computer code system used for this study is WIMS-AECL/DRAGON/RFSP. The results simulated for the case of 4-bundle shift refueling scheme shows that the peak maximum channel power and peak maximum CPPF(Channel Power Peaking Factor)of 7228 kW and 1.175, respectively, seems too high to maintain the available operating margins, because some data of the maximum channel power exceed the operating limit(7070 kW based on the Technical Specifications of Wolsong 3 and 4 Units). Whereas, the results simulated for the case of 2-bundle shift refuelling scheme shows that sufficient operating margin could be secured where the peak maximum channel power and peak maximum CPPF were 6889 kW and 1.094, respectively. However, the channel refuelling rate (channels/day) of the 2-bundle shift refuelling scheme is twice that of the 4-bundle shift refuelling scheme, and hence the 2-bundle shift refuelling would not be an economical refuelling scheme for the RUFIC fuel bundles. Therefore, a 3-bundle shift refuelling scheme for the RUFIC fuel in CANDU 6 reactor was also studied by the 1200 FPD refuelling simulation. As a result, it is found that all the operating parameters in the 3-bundle shift case are achivable for the CANDU 6 reactor operation, and the channel refuelling rate of 2.88 channels/day seems to be attractive compared to the refuelling rate of 4.32 channels/day in the 2-bundle shift case.

  9. Diagnostic Technology Development for Core Internal Structure in CANDU reactor

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

    2005-04-01

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods 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 CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

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

    International Nuclear Information System (INIS)

    Sood, S.K.; Fong, C.; Kalyanam; Woodall, K.B.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

    2000-09-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

    2000-09-01

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

    International Nuclear Information System (INIS)

    Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan

    2008-01-01

    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

  14. CANDU lectures

    International Nuclear Information System (INIS)

    Rouben, B.

    1984-06-01

    This document is a compilation of notes prepared for two lectures given by the author in the winter of 1983 at the Institut de Genie Nucleaire, Ecole Polytechnique, Montreal. The first lecture gives a physical description of the CANDU reactor core: the nuclear lattice, the reactivity mechanisms, their functions and properties. This lecture also covers various aspects of reactor core physics and describes different calculational methods available. The second lecture studies the numerous facets of fuel management in CANDU reactors. The important variables in fuel management, and the rules guiding the refuelling strategy, are presented and illustrated by means of results obtained for the CANDU 600

  15. Leak before break experience in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Moan, G.D.; Coleman, C.E.

    1988-04-01

    The paper describes how the requirements for Leak-Before-Break are met in CANDU reactors. The requirements are based on operational and laboratory experience. After the onset of leakage in a fuel channel from a delayed hydride crack, time is available to the operator to take action before the crack grows to an unstable length. The time available is calculated using different models which use crack growth data from small specimen tests. When the results from crack growth behaviour experiments, carried out on components removed from reactor are used in the model, the time available for operator response is about 100 hours

  16. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    Smith, A.I.

    1984-01-01

    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

  17. CANDU reactor core simulations using fully coupled DRAGON and DONJON calculations

    International Nuclear Information System (INIS)

    Varin, E.; Marleau, G.

    2006-01-01

    The operating CANDU-6 reactors are refueled on-power to compensate for the reactivity loss due to fuel burnup. In order to predict the core behavior, fuel bundle burnups and local parameter information need to be tracked. The history-based approach has been developed to follow local parameter as well as history effect in CANDU reactors. The finite reactor diffusion code DONJON and the lattice code DRAGON have been coupled to perform reactor follow-up calculations using a history-based approach. A coupled methodology that manages the transfer of information between standard DONJON and DRAGON data structures has been developed. Push-through refueling can be taken into account directly in cell calculations. Using actual on-site information, an isotopic core content database has been generated with coupled DONJON and DRAGON calculations. Moreover calculations have been performed for different local parameters. Results are compared with those obtained using standard cross section generation approaches

  18. CANDU design options with detritiation

    International Nuclear Information System (INIS)

    Wren, D.J.; Hart, R.S.

    1997-01-01

    CANDU reactors include a number of auxiliary systems to manage the inventory, purification, clean-up and isotopic purity of the heavy water used in the moderator and heat transport system. These systems are designed and installed to treat the moderator and heat transport water in separate parallel systems. One of the reasons for this parallel approach to heavy water management is the tritium inventory in the heavy water. Different levels of tritium accumulate in the moderator and heat transport system during reactor operation, with the moderator water having a much higher tritium concentration. Strict separation of the high- tritium-concentration moderator water from the low-tritium-concentration heat transport system water is an integral component of the CANDU design and operating strategy to limit potential releases of tritium to the containment building atmosphere. AECL is developing a new cost-effective technology for the detritiation of heavy water based on the Combined Electrolysis and Catalytic Exchange (CECE) process. This detritiation technology has the potential to be integrated into the heavy water management systems of a CANDU reactor. On-line detritiation could be used to limit the concentration of tritium in the moderator and also to detritiate any water collected within the containment building from other sources. The availability of economic detritiation technology would provide a flexibility to redesign some of the auxiliary heavy water management systems. In particular, there is potential to eliminate some of the duplication in the current management systems and also reduce costs by reclassifying some reactor systems that would have lower maximum tritium concentrations. This paper discusses some of the advantages of detritiation and some of the conceptual design options that detritiation would provide. The goal would be to lower the overall reactor cost with detritiation, but it is premature to assess whether this goal can be achieved. (author)

  19. A Comparative Study on the Refueling Simulation Method for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Do, Quang Binh; Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The Canada deuterium uranium (CANDU) reactor calculation is typically performed by the RFSP code to obtain the power distribution upon a refueling. In order to assess the equilibrium behavior of the CANDU reactor, a few methods were suggested for a selection of the refueling channel. For example, an automatic refueling channel selection method (AUTOREFUEL) and a deterministic method (GENOVA) were developed, which were based on a reactor's operation experience and the generalized perturbation theory, respectively. Both programs were designed to keep the zone controller unit (ZCU) water level within a reasonable range during a continuous refueling simulation. However, a global optimization of the refueling simulation, that includes constraints on the discharge burn-up, maximum channel power (MCP), maximum bundle power (MBP), channel power peaking factor (CPPF) and the ZCU water level, was not achieved. In this study, an evolutionary algorithm, which is indeed a hybrid method based on the genetic algorithm, the elitism strategy and the heuristic rules for a multi-cycle and multi-objective optimization of the refueling simulation has been developed for the CANDU reactor. This paper presents the optimization model of the genetic algorithm and compares the results with those obtained by other simulation methods.

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

    International Nuclear Information System (INIS)

    Kuran, Sermet; Yang, Dezi

    2012-01-01

    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

  1. A three-dimensional operational transient simulation of the CANDU core with typical reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae; Park, Kyung Seok; Park, Jong Woon [Institute for Advanced Engineering, Taejon (Korea, Republic of)

    1995-07-01

    This paper describes the results of simulation of a CANDU operational transient problem (re-startup after short shutdown) using the Coupled Reactor Kinetics(CRKIN) code developed previously with CANDU Reactor Regulating System (RRS) logic. The performance in the simulation is focused on investigating the behaviours of neutron power and regulating devices in accordance with the changes of xenon concentration following the operation of the RRS.

  2. 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)

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

    International Nuclear Information System (INIS)

    Catana, Alexandru; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D 2 O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D 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)

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

    International Nuclear Information System (INIS)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  6. Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

    International Nuclear Information System (INIS)

    Shen, W.

    2012-01-01

    Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)

  7. Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shen, W. [Candu Energy Inc., 2285 Speakman Dr., Mississauga, ON L5B 1K (Canada)

    2012-07-01

    Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)

  8. Transport-diffusion coupling for Candu reactor core follow-Up

    International Nuclear Information System (INIS)

    Varin, E.; Marleau, G.; Chambon, R.

    2003-01-01

    We couple the finite reactor diffusion code DONJON and the lattice code DRAGON, called for simplicity DD, to perform reactor follow-up calculations using a history-based approach. In order to do this, a new DD module is developed. This module manages the transfer of information between standard DONJON and DRAGON data structures. Moreover, it stores in a history data structure the global and local parameters required for cell calculations as well as the isotopic composition of the various materials present in each cell of the reactor. We then implement in DD a parallel algorithm to perform history-based Candu reactor calculations. Here, we assign to each processor a specific number of fuel channels to be analyzed. The DRAGON cell calculations for each of the fuel bundles associated with the specified channels are performed on the same processor in order to minimize communication time. Only the macroscopic cross section libraries are exchanged between the processor. Since the amount of data exchanged is relatively small, we expect to obtain an ideal speed-up. The coupling is tested for the analysis of a simplified Candu reactor model with 4 x 4 channels each containing 4 bundles. A 100 full-power days core tracking sequence with 16 refueling steps is studied. Results are coherent with those obtained using more approximate approaches. Parallel speed-up is near optimal indicating that the use of this approach for more realistic reactor calculations should be pursued. (authors)

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

    Indian Academy of Sciences (India)

    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 ...

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

    International Nuclear Information System (INIS)

    Wren, D.J.; Popov, N.; Snell, V.G.

    2004-01-01

    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

  11. Optimal Refueling Pattern Search for a CANDU Reactor Using a Genetic Algorithm

    International Nuclear Information System (INIS)

    Quang Binh, DO; Gyuhong, ROH; Hangbok, CHOI

    2006-01-01

    This paper presents the results from the application of genetic algorithms to a refueling optimization of a Canada deuterium uranium (CANDU) reactor. This work aims at making a mathematical model of the refueling optimization problem including the objective function and constraints and developing a method based on genetic algorithms to solve the problem. The model of the optimization problem and the proposed method comply with the key features of the refueling strategy of the CANDU reactor which adopts an on-power refueling operation. In this study, a genetic algorithm combined with an elitism strategy was used to automatically search for the refueling patterns. The objective of the optimization was to maximize the discharge burn-up of the refueling bundles, minimize the maximum channel power, or minimize the maximum change in the zone controller unit (ZCU) water levels. A combination of these objectives was also investigated. The constraints include the discharge burn-up, maximum channel power, maximum bundle power, channel power peaking factor and the ZCU water level. A refueling pattern that represents the refueling rate and channels was coded by a one-dimensional binary chromosome, which is a string of binary numbers 0 and 1. A computer program was developed in FORTRAN 90 running on an HP 9000 workstation to conduct the search for the optimal refueling patterns for a CANDU reactor at the equilibrium state. The results showed that it was possible to apply genetic algorithms to automatically search for the refueling channels of the CANDU reactor. The optimal refueling patterns were compared with the solutions obtained from the AUTOREFUEL program and the results were consistent with each other. (authors)

  12. Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

    2017-01-01

    Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

  13. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    Slater, J.B.

    1986-03-01

    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

  14. Advanced CANDU reactor pre-licensing progress

    International Nuclear Information System (INIS)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C.

    2005-01-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  15. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

    1984-03-01

    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

  16. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

    1982-03-01

    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

  17. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    Jackson, H.A.; Horton, E.P.; Woodhead, L.W.; Fanjoy, G.R.

    1985-03-01

    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

  18. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

    1996-04-01

    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

  19. Response characteristics of self-powered flux detectors in CANDU reactors

    International Nuclear Information System (INIS)

    Allan, C.J.

    1978-05-01

    As part of the development of a new flux-detector assembly for future CANDU reactors, the sensitivities of a variety of vanadium, cobalt and platinum self-powered detectors have been determined in a simulated CANDU core installed in the ZED-2 test reactor at CRNL. While the vanadium and cobalt detectors had solid emitters, the platinum detectors were of two types, having either solid platinum emitters, or emitters consisting of a platinum sheath over an Inconel core. Almost all of the signal from the cobalt and vanadium detectors is due to neutron events in the emitters. For these detectors we have measured the total sensitivities per unit length. For the platinum detectors, reactor γ-rays and neutrons both contribute appreciably to the output signal, and in addition to the total sensitivity, we have determined the individual neutron and γ-ray sensitivities for these detectors. It was found that the detector sensitivities depend primarily on emitter diameter and that the observed variations can be fitted by means of power laws. (author)

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

    International Nuclear Information System (INIS)

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

    1980-11-01

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

  1. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Yu, S.; Pakan, M.; Soulard, M.

    2002-01-01

    AECL has developed a suite of technologies for Candu R 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. An emergency water injection system (EWIS) for future CANDU reactors

    International Nuclear Information System (INIS)

    Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

    2000-01-01

    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 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)

  3. Cobalt-60 production in CANDU reactors

    International Nuclear Information System (INIS)

    Ross, Michel; Lemire, Christian

    2002-01-01

    CANDU reactors can produce cobalt-60 very efficiently and with an interesting return on investment. This paper discusses what is needed to convert a CANDU reactor into a cobalt-60 producer: what are the different phases, the safety studies required, the physical modifications needed, and what is the minimum involvement of the utility owning the plant. The past ten years of experience of Hydro-Quebec as a cobalt-60 producer will be reviewed, including the management of the risk of both incident and electricity generation loss, and including the benefits for the utility and its personnel. Originally a simple metal used for centuries as a pigment, cobalt-59 today is transformed into cobalt-60, a radioactive element of unprecedented value. Well known in medicine for cancer treatment, cobalt-60 is also used to sterilize a wide range of disposable medical products used in hospitals and to sanitize pharmaceutical and cosmetic products. Cobalt-60 is proving to be a new and effective solution, in the food sector, for preserving harvests and controlling food-borne diseases, or to advantageously replace certain gases and chemical products which are suspected of being harmful or carcinogenic. There are also other applications, such as: hardening of some plastics, treatment of sewage sludge and elimination of harmful insect populations. With a half-life of 5,3 years, cobalt-60 is a metal not found in nature. It is a radioactive isotope produced by exposing stable nuclei of cobalt-59 to neutrons. One of the best places to find such an important neutron source is a nuclear reactor. High energy gamma rays are then emitted during the process of radioactive decay, where cobalt-60 seeks again its stable state

  4. The final report of ''on-the-job training'' on the CANDU reactor

    International Nuclear Information System (INIS)

    Kim, D.H.; Koh, B.J.

    1983-01-01

    This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capability for the nuclear safety evaluation in order to contribute the future safe operation of the CANDU nuclear power plant. The training has been excuted through three level courses as elementary, intermediate and ''on-the-job training'' at Wolsung power plant. The elementary course was introduction to the CANDU basics and fundamentals. The intermediate course was the more advanced course, and the detailed concepts and engineering explanations of the CANDU system had been instructed. The third course was the ''on-the-job training'' at the Wolsung plant site, which was the most emphasized course during the project. (Author)

  5. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.

  6. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu

    2016-01-01

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely

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

    International Nuclear Information System (INIS)

    Stratton, D.; Butt, C.

    1982-04-01

    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

  8. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

    1996-01-01

    There are many reasons for countries embarking on a CANDU R program to start with the natural uranium fuel cycle. Simplicity of fuel design, ease of fabrication, and ready availability of natural uranium all help to localize the technology and to reduce reliance on foreign technology. Nonetheless, at some point, the incentives for using natural uranium fuel may be outweighed by the advantages of alternate fuel cycles. The excellent neutron economy, on-line refuelling, and simple fuel-bundle design provide an unsurpassed degree of fuel-cycle flexibility in CANDU reactors. 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 two- to 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 dose 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

  9. Development of Operational Safety Monitoring System and Emergency Preparedness Advisory System for CANDU Reactors (I)

    International Nuclear Information System (INIS)

    Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho; Son, Han Seong; Song, Deok Yong

    2007-01-01

    As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code

  10. Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Richinson, P.J.

    1978-08-01

    The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

  11. The extension of the SWS period or CANDU reactors with particular reference to Douglas Point

    International Nuclear Information System (INIS)

    Bennett, C.R.

    1985-01-01

    The foregoing approach to the determination of the fate of a concrete containment building is worth much consideration. The expenditure of $10 8 or its escalated equivalent is too much to pay for the probable saving of fraction of a statistical life. The unquestioning adoption of the dogma of reactor dismantlement displays a complete misunderstanding of the numerics of ''risk'', even the place of reactor dismantling in the spectrum of nuclear risk. The position of the risk of reactor dismantling is more than an order of magnitude lower than the former of these. The most altruistic criterion for any engineering activity is the achievement of the greatest expected net benefit (or the least expected net detriment) when all the consequences of the activity are taken into account. As has been shown this criterion leads to the conclusion that, at least in CANDU reactors and particularly Douglas Point, there is apparently no reason why the S.W.S. period should not be extended indefinitely

  12. CANDU nuclear power system

    International Nuclear Information System (INIS)

    1981-01-01

    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

  13. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    Langstaff, J.H.

    1980-06-01

    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.)

  14. Some aspects of primary and secondary water chemistry in CANDU reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-09-01

    A brief review of the water chemistry in various circuits of CANDU reactors is given. Then, five particular aspects of recent work are highlighted: (i) Radiation Field Growth: in-reactor and out-reactor studies have related water chemistry to corrosion product deposition on fuel sheaths and subsequent contamination of out-core surfaces. (ii) Metal Oxide Solubility: novel techniques are being used to measure the solubilities of metal oxides at primary circuit conditions. (iii) Decontamination: the use of heavy water as coolant in CANDU reactors led to the development of a unique decontamination strategy and technique, called CAN-DECON, which has attracted the attention of operators of light-water reactors. (iv) Steam Generator Corrosion: mathematical modelling of the water chemistry in the bulk and crevice regions of nuclear steam generators, supported by chemical experiments, has shown why sea water ingress from leaking condensers can be damaging, and has provided a rapid way to evaluate alternative boiler water chemistries. (v) Automatic Control of Feedwater Chemistry: on-line automatic chemical analysis and computer control of feedwater chemistry provides All Volatile Treatment for normal operation with pure feedwater, and carefully controlled sodium phosphate addition when there is detectable sea-water ingress from leaking condensers. (author)

  15. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.

    1994-05-01

    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

  16. Mathematical modeling of CANDU-PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

    2003-07-01

    The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR). This study involved mathematical modeling of CANDU-PHWR to study its thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique. The reactor model includes the neutronic, reactivity, and fuel channel heat transfer. The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and the reactivity feed back due to the changes in the fuel temperature and coolant temperature. The CANDU-PHWR model was coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

  17. Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

    2014-07-01

    A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm{sup 2}s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

  18. Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

    2014-01-01

    A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm 2 s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

  19. Hydride blister formation simulation in Candu type reactors

    International Nuclear Information System (INIS)

    Otero, D.; Bollini, C.; Sangregorio, D.

    1992-01-01

    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)

  20. Improvement of top shield analysis technology for CANDU 6 reactor

    International Nuclear Information System (INIS)

    Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

    1996-07-01

    As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

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

    International Nuclear Information System (INIS)

    1987-05-01

    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

  2. Once-through thorium cycles in Candu reactors

    International Nuclear Information System (INIS)

    Milgram, M.S.

    1982-01-01

    In once-through thorium cycles pure thorium fuel bundles can be irradiated conjointly with uranium fuel bundles in a CANDU reactor with parameters judiciously chosen such that the overall fuel cycle cost is competitive with other possibilities - notably low-enriched uranium. Uranium 233 can be created and stockpiled for possible future use with no imperative that it be used unless future conditions warrant, and a stockpile can be begun independently of the state of reprocessing technology. The existence and general properties of these cycles are discussed

  3. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-15

    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.

  4. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-01

    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. CANDU 9 fuelling machine carriage

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, D J; Slavik, J F [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1997-12-31

    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.

  6. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Ullrich, D.J.; Slavik, J.F.

    1996-01-01

    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

  7. Objectives and techniques of an advanced safeguards system for the CANDU reactor

    International Nuclear Information System (INIS)

    Smith, R.M.; Zarecki, C.W.; Head, D.A.

    1981-01-01

    In 1975, Canada began to actively assist the IAEA with manpower and research and development efforts to meet this requirement for CANDU reactors. This paper describes various aspects of the CANDU safeguards scheme, including the containment and surveillance equipment that has been developed. It includes consideration of the following: objectives of the safeguards system, role of equipment in meeting system objectives, cost and maintenance of equipment, capabilities and limitations of equipment, and effectiveness of the scheme and equipment in providing assurance of diversion detection. 11 refs

  8. CANDU: study and review

    International Nuclear Information System (INIS)

    Morad, César M.; Santos, Thiago A. dos

    2017-01-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  9. CANDU: study and review

    Energy Technology Data Exchange (ETDEWEB)

    Morad, César M., E-mail: cesar.morad@usp.br [Universidade de São Paulo (POLI/USP), SP (Brazil). Escola Politécnica; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  10. Safety research for CANDU reactors

    International Nuclear Information System (INIS)

    Hancox, W.T.

    1982-10-01

    Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include: (i) fluid-dynamic and heat-transfer processes in the heat transport system during the blowdown and refilling phases of LOCAs; (ii) thermal and mechanical behaviour of fuel elements; (iii) thermal and mechanical behaviour of the fuel and the fuel-channel assembly in situations where the heavy-water moderator is the sink for decay heat produced in the fuel; (iv) chemical behaviour of fission gases that might be released into the reactor coolant and transported to the containment system; and (v) combustion of hydrogen-air-steam mixtures that would be produced if fuel temperatures were sufficiently high to initiate the zirconium-water reaction. The current status of the research on each of these topics is highlighted with particular emphasis on the conclusions reached to date and their impact on the continuing program

  11. CANDU - Canadian experience and expectations with the heavy-water reactor

    International Nuclear Information System (INIS)

    Foster, J.S.; Russell, S.H.

    1977-05-01

    The paper describes the evolution of the CANDU nuclear-power plants with particular reference to the objectives of safety, reliability and economy; the development of industrial capacity for the supply of fuel, components and heavy water; and the prospective development of advanced fuel cycles and the projected results. It provides data on radiation, releases, and exposures, internal and external to the power plants; plant availability, capacity factors and other performance data; heavy water production data with reference to safety, reliability, and economics; projections of the performance of CANDU reactors operating on a thorium-U-233 cycle and the development required to establish this cycle; and intent with respct to spent-fuel management and radioactive-waste storage. (author)

  12. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    International Nuclear Information System (INIS)

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

    2012-01-01

    The Enhanced CANDU 6 R (ECo R ) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  13. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    International Nuclear Information System (INIS)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin

    2016-01-01

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

  14. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

  15. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Charlebois, P.; Hart, R.S.; Hopkins, J.R.

    1998-01-01

    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

  16. Qinshan CANDU project open top construction method

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Wittann, K.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The significant schedule reductions achieved on the Qinshan CANDU Project were due in large part to the incorporation of advanced construction technologies in project design and delivery. For the Qinshan Project, a number of key advantages were realized through the use of the 'Open Top' construction method. This paper discusses the Qinshan Phase III CANDU Project Open Top implementation method. The Open Top method allowed major equipment to be installed simply, via the use of a Very Heavy Lift (VHL) crane and permitted the use of large-scale modularization. The advantages of Open Top construction, such as simplified access, more flexible project scheduling, improved construction safety and quality, and reduced labours are presented. The large-scale modularization of the Reactor Building Dousing System and the Open Top installation method and advantages relative to traditional CANDU 6 construction practices are also presented. Finally, major improvements for future CANDU plant construction using the Open Top method are discussed. (author)

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

    International Nuclear Information System (INIS)

    Burrill, K.A.; Guzonas, D.A.

    2002-01-01

    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)

  18. 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)

    1999-12-31

    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)

  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. Team CANDU : ready for the marketplace

    International Nuclear Information System (INIS)

    Howieson, J.Q.

    2007-01-01

    This paper outlines the partnership between AECL and a number of leading global nuclear suppliers to market the Candu power reactor. The mission of the CANDU team is to develop market opportunities for CANDU technology and deliver successful CANDU projects

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

    International Nuclear Information System (INIS)

    Stirling, A.J.; Payne, E.

    1985-01-01

    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

  2. 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.)

  3. Examination of core components removed from CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

    1988-11-01

    Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

  4. Modeling and simulation of CANDU reactor and its regulating system

    Science.gov (United States)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

  5. Evolution of CANDU reactor design

    International Nuclear Information System (INIS)

    Pon, G.A.

    1978-08-01

    The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

  6. 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.

  7. Leak detection capability in CANDU reactors

    International Nuclear Information System (INIS)

    Azer, N.; Barber, D.H.; Boucher, P.J.

    1997-01-01

    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

  8. Improvements of the reactivity devices modeling for the advanced CANDU reactor

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Dahmani, M.; Hebert, A.

    2008-01-01

    In the context of the ACR TM (Advanced CANDU Reactor), 3D transport calculations are required in order to simulate the reactivity devices located perpendicularly to the fuel channels. The computational scheme that is usually used for CANDU-6 and ACR reactors is based on a simplified supercell geometry in which the fuel clusters and devices are replaced by annuli. Recently, an exact modeling of 3D supercell configurations was introduced within the framework of the ACR calculations. However, with such a model, fine meshing requirements lead to problems that are very demanding in terms of computational resources. In this paper, we present improvements introduced in the ACR context to reduce the cost of the 3D supercell calculations. Two avenues of investigations are reported. First, the introduction of an accelerated characteristics method permits to reduce the computational burden of such calculations involving a large number of regions. In addition, contrarily to CANDU-6 supercell configurations, the ACR 3D geometry is prismatic and consequently a special tracking procedure can be used. This approach introduces no approximation and is significantly faster than the general 3D tracking technique. Thanks to these modifications in the computational procedure, 3D supercell calculations with a level of mesh discretization comparable to 2D cell configurations become affordable for industrial applications

  9. CANDU flexible and economical fuel technology in China

    Energy Technology Data Exchange (ETDEWEB)

    Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

    2014-07-01

    Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

  10. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  11. CANDU-PHW fuel management

    International Nuclear Information System (INIS)

    Frescura, G.M.; Wight, A.L.

    1982-01-01

    This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

  12. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Duffey, R.B.; Torgerson, D.F.

    2001-01-01

    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. Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

    International Nuclear Information System (INIS)

    Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

    2005-01-01

    In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

  14. Hybrid simulation of reactor kinetics in CANDU reactors using a modal approach

    International Nuclear Information System (INIS)

    Monaghan, B.M.; McDonnell, F.N.; Hinds, H.W.T.; m.

    1980-01-01

    A hybrid computer model for simulating the behaviour of large CANDU (Canada Deuterium Uranium) reactor cores is presented. The main dynamic variables are expressed in terms of weighted sums of a base set of spatial natural-mode functions with time-varying co-efficients. This technique, known as the modal or synthesis approach, permits good three-dimensional representation of reactor dynamics and is well suited to hybrid simulation. The hybrid model provides improved man-machine interaction and real-time capability. The model was used in two applications. The first studies the transient that follows a loss of primary coolant and reactor shutdown; the second is a simulation of the dynamics of xenon, a fission product which has a high absorption cross-section for neutrons and thus has an important effect on reactor behaviour. Comparison of the results of the hybrid computer simulation with those of an all-digital one is good, within 1% to 2%

  15. Reliability assessment of the fueling machine of the CANDU reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.

    1985-01-01

    Fueling of CANDU-reactors is carried out by two fueling machines, each serving one end of the reactor. The fueling machine becomes a part of the primary heat transport system during the refueling operations, and hence, some refueling machine malfunctions could result in a small scale-loss-of-coolant accident. Fueling machine failures and the failure sequences are discussed. The unavailability of the fueling machine is estimated by using fault tree analysis. The probability of mechanical failure of the fueling machine interface is estimated as 1.08 x 10 -5 . (orig.) [de

  16. Thermal Hydraulic Assessment for Loss of SDCS Event During the Outage of CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [Gnest, Inc. Taejon (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    During the outage(overhaul) of the nuclear power plant, there are several operating states other than the full power state, that is 'Hot-Zero Power', 'Depressurized-Cooldown', and 'Partially Drained'. Until now safety assessment has not been done much for this operating state of CANDU type reactor worldwide. For the accuracy and confidence of PSA for the CANDU outage, the safety analysis is necessary. At the first stage, we analyzed the thermal hydraulic characteristics and safety of the postulated event of loss of shutdown cooling system (SDCS) during the partially drained state which is the longest one in the middle of outage period. As an analysis tool, this study uses the best estimate thermal hydraulic code, RELAP5/CANDU which was modified according to the CANDU specific characteristics and based on RELAP5.Mod3.

  17. A study concerning tritium concentration evolution in the moderator of a CANDU reactor connected on-line to a detritiation facility

    International Nuclear Information System (INIS)

    Bidica, Nicolae; Bornea, Anisia

    2005-01-01

    The present work is a theoretical study on the tritium concentration evolution in the CANDU reactor moderator connected on-line with a detritiation facility. This study is based on a calculation model which takes into account the evolution curve of the tritium concentration in the absence of detritiation process in both the moderator and SPTC of the Unit 1 CANDU reactor at Cernavoda NPP. This study leads to determination of the tritium concentration evolution in the moderator in the presence of the detritiation process for both a range of intake flows and initial concentration. Also, the intake flow change will be analyzed for a detritiation facility as a function of tritium initial concentration existing in the moderator in the case of a survey of the detritiation over a given period of time. The conclusions of this study were the following: - an optimum of the detritiation factor can be determined; - detritiation starts at a lower value for the tritium concentration in moderator which reduces the strain upon the detritiation facility and therefore the costs of its building, maintenance and operation. (authors)

  18. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

    Baron, J.; Jarvis, G.N.; Dolbey, M.P.; Hayter, D.M.

    1986-01-01

    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) [pt

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

    International Nuclear Information System (INIS)

    1996-09-01

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

  20. Development of Realistic Safety Analysis Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Rho, G. H.

    2010-04-01

    The following 3 research items have been studied to develop and establish the realistic safety analysis and the associated technologies for a CANDU reactor. At the first, WIMS-CANDU which is physics cell code for a CANDU has been improved and validated against the physics criticality experiment data transferred through the international cooperation programs. Also an improved physics model to take into account the pressure tube creep was developed and utilized to assess the effects of the pressure tube creep of 0%, 2.5% and 5% diametral increase of pressure tube on core physics parameters. Secondly, the interfacing module between physics and thermal-hydraulics codes has been developed to provide the enhancement of reliability and convenience of the calculation results of the physics parameters such as power coefficient which was calculated by independent code systems. Finally, the important parameters related to the complex heat transfer mechanisms in the crept pressure tubes were identified to find how to improve the existing fuel channel models. One of the important parameters such as the oxidation model of Zr-steam reaction was identified, implemented and verified with the experimental data of the high pressure and temperature fuel channel and its model was utilized for CFD analysis of the crept pressure tube effect on the reactor safety. The results were also provided to validate the CATNENA models of the crept pressure tube and the effects of the pressure tube creep on the blowdown and post-blowdown phase during LOCA was assessed. The results of this study can be used to assess the uncertainty analysis of coolant void reactivity and the effects of the creep deformed pressure tubes on physics/TH/safety issues. Also, those results will be used to improve the current design and operational safety analysis codes, and to technically support the related issues to resolve their problems

  1. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

    1992-05-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

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

    International Nuclear Information System (INIS)

    Xie Zhongsheng; Huo Xiaodong

    2002-01-01

    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

  3. Severe Accident R and D for Enhanced CANDU-6 Reactors

    International Nuclear Information System (INIS)

    Nitheanandan, Thambiayah

    2012-01-01

    CANDU reactors possess a number of inherent of inherent and designed safety features that make them resistant to core damage accidents. The unique feature is the low temperature moderator surrounding the fuel channels, which can serve as an alternate heat sink. The fuel is surrounded by three water systems: heavy water primary coolant, heavy water moderator, and light water calandria vault and shield water. In addition, the liquid inventory in the steam generators is a fourth indirect heat sink, able to cool the primary coolant. The water inventories in the emergency core cooling system and the reserve water tank at the dome of the containment can also provide fuel cooling and water makeup to prevent severe core damage or mitigate the consequences of a severe core damage accident. An assessment of the adequacy of the existing severe accident knowledge base, to confidently perform consequence analyses for the Enhanced CANDU-6 reactor in compliance with regulatory requirements, was recently completed. The assessment relied on systematic Phenomena Identification and Ranking Tables (PIRT) studies completed domestically and internationally. The assessment recommends cost-effective R and D to mitigate the consequences of severe accidents and associated risk vulnerabilities

  4. Prospects of Using Reprocessed Uranium in CANDU Reactors, in the U.S. GNEP Program

    International Nuclear Information System (INIS)

    Ellis, Ronald James

    2007-01-01

    Current Global Nuclear Energy Partnership (GNEP) plans envision reprocessing spent fuel (SF) with view to minimizing high-level waste (HLW) repository use and recovering actinides (U, Np, Pu, Am, and Cm) for transmutation in reactors as fuel and targets. The reprocessed uranium (RU), however, is to be disposed of. This paper presents a limited-scope analysis of possible reuse of RU in CANDU (Canada Deuterium Uranium) Reactors, within the context of the US GNEP program. Other papers on this topic submitted to this conference discuss the possibility of RU reuse in light-water reactors (LWRs) (with enrichment) and offer an independent economic analysis of RU reuse. A representative RU uranium 'vector', from reprocessed spent LWR fuel, comprises 98.538 wt% 238U, 0.46 wt% 236 U, 0.986 wt% 235 U, and 0.006 wt% 234 U. After multiple recyclings, the concentration of 234 U can approach 0.02 wt%. The presence of 234 U and 236 U in RU reduces the reactivity and fuel lifetime (exit burnup), which is particularly an issue in LWRs. While in PWR analyses, the burnup penalty caused by the concentration of 236 U in RU needs to be offset by additional 235 U enrichment in the amount of ∼25% to 30% of the weight percentage of the 236 U; however, the effect in CANDU is much smaller. Furthermore, since the 235 U content in RU exceeds that of natural uranium, CANDU offers the advantageous option of uranium recycling without reenrichment. The exit burnup of CANDU RU-derived fuel is considerably larger than that for natural uranium-fueled scenario, despite the presence of 234 U and 236 U.

  5. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-05-01

    Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) a thorough understanding of the fundamental behaviour of CANDU fuel; (b) data showing that the predicted high utilization of uranium has been achieved; (c) criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to 'CANLUB' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases; (d) proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). As of mid-1976 over 3 x 10 6 individual elements have been built and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification

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

    International Nuclear Information System (INIS)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan

    2006-01-01

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

  7. 3D simulation of CANDU reactor regulating system

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.

    2013-01-01

    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)

  8. Use of enriched uranium as a fuel in CANDU reactors

    International Nuclear Information System (INIS)

    Zech, H.J.

    1976-08-01

    The use of slightly enriched uranium as a fuel in CANDU-reactors is studied in a simple parametric way. The results show the possibility of 1) about 30% savings in natural uranium consumption 2) about 35% increase in the utilization of the natural uranium 3) a decrease in fuelling costs to about 70 - 80% of the normal case of natural uranium fuelling. (orig.) [de

  9. Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

    International Nuclear Information System (INIS)

    Kim, Tae H.; Yang, Chae Y.

    2013-01-01

    The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

  10. Environmental effects on the response of self-powered flux detectors in CANDU reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.; Shields, R.B.; Joslin, C.W.

    1976-01-01

    Self-powered flux detectors are playing an increasingly important role in the control and safety systems of CANDU-type reactors. In this paper we report on recent experiments to determine how local reactor conditions affect the output signals from self-powered detectors with vanadium, platinum and cobalt emitters. The results are interpreted in terms of variations in the local neutron, γ-ray and electron fluxes. (author)

  11. 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

  12. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

    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

  13. An overview of the potential of the CANDU reactor as a thermal breeder

    International Nuclear Information System (INIS)

    Slater, J.B.

    1977-02-01

    This paper is concerned with the use of thorium as a fuel in the existing CANDU concept. The neutron balance of the reactor core is analyzed and an assessment is made of the potential for development of a thermal 'breeder' reactor system. It is concluded that while the SSET cycle (i.e. self-sufficient equilibrium thorium cycle) appears feasible, there is little potential for developing a significant 'breeding' fuel cycle if current reactor operating capability and capital costs are to be maintained. (author)

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

    International Nuclear Information System (INIS)

    Tingle, C.P.; Bonin, H.W.

    1999-01-01

    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 UO 2 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 UO 2 . The model was initially tested and the average discharge burnup for natural UO 2 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

  15. The Thermal-hydraulic Analysis for the Aging Effect of the Component in CANDU-6 Reactor

    International Nuclear Information System (INIS)

    Bae, Jun Ho; Jung, Jong Yeob

    2014-01-01

    CANDU reactor consists of a lot of components, including pressure tube, reactor pump, steam generator, feeder pipe, and so on. These components become to have the aging characteristics as the reactor operates for a long time. The aging phenomena of these components lead to the change of operating parameters, and it finally results to the decrease of the operating safety margin. Actually, due to the aging characteristics of components, CANDU reactor power plant has the operating license for the duration of 30 years and the plant regularly check the plant operating state in the overhaul period. As the reactor experiences the aging, the reactor operators should reduce the reactor power level in order to keep the minimum safety margin, and it results to the deficit of economical profit. Therefore, in order to establish the safety margin for the aged reactor, the aging characteristics for components should be analyzed and the effect of aging of components on the operating parameter should be studied. In this study, the aging characteristics of components are analyzed and revealed how the aging of components affects to the operating parameter by using NUCIRC code. Finally, by scrutinizing the effect of operating parameter on the operating safety margin, the effect of aging of components on the safety margin has been revealed

  16. Fuel cycles - a key to future CANDU success

    International Nuclear Information System (INIS)

    Kuran, S.; Hopwood, J.; Hastings, I.J.

    2011-01-01

    Globally, fuel cycles are being evaluated as ways of extending nuclear fuel resources, addressing security of supply and reducing back-end spent-fuel management. Current-technology thermal reactors and future fast reactors are the preferred platform for such fuel cycle applications and as an established thermal reactor with unique fuel-cycle capability, CANDU will play a key role in fulfilling such a vision. The next step in the evolution of CANDU fuel cycles will be the introduction of Recovered Uranium (RU), derived from conventional reprocessing. A low-risk RU option applicable in the short term comprises a combination of RU and Depleted Uranium (DU), both former waste streams, giving a Natural Uranium Equivalent (NUE) fuel. This option has been demonstrated in China, and all test bundles have been removed from the Qinshan 1 reactor. Additionally, work is being done on an NUE full core, a Thorium demonstration irradiation and an Advanced Fuel CANDU Reactor(AFCR). AECL is developing other fuel options for CANDU, including actinide waste burning. AECL has developed the Enhanced CANDU 6 (EC6) reactor, upgraded from its best-performing CANDU 6 design. High neutron economy, on-power refueling and a simple fuel bundle provide the EC6 with the flexibility to accommodate a range of advanced fuels, in addition to its standard natural uranium. (author)

  17. CANDU 3 - Modularization

    International Nuclear Information System (INIS)

    McAskie, M.J.

    1991-01-01

    The CANDU 3 Heavy Water Reactor is the newest design developed by AECL CANDU. It has set as a major objective, the achievement of significant reductions in both cost and schedule over previous designs. The basic construction strategy is to incorporate extensive modularization of the plant in order to parallel the civil and mechanical installation works. This results in a target 38 month construction schedule from first concrete to in-service compared to 68 months for the Wolsong-1 CANDU 6 actually achieved and the 54 months envisaged for an improved CANDU 6. This paper describes the module concepts that have been developed and explains how they contribute to the overall construction program and achieve the desired cost and schedule targets set for the CANDU 3. (author). 7 figs, 2 tabs

  18. CANDU 3000

    International Nuclear Information System (INIS)

    Keillor, Mac

    1987-01-01

    In this article, the CANDU 300 design, and the team that designed it, are featured. The CANDU 300 will operate at an energy cost similar to that of the larger CANDU units, but is sized for emerging markets. Ease of construction is an important feature: for example, full 360-degree access is available to each of the five buildings during construction; and the whole plant consists of about 90 modules, which can be built in separate locations, and hoisted into place

  19. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

    1989-01-01

    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)

  20. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  1. Temporary solutions for a conservative estimation of void reactivity insertion in CANDU reactor

    International Nuclear Information System (INIS)

    Dumitrache, I.

    1997-01-01

    One of the most difficult task of the CANDU Reactor Physics Analysis is related to the correct treatment of the deviations from the reference coolant properties. The most significant problem is the reactivity inserted by a given coolant density variation. From the practical Nuclear Safety Analysis point of view, the solution must be not only conservative, but also adaptable to the current chain of codes utilized for accident simulation. The first set of experimental data was obtained by AECL many years ago. The fuel was fresh, clean and cold. Some of the currently used computer codes offer accurate predictions of the measured void reactivities. Unfortunately, the existing experimental data do not cover and are not significant for the burned CANDU fuel. A specific benchmark problem was suggested by the Institute for Nuclear Research (ICN) Pitesti. The problem was analysed and slightly modified during an IAEA Vienna RCM (Research Coordinating Meeting), Buenos Aires, 1990. Afterwards, the problem was independently solved in several countries, interested by the CANDU reactor. The results were presented and analysed at the Bombay RCM, 1992. It was clear that the interval defined by the code predictions is much too broad. New experimental data are necessary. They must cover the fuel isotopic composition specific for the burned CANDU fuel. The work is in progress at the Chalk River Laboratory. Temporary solutions have been analysed at the ICN Pitesti. The first aim was to identify the reactivity numerical values that are conservative, but not too inaccurate. The WIMS code predictions have been compared against other estimations, including the Monte-Carlo based ones. The second aim was to force the currently used code, PPV, to offer cell cross sections that are correct from the Reactor Physics point of view, and compatible with the imposed reactivity. Physical and mathematical procedures were proposed and evaluated. An additional solution was also taken into account: to

  2. Methodologies for optimizing ROP detector layout for CANDU (registered) reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kastanya, Doddy, E-mail: kastanyd@aecl.c [Reactor Core Physics Branch, Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, ON, L5K 1B2 (Canada); Caxaj, Victor [Reactor Core Physics Branch, Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, ON, L5K 1B2 (Canada)

    2011-01-15

    The regional overpower protection (ROP) systems protect CANDU (registered) reactors against overpower in the fuel that would reduce the safety margin-to-dryout. Both a localized power peaking within the core (for example, as a result of certain reactivity device configuration) or a general increase in the core power level during a slow-loss-of-regulation (SLOR) event could cause overpower in the fuel. This overpower could lead to fuel sheath dryout. In the CANDU (registered) 600 MW (CANDU 6) design, there are two ROP systems in the core, one for each fast-acting shutdown systems. Each ROP system includes a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal assemblies. Traditionally, the placement of these detectors was done using a method called the detector layout optimization (DLO). A new methodology for designing the detector layout for the ROP system has been developed recently. The new method, called the DETPLASA algorithm, utilizes the simulated annealing (SA) technique to optimize the placement of the detectors in the core. Both methodologies will be discussed in detail in this paper. Numerical examples are employed to better illustrate how each method works. Results from some sensitivity studies on three SA parameters are also presented.

  3. Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

    The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

  4. Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

    1978-12-01

    This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

  5. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

    Hopwood, J. M.; Saroudis, J.

    2013-01-01

    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, D 2 O, 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

  6. Privatize Candu (question mark)

    International Nuclear Information System (INIS)

    Kelly, Thomas.

    1981-01-01

    A report sponsored by a group of nuclear suppliers and the Royal Bank suggested that the Candu reactor system would sell better if it were owned by a private company. Licensing of a Candu reactor in the U.S.A. was also suggested. The author of this article agrees with these points, but disagrees with the suggestion that safeguards should be relaxed. He suggests that contracts should stipulate that instrumentation should be supplied as much as possible from Canadian sources

  7. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  8. Plutonium dispositioning in CANDU

    International Nuclear Information System (INIS)

    Boczar, P.G.; Feinroth, H.; Luxat, J.C.

    1995-07-01

    Recently, the U.S. Department of Energy (DOE) sponsored Atomic Energy of Canada Limited (AECL) to evaluate salient technical, strategic, schedule, and cost-related parameters of using CANDU reactors for dispositioning of weapons-grade plutonium in the form of Mixed OXide (MOX) fuel. A study team, consisting of key staff from the CANDU reactor designers and researchers (AECL), operators (Ontario Hydro) and fuel suppliers, analyzed all significant factors involved in such application, with the objective of identifying an arrangement that would permit the burning of MOX in CANDU at the earliest date. One of Ontario Hydro's multi-unit stations, Bruce A nuclear generating station (4x769 MW(e)), was chosen as the reference for the study. The assessment showed that no significant modifications of reactor or process systems are necessary to operate with a full MOX core. Plant modifications would be limited to fuel handling and modifications necessary to accommodate enhanced security and safeguards requirements. No safety limitations were identified

  9. New build CANDU in Canada - development and application of information management systems for latest construction technology

    International Nuclear Information System (INIS)

    Asakura, S.; Akane, N.; Byrne, J.; Canas, B.; Kereliuk, S.; Akahori, S.

    2007-01-01

    In the wake of recent events representing growing nuclear energy interest in Canada, AECL and Hitachi are sharing each other's expertise as each company has its own unique and unrivalled capability in managing new build construction projects. This paper addresses some of the key strategies used to execute successfully a New CANDU build project, by focusing on recent developments and implementations in the construction and project management fields and by highlighting the use of cutting edge information technology. These strategies are designed to achieve and maximize their benefit to the New Build CANDU project team, to AECL/Hitachi as well as customers, partners and suppliers. (author)

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

    International Nuclear Information System (INIS)

    Park, S. Y.; Song, Y. M.

    2015-01-01

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

  11. 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.

  12. 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)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Cho, Y. J.; Kim, M. W.

    2008-01-01

    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

  13. Use of dwell time concept in fission product inventory assessment for CANDU reactors

    International Nuclear Information System (INIS)

    Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

    2003-01-01

    A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

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

    International Nuclear Information System (INIS)

    Suripto, A.; Badruzzaman, M.; Latief, A.

    1997-01-01

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

  15. Next generation CANDU plants

    International Nuclear Information System (INIS)

    Hedges, K.R.; Yu, S.K.W.

    1998-01-01

    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

  16. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  17. Operating performance and reliability of CANDU PHWR fuel channels in Canada

    International Nuclear Information System (INIS)

    Strachan, B.; Brown, D.R.

    1983-03-01

    CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

  18. Design of a Multi-Spectrum CANDU-based Reactor, MSCR, with 37-element fuel bundles using SERPENT code

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.; Chan, P.

    2015-01-01

    The burning of highly-enriched uranium and plutonium from dismantled nuclear warhead material in the new design nuclear power plants represents an important step towards nonproliferation. The blending of these highly enriched uranium and plutonium with with uranium dioxide from the spent fuel of CANDU reactors, or mixing it with depleted uranium would need a very long time to dispose of this material. Consequently, considering that more efficient transmutation of actinides occurs in fast neutron reactors, a novel Multi-Spectrum CANDU Reactor, has been designed on the basis of the CANDU6 reactor with two concentric regions. The simulations of the MSCR were carried out using the SERPENT code. The inner or fast neutron spectrum core is fuelled by different levels of enriched uranium oxides. The helium is used as a coolant in the fast neutron core. The outer or the thermal neutron spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. Both cores use 37- element fuel bundles. The size of the two cores and the percentage level of enrichment of the fresh fuel in the fast core were optimized according to the criticality safety of the whole reactor. The excess reactivity, the regeneration factor, radial and axial flux shapes of the MSCR reactor were calculated at different of the concentration of fissile isotope 235 U of uranium fuel at the fast neutron spectrum core. The effect of variation of the concentration of the fissile isotope on the fluxes in both cores at each energy bin has been studied. (author)

  19. Design of a Multi-Spectrum CANDU-based Reactor, MSCR, with 37-element fuel bundles using SERPENT code

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.; Chan, P., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca, E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada)

    2015-07-01

    The burning of highly-enriched uranium and plutonium from dismantled nuclear warhead material in the new design nuclear power plants represents an important step towards nonproliferation. The blending of these highly enriched uranium and plutonium with with uranium dioxide from the spent fuel of CANDU reactors, or mixing it with depleted uranium would need a very long time to dispose of this material. Consequently, considering that more efficient transmutation of actinides occurs in fast neutron reactors, a novel Multi-Spectrum CANDU Reactor, has been designed on the basis of the CANDU6 reactor with two concentric regions. The simulations of the MSCR were carried out using the SERPENT code. The inner or fast neutron spectrum core is fuelled by different levels of enriched uranium oxides. The helium is used as a coolant in the fast neutron core. The outer or the thermal neutron spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. Both cores use 37- element fuel bundles. The size of the two cores and the percentage level of enrichment of the fresh fuel in the fast core were optimized according to the criticality safety of the whole reactor. The excess reactivity, the regeneration factor, radial and axial flux shapes of the MSCR reactor were calculated at different of the concentration of fissile isotope {sup 235}U of uranium fuel at the fast neutron spectrum core. The effect of variation of the concentration of the fissile isotope on the fluxes in both cores at each energy bin has been studied. (author)

  20. Fuelling study of CANDU reactors using neutron absorber poisoned fuel

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.J.; Chan, P.K.; Bonin, H.W., E-mail: s25815@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2014-07-01

    A comparative fuelling study is conducted to determine the potential gain in operating margin for CANDU reactors incurred by implementing a change to the design of the conventional 37-element natural uranium (NU) fuel. The change involves insertion of minute quantities of neutron absorbers, Gd{sub 2}O{sub 3} and Eu{sub 2}O{sub 3}, into the fuel pellets. The Reactor Fuelling Simulation Program (RFSP) is used to conduct core-following simulations, for the regular 37-element NU fuel, which is to be used as control for comparison. Preliminary results are presented for fuelling with the regular 37-element NU fuel, which indicate constraints on fuelling that may be relaxed with addition of neutron absorbers. (author)

  1. Analysis of log rate noise in Ontario's CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hinds, H.W. [Dynamic Simulation and Analysis Corp., Deep River, Ontario (Canada); Banica, C.; Arguner, D. [Ontario Power Generation, Ajax, Ontario (Canada); Scharfenberg, R. [Bruce Power, Tiverton, Ontario (Canada)

    2007-07-01

    In the fall of 2003, the operators noticed that in the recently-refurbished Bruce A Shutdown System no. 1 (SDS1) the noise level in Log Rate signals were much larger than before. At the request of the Canadian Nuclear Safety Commission (CNSC), all Canadian CANDU reactors took action to characterize their Log Rate noise. Staff of the Inspection and Maintenance Services division of Ontario Power Generation (OPG) has collected high-speed high-accuracy noise data from nearly all 16 Ontario reactors, either as part of routine measurements before planned outages or as a dedicated noise recording. This paper gives the results of examining a suitable subset of this data, with respect to the characteristics and possible causes of Log Rate noise. The reactor and instrumentation design is different at each station: the locations of the moderator injection nozzles, the location of the ion chambers for each system, and the design of the Log Rate amplifiers. It was found that the Log noise (source of Log Rate noise) was much larger for those ion chambers in the path of the moderator injection nozzles, compared to those which were not in the path. This 'extra' Log noise would then be either attenuated or amplified depending on the transfer function (time constants) of the Log Rate amplifier. It was also observed that most of the Log and Log Rate noise is independent of any other signal measured. Although all CANDU reactors in Ontario have Log and Log Rate noise, the Bruce A SDS1 system has the largest amount of Log Rate noise, because (a) its SDS1 (and RRS) ion chambers are at the top of the reactor in the path of the moderator injection nozzles, and (b) its SDS1 Log Rate amplifiers have the smallest time constants. (author)

  2. CANDU development

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1981-06-01

    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 32 0 C. 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

  3. CANDU passive shutdown systems

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

    1996-12-01

    CANDU incorporates two diverse, passive shutdown systems, 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 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 description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

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

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    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

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

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-01

    The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2 UO 2 and ThO 2 UO 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

  6. 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.

  7. 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)

    1998-12-31

    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)

  8. 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)

  9. The CANDU 9 distributed control system design process

    International Nuclear Information System (INIS)

    Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

    1997-01-01

    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)

  10. Impact of aging and material structure on CANDU plant performance

    International Nuclear Information System (INIS)

    Nadeau, E.; Ballyk, J.; Ghalavand, N.

    2011-01-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

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

    International Nuclear Information System (INIS)

    Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Park, Kwang Seok; Kim, Bong Ki; Lee, Yeong Uk; Jeong, Chang Joon; Oh, Deok Joo; Lee, Ui Joo; Park, Joo Hwan; Lee, Sang Yong; Jeong, Beop Dong; Choi, Han Rim; Lee, Yeong Jin; Choi, Cheol Jin; Choi, Jong Ho; Lee, Kwang Won; Cho, Cheon Hyi; On, Myeong Ryong; Kim, Taek Mo; Lim, Hong Sik; Lee, Kang Moon; Lee, Nam Ho; Lee, Kyu Hyeong

    1994-07-01

    It has been projected that a total of 5 pressurized heavy water reactors (PHWR) including Wolsong 1 under operation and Wolsong 2, 3 and 4 under construction will be operated by 2006, and so about 500 ton of natural uranium will be consumed every year and a lot of spent fuels will be generated. Therefore, the ultimate goal of this R and D project is to develop the CANDU advanced fuel having the following capabilities compared with existing standard fuel: (1) To reduce linear heat generation rating by more than 15% (i.e., less than 50 kW/m), (2) To extend fuel burnup by more than 3 times (i.e., higher than 21,000 MWD/MTU), and (3) To increase critical channel power by more than 5%. In accordance, the followings are performed in this fiscal year: (1) Undertake CANFLEX-NU design and thermalmechanical performance analysis, and prepare design documents, (2) Establish reactor physics analysis code system, and investigate the compativility of the CANFLEX-NU fuel with the standard 37-element fuel in the CANDU-6 reactor. (3) Establish safety analysis methodology with the assumption of the CANFLEX-NU loaded CANDU-6 reactor, and perform the preliminary thermalhydraulic and fuel behavior for the selected DBA accidents, (4) Investigate reactor physics analysis code system as pre-study for CANFLEX-SEU loaded reactors

  12. Jet flow analysis of liquid poison injection in a CANDU reactor using source term

    International Nuclear Information System (INIS)

    Chae, Kyung Myung; Choi, Hang Bok; Rhee, Bo Wook

    2001-01-01

    For the performance analysis of Canadian deuterium uranium (CANDU) reactor shutdown system number 2 (SDS2), a computational fluid dynamics model of poison jet flow has been developed to estimate the flow field and poison concentration formed inside the CANDU reactor calandria. As the ratio of calandria shell radius over injection nozzle hole diameter is so large (1055), it is impractical to develop a full-size model encompassing the whole calandria shell. In order to reduce the model to a manageable size, a quarter of one-pitch length segment of the shell was modeled using symmetric nature of the jet; and the injected jet was treated as a source term to avoid the modeling difficulty caused by the big difference of the hole sizes. For the analysis of an actual CANDU-6 SDS2 poison injection, the grid structure was determined based on the results of two-dimensional real- and source-jet simulations. The maximum injection velocity of the liquid poison is 27.8 m/s and the mass fraction of the poison is 8000 ppm (mg/kg). The simulation results have shown well-established jet flow field. In general, the jet develops narrowly at first but stretches rapidly. Then, the flow recirculates a little in r-x plane, while it recirculates largely in r-θ plane. As the time goes on, the adjacent jets contact each other and form a wavy front such that the whole jet develops in a plate form. his study has shown that the source term model can be effectively used for the analysis of the poison injection and the simulation result of the CANDU reactor is consistent with the model currently being used for the safety analysis. In the future, it is strongly recommended to analyze the transient (from helium tank to injection nozzle hole) of the poison injection by applying Bernoulli equation with real boundary conditions

  13. Managing the water chemistry of a CANDU reactor with an expert system

    International Nuclear Information System (INIS)

    Lamirande, S.; Roberge, P.R.

    1990-01-01

    The aim of this project was to capture the expertise of Ontario Hydro in the water chemistry of the heat transport system (HTS) of the CANDU nuclear reactor and transform it into an Expert System prototype. The end product is an Expert System which can realistically diagnose situations and recommend proper courses of action based on the user's water chemistry analysis

  14. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G.

    1999-01-01

    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

  15. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G

    1998-05-01

    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

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

    International Nuclear Information System (INIS)

    Park, Tae Keun; Yeo, Ji Won

    1996-03-01

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

  17. Incorporating single detector failure into the ROP detector layout optimization for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kastanya, Doddy, E-mail: Doddy.Kastanya@snclavalin.com

    2015-12-15

    Highlights: • ROP TSP value needs to be adjusted when any detector in the system fails. • Single detector failure criterion has been incorporated into the detector layout optimization as a constraint. • Results show that the optimized detector layout is more robust with respect to its vulnerability to a single detector failure. • An early rejection scheme has been introduced to speed-up the optimization process. - Abstract: In CANDU{sup ®} reactors, the regional overpower protection (ROP) systems are designed to protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. In the CANDU{sup ®} 600 MW (CANDU 6) design, there are two ROP systems in the core, each of which is connected to a fast-acting shutdown system. Each ROP system consists of a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal flux detector assemblies. The placement of these ROP detectors is a challenging discrete optimization problem. In the past few years, two algorithms, DETPLASA and ADORE, have been developed to optimize the detector layout for the ROP systems in CANDU reactors. These algorithms utilize the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The objective of the optimization process is typically either to maximize the TSP value for a given number of detectors in the system or to minimize the number of detectors in the system to obtain a target TSP value. One measure to determine the robustness of the optimized detector layout is to evaluate the maximum decrease (penalty) in TSP value when any single detector in the system fails. The smaller the penalty, the more robust the design is. Therefore, in order to ensure that the optimized detector layout is robust, the single detector failure (SDF) criterion has been incorporated as an additional constraint into the ADORE algorithm. Results from this study indicate that there

  18. The improved quasi-static method vs the direct method: a case study for CANDU reactor transients

    International Nuclear Information System (INIS)

    Kaveh, S.; Koclas, J.; Roy, R.

    1999-01-01

    Among the large number of methods for the transient analysis of nuclear reactors, the improved quasi-static procedure is one of the most widely used. In recent years, substantial increase in both computer speed and memory has motivated a rethinking of the limitations of this method. The overall goal of the present work is a systematic comparison between the improved quasi-static and the direct method (mesh-centered finite difference) for realistic CANDU transient simulations. The emphasis is on the accuracy of the solutions as opposed to the computational speed. Using the computer code NDF, a typical realistic transient of CANDU reactor has been analyzed. In this transient the response of the reactor regulating system to a substantial local perturbation (sudden extraction of the five adjuster rods) has been simulated. It is shown that when updating the detector responses is of major importance, it is better to use a well-optimized direct method rather than the improved quasi-static method. (author)

  19. Fuel cycle model and the cost of a recycling thorium in the CANDU reactor

    International Nuclear Information System (INIS)

    Choi, Hangbok; Park, Chang Je

    2005-01-01

    The dry process fuel technology has a high proliferation-resistance, which allows applications not only to the existing but also to the future nuclear fuel cycle systems. In this study, the homogeneous ThO 2 -UO 2 recycling fuel cycle in a Canada deuterium uranium (CANDU) reactor was assessed for a fuel cycle cost evaluation. A series of parametric calculations were performed for the uranium fraction, enrichment of the initial uranium fuel, and the fission product removal rated of the recycled fuel. The fuel cycle cost was estimated by the levelized lifetime cost model provided by the Organization for Economic Cooperation and Development/Nuclear Energy Agency. Though it is feasible to recycle the homogeneous ThO 2 -UO 2 fuel in the CANDU reactor from the viewpoint of a mass balance, the recycling fuel cycle cost is much higher than the conventional natural uranium fuel cycle cost for most cases due to the high fuel fabrication cost. (author)

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

    International Nuclear Information System (INIS)

    Park, T-K.; Jung, S-H.

    1996-01-01

    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 D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

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

    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.

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

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.

    2009-01-01

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

  3. Tritium inventory prediction in a CANDU plant

    International Nuclear Information System (INIS)

    Song, M.J.; Son, S.H.; Jang, C.H.

    1995-01-01

    The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983--1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors

  4. Effect of scaling on the thermal hydraulics of the moderator of a CANDU reactor

    International Nuclear Information System (INIS)

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2011-01-01

    Three dimensional numerical simulations are conducted on the CANDU Moderator Test Facility (MTF) and the actual size CANDU reactor. Moderator test facility is ¼ scale of the actual reactor. The heat input and other operating conditions are scaled down from the real reactor to the MTF using constant Archimedes number (as considered in MTF experiments performed by Atomic Energy of Canada Ltd.). The heat generations inside both tanks are applied through volumetric heating. In this method, heat is added to the fluid throughout the volume as it occurs in real reactor through fission heat generation and gamma rays from radioactive materials. The temperatures in actual reactor simulation are about 10 deg C greater than in MTF simulations. The separation between high and low temperature zones are more visible in real reactor simulation comparing to MTF simulation. The result indicates that the MTF has better mixing and weaker buoyancy forces comparing to real reactor. The velocity distribution in both cases seems similar with impingement point for inlet jets in both cases at the right hand side of the tank. Although the velocities are considerably higher (about 40%) in the case of real reactor, but as we go toward inner core of the tanks, the velocities are similar and very low. Several points inside the tank are monitored for their temperature and velocity with time. The results for these points show fluctuations in both temperature and velocity inside the tank. The fluctuations frequency seems higher in the case of real reactor while the amplitude of fluctuations is smaller in real reactor in most of the points. Here, in this research we have shown that Archimedes number alone cannot be a good scaling parameter (as used in MTF experiments) and it should be used along with Rayleigh number for scaling purposes. (author)

  5. Numerical simulator of the CANDU fueling machine driving desk

    International Nuclear Information System (INIS)

    Doca, Cezar

    2008-01-01

    As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

  6. Fuel management in CANDU reactors: Daniel Rozon's contribution

    International Nuclear Information System (INIS)

    Rozon, D.; Varin, E.; Chambon, R.

    2010-01-01

    The CANDU fuel management optimization problem is in many ways different from LWRs fuel management, because of the on-line refueling and the complete 3-D geometry problem. Daniel Rozon was an outstanding leader in the understanding and resolution of this optimization problem and remained during his entire career. Daniel Rozon and his students have used the generalized adjoint formalism implemented in standard mathematical programming methods to solve the optimization of the exit burnup in the reactor as well as the optimization of control rod worth or fuel enrichment. We have summarized here the theoretical basis of fuel management and resolution methods, the latest approaches of optimization and results as obtained using the OPTEX code. (author)

  7. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1988-01-01

    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

  8. Construction of dynamic model of CANDU-SCWR using moving boundary method

    International Nuclear Information System (INIS)

    Sun Peiwei; Jiang Jin; Shan Jianqiang

    2011-01-01

    Highlights: → A dynamic model of a CANDU-SCWR is developed. → The advantages of the moving boundary method are demonstrated. → The dynamic behaviours of the CANDU-SCWR are obtained by simulation. → The model can predict the dynamic behaviours of the CANDU-SCWR. → Linear dynamic models for CANDU-SCWR are derived by system identification techniques. - Abstract: CANDU-SCWR (Supercritical Water-Cooled Reactor) is one type of Generation IV reactors being developed in Canada. Its dynamic characteristics are different from existing CANDU reactors due to the supercritical conditions of the coolant. To study the behaviours of such reactors under disturbances and to design adequate control systems, it is essential to have an accurate dynamic model to describe such a reactor. One dynamic model is developed for CANDU-SCWR in this paper. In the model construction process, three regions have been considered: Liquid Region I, Liquid Region II and Vapour Region, depending on bulk and wall temperatures being higher or lower the pseudo-critical temperature. A moving boundary method is used to describe the movement of boundaries across these regions. Some benefits of adopting moving boundary method are illustrated by comparing with the fixed boundary method. The results of the steady-state simulation based on the developed model agree well with the design parameters. The transient simulations demonstrate that the model can predict the dynamic behaviours of CANDU-SCWR. Furthermore, to investigate the responses of the reactor to small amplitude perturbations and to facilitate control system designs, a least-square based system identification technique is used to obtain a set of linear dynamic models around the design point. The responses based on the linear dynamic models are validated with simulation results from nonlinear CANDU-SCWR dynamic model.

  9. The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Zhang, J.; Sills, H.E.; Flatt, L.; Jenkins, D.; Wallace, D.J.; Popov, N.

    2002-01-01

    The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its proto-typing application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX R fuel. The methodology is consistent with and builds on world practice. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions. (authors)

  10. Economic case for CANDU life extension projects

    International Nuclear Information System (INIS)

    Qureshi, S.; Tenev, T.; Lewi, M.

    2014-01-01

    As CANDU reactors approach their original end of design life utilities are faced with two options: to extend the operating life of the reactor by undergoing a life extension project (LEP), or to commence decommissioning activities. Recent project experience has shown that there is economic merit in extending the life of the operating reactor. There are many benefits to such a decision, the most obvious being the revenue that will be generated from the additional years of electricity production by the utility. Delays in decommissioning are also advantageous since the large costs associated with such a long-term activity are pushed into the future, thereby decreasing the net present value (NPV) of the investment. In addition, relatively few power reactors have been fully decommissioned to date and deferring this activity transfers the associated risks to others that are currently obligated to undertake decommissioning activities sooner. Candu Energy has been involved with the life extension projects of the following CANDU reactors: Point Lepreau (New Brunswick, Canada), Bruce Unit 1 and Unit 2 (Ontario, Canada), and Wolsong Unit 1 (South Korea). These reactors underwent fuel channel replacement programs in addition to replacement of major reactor components. Most recently, both Ontario Power Generation (OPG) and Nucleoelectrica Argentina Sociedad Anonima (NA-SA) have commenced work on life extension projects at the Darlington (Canada) and Embalse (Argentina) sites respectively. The experience gained from previous LEP projects allows Candu Energy to deliver future projects in a timely, efficient, and cost effective manner. (author)

  11. Development and applications of reactor noise analysis at Ontario Hydro's CANDU reactors

    International Nuclear Information System (INIS)

    Gloeckler, O.; Tulett, M.V.

    1995-01-01

    In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro's CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

  12. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

    1999-01-01

    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)

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

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1998-01-01

    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

  14. Valve maintainability in CANDU-PHW nuclear generating stations

    International Nuclear Information System (INIS)

    Pothier, N.E.; Crago, W.A.

    1977-09-01

    Design, application, layout and administrative factors which affect valve maintainability in CANDU-PHW power reactors are identified and discussed. Some of these are illustrated by examples based on prototype reactor operation experience. Valve maintainability improvements resulting from laboratory development and maintainability analysis, have been incorporated in commercial CANDU-PHW nuclear generating stations. These, also, are discussed and illustrated. (author)

  15. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

    2013-01-01

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

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

    International Nuclear Information System (INIS)

    Carlucci, L.N.

    1982-10-01

    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

  17. CANDU reg-sign -- A Canadian energy system

    International Nuclear Information System (INIS)

    Sejnoha, R.

    1994-01-01

    Uranium is one of Canada's important natural resources. It is perhaps not surprising that a country with such an abundance of uranium developed its own inexpensive, safe, and environmentally friendly system of energy generation, based on uranium: CANDU reg-sign (CANada-Deuterium-Uranium, a registered trademark of AECL). The objective of this paper is to describe briefly the main features of the CANDU system and explain methods used to assure the compatibility with the requirements for a clean and safe environment. The paper describes the CANDU reactor, and discusses storage of spent fuel, reactor performance, normal operating conditions, safety under accident conditions, and quality assurance in design, manufacturing, and operation

  18. Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.

    1998-01-01

    Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

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

    International Nuclear Information System (INIS)

    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

  20. Cross section homogenization analysis for a simplified Candu reactor

    International Nuclear Information System (INIS)

    Pounders, Justin; Rahnema, Farzad; Mosher, Scott; Serghiuta, Dumitru; Turinsky, Paul; Sarsour, Hisham

    2008-01-01

    The effect of using zero current (infinite medium) boundary conditions to generate bundle homogenized cross sections for a stylized half-core Candu reactor problem is examined. Homogenized cross section from infinite medium lattice calculations are compared with cross sections homogenized using the exact flux from the reference core environment. The impact of these cross section differences is quantified by generating nodal diffusion theory solutions with both sets of cross sections. It is shown that the infinite medium spatial approximation is not negligible, and that ignoring the impact of the heterogeneous core environment on cross section homogenization leads to increased errors, particularly near control elements and the core periphery. (authors)

  1. Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

    OpenAIRE

    Shenton B.; Philipose K.

    2011-01-01

    The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in th...

  2. Impact of aging and material structure on CANDU plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Nadeau, E.; Ballyk, J.; Ghalavand, N. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  3. CANDU, building the future

    International Nuclear Information System (INIS)

    Stern, F.

    1997-01-01

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

  4. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

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

  5. Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

    International Nuclear Information System (INIS)

    Hazen Hezhi Fan; Zoran Bilanovic

    2005-01-01

    The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

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

    International Nuclear Information System (INIS)

    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

    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

  7. Applicability of ANSYS ELBOW290 element for flexibility calculation of tight radius bends on feeder pipes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, X., E-mail: Xuan.Zhang@candu.com [Candu Energy Inc, Mississauga, ON (Canada)

    2015-07-01

    A curved pipe element, ELBOW290, became available in ANSYS 12. This element was developed based on a simplified shell theory, and maintains the ability to capture cross-sectional deformations of elbows. Numerical testing on the applicability of this element for the flexibility calculation of the tight radius bends in CANDU reactors is carried out to determine the usability of this element in completing stress analyses for feeder pipes. Comparisons are made between the ELBOW290 and the shell element for various feeder bend types found in domestic and overseas CANDU reactors. The comparisons show that the ELBOW290 element is suitable for calculating the flexibility of the tight radius bends. (author)

  8. Applications of perturbation theory to the study of CANDU reactors

    International Nuclear Information System (INIS)

    Rozon, D.; Beaudet, M.

    1990-01-01

    The use of Generalized Perturbation Theory (GPT) in the computer code OPTEX-4 is described. This code can be used to simultaneously optimize the fuel management and the control absorber distribution in a CANDU reactor at equilibrium refueling. The gradient of the characteristic functionals are obtained using two independent approaches, requiring the solution of a fixed source eigenvalue problem (direct for the explicit approach. adjoint for the implicit approach). These solutions, as well as the solution of the diffusion problem is obtained in 3D by calling the diffusion module TRIVAC-2. The equivalence of the two approaches is demonstrated [fr

  9. CANDU reactors and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

    1999-01-01

    This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

  10. CANDU reactors and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

    1998-01-01

    This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

  11. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Hopkins, J.R.

    1998-01-01

    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

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

    International Nuclear Information System (INIS)

    Hedges, K. R.; Hopwood, J. M.

    2001-01-01

    margins and performance. The CANDU 6 design has evolved in both design and project delivery, so that the current CANDU 6 construction project at Qinshan Phase 3, in China, represents the nuclear project state-of-the-art. Experience with current plants, such as Cernavoda, is an important input to the development of improved designs for the future. Building on the strong CANDU knowledge base, AECL is continuing to adapt the CANDU design to develop the Next Generation CANDU medium-sized fuel channel reactor. This paper discusses the next generation of CANDU designs featuring major improvements in economics, inherent safety characteristics and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. This leads to a seamless evolution which ensures that innovation is introduced in a way which minimizes project or operations uncertainties. (authors)

  13. Development and applications of reactor noise analysis at Ontario Hydro`s CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gloeckler, O [Ontario Hydro, Toronto, ON (Canada); Tulett, M V [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

    1996-12-31

    In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro`s CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

  14. Mathematical modeling of CANDU-PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

    2001-07-01

    The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR) system. This study involved mathematical modeling of CANDU PHWR major system components and the developments of software to study the thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique.The integrated CANDU-PHWR model includes the neutronic, reactivity, fuel channel heat transfer, piping and the preheater type U-tube steam generator (PUTSG). The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and reactivity feed back due to the changes in fuel temperature and coolant temperature. The complex operation of the preheater type U-tube steam generator (PUTSG) is represented by a non-linear dynamic model using a state variable, moving boundary and lumped parameter techniques. The secondary side of the PUTSG model has six separate lumps including a preheater region, a lower boiling section, a mixing region, a riser, a chimmeny section, and a down-corner. The tube side of PUTSG has three main thermal zones. The PUTSG model is based on conservation of mass, energy and momentum relation-ships. The CANDU-PHWR integrated model are coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

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

    International Nuclear Information System (INIS)

    Duffey, R. B.

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

    van Erp, J.B.

    1977-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    van Erp, J B

    1977-12-01

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

  18. Thermosyphon Phenomenon as an alternate heat sink of Shutdown Cooling System for the CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [GNEST, Seoul (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    During the outage(overhaul) of the CANDU plant, there is a period when the coolant is partially drained to the reactor header level and the coolant is cooled and depressurized by Shutdown Cooling System(SDCS) other than PHTS pump. In the postulated accident of the loss of SDCS-the PHTS pump failure, the primary coolant system should be cooled by the alternate heat sink using the thermosyphon pheonomenon(TS) through the steam generator(SG) This study was aimed at verification and analyzing the core cooling ability of the TS. And the sensitivity analysis was done for the number of SGs used in the TS. As an analysis tool, RELAP5/CANDU was used.

  19. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    Ingolfsrud, L. John

    1989-01-01

    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

  20. The CANDU 3 containment structure

    International Nuclear Information System (INIS)

    1994-01-01

    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

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

    Directory of Open Access Journals (Sweden)

    W. Shen

    2015-01-01

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

  2. Reactor building pressure proof test (PPT) and leak rate test (LRT) of Qinshan phase III (CANDU) project

    International Nuclear Information System (INIS)

    Gu Jun; Shi Jinqi; Fan Fuping

    2004-12-01

    As the first reactor building (R/B) without stainless steel liner in china, TQNPC studied the containment characteristics, such as strong concrete absorb/release air effect, poor containment penetration. etc. And carefully prepared test scheme and emergency response, creatively introduced the instrument air self-supply system in reactor building, developed the special measurement and analysis system for PPT and LRT, organized work under high-pressure on large-scale in the test. Finally got the containment leak rate result and the test-cost-time value is the best in all same type tests. (authors)

  3. The feasibility study of using deuterated gadolinium nitrate for moderator-poisoned shutdown and excess reactivity control in CANDU reactors

    International Nuclear Information System (INIS)

    Li, J.; Everatt, A.

    2006-01-01

    Gadolinium nitrate is used in CANDU stations as moderator poison for reactor shutdowns and excess reactivity control. The use of the light-water hydrate introduces significant quantities of light water into the moderator system, which must be removed from the moderator by periodically upgrading the moderator (isotopic maintenance). The benefit of using a deuterated gadolinium nitrate would be a higher moderator isotopic and/or a lesser isotopic maintenance requirement. This study evaluated the economics of using deuterated gadolinium nitrate, as opposed to the light-water hydrate, for moderator-poisoned shutdowns and excess reactivity control in CANDU-6 reactors. Normal gadolinium nitrate (i.e., the light-water hydrate) is available from suppliers at ∼125 $/kg. Supplier quotes for deuterated gadolinium nitrate ranged from 1900 to 4000 $/kg. To examine the possibility of producing deuterated gadolinium nitrate in-house at a lower cost than commercially available, a three-stage dissolution/evaporation manufacturing process was conceived and costed. Depending on the assumed demand for the product (i.e., the number of reactors adopting the use of the product) and the capital recovery period, the estimated unit cost for the dissolution/evaporation process ranged from 730 to 2500 $/kg. The determination of economic benefit from using deuterated gadolinium nitrate in existing CANDU stations was based on the cost savings resulting from a higher fuel burn-up (i.e., the higher moderator isotopic would give a higher fuel burn-up). The net benefit of using deuterated gadolinium nitrate for most CANDU stations was determined to be marginal (i.e., <20 k$/a). Only for those CANDU stations where the moderator isotopic was relatively low (e.g., 99.85 wt%) was there a potential significant benefit (20-100 k$/a). However, if the reason for the low moderator isotopic is a relatively high moderator light-water ingress rate from sources other than the use of the light-water hydrate

  4. CANDU-PHW fuel channel replacement experience

    International Nuclear Information System (INIS)

    Dunn, J.T.; Kakaria, B.K.

    1982-09-01

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

  5. Distributed control system for CANDU 9 nuclear power plant

    International Nuclear Information System (INIS)

    Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

    1996-01-01

    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. The CANDU 9 plant monitoring, annunciation, and control functions are implemented in two evolutionary systems; the distributed control system (DCS) and the plant display system (PDS). The CDS implements most of the plant control functions in a single hardware platform. The DCS communicates with the PDS to provide the main operator interface and annunciation capabilities of the previous control computer designs along with human interface enhancements required in a modern control system. (author)

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

    International Nuclear Information System (INIS)

    Lipsett, M. G.; Rody, K.H.

    1995-01-01

    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)

  7. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

    The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

  8. Possibility of plutonium burning out and minor actinides transmutation in CANDU type reactor

    International Nuclear Information System (INIS)

    Gerasimov, A.S.; Kiselev, G.V.; Myrtsymova, L.A.

    2000-01-01

    The possibility of power or weapon-grade plutonium use as nuclear fuel in CANDU type reactor with simultaneous minor actinides burn-out is studied. Total thermal power is 1900 MW. The fuel lifetime makes 0.24 years, neutron flux density 10 14 neutr/cm 2 s. About 40-45 % of plutonium is incinerated during fuel lifetime. If weapon-grade plutonium is used in fuel channels instead of power one, its consumption is 40% lower. (author)

  9. Physics characteristics of CANDU cores with advanced fuel cycles

    International Nuclear Information System (INIS)

    Garvey, P.M.

    1985-01-01

    The current generation of CANDU reactors, of which some 20 GWE are either in operations or under construction worldwide, have been designed specifically for the natural uranium fuel cycle. The CANDU concept, due to its D 2 O coolant and moderator, on-power refuelling and low absorption structural materials, makes the most effective utilization of mined uranium of all currently commercialized reactors. An economic fuel cycle cost is also achieved through the use of natural uranium and a simple fuel bundle design. Total unit energy costs are achieved that allow this reactor concept to effectively compete with other reactor types and other forms of energy production. There are, however, other fuel cycles that could be introduced into this reactor type. These include the slightly enriched uranium fuel cycle, fuel cycles in which plutonium is recycled with uranium, and the thorium cycle in which U-233 is recycled. There is also a special range of fuel cycles that could utilize the spent fuel from LWR's. Two specific variants are a fuel cycle that only utilizes the spent uranium, and a fuel cycle in which both the uranium and plutonium are recycled into a CANDU. For the main part these fuel cycles are characterized by a higher initial enrichment, and hence discharge burnup, than the natural uranium cycle. For these fuel cycles the main design features of both the reactor and fuel bundle would be retained. Recently a detailed study of the use in a CANDU of mixed plutonium and uranium oxide fuel from an LWR has been undertaken by AECL. This study illustrates many of the generic technical issues associated with the use of Advanced Fuel Cycles. This paper will report the main findings of this evaluation, including the power distribution in the reactor and fuel bundle, the choice of fuel management scheme, and the impact on the control and safety characteristics of the reactor. These studies have not identified any aspects that significantly impact upon the introduction of

  10. Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Ganguly, C.

    2005-01-01

    Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

  11. An optimum fuel management method based on CANDU in-core detector readings

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok

    2001-01-01

    In this study, a new optimal fuel management method is developed for a CANDU 600 MWe (CANDU-6) reactor. At first, an efficient power mapping method has been developed, which provides an accurate core status of an operating CANDU reactor. Secondly, an optimum refueling channel selection method has been developed by an optimization theory. For the power mapping method, the measured detector readings are used as boundary conditions of the diffusion theory calculation with the Kalman filtering (DIKAL) method. The performance of the DIKAL method was assessed for various core states and applied to the calculation of power and flux distribution in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL method is insensitive to the detector random and systematic errors. An optimal refueling simulation method (OPTIMA), practically applicable to a CANDU 6 reactor, has also been developed. The objective of the optimization is to reproduce the reference core performance during refueling simulation, while satisfying the operation limits of channel and bundle powers. The optimization process consists of two stages: i) elimination of candidate refueling channels by several constraints and ii) selection of refueling channels by a direct search method that uses sensitivity coefficients of channel power generated for the reference core. The elimination process sorts out an appropriate number of fuel channels suitable for refueling, considering the channel power, bundle power and fuel burnup. The optimum refueling channels are then selected such that the difference of power distribution from the reference is minimized. In order to demonstrate the applicability of the overall fuel management methodology developed in this study, the DIKAL-OPTIMA method was applied to Wolsong-3 reactor refueling simulation, which is a typical CANDU-6 reactor. The results of refueling simulation have shown that the method can be efficiently used for the performance analysis of the operating

  12. An optimum fuel management method based on CANDU in-core detector readings

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Choi, Hang Bok

    2001-01-01

    In this study, a new optimal fuel management method is developed for a CANDU 600 MWe (CANDU-6) reactor. At first, an efficient power mapping method has been developed, which provides an accurate core status of an operating CANDU reactor. Secondly, an optimum refueling channel selection method has been developed by an optimization theory. For the power mapping method, the measured detector readings are used as boundary conditions of the diffusion theory calculation with the Kalman filtering (DIKAL) method. The performance of the DIKAL method was assessed for various core states and applied to the calculation of power and flux distribution in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL method is insensitive to the detector random and systematic errors. An optimal refueling simulation method (OPTIMA), practically applicable to a CANDU 6 reactor, has also been developed. The objective of the optimization is to reproduce the reference core performance during refueling simulation, while satisfying the operation limits of channel and bundle powers. The optimization process consists of two stages: i) elimination of candidate refueling channels by several constraints and ii) selection of refueling channels by a direct search method that uses sensitivity coefficients of channel power generated for the reference core. The elimination process sorts out an appropriate number of fuel channels suitable for refueling, considering the channel power, bundle power and fuel burnup. The optimum refueling channels are then selected such that the difference of power distribution from the reference is minimized. In order to demonstrate the applicability of the overall fuel management methodology developed in this study, the DIKAL-OPTIMA method was applied to Wolsong-3 reactor refueling simulation, which is a typical CANDU-6 reactor. The results of refueling simulation have shown that the method can be efficiently used for the performance analysis of the operating

  13. Qinshan CANDU NPP outage performance improvement through benchmarking

    International Nuclear Information System (INIS)

    Jiang Fuming

    2005-01-01

    With the increasingly fierce competition in the deregulated Energy Market, the optimization of outage duration has become one of the focal points for the Nuclear Power Plant owners around the world. People are seeking various ways to shorten the outage duration of NPP. Great efforts have been made in the Light Water Reactor (LWR) family with the concept of benchmarking and evaluation, which great reduced the outage duration and improved outage performance. The average capacity factor of LWRs has been greatly improved over the last three decades, which now is close to 90%. CANDU (Pressurized Heavy Water Reactor) stations, with its unique feature of on power refueling, of nuclear fuel remaining in the reactor all through the planned outage, have given raise to more stringent safety requirements during planned outage. In addition, the above feature gives more variations to the critical path of planned outage in different station. In order to benchmarking again the best practices in the CANDU stations, Third Qinshan Nuclear Power Company (TQNPC) have initiated the benchmarking program among the CANDU stations aiming to standardize the outage maintenance windows and optimize the outage duration. The initial benchmarking has resulted the optimization of outage duration in Qinshan CANDU NPP and the formulation of its first long-term outage plan. This paper describes the benchmarking works that have been proven to be useful for optimizing outage duration in Qinshan CANDU NPP, and the vision of further optimize the duration with joint effort from the CANDU community. (authors)

  14. Development of the Advanced CANDU Reactor control centre

    International Nuclear Information System (INIS)

    Malcolm, S.; Leger, R.

    2004-01-01

    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

  15. Development of the advanced CANDU reactor control centre

    International Nuclear Information System (INIS)

    Malcolm, S.; Leger, R.

    2004-01-01

    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

  16. Evaluation of fuel performance for fresh and aged CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

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

    International Nuclear Information System (INIS)

    Cross, W.G.

    1992-08-01

    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.)

  18. 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)

  19. CANDU project development

    International Nuclear Information System (INIS)

    Hedges, K.R.

    1995-01-01

    Advanced CANDU reactor design strategy follows an evolutionary approach, taking manageable steps in the development of power plants from today's available designs, and in parallel carrying out longer-term studies to develop future-generation reactor concepts. The major emphasis is on safety, on on reducing cost and schedule. New features are developed and thoroughly proof-tested before introduction into designs, in order to maximize owner confidence. (author). 4 figs

  20. CANDU project development

    Energy Technology Data Exchange (ETDEWEB)

    Hedges, K R [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Advanced CANDU reactor design strategy follows an evolutionary approach, taking manageable steps in the development of power plants from today`s available designs, and in parallel carrying out longer-term studies to develop future-generation reactor concepts. The major emphasis is on safety, on on reducing cost and schedule. New features are developed and thoroughly proof-tested before introduction into designs, in order to maximize owner confidence. (author). 4 figs.

  1. The next generation CANDU 6

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    1999-01-01

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

  2. Effect of lattice-level adjoint-weighting on the kinetics parameters of CANDU reactors

    International Nuclear Information System (INIS)

    Nichita, Eleodor

    2009-01-01

    Space-time kinetics calculations for CANDU reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the 'best' effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the 'best' effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the

  3. AUTOSORO: A fuel management study program for Ontario Hydro CANDU reactors

    International Nuclear Information System (INIS)

    Wilk, L.

    1988-01-01

    A computer program, AUTOSORO, has been developed to automatically simulate an Ontario Hydro CANDU reactor core for any time duration according to user-defined on-power refuelling criteria. It is a three-dimensional two-group diffusion code coupled to refuelling decision logic at three screening levels: burnup, coupled neighbor, full-core. A central feature is a projected local-iteration scheme for predicting fuelling-induced local neutron flux changes. Comparisons of AUTOSORO results with actual histories demonstrate that it will be an excellent productivity tool for future in-core fuel management studies, reducing several man-months of effort to several man-hours

  4. Parametric study of moderator heat exchanger for Candu 6 advanced reactor

    International Nuclear Information System (INIS)

    Umar, Efrizon; Vecchiarelli, Jack

    2000-01-01

    The passive moderator system for Candu 6 advanced reactor require moderator heat exchanger with the small size and the low resistance coefficient of the shell-side. The study is to determine the required size of moderator heat exchanger, and to calculate the shell side of resistance coefficient have been done. Using computer code CATHENA, it is concluded that the moderator heat exchanger can be used at full power-normal operation condition, especially for the cases with 3600 to 8100 number of tube and 15.90 mm tube diameter. This study show that the proposed moderator heat exchanger have given satisfactory results

  5. Applying operating experience to design the CANDU 3 process

    International Nuclear Information System (INIS)

    Harris, D.S.; Hinchley, E.M.; Pauksens, J.; Snell, V.; Yu, S.K.W.

    1991-01-01

    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

  6. Applying operating experience to design the CANDU 3 process

    Energy Technology Data Exchange (ETDEWEB)

    Harris, D S; Hinchley, E M; Pauksens, J; Snell, V; Yu, S K.W. [AECL-CANDU, Ontario (Canada)

    1991-04-01

    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

  7. Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

    Energy Technology Data Exchange (ETDEWEB)

    Catana, A.; Prodea, L. [RAAN, Institute for Nuclear Research, Arges (Romania); Danila, N.; Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica(Romania)

    2007-07-01

    The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up.

  8. Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

    International Nuclear Information System (INIS)

    Catana, A.; Prodea, L.; Danila, N.; Prisecaru, I.; Dupleac, D.

    2007-01-01

    The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up

  9. The use of graphite for the reduction of void reactivity in CANDU reactors

    International Nuclear Information System (INIS)

    Min, B.J.; Kim, B.G.; Sim, K-S.

    1995-01-01

    Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

  10. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    Hilborn, J.S.; Seddon, W.A.; Barnstaple, A.G.

    1980-08-01

    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)

  11. CANDU market prospects

    International Nuclear Information System (INIS)

    Kakaria, B.K.

    1994-01-01

    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

  12. The future role of thorium in assuring CANDU fuel supplies

    International Nuclear Information System (INIS)

    Slater, J.B.

    1985-01-01

    Atomic Energy of Canada Limited (AECL), in partnership with Canadian industry and power utilities, has developed the CANDU reactor as a safe, reliable and economic means of transforming nuclear fuel into useable power. The use of thorium/uranium-233 recycle gives the possibility of a many-fold increase in energy yield over that which can be obtained from the use of uranium in once-through cycles. The neutronic properties of uranium-233 combine with the inherent neutron economy of the CANDU reactor to offer the possibility of near-breeder cycles in which there is no net consumption of fissile material under equilibrium fuelling conditions. Use of thorium cycles in CANDU will limit the impact of higher uranium prices. When combined with the potential for significant reductions in CANDU capital costs, then the long-term prospect is for generating costs near to current levels. Development of thorium cycles in CANDU will safeguard against possible uranium shortages in the next century, and will maintain and continue the commercial viability of CANDU as a long-term energy technology. (author)

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

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Bock, D.; Miller, A.; Kuran, S.; Keil, H.; Fiorino, L.; Duffey, R.; Dunbar, R.B.

    2003-01-01

    Atomic Energy of Canada Limited (AECL) is developing the ACR-700 TM (Advanced CANDU Reactor-700 TM ) 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

  14. The evolution of CANDU containment design

    International Nuclear Information System (INIS)

    Pendergast, Duane R.; Meneley, Daniel A.

    1995-01-01

    This paper reviews Canadian containment design, reflects on forces which have shaped the evolving designs and contemplates future containment design subject to existing constraints of Canadian and International nuclear power regulations. The discussion mentions modifications which could play a role in easing customary restrictions on siting while meeting the intent and literal definition of Canadian licensing requirements. Many different containment concepts have been considered and deployed over the years. Professor Birkhofer points out that the earliest systems installed were 'dry containment'. These rely on large containment buildings designed to contain all the the steam from a loss of coolant. It was soon realized the large water inventory of boiling water reactors would require very large and expensive containment buildings. This led to the development of pressure suppression concepts. Various pressure suppression schemes have been devised to condense steam and thus reduce design pressure and volume requirements. Examples include the forcing of steam into condensing water pools of BWR's steam condensation by passage into arrays of ice, dousing spray systems in CANDU reactors, and the spray condenser systems of late model Soviet designed VVER - 440 reactors. The fundamental concept of energy removal by steam condensation can result in a small containment volume requiring low design pressure capability

  15. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  16. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

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

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    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)

  18. The Intelligent Safety System: could it introduce complex computing into CANDU shutdown systems

    International Nuclear Information System (INIS)

    Hall, J.A.; Hinds, H.W.; Pensom, C.F.; Barker, C.J.; Jobse, A.H.

    1984-07-01

    The Intelligent Safety System is a computerized shutdown system being developed at the Chalk River Nuclear Laboratories (CRNL) for future CANDU nuclear reactors. It differs from current CANDU shutdown systems in both the algorithm used and the size and complexity of computers required to implement the concept. This paper provides an overview of the project, with emphasis on the computing aspects. Early in the project several needs leading to an introduction of computing complexity were identified, and a computing system that met these needs was conceived. The current work at CRNL centers on building a laboratory demonstration of the Intelligent Safety System, and evaluating the reliability and testability of the concept. Some fundamental problems must still be addressed for the Intelligent Safety System to be acceptable to a CANDU owner and to the regulatory authorities. These are also discussed along with a description of how the Intelligent Safety System might solve these problems

  19. Load-following performance and assessment of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Floyd, M.; Rattan, D.; Xu, Z.; Manzer, A.; Lau, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Power Generation, Fuel and Fuel Channel Analysis Dept., Toronto, Ontario (Canada)

    1999-09-01

    Load following of nuclear reactors is now becoming an economic necessity in some countries. When nuclear power stations are operated in a load-following mode, the reactor and the fuel may be subjected to step changes in power on a weekly, daily, or even hourly basis, depending on the grid's needs. This paper updates the previous surveys of load-following capability of CANDU fuel, focusing mainly on the successful experience at the Bruce B station. As well, initial analytical assessments are provided that illustrate the capability of CANDU fuel to survive conditions other than those for which direct in-reactor evidence is available. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  1. The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

    1995-01-01

    This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

  2. Candu 600 fuelling machine testing, the romanian experience

    International Nuclear Information System (INIS)

    Valeca, S.; Doca, C.; Iorga, C.

    2013-01-01

    The Candu 600 Fuelling Machine is a complex mechanism which must run in safety conditions and with high reliability in the Candu Reactor. The testing and commissioning process of this nuclear equipment meets the high standards of NPPs requirements using special technological facilities, modern measurement instruments as well the appropriate IT resources for data acquisition and processing. The paper presents the experience of the Institute for Nuclear Research Pitesti, Romania, in testing Candu 600 Fuelling Machines, inclusive the implied facilities, and in development of four simulators: two dedicated for the training of the Candu 600 Fuelling Machine Operators, and another two to simulate some process signals and actions. (authors)

  3. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    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)

  4. An integrated CANDU system

    International Nuclear Information System (INIS)

    Donnelly, J.

    1982-09-01

    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

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

    International Nuclear Information System (INIS)

    Veeder, J.; Didsbury, R.

    1985-06-01

    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

  6. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.

    2003-01-01

    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)

  7. The future for CANDU

    International Nuclear Information System (INIS)

    Foster, J.S.

    1977-06-01

    Canada could have 60,000 MW(e) of installed nuclear-electric generating capacity by the year 2000 and have exported the plan to generate a further 5,000 MW(e). While the CANDU reactor can readily be scaled up to larger unit sizes, its real potential lies in the even greater efficiency that can be obtained by using alternative fuel cycles. The thorium - uranium-233 fuel cycle, for instance, makes it possible to attain a conversion factor of unity, or a little better, on a feed of pure thorium in a substantially unmodified CANDU reactor. Further developments, such as spallation, offer means of converting fertile to fissile material to provide a fissile inventory for an expanding system. The coincidence of expected future shortages of other energy supplies with continuing good experience in the nuclear field should assist in creating a climate that will permit accelerated nuclear power development. (author)

  8. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  9. Steps to Advanced CANDU 600

    International Nuclear Information System (INIS)

    Oh, Yongshick; Brooks, G. L.

    1988-01-01

    The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

  10. Advanced instrumentation and control systems for CANDU refurbishment

    International Nuclear Information System (INIS)

    Sklyar, V.; Bakhmach, I.; Kharchenko, V.; Andrashov, A.; Baranova, O.

    2011-01-01

    The purpose of the work is to discuss opportunities to modernize I and C systems of CANDU reactors on the base of Radiy's digital safety platform. This paper discusses the following topics: a business model for CANDU, I and C systems refurbishment, FPGA technology issues, comparison of different approaches to refurbish obsolete I and C systems. (author)

  11. CANDU-6 fuel optimization for advanced cycles

    Energy Technology Data Exchange (ETDEWEB)

    St-Aubin, Emmanuel, E-mail: emmanuel.st-aubin@polymtl.ca; Marleau, Guy, E-mail: guy.marleau@polymtl.ca

    2015-11-15

    Highlights: • New fuel selection process proposed for advanced CANDU cycles. • Full core time-average CANDU modeling with independent refueling and burnup zones. • New time-average fuel optimization method used for discrete on-power refueling. • Performance metrics evaluated for thorium-uranium and thorium-DUPIC cycles. - Abstract: We implement a selection process based on DRAGON and DONJON simulations to identify interesting thorium fuel cycles driven by low-enriched uranium or DUPIC dioxide fuels for CANDU-6 reactors. We also develop a fuel management optimization method based on the physics of discrete on-power refueling and the time-average approach to maximize the economical advantages of the candidates that have been pre-selected using a corrected infinite lattice model. Credible instantaneous states are also defined using a channel age model and simulated to quantify the hot spots amplitude and the departure from criticality with fixed reactivity devices. For the most promising fuels identified using coarse models, optimized 2D cell and 3D reactivity device supercell DRAGON models are then used to generate accurate reactor databases at low computational cost. The application of the selection process to different cycles demonstrates the efficiency of our procedure in identifying the most interesting fuel compositions and refueling options for a CANDU reactor. The results show that using our optimization method one can obtain fuels that achieve a high average exit burnup while respecting the reference cycle safety limits.

  12. Technologies in support of CANDU development

    International Nuclear Information System (INIS)

    Turner, C.; Tapping, B.

    2005-01-01

    Atomic Energy of Canada, Ltd. (AECL) has significant research and development (R and D) programs designed to meet the needs of both existing CANDU reactors and new and evolving CANDU plant designs. These R and D programs cover a wide range of technology, from chemistry and materials support through to inspection and life management tools. Emphasis is placed on effective technology development programs for fuel channels, feeders and steam generators to ensure their operation through design life, and beyond. This paper specifically addresses how the R and D has been applied in the production of longer-lived pressure tubes for the most recent CANDU 6 reactors, and how this technology forms the basis for the pressure tubes of the Advanced CANDU Reactor (ACR). Similarly, AECL has developed solutions for other critical components such as calandria tubes, feeder pipe and steam generators. The paper also discusses how the R and D knowledge has been integrated into aging management databases and health monitoring tools. Since 1997, AECL has been working with CANDU utilities on comprehensive and integrated CANDU Plant Life Management (PLiM) programs for successful and reliable plant operation through design life and beyond. AECL has developed and implemented an advanced chemistry monitoring and diagnostic system, called ChemAND which allows on-line access by the operators to current and past chemistry conditions enabling appropriate responses and facilitating planning of shutdown maintenance actions. An equivalent tool for monitoring, trending and diagnosing thermal and mechanical data has also been developed; this tool is called ThermAND. AECL is developing the Maintenance Information, Monitoring, and Control (MIMC) system, which provide information to the user for condition-based decision-making in maintenance. To enable more effective inspections, surveillance and data collection, AECL has developed unique one-off tooling to carry out unanticipated inspection and repair

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

    International Nuclear Information System (INIS)

    Erp, J.B. van

    1978-01-01

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

  14. Verification tests for CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-15

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

  16. Advancing CANDU technology AECL's Development program

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1997-01-01

    AECL has a comprehensive product development program that is advancing all aspects of CANDU technology including fuel and fuel cycles, fuel channels, heavy water and tritium technology, safety technology, components and systems, constructability, health and environment, and control and instrumentation. The technology arising from these programs is being incorporated into the CANDU design through an evolutionary process. This evolutionary process is focused on improving economics, enhancing safety and ensuring fuel cycle flexibility to secure fuel supply for the foreseeable future. This strategic thrusts are being used by CANDU designers and researchers to set priorities and goals for AECL's development activities. The goals are part of a 25-year development program that culminates in the 'CANDU X'. The 'CANDU X' is not a specific design - it is a concept that articulates our best extrapolation of what is achievable with the CANDU design over the next 25 years, and includes the advanced features arising from the R and D and engineering to be done over that time. AECL's current product, the 700 MWe class CANDU 6 and the 900 MWe class CANDU 9, both incorporate output from the development programs as the technology become available. A brief description of each development areas is given below. The paper ends with the conclusion that AECL has a clear vision of how CANDU technology and products will evolve over the next several years, and has structured a comprehensive development program to take full advantage of the inherent characteristics of heavy water reactors. (author)

  17. Tritium resources available for fusion reactors

    Science.gov (United States)

    Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

    2018-02-01

    The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

  18. Recent advances in self-powered flux detector development for CANDU reactors

    International Nuclear Information System (INIS)

    Allan, C.J.; Drewell, N.H.; Hall, D.S.

    1983-01-01

    The characteristics of self-powered flux detectors used in CANDU reactors are reviewed. Detectors with emitters of vanadium, platinum, platinum-clad Inconel and Inconel are used. Data on dynamic response, relative neutron and gamma-ray sensitivities, and burnout, obtained both from experiments and from the Monte Carlo code ICARES, are presented. Since the response of a detector depends on the relative magnitudes of the various current-producing mechanisms, the operating principles of self-powered detectors are briefly reviewed. Current research programmes are discussed. These include modifying the design of the platinum-clad Inconel detector in order to match its dynamic response to that of the fuel power and developing a prompt-responding flux-mapping detector. (author)

  19. CANDU-9/480-SEU fuel handling system assessment document

    International Nuclear Information System (INIS)

    Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.; Morikawa, D. T.

    1996-11-01

    This report summarize the rationale for the CANDU 9 fuel handling system, and the design choices recommended for components of the system. Some of the design requirements applicable to the CANDU 9 480-SEU fuel handling design choices are described. These requirements imposed by the CANDU 9 project. And the design features for the key components of fuel handling system, such as the fuelling machine, the carriage, the new fuel transfer system and the irradiated fuel transfer system, are described. The carriage seismic load evaluations relevant to the design are contained in the appendices. The majority of the carriage components are acceptable, or will likely be acceptable with some redesign. The concept for the CANDU 9 fuel handling system is based on proven CANDU designs, or on improved CANDU technology. Although some development work must be done, the fuel handling concept is judged to be feasible for the CANDU 9 480-SEU reactor. (author). 2 refs

  20. Material and fabrication considerations for the CANDU-PHWR heat transport system

    International Nuclear Information System (INIS)

    Filipovic, A.; Price, E.G.; Barber, D.; Nickerson, J.

    1987-03-01

    CANDU PHWR nuclear systems have used carbon steel material for over 25 years. The accumulated operating experience of over 200 reactor years has proven this unique AECL approach to be both technically and economically attractive. This paper discusses design, material and fabrication considerations for out-reactor heat transport system major components. The contribution of this unique choice of materials and equipment to the outstanding CANDU performance is briefly covered

  1. Fuel Management in Candu Reactors Using Tabu Search

    International Nuclear Information System (INIS)

    Chambon, R.; Varin, E.

    2008-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  4. Development of a new bundle welding technology for CANDU fuels

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, D. Y.; Goo, D. S.

    2010-01-01

    The new technology of welding process for fuel bundle of CANDU nuclear fuels is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments of fuel bundles, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. The fuel bundles of CANDU nuclear fuels are welded by the electrical resistance method, connecting the endplates and endcaps with fuel rods. Therefore, the purpose of this study of the 2nd year is to select the proper welding parameters and to investigate the characteristics of the full-sized samples using the projection endplates and make some prototype samples for the endplate welding of CANDU nuclear fuels. This study will be also provide the fundamental data for the new design and fabrications of CANDU nuclear fuel bundles

  5. A Preliminary Study on the Reuse of the Recovered Uranium from the Spent CANDU Fuel Using Pyroprocessing

    International Nuclear Information System (INIS)

    Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.

    2009-01-01

    During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated

  6. Development of the CANDU 66-group SN transport library

    International Nuclear Information System (INIS)

    Tsang, K.T.

    2001-01-01

    The design of the shield configuration around a nuclear reactor is strongly dependent on the neutron and photon spatial and energy distributions. The nuclear heat deposition and material damage in and surrounding the reactor core are also a function of the neutron and photon distributions. Therefore, to ensure a suitable configuration of materials for shielding or heat transfer, an accurate calculation of the particle fluxes in the reactor systems is essential. The CANDU 66-group library was developed to update the cross sections that are needed to assess the performance of CANDU bulk shields. Since about 1980, shielding analysts at Atomic Energy of Canada Limited (AECL) and Ontario Power Generation Inc. (OPGI) have been using a 38-group CANDU-specific library to perform S N transport calculations. In 1994, a new CANDU 67-group cross-section library was developed. The 67-group cross-section library was developed to provide radiation-physics analysts with up-to-date nuclear data to correct deficiencies with documentation of the old library. Although there were improvements over the 38-group library, initial use showed there were some deficiencies in the 67-group library. To correct these deficiencies, the CANDU 66-group S N transport cross-section library was developed. The 66-group library is based on the 241-group cross-section library VITAMIN-B6. Collapsing and weighting of the 241-group cross sections into 66 groups were performed using the modular code system SCALE 4.4. This paper describes how the modules in the SCALE system were applied to generate the 66-group library. The CANDU 66-group library includes both core-weighted and lattice-weighted cross sections of 235 U, 238 U, and 239 Pu with, and without, delayed fission-product photons. In addition, the 66-group library contains more response functions than did the 67-group library. Finally, the CANDU 66-group library has been validated against one-dimensional benchmark problems. The results generated with

  7. Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun

    2001-08-01

    This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given.

  8. Nuclear energy in Canada: the CANDU system

    International Nuclear Information System (INIS)

    Robertson, J.A.L.

    1979-10-01

    Nuclear electricity in Canada is generated by CANDU nuclear power stations. The CANDU reactor - a unique Canadian design - is fuelled by natural uranium and moderated by heavy water. The system has consistently outperformed other comparable nuclear power systems in the western world, and has an outstanding record of reliability, safety and economy. As a source of energy it provides the opportunity for decreasing our dependence on dwindling supplies of conventional fossil fuels. (auth)

  9. Improved locations of reactivity devices in future CANDU reactors fuelled with natural uranium or enriched fuels

    International Nuclear Information System (INIS)

    Boczar, P.G.; Van Dyk, M.T.

    1987-02-01

    A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg

  10. CANDU operating experience

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.; Thurygill, E.W.

    1980-05-01

    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)

  11. Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  12. Regional overpower protection system analysis for a DUPIC fuel CANDU core

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok; Park, Jee Won

    2003-06-01

    The regional overpower protection (ROP) system was assessed a CANDU 6 reactor with the DUPIC fuel, including the validation of the WIMS/RFSP/ROVER-F code system used for the estimation of ROP trip setpoint. The validation calculation has shown that it is valid to use the WIMS/RFSP/ROVER-F code system for ROP system analysis of the CANDU 6 core. For the DUPIC core, the ROP trip setpoint was estimated to be 125.7%, which is almost the same as that of the standard natural uranium core. This study has shown that the DUPIC fuel does not hurt the current ROP trip setpoint designed for the natural uranium CANDU 6 reactor

  13. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  14. Development of modern CANDU PHWR cross-section libraries for SCALE

    International Nuclear Information System (INIS)

    Shoman, Nathan T.; Skutnik, Steven E.

    2016-01-01

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of "1"5"4Eu and "1"5"5Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the "1"5"4Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a typical

  15. Plating end fittings to reduce hydrogen ingress at rolled joints in CANDU reactors

    International Nuclear Information System (INIS)

    White, A.J.; Urbanic, V.F.; Bahurmuz, A.A.; Clendening, W.R.; Joynes, R.; McDougall, G.M.; Skinner, B.C.; Venkatapathi, S.

    1993-10-01

    Zr-2.5Nb pressure tubes in CANDU nuclear reactors absorb hydrogen at a low rate from the primary heat transport water circulated through the tubes. Extra hydrogen is picked up at the rolled joints that connect the pressure tubes to out-of-core steel piping. This enhanced ingress may contribute to pressure-tube cracking at incorrectly assembled joints. The risk of pressure-tube failure has been decreased by ensuring correct joint assembly, and could be further decreased by reducing hydrogen ingress at rolled joints. This paper reviews progress toward using plated end fittings to reduce rolled-joint hydrogen ingress

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

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    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 reg-sign 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

  17. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  18. Advancing CANDU Technology Through R and D

    International Nuclear Information System (INIS)

    Torgerson, David F.

    1993-01-01

    CANDU reactors are evolving to meet future requirements using incremental changes as opposed to revolutionary design changes. The main elements for advancing the technology reducing capital and operating, increasing capacity factors, increasing passive safety, and enhancing fuel/fuel cycle flexibility. These elements are being addressed by carrying out research and development in the areas of safety, plant systems and components, heavy water production, information technology, fuel channels, and fuel/fuel cycle technology. In safety, the focus is on using the inherent features of CANDU to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of advanced fuels to improve critical heat flux and to reduce source terms. Plant systems and components work includes improvements to plant systems such as steam generators, heat exchangers, pump seals, and advanced control room technology. Heavy water processes are being developed that can be used with existing hydrogen production plants, or that can be used in a stand-alone mode. Information technology is being developed to cover all aspects of CANDU design, construction, and operation. Fuel channel improvements include elucidation and application of basic materials science for life extension, and the development of advanced non-destructive examination methods. Fuel and fuel cycle work is focusing on LWR/CANDU synergy, such as the use of recovered uranium and the direct use of spent PWR fuel in CANDU reactor, advanced fuels to improve burnup and economics (e. g., the joint AECB/KAERI Conflux program), and low void reactivity fuel to enhance passive safety. This paper gives an overview of some of the R and D supporting these activities, with particular emphasis on Alice's vision for advancing CANDU technology over the next 10 years

  19. Future development in heavy water reactors in Canada

    International Nuclear Information System (INIS)

    Donnelly, J.; Hart, R.G.

    1982-01-01

    1982 marks the 35th anniversary of the start-up of Canada's first research and test reactor, NRX. Its first power reactor has been operating successfully for the past 20 years. With 5,000 MWe of domestic capacity installed, Canada's major CANDU (Canada Deuterium, Uranium) nuclear program has a further 9,500 MWe under construction in Canada for completion by 1990 as well as committed offshore projects in Argentina, Korea and Romania. The CANDU operating record, by any measure of performance, has been outstanding. This performance is largely due to the discipline imposed on the development, design, construction and operation by two fundamental choices: natural uranium and heavy water. The impact of these two choices on availability, fuel utilization, safety and economics is discussed. Future plans call for building on those characteristics which have made CANDU so successful. When time for change comes, current assessments indicate that it will be possible to convert to more efficient advanced fuel cycles without major changes to the basic CANDU design. Primary attention is being focussed on thorium fuel cycles to ensure an abundant and continuing supply of low cost energy for the long term. The resource savings available from these fuel cycles in expanding systems are reviewed and compared with those available from LWR's and Fast Breeders. The results clearly illustrate the versatility of the CANDU reactor. It can benefit from enrichment plants or get along without them. It can complement LWR's or compete with them. It can complement Fast Breeder Reactors or compete with them as well. In the very long term CANDU's and Fast Breeders combined offer the potential of burning all the world's uranium and all the world's thorium. (author)

  20. CANDU channel flow verification

    International Nuclear Information System (INIS)

    Mazalu, N.; Negut, Gh.

    1997-01-01

    The purpose of this evaluation was to obtain accurate information on each channel flow that enables us to assess precisely the level of reactor thermal power and, for reasons of safety, to establish which channel is boiling. In order to assess the channel flow parameters, computer simulations were done with the NUCIRC code and the results were checked by measurements. The complete channel flow measurements were made in the zero power cold condition. In hot conditions there were made flow measurements using the Shut Down System 1 (SDS 1) flow devices from 0.1 % F.P. up to 100 % F.P. The NUCIRC prediction for CANDU channel flows and the measurements by Ultrasonic Flow Meter at zero power cold conditions and SDS 1 flow channel measurements at different reactor power levels showed an acceptable agreement. The 100 % F.P. average errors for channel flow of R, shows that suitable NUCIRC flow assessment can be made. So, it can be done a fair prediction of the reactor power distribution. NUCIRC can predict accurately the onset of boiling and helps to warn at the possible power instabilities at high powers or it can detect the flow blockages. The thermal hydraulic analyst has in NUCIRC a suitable tool to do accurate predictions for the thermal hydraulic parameters for different steady state power levels which subsequently leads to an optimal CANDU reactor operation. (authors)

  1. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    Yu, A.M.

    1982-09-01

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

  2. CANDU-BLW-250

    Energy Technology Data Exchange (ETDEWEB)

    Pon, G A

    1967-09-15

    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. Challenges associated with the current processes for ultrasonic inspection of CANDU reactor feeder piping

    Energy Technology Data Exchange (ETDEWEB)

    Machowski, C. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

    2012-07-01

    CANDU® PHT Feeder Piping is generally constructed from SA106 Grade B carbon steel, which is known to be susceptible to flow accelerated corrosion when exposed to certain environmental conditions. The configuration of the CANDU reactor promotes thinning of the inside surface of the pipe walls, predominantly at the outlet feeders. Inspection of this piping is currently conducted using ultrasonic techniques and is governed by the requirements established by the CANDU Owners Group (COG). There are many challenges associated with these inspections as a result of the complexity of the reactor piping configuration. Geometrical anomalies on the surface of the pipe and non-circular geometries at the tight radius bends hinder the performance of conventional ultrasonic techniques. This can cause lost signals in areas of interest, which in turn often results in rework in order to satisfy the inspection requirements and justify fitness for service of these components. There are also many inspection sites which have limited access due to physical restrictions on the reactor face; therefore in order to maximize the performance of an inspection campaign, it is paramount that the inspection personnel and the inspection technology be well integrated through training simulations prior to execution. These inspection challenges increase the complexity of the analysis process as ultrasonic signals get distorted and lost as a result of non-circular pipe geometries. In order to ensure a high level of integrity in the analysis results, a conservative process is utilized in which two analysts independently examine the data, and a third analyst reviews their results and submits the final call. A Data Management Software application (DMS) is used to input and store the three analysis results. Another important function of the DMS is to provide a communication link between the different work-groups associated with the inspection activities. The focus of this presentation discusses:

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

    International Nuclear Information System (INIS)

    Lee, D.

    1995-02-01

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

  5. CANDU physics considerations for the disposition of weapons-grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Pitre, J; Chan, P; Dastur, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    At the request of the US Department of Energy AECL has examined the feasibility of using CANDU for the disposition of weapons grade plutonium. Utilizing existing CANDU technology, the feasibility of using MOX (mixed oxide) fuel in an existing CANDU reactor was studied. The results of this study indicate that the target disposition for disposal of weapons grade plutonium can be met without the requirement of any major modifications to existing plant design. (author). 3 refs., 4 tabs., 5 figs.

  6. CANDU physics considerations for the disposition of weapons-grade plutonium

    International Nuclear Information System (INIS)

    Pitre, J.; Chan, P.; Dastur, A.

    1995-01-01

    At the request of the US Department of Energy AECL has examined the feasibility of using CANDU for the disposition of weapons grade plutonium. Utilizing existing CANDU technology, the feasibility of using MOX (mixed oxide) fuel in an existing CANDU reactor was studied. The results of this study indicate that the target disposition for disposal of weapons grade plutonium can be met without the requirement of any major modifications to existing plant design. (author). 3 refs., 4 tabs., 5 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  8. Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

    International Nuclear Information System (INIS)

    Torabi, T.; Bettig, R.; Iliescu, P.; Robinson, J.; Santamaura, P.; Skorupska, B.; Tyagi, A.K.; Vencel, I.

    2013-01-01

    The Enhanced CANDU 6(EC6) is a 700 MWe reactor, which has evolved from the well-established CANDU line of reactors, which are heavy-water moderated, and heavy-water cooled horizontal pressure tube reactors, using natural uranium fuel. The EC6 design retains the generic CANDU design features, while incorporating innovations and state-of-the-art technologies to ensure competitiveness with other design with respect to operation, performance and economics. A design assist probabilistic safety assessment (PSA) was conducted during the design change phase of the project. The purpose of the assessment was to assess internal events during at-power operation and identify the design improvements and additional features needed to comply with the latest regulatory requirements in Canada and compete with other reactor designs, internationally. The PSA results show that the EC6 plant response to the postulated initiating events is well balanced, and the design meets its safety objectives. This paper summarizes the results and insights gained during the development of the PSA models for at-power internal events. (author)

  9. 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).

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

    International Nuclear Information System (INIS)

    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

    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)

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

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

    1989-01-01

    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/dm 2 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)

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

    International Nuclear Information System (INIS)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  14. Natural uranium equivalent fuel an innovative design for proven CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)

    2015-07-01

    The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

  15. The Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the two-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These two programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

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

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    2001-01-01

    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

  17. Transmutation of minor actinides in a Candu thorium borner

    International Nuclear Information System (INIS)

    Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Yildiz, K.; Sahin, N.; Altinok, T.; Alkan, M.

    2007-01-01

    The paper investigates the prospects of exploitation of rich world thorium reserves in CANDU reactors. 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 form of mixed ThO 2 /PuO 2 fuel in a CANDU fuel bundle in order to assure reactor criticality. Two different fuel compositions have been selected for investigations: 1) 96% thoria (ThO 2 ) + 4% PuO 2 and 2) 91% ThO 2 + 5% UO 2 + 4 PuO 2 . The latter is used for the purpose of denaturing the new 2 33U fuel with 2 38U. The behavior of the 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 the criticality drops down asymptotically to values k ∞ > 1.06, still tolerable and usable in a CANDU reactor. Reactor criticality k ∞ remains nearly constant between the 4th year and 7th year of plant operation and then a slight increase is observed thereafter, along with a continuous depletion of 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 burnt 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. There is a great quantity of weapon grade plutonium accumulated in nuclear stockpiles. In the second phase of investigations, weapon grade plutonium is used as a booster fissile fuel material in form of mixed ThO 2 /PuO 2 fuel in a CANDU fuel bundle in order to assure the initial criticality at startup. Two different fuel compositions have been used: 1) 97% thoria (ThO 2 ) + 3% PuO 2 and 2) 92% ThO 2 + 5% UO 2 + 3% PuO 2 . The

  18. Development of a Web-based CANDU Core Management Procedure Automation System

    International Nuclear Information System (INIS)

    Lee, Sanghoon; Kim, Eunggon; Park, Daeyou; Yeom, Choongsub; Suh, Hyungbum; Kim, Sungmin

    2006-01-01

    CANDU reactor core needs efficient core management to increase safety, stability, high performance as well as to decrease operational cost. The most characteristic feature of CANDU is so called 'on-power refueling' i.e., there is no shutdown during refueling in opposition to that of PWR. Although this on-power refueling increases the efficiency of the plant, it requires heavy operational task and difficulties in real time operation such as regulating power distribution, burnup distribution, LZC statistics, the position of control devices and so on. To enhance the CANDU core management, there are several approaches to help operator and reduce difficulties, one of them is the COMOS (CANDU Core On-line Monitoring System). It has developed as an online core surveillance system based on the standard incre instrumentation and the numerical analysis codes such as RFSP (Reactor Fueling Simulation Program). As the procedure is getting more complex and the number of programs is increased, it is required that integrated and cooperative system. So, KHNP and IAE have been developing a new web-based system which can support effective and accurate reactor operational environment called COMPAS that means CANDU cOre Management Procedure Automation System. To ensure development of successful system, several steps of identifying requirements have been performed and Software Requirement Specification (SRS) document was developed. In this paper we emphasis on the how to keep consistency between the requirements and system products by applying requirement traceability methodology

  19. Highlights of the metallurgical behaviour of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Price, E.G.

    1984-10-01

    This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

  20. MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS

    International Nuclear Information System (INIS)

    Kwee, M.T.; Choi, M.H.; Leung, R.K.

    1996-01-01

    Severe accidents have been the subject of a great deal of analysis and research, particularly in the light water reactor community. Although severe accident analysis in Canada deuterium-uranium (CANDU) reactors has not been published abundantly, a significant body of research and analysis has been accumulated. This has occurred because CANDU has directly taken into consideration a set of severe accidents [e.g loss-of-coolant accidents (LOCAs) coincident with a loss-of-emergency-coolant injection (LOECI)] in the design basis. These accidents have served to define the design requirements that ensure the integrity of the heat transport system. The CANDU reactor design has inherent heat sinks such as the primary heat transport system, the secondary side, moderator system, and shielding system (shield tank and end shields). These heat sinks are significant and are able to moderate or terminate the progression of severe accidents that go beyond the design base cases. These types of accidents are typically analyzed at Ontario Hydro in conjunction with probabilistic safety analysis (PSA), where the severe accident consequences are analyzed by a series of conservative hand-calculation methods

  1. The status of safeguarding 600 MW(e) CANDU reactors

    International Nuclear Information System (INIS)

    Von Baeckmann, A.; Rundquist, D.E.; Pushkarjov, V.; Smith, R.M.; Zarecki, C.W.

    1982-09-01

    There has been extensive work in the development of CANDU safeguards since the last International Conference on Nuclear Power, and this has resulted in the development of improved equipment for the safeguards system now being installed in the 600 MW(e) CANDU generating stations. The overall system is designed to improve on the existing IAEA safeguards and to provide adequate coverage for each plausible nuclear material diversion route. There is sufficient sensitivity and redundancy to enable the timely detection of the possible diversion of significant quantities of nuclear material

  2. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

    Collingwood, B.R.

    1997-01-01

    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

  3. Overview of current RFSP-code capabilities for CANDU core analysis

    International Nuclear Information System (INIS)

    Rouben, B.

    1996-01-01

    RFSP (Reactor Fuelling Simulation Program) is the major finite-reactor computer code in use at the Atomic Energy of Canada Limited for the design and analysis of CANDU reactor cores. An overview is given of the major computational capabilities available in RFSP. (author) 11 refs., 29 figs

  4. A description of the Canadian irradiation-research facility proposed to replace the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, A G; Lidstone, R F; Bishop, W E; Talbot, E F; McIlwain, H [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MW{sub t}, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 10{sup 18} n{center_dot}m{sup -2}{center_dot}s{sup -1} in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs.

  5. Advanced CANDU reactor design for operability

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Lalonde, R.; Soulard, M.

    2003-01-01

    This paper outlines design features and engineering processes in the ACR TM development program which contribute to excellence in performance and low operating cost. AECL recognizes that future plant owners will place a high priority in these operational characteristics. A successful next generation plant will have a best-in-class capability, both in its design characteristics, in the engineering philosophy and program adopted during the product development, and in the vendor's approach to operating station support. The ACR program addresses each of these drivers. Operability considerations are built-in to the design at an overall, plant wide level. For example, based on the strong CANDU 6 operating record, targets for standard outage duration, time between outages and component durability are set, while the design engineering is managed to achieve these targets. The ultimate maintenance target for the ACR, once initial operating experience has been gained, is to operate with a 21-day standard maintenance outage at an interval of once every three years. At the detailed design level, close attention is paid to space allocation, to enable good maintenance access. Selection of components also places emphasis on maintainability based on the extensive and current experience with CANDU projects. (author)

  6. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

  7. CANDU-BLW-250

    Energy Technology Data Exchange (ETDEWEB)

    Pon, G. A. [Atomic Energy of Canada Ltd, Sheridan Park, ON (Canada)

    1968-04-15

    The plant ''La Centrale nucleaire de Gentiliy'' is located between Montreal and Quebec City on the south shore of the St. Lawrence River. Startup is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. This 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: (i) Heat transfer. It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. Some experiments and backup calculations are presented to support this specification. (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 is detailed. (iv) Safety. The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analyses are presented. (author)

  8. Conceptual design of a two-phase flow absorber system for neutron flux regulation in a CANDU-PHW-1250 reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Moeck, E.O.

    1979-07-01

    A two-phase absorber control (TOPAC) system has been under development at the Chalk River Nuclear Laboratories to meet the need for improved spatial neutron flux control for future CANDU power reactors. Aspects of the conceptual design study presented in this paper include system controllability, in-reactor noise sensitiity, the effect of equipment malfunctions on plant operation, and a comparison with competing systems. The TOPAC system is shown to be a viable alternative to existing and future neutron flux regulating systems based on liquid H 2 O zone compartments. (auth)

  9. Prediction of Axial and Radial Creep in CANDU 6 Pressure Tubes

    International Nuclear Information System (INIS)

    Radu, Vasile S.

    2013-01-01

    Status and proposals: 1. A review of literature concerning on the in-reactor deformation of PTs has been carried ouţ. 2. A model based on MFNN has been proposed to assess the radial and axial creep of CANDU 6 PTs. 3. Preliminary discussion with Cernavoda NPP (Romania) has been lunched, and now the preparation of official documents (collaboration in providing the inspection data from fuel channel in Unit 1 and 2) are in progress. 4. Further activities: • Improvement MFNN to accommodate complex data base (eventually with many variables) for radial and axial in-reactor deformation PT, and to satisfy the requirements from NPP Cernavoda and hopefully from present CRP database; • To build-up a database by running the creep equations (if the creep constants are provided by AECL); training of MFNN on them and to qualify it as a tool for PT in-reactor deformation prediction

  10. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gheorghe [National Agency for Radioactive Waste, 1, Campului Str., 115400 Mioveni (Romania)], E-mail: gheorghe.negut@andrad.ro; Catana, Alexandru [Institute for Nuclear Research Pitesti, 1, Campului Str., Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie; Dupleac, Daniel [Politehnica University Bucharest, 313, Splaiul Independentei, Sect. 6, 060042 Bucharest (Romania)

    2009-09-15

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  11. Severe accident development modeling and evaluation for CANDU

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2009-01-01

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

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

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, J K; Wollin, S; Selvadurai, S; Saudy, A M [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    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.

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

    International Nuclear Information System (INIS)

    Biswas, J.K.; Wollin, S.; Selvadurai, S.; Saudy, A.M.

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    Dolbey, M.P.; Kupcis, O.A.

    1979-01-01

    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 (10 6 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  17. Preliminary analysis for u tube degradation in CANDU steam generator using CATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Shin, So Eun; Lee, Jeong Hun; Park, Tong Kyu; Hwang, Su Hyun [FNC Technology Co., Seoul (Korea, Republic of); Jung, Jong Yeo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The interest in plant safety and integrity has been increasing due to long term operation of nuclear power plants (NPPs) and lots of efforts have been devoted to developing the degradation evaluation model for all the Structure, System, and Components (SSCs) of NPPs in these days. The efforts, however, were mainly concentrated on pressurized light water reactors (PWRs) in domestic. In contrast, the study for the aging degradation of counterparts of CANDU (CANada Deuterium Uranium) reactors has been rarely performed, even though Wolsong unit 1 (WS1), that is a CANDU 6 NPP in Korea, has been operating for almost 30 years. Therefore, the assessment of the aging degradation is required and the proper and exact evaluation model for the aging degradation of SCCs of CANDU, especially WS1, is urgently needed. In this study, the aging degradation of steam generators (SGs) in WS1 was mainly discussed. Based on cases of the aging degradation of SGs in overseas CANDU reactors, the major potential aging mechanisms of SGs were estimated since there has been no case of accident due to degradation in CANDU NPPs in Korea . Some core parameters which are indicators of the degree of degradation were calculated by CATHENA (Canadian algorithm for thermal hydraulic network analysis). In the result of comparing two calculation cases; core parameters for only aged SGs in fresh plant and those for all the aged component, it can be concluded that aging of SGs is a main component in the degradation assessment of CANDU NPPs, and keeping the integrity of steam generator (SG) tubes is important to guarantee the safety of the NPPs.

  18. Chances dim for Sask. reactor

    International Nuclear Information System (INIS)

    1992-01-01

    It now appears quite unlikely that a new-generation CANDU 3 reactor will be build in Saskatchewan, as the minister responsible for such matters in the province backed away from Sask. Power's participation in a $50 million joint venture with Atomic Energy of Canada Ltd. Dwain Lingenfelter, Saskatchewan's economic diversification minister and the minister responsible for Sask. Power, said last week his government has a number of reservations about going ahead with the joint venture agreement, which flowed from a 1991 memorandum of understanding between then premier Grant Devine and federal Energy Ministry Jake Epp which would see Ottawa and Regina each spend $25 million to research various energy alternatives for the province. But, Lingenfelter said last week, the deal apparently hinged on Saskatchewan agreeing to provide a site for AECL CANDU's new CANDU 3 reactor and developing storage facilities for nuclear waste. 'It looks like we are putting $25 million into an agreement on nuclear well in advance of a decision by the government that this is the right way to be going.,' he said. 'We are spending the money on nuclear, and then saying we are going to study the options.'

  19. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    Lee, Ikhwan; Yu, S. K. W.

    1994-01-01

    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. Optimization of a fuel bundle within a CANDU supercritical water reactor

    International Nuclear Information System (INIS)

    Schofield, M.E.

    2009-01-01

    The supercritical water reactor is one of six nuclear reactor concepts being studied under the Generation IV International Forum. Generation IV nuclear reactors will improve the metrics of economics, sustainability, safety and reliability, and physical protection and proliferation resistance over current nuclear reactor designs. The supercritical water reactor has specific benefits in the areas of economics, safety and reliability, and physical protection. This work optimizes the fuel composition and bundle geometry to maximize the fuel burnup, and minimize the surface heat flux and the form factor. In optimizing these factors, improvements can be achieved in the areas of economics, safety and reliability of the supercritical water reactor. The WIMS-AECL software was used to model a fuel bundle within a CANDU supercritical water reactor. The Gauss' steepest descent method was used to optimize the above mentioned factors. Initially the fresh fuel composition was optimized within a 43-rod CANFLEX bundle and a 61-rod bundle. In both the 43-rod and 61-rod bundle scenarios an online refuelling scheme and non-refuelling scheme were studied. The geometry of the fuel bundles was then optimized. Finally, a homogeneous mixture of thorium and uranium fuel was studied in a 60-rod bundle. Each optimization process showed definitive improvements in the factors being studied, with the most significant improvement being an increase in the fuel burnup. The 43-rod CANFLEX bundle was the most successful at being optimized. There was little difference in the final fresh fuel content when comparing an online refuelling scheme and non-refuelling scheme. Through each optimization scenario the ratio of the fresh fuel content between the annuli was a significant determining cause in the improvements in the factors being optimized. The geometry optimization showed that improvement in the design of a fuel bundle is indeed possible, although it would be more advantageous to pursue it

  1. Results of the CANDU 3 probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaitly, R K [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10{sup -6}/year. 4 refs., 1 fig.

  2. Results of the CANDU 3 probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    1995-01-01

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10 -6 /year. 4 refs., 1 fig

  3. Root-cause Investigation for No Setback Initiation at Liquid Zone Control Unit Perturbation in CANDU6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Liquid zone control system (LZCS) is one of the indigenous systems in CANDU type reactor for reactor reactivity control. The LZCS is filled with light water and used to provide a continuous fine control of the reactivity and the reactor power level. This system is also designed to accomplish spatial control of the power distribution, automatically, which prevents xenon induced power oscillations. As the tilt control term is phased out, it is replaced by a level control term, which tends to drive the individual zone levels towards the average level of all the zones. Most of CANDU reactors have been experienced these events. Generally setback or stepback conditions are on when variables of spatial control off, high zone power, etc. are reached to the initiating conditions before ROP trip. But the condition of setback or stepback is not initiated before ROP trip sometime. In this study the root-causes for this event are investigated, and the impact assessment is performed by physics computational modeling. To investigate the root-cause of ROP trip before initiating setback at abnormal operating condition, some LZC perturbation models were simulated and investigated the neutron flux readings of zone detector and ROP detector. Two root-causes were founded. The first, flux variation by water level change is more gradual than other zones due to design characteristics in zone 03. The second, ROP detector (SDS no. 2 3G) in the near zone 03 is very sensitive below 40% of water level due to ROP detector installed position. Even though setback is initiated earlier than ROP trip in case of zone 03 perturbation, ROP trip will be occurred because power decreasing rate is very slow(0.1%/sec) on setback condition.

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

    International Nuclear Information System (INIS)

    Whitlock, J.

    2001-01-01

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

  5. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-01-01

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

  6. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

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

  7. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

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

  8. Safety Parameters for the Recycled Uranium Loaded into a CANDU Reactor

    International Nuclear Information System (INIS)

    Park, Chang Je; Kang, Kweon Ho; Na, Sang Ho; Kim, Young Hee; Ryu, Ho Jin; Park, Geun Il; Song, Kee Chan

    2008-01-01

    In order to recover uranium and TRU from spent nuclear fuels, a pyroprocessing has been developed through a dry and metallurgical reprocess technology using a series of electrolyses such as an electro-reduction, an electro-refining, and an electro-winning. When the spent fuel is being fed into the pyroprocess, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process. It is expected that the recovered uranium will be sent to a spent fuel storage site after converting it into a metal ingot form to reduce its storage space and transportation burden. However, the weight percent of U-235 in the recovered uranium is about 0.9 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economical profit and save of uranium resources but also an alleviation of burden on the management and disposal of the spent fuel. A previous research on recycling of recovered uranium was carried out and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is a sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. And the DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) program has also been performed and demonstrated the fundamental technologies. The recovered uranium from a pyroprocess contains some TRU as an impurity and it will exhibit a slightly different behavior from the previous recycling options. In this paper, the reactor's characteristics including safety parameters are investigated based on the lattice calculations which are performed for the CANFELX bundle

  9. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-02-01

    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

  10. Candu reactors - experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-01-01

    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

  11. The integrity of CANDU fuel during load following

    International Nuclear Information System (INIS)

    Tayal, M.; Manzer, A.M.; Sejnoha, R.; Hains, A.J.

    1989-08-01

    This paper summarizes data and analyses of integrity and of physics of CANDU fuel during load following. Measurements of irradiated fuel show that power cycles do not enhance release of fission gas. Data from research reactors show that the power cycles cause cyclic strains in the sheath. Finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath. The stresses and the strains are well into the plastic range. The cyclic loads 'use up' some fraction of the sheath's resistance to environmentally-assisted cracking (EAC), depending on the details of the fuel design and of then power cycles. The balance of the sheath's resistance to EAC continues to be available to counteract static loads. Thousands of fuel bundles have experienced many power cycles in research and in commercial reactors. Overall integrity of fuel bundles is well over 99%. Thus, CANDU fuel continues to show good performance in both base-load and load-following reactors

  12. 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).

  13. 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)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; Fong, R.W.L.; Coleman, C.E.

    1998-08-01

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

  14. Nuclear reactor building

    International Nuclear Information System (INIS)

    Oshima, Nobuaki.

    1991-01-01

    The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

  15. Study of candu fuel elements irradiated in a nuclear power plant

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

    2015-01-01

    The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of 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) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

  16. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

    2013-01-01

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

  17. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

  18. Integrated aerosol and thermalhydraulics modelling for CANDU safety analysis

    International Nuclear Information System (INIS)

    McDonald, B.H.; Hanna, B.N.

    1990-08-01

    Analysis of postulated accidents in CANDU reactors that could result in severe fuel damage requires the ability to model the formation of aerosols containing fission product materials and the transport of these aerosols from the fuel, through containment, to any leak to the atmosphere. Best-estimate calculations require intimate coupling and simultaneous solution of all the equations describing the entire range of physical and chemical phenomena involved. The prototype CATHENA/PACE-3D has been developed for integrated calculation of thermalhydraulic and aerosol events in a CANDU reactor during postulated accidents. Examples demonstrate the ability of CATHENA/PACE-3D to produce realistic flow and circulation patterns and reasonable accuracy in solution of two simple fluid-flow test cases for which analytical solutions exist

  19. The design and installation of a core discharge monitor for CANDU-type reactors

    International Nuclear Information System (INIS)

    Halbig, J.K.; Monticone, A.C.; Ksiezak, L.; Smiltnieks, V.

    1990-01-01

    A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses (γ,n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs

  20. Natural uranium equivalent fuel. An innovative design for proven CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S. [Candu Energy Inc., Mississauga, Ontario (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Co., Haiyan, Zhejiang (China)

    2015-09-15

    The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU® reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

  1. Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

    International Nuclear Information System (INIS)

    Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee

    2015-01-01

    Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

  2. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  3. Hitachi's proposed DCS solution for new build CANDU EC6 using the G-HIACS unified platform

    Energy Technology Data Exchange (ETDEWEB)

    Tan, D.; Ishii, K.; Otsuka, Y.; Uemura, K., E-mail: daisuke.tan.ye@hitachi.com [Hitachi Ltd., Infrastructure Systems Co., Ibaraki (Japan); Marko, P.E. [Hitachi Power Systems Canada Ltd., Power and Industry Div., Ontario (Canada)

    2013-07-01

    Hitachi Ltd. has developed the safe and secure functional safety DCS controller for potential new build NPP projects in the global market. Hitachi has improved the availability, maintainability, and reliability for its latest DCS systems named G-HIACS. In this latest paper on its DCS product development program, Hitachi would like to report a proposed DCS solution for new build CANDU NSP and BOP based on the G-HIACS Unified Architecture (R800FS/HSC800FS vSAFE Functional Safety Controller and R900/HSC900 General Purpose Controller) hybrid control system. (author)

  4. CANDU fuel cycles - present and future

    International Nuclear Information System (INIS)

    Mooradian, A.J.

    1976-05-01

    The present commercially proven Canadian nuclear power system is based on a once-through natural uranium fuel cycle characterized by high uranium utilization and a high conversion efficiency. The cycle closes with secure retrievable storage of spent fuel. This cycle is based on a CANDU reactor concept which is now well understood. Both active and passive fuel storage options have been investigated and will be described in this paper. Future development of the CANDU system is focussed on conservation of uranium by plutonium and thorium recycle. The full exploitation of these options requires continued emphasis on neutron conservation, efficiency of extraction and fuel refabrication processes. The results of recent studies are discussed in this paper. (author)

  5. Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

    International Nuclear Information System (INIS)

    Crespi, J.C.

    1992-01-01

    This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

  6. Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

    2003-12-01

    Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

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

    International Nuclear Information System (INIS)

    Park, D.H.; Lee, E.K.; Shin, H.C.; Bae, S.M.; Hong, S.Y.

    2013-01-01

    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

  8. Towards more efficient safeguards approach of transfer campaign in Wolsong CANDU reactor

    International Nuclear Information System (INIS)

    Park, S. K.; Na, W. W.; Park, W. S.; Jung, S. T.; Park, S. J.

    2002-01-01

    Due to the unique character of the CANDU reactor, a transfer campaign has been carried out every year for 2 months or so in Wolsong unit 1. It is expected to require performance of the transfer campaign for 4 units of Wolsong site in 2006 and more than 50% of the ROK safeguards inspection efforts. The IAEA and TCNC have gained several years of experience in safeguards approach during transfer campaign. Occasionally the deterrence and interference of operator transfer work have been occurred in order to attain safeguards inspection goal. These could be minimized using optimum C/S applications, NDA and RDT techniques. This paper shows the interrelationship and combinations of these resources and proposes new safeguards approach to maintain the continuity of knowledge from fuel loading in the spent fuel pond to the point of canister loading and closure during transfer campaign

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    International Nuclear Information System (INIS)

    Yilmaz, S.

    2000-01-01

    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. Used fuel packing plant for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

    2015-07-01

    Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

  12. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  13. GENOVA: a generalized perturbation theory program for various applications to CANDU core physics analysis (I)-theory and application

    International Nuclear Information System (INIS)

    Kim, Do Heon; Choi, Hang Bok

    2001-01-01

    A generalized perturbation theory (GPT) program, GENOVA, has been developed for the purpose of various applications to Canadian deuterium uranium (CANDU) reactor physics analyses. GENOVA was written under the framework of CANDU physics design and analysis code, RFSP. A sensitivity method based on the GPT was implemented in GENOVA to estimate various sensitivity coefficients related to the movement of zone controller units (ZCUs) existing in the CANDU reactor. The numerical algorithm for the sensitivity method was verified by a simple 2 x 2 node problem. The capability of predicting ZCU levels upon a refueling perturbation was validated for a CANDU-6 reactor problem. The applicability of GENOVA to the CANDU-6 core physics analysis has been demonstrated with the optimum refueling simulation and the uncertainty analysis problems. For the optimum refueling simulation, an optimum channel selection strategy has been proposed, using the ZCU level predicted by GENOVA. The refueling simulation of a CANDU-6 natural uranium core has shown that the ZCU levels are successfully controlled within the operating range while the channel and bundle powers are satisfying the license limits. An uncertainty analysis has been performed for the fuel composition heterogeneity of a CANDU DUPIC core, using the sensitivity coefficients generated by GENOVA. The results have shown that the uncertainty of the core performance parameter can be reduced appreciably when the contents of the major fissile isotopes are tightly controlled. GENOVA code has been successfully explored to supplement the weak points of the current design and analysis code, such as the incapacity of performing an optimum refueling simulation and uncertainty analysis. The sample calculations have shown that GENOVA has strong potential to be used for CANDU core analysis combined with the current design and analysis code, RFSP, especially for the development of advanced CANDU fuels

  14. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  15. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Choi, H B; Roh, G H; Jeong, C J; Rhee, B W; Choi, J W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

  16. 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)

    1998-12-31

    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)

  17. 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)

  18. Accelerated reduction of used CANDU fuel waste with fast-neutron reactors: fuel cycle strategy cuts TRU waste lifespan from 400,000 years to less than 80 years

    International Nuclear Information System (INIS)

    Ottensmeyer, P.

    2013-01-01

    Canada's 45,000 tonnes of nuclear fuel waste contain over 99% heavy atoms whose nuclear energy can provide $50 trillion of non-carbon electricity in fast-neutron reactors (FNRs), equivalent to 4000 years of nuclear power at present levels. FNRs can utilize the 98.9% uranium in CANDU fuel waste and also the 0.38% transuranic actinides which impose its 400,000-year radiotoxicity. Separation of uranium from CANDU nuclear fuel waste would permit refueling of FNRs primarily with transuranics, hugely accelerating the elimination of the long-term radiotoxicity of the CANDU fuel waste. Practicable separations of uranium would result in the complete elimination of the transuranics in about 80 years using FNRs at current Canadian nuclear energy output, while ideal separations could lower this to 16 years. (author)

  19. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

    1991-12-31

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

  20. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

    1990-01-01

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

  1. Remote handling equipment for CANDU retubing

    International Nuclear Information System (INIS)

    Crawford, G.S.; Lowe, H.

    1993-01-01

    Numet Engineering Ltd. has designed and supplied remote handling equipment for Ontario Hydro's retubing operation of its CANDU reactors at the Bruce Nuclear Generating Station. This equipment consists of ''Retubing Tool Carriers'' an'' Worktables'' which operate remotely or manually at the reactor face. Together they function to transport tooling to and from the reactor face, to position and support tooling during retubing operations, and to deliver and retrieve fuel channels and channel components. This paper presents the fundamentals of the process and discusses the equipment supplied in terms of its design, manufacturing, components and controls, to meet the functional and quality requirements of Ontario Hydro's retubing process. (author)

  2. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

    1995-01-01

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

  3. Seismic analysis during development stage of CANDU Model 2 fueling machine design

    International Nuclear Information System (INIS)

    Lee, L.S.S.; Mansfield, R.A.

    1989-01-01

    The CANDU Model 3 is a new small reactor presently being designed. This reactor is 450 MWe, and as with current operating CANDU's, is based on a heavy water moderated and cooled system using on-power fuelling for the once-through natural uranium fuel cycle. The CANDU 3 Standard plant is designed to be adaptable to a range of world-wide site conditions, i.e. for a peak ground acceleration of 0.3 g and a wide range of soft, medium and hard foundation medium properties. Consequently, a conservatism in the design of structure and equipment is accounted by using enveloped floor response spectra generated by the soil-structure interaction analysis. Seismic qualification of the fuelling machine (F/M) and its support structure are an essential design requirement for maintaining the integrity of the reactor coolant heat transport system (HTS) pressure boundary and the service ports penetrating the containment structure during on-power fueling. This paper deals with the initial conceptual phase of design where the details of the design are in fundamental outline form only and basic mass distribution plus layout geometry is defined

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-04-01

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

  5. The Difference between Flux Spectrums of WH-type Assembly and CANDU-type Lattice

    International Nuclear Information System (INIS)

    Ryu, Eun Hyun; Song, Yong Mann

    2014-01-01

    The nuclear reactors are categorized by the material of the moderator because of its importance. The representative materials of the moderator are light water (H 2 O) and heavy water (D 2 O). Also, it is well known that the slowing-down ratio of D 2 O is hundreds of times larger than that of H 2 O while the slowing-down power of H 2 O is several times larger than that of D 2 O. This means that the H 2 O sometimes plays a role of an absorber such as the liquid zone controller (LZC) in a CANDU-type reactor. It is thought that the flux spectrums in a different reactor can differ from each other. In this research, two representative assemblies (the Westinghouse (WH)-type fuel assembly of PWR and the CANDU-type fuel lattice of PHWR) are selected and the flux results for each group are extracted. Although there are many codes for the lattice transport calculation, the WIMS code and the HELIOS code are used for the calculation of the WH-type fuel lattice and the CANDU-type fuel lattice. A clear difference in spectrum between the CANDU-type lattice and WH 16GD-type lattice is confirmed. Because of the superior moderating ratio of the heavy water, the thermal flux ratio of the CANDU-type lattice is almost 82%, while that of the WH 16 GD-type lattice is around 23%. Because of the large portion of the thermal flux in the CANDU-type lattice, the boron effect is maximized with the result from variations of boron. Thus it can be said that the spectrum largely depends on the moderator material, and the boron effect and sensitivity largely depends on the flux spectrum. Because of the dominant effect of the moderator material on the flux spectrum in a nuclear reactor, in the future, a comparison of the spectra of SFR, HTGR, PWR, and PHWR are also an interesting subject to study. Over-moderation in PHWR lattice and under-moderation in PWR lattice can be explained by the investigation about flux spectrums with variations of moderator density in each lattice

  6. The Difference between Flux Spectrums of WH-type Assembly and CANDU-type Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Eun Hyun; Song, Yong Mann [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The nuclear reactors are categorized by the material of the moderator because of its importance. The representative materials of the moderator are light water (H{sub 2}O) and heavy water (D{sub 2}O). Also, it is well known that the slowing-down ratio of D{sub 2}O is hundreds of times larger than that of H{sub 2}O while the slowing-down power of H{sub 2}O is several times larger than that of D{sub 2}O. This means that the H{sub 2}O sometimes plays a role of an absorber such as the liquid zone controller (LZC) in a CANDU-type reactor. It is thought that the flux spectrums in a different reactor can differ from each other. In this research, two representative assemblies (the Westinghouse (WH)-type fuel assembly of PWR and the CANDU-type fuel lattice of PHWR) are selected and the flux results for each group are extracted. Although there are many codes for the lattice transport calculation, the WIMS code and the HELIOS code are used for the calculation of the WH-type fuel lattice and the CANDU-type fuel lattice. A clear difference in spectrum between the CANDU-type lattice and WH 16GD-type lattice is confirmed. Because of the superior moderating ratio of the heavy water, the thermal flux ratio of the CANDU-type lattice is almost 82%, while that of the WH 16 GD-type lattice is around 23%. Because of the large portion of the thermal flux in the CANDU-type lattice, the boron effect is maximized with the result from variations of boron. Thus it can be said that the spectrum largely depends on the moderator material, and the boron effect and sensitivity largely depends on the flux spectrum. Because of the dominant effect of the moderator material on the flux spectrum in a nuclear reactor, in the future, a comparison of the spectra of SFR, HTGR, PWR, and PHWR are also an interesting subject to study. Over-moderation in PHWR lattice and under-moderation in PWR lattice can be explained by the investigation about flux spectrums with variations of moderator density in each

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

    International Nuclear Information System (INIS)

    Krause, T.W.; Schankula, J.; Sullivan, S.P.

    2002-01-01

    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)

  8. Determination of filling-up time of the dousing system of CANDU-600 reactor

    International Nuclear Information System (INIS)

    Stefanescu, P.; Damian, C.

    1997-01-01

    The dousing system belongs to the envelope system being one of the fourth special safety system provided in the plant design in order to limit the release of the radioactive products to the population in case of serious damages. It is designed to reduce the top pressure and overpressure period within the envelope by taking over by the dousing water of the steam energy as resulted from an accident. Such an accident can be caused either by the loss of the primary cooling agent or the damage of the main steam pipe of the section situated in the envelope. At the same time this system is designed to achieve the function of maintaining sufficient water supply for the meeting of the dousing water necessities in the envelope and for the cooling of the reactor core in case of an accident with the loss of primary cooling agent. The paper has its target the calculation of the fill-up time of the dousing system of the CANDU-600 reactor, of the time period between the setting-up of the system and the moment the farthest nozzle starts dousing, respectively (the system douses at nominal flow within the envelope. (authors)

  9. A CANDU Severe Accident Analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

    2006-01-01

    As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

  10. Seismic retrofitting of Apsara reactor building

    International Nuclear Information System (INIS)

    Reddy, G.R.; Parulekar, Y.M.; Sharma, A.; Rao, K.N.; Narasimhan, Rajiv; Srinivas, K.; Basha, S.M.; Thomas, V.S.; Soma Kumar, K.

    2006-01-01

    Seismic analysis of Apsara Reactor building was carried out and was found not meeting the current seismic requirements. Due to the building not qualifying for seismic loads, a retrofit scheme using elasto-plastic dampers is proposed. Following activities have been performed in this direction: Carried out detailed seismic analysis of Apsara reactor building structure incorporating proposed seismic retrofit. Demonstrating the capability of the retrofitted structure to with stand the earth quake level for Trombay site as per the current standards by analysis and by model studies. Implementation of seismic retrofit program. This paper presents the details of above aspects related to Seismic analysis and retrofitting of Apsara reactor building. (author)

  11. Coolability of severely degraded CANDU cores

    International Nuclear Information System (INIS)

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

    1995-07-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

  12. Passive heat removal in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R S; Snell, V G [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

    1996-12-01

    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.

  13. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.; Snell, V.G.

    1996-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  15. Drift effects in CANDU reactors

    International Nuclear Information System (INIS)

    Koclas, J.; Roy, R.; Marleau, G.

    1993-01-01

    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

  16. Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

    International Nuclear Information System (INIS)

    Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

    1981-05-01

    The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

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

    International Nuclear Information System (INIS)

    Whitlock, J.

    2000-01-01

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

  18. Assessment of reactivity devices for CANDU-6 with DUPIC fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok

    1998-01-01

    Reactivity device characteristics for a CANDU-6 reactor loaded with DUPIC fuel have been assessed. A transport code WIMS-AECL and a three-dimensional diffusion code RFSP were used for the lattice parameter generation and the core calculation, respectively. Three major reactivity devices have been assessed for their inherent functions. For the zone controller system, damping capability for spatial oscillation was investigated. The restart capability of the adjuster system was investigated. The shim operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster rod system. The mechanical control absorber was assessed for the capability to compensate the temperature reactivity feedback following a power reduction. This study has shown that the current reactivity device systems retain their functions when used in a DUPIC fuel CANDU reactor

  19. CANDU 9 nuclear power plant simulator

    International Nuclear Information System (INIS)

    Kattan, M.; MacBeth, M.J.; Lam, K.

    1995-01-01

    Simulators are playing, an important role in the design and operations of CANDU reactors. They are used to analyze operating procedures under standard and upset conditions. The CANDU 9 nuclear power plant simulator is a low fidelity, near full scope capability simulator. It is designed to play an integral part in the design and verification of the control centre mock-up located in the AECL design office. It will also provide CANDU plant process dynamic data to the plant display system (PDS), distributed control system (DCS) and to the mock-up panel devices. The simulator model employs dynamic mathematical models of the various process and control components that make up a nuclear power plant. It provides the flexibility to add, remove or update user supplied component models. A block oriented process input is provided with the simulator. Individual blocks which represent independent algorithms of the model are linked together to generate the required overall plant model. As a design tool the simulator will be used for control strategy development, human factors studies (information access, readability, graphical display design, operability), analysis of overall plant control performance, tuning estimates for major control loops and commissioning strategy development. As a design evaluation tool, the simulator will be used to perform routine and non-routine procedures, practice 'what if' scenarios for operational strategy development, practice malfunction recovery procedures and verify human factors activities. This paper will describe the CANDU 9 plant simulator and demonstrate its implementation and proposed utility as a tool in the control system and control centre design of a CANDU 9 nuclear power plant. (author). 2 figs

  20. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  1. 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).

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

    International Nuclear Information System (INIS)

    Yoon, Moon Young; Kwon, O Hwan; Kim, Kyung Hwa; Yeom, Choong Sub

    2004-01-01

    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

  3. Development of an automated system for CANDU secondary coolant circuit chemistry control

    International Nuclear Information System (INIS)

    Dean, J.R.; Stewart, R.B.

    1978-04-01

    This report is a summary of work done to develop a means for automated control of the secondary coolant chemistry of CANDU 600 MW(e) power reactors using on-line analyzers and a minicomputer. The development work was carried out in cooperation with Saskatchewan Power Corporation at Estevan. Results and conclusions of the program are included, as are recommendations for a prototype installation in a domestic CANDU 600 MW steam generator. (author)

  4. Mechanical development for reliable reactor components

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Metcalfe, R.

    1983-09-01

    The CANDU reactor has achieved worldwide distinction because of its reliable performance. To achieve this, special attention was given to the reliability and maintainability of components in the heavy water circuits. Development programs were initiated early in the history of the CANDU reactor to improve the effectiveness of pump seals, valves, and static seals because of unacceptable performance of the commercial equipment then available. As a result, pump seals with a five year life now appear achievable, and valves and static seals are no longer a significant concern in CANDU reactors. Increasing effort is being given remotely operated tools and fabrication systems for radioactive environments

  5. CANDU fuel behaviour under transient conditions

    International Nuclear Information System (INIS)

    Segel, A.W.L.

    1979-04-01

    The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO 2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

  6. Primary side crud sampling and characterization: how it may help to better maintain CANDU reactors

    International Nuclear Information System (INIS)

    Sawicki, J.A.; Dymarski, M.; Sawicka, B.D.

    1997-01-01

    The paper reports a corrosion-product transport study in the primary-side of CANDU reactors. The study is based on systematic crud sampling examinations at Darlington NGS. Emphasis is placed on the corrosion-product transport and oxidation state as a monitor of primary water chemistry state, and as a monitor of system corrosion effects. The D 2 O was sampled at reactor outlet headers and the deposits collected on 0.45 μm membrane filters were analyzed by using γ-ray spectroscopy, X-ray fluorescence and Moessbauer effect techniques. In the beginning of this study in 1994 and 1995, the steady state crude concentration used to be as high as 20 to 30 μg/kg D 2 O and the crud contained large fractions of highly oxidized species (40 to 60% Fe), in the form of ferrihydrite, lepidocrocite, goethite and hematite. During start-up operations, the crud concentrations were as high as 8 mg/kg D 2 O and contained 70 to 80% Fe in the form of oxidized iron species. However, as a result of various improvements, and especially because of better oxygen control during lay-ups (nitrogen blanketing) and better pH control (lower and more stable values), in recent steady-state runs, the crud concentrations were found to be lower than 3 to 5 μg/kg D 2 O. At present, the steady-state crud consists mostly of Fe, largely in the form of magnetite (60 to 80% Fe), and hematite (20 to 40% Fe). The steady-state concentration of crud was found to increase markedly with pH a in the range from 10.3 to 10.7. This seems to be in accord with the trend predicted by flow-assisted solubility of magnetite in lithiated water at ∼ 300 degrees C, and is generally in accord with the model of dissolution, transport, and redeposition of iron in the primary heat transport system of CANDU reactors. The crud contains also significant amounts of Zr (∼ 5 to 30 wt. % Fe), apparently in the form of zirconium oxide. Zirconium oxide particles may originate largely from the wear of fuel bundle pads and pressure tube

  7. Three dimensional numerical simulation of a full scale CANDU reactor moderator to study temperature fluctuations

    International Nuclear Information System (INIS)

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2014-01-01

    Highlights: • 3D model of a Candu reactor is modeled to investigate flow distribution. • The results show the temperature distribution is not symmetrical. • Temperature contours show the hot regions at the top left-hand side of the tank. • Interactions of momentum flows and buoyancy flows create circulation zones. • The results indicate that the moderator tank operates in the buoyancy driven mode. -- Abstract: Three dimensional numerical simulations are conducted on a full scale CANDU Moderator and transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows in the center region of the tank, results in the formation of circulation zones. The moderator tank operates in the buoyancy driven mode and any small disturbances in the flow or temperature makes the system unstable and asymmetric. Different types of temperature fluctuations are noted inside the tank: (i) large amplitude are at the boundaries between the hot and cold; (ii) low amplitude are in the core of the tank; (iii) high frequency fluctuations are in the regions with high velocities and (iv) low frequency fluctuations are in the regions with lower velocities

  8. Next Generation CANDU: Conceptual Design for a Short Construction Schedule

    International Nuclear Information System (INIS)

    Hopwood, Jerry M.; Love, Ian J.W.; Elgohary, Medhat; Fairclough, Neville

    2002-01-01

    Atomic Energy of Canada Ltd. (AECL) has very successful experience in implementing new construction methods at the Qinshan (Phase III) twin unit CANDU 6 plant in China. This paper examines the construction method that must be implemented during the conceptual design phase of a project if short construction schedules are to be met. A project schedule of 48 months has been developed for the nth unit of NG (Next Generation) CANDU with a 42 month construction period from 1. Concrete to In-Service. An overall construction strategy has been developed involving paralleling project activities that are normally conducted in series. Many parts of the plant will be fabricated as modules and be installed using heavy lift cranes. The Reactor Building (RB), being on the critical path, has been the focus of considerable assessment, looking at alternative ways of applying the construction strategy to this building. A construction method has been chosen which will result in excess of 80% of internal work being completed as modules or as very streamlined traditional construction. This method is being further evaluated as the detailed layout proceeds. Other areas of the plant have been integrated into the schedule and new construction methods are being applied to these so that further modularization and even greater paralleling of activities will be achieved. It is concluded that the optimized construction method is a requirement, which must be implemented through all phases of design to make a 42 month construction schedule a reality. If the construction methods are appropriately chosen, the schedule reductions achieved will make nuclear more competitive. (authors)

  9. Assessment of core characteristics during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Suk, Ho Chun

    2002-01-01

    A transition from 37-element natural uranium fuel to CANFLEX-NU fuel has been modeled in a 1200-day time-dependent fuel management simulation for a CANDU 6 reactor. The simulation was divided into three parts. The pre-transition period extended from 0 to 300 FPD, in which the reactor was fuelled only with standard 37-element fuel bundles. In the transition period, refueling took place only with the CANFLEX-NU fuel bundle. The transition stage lasted from 300 to 920 FPD, at which point all of the 37-element fuel in the core had been replaced by CANFLEX-NU fuel bundle. In the post-transition phase, refueling continued with CANFLEX-NU fuel until 1200 FPD, to arrive at estimate of the equilibrium core characteristics with CANFLEX-NU fuel. Simulation results show that the CANFLEX-NU fuel bundle has a operational compatibility with the CANDU 6 reactor during the transition core, and also show that the transition core from 37-element natural uranium fuel to CANFLEX-NU can be operated without violating any license limit of the CANDU 6 reactor

  10. A plan for the modification and assessment of TRAC-PF1/MOD2 for use in analyzing CANDU 3 transient thermal-hydraulic phenomena

    International Nuclear Information System (INIS)

    Siebe, D.A.; Boyack, B.E.; Giguere, P.T.

    1994-11-01

    This report presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-hydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PF1/MOD2