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Sample records for candu power reactor

  1. 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.)

  2. 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

  3. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

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

    2003-01-01

    therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

  4. 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

  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. 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

  7. 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

  8. 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)

  9. 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

  10. 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 ...

  11. 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)

  12. 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)

  13. 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.)

  14. 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

  15. 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)

  16. 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.

  17. 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

  18. 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)

  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. 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

  1. 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)

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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)

  9. The formation, composition and structure of corrosion products in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Rummery, T.E.

    1978-01-01

    To gain a better understanding of the formation and transport of corrosion products in CANDU-PHW power reactors, and the role played by these products in the generation and subsequent fixation of radioactive species, we have examined in detail several surfaces removed from the Douglas Point Generating Station (Douglas Point, Ontario). Results are given for the surface of the primary-side of a Monel-400 boiler tube, and surfaces of carbon steel piping at the inlet and outlet of the boiler. The experimental techniques that were used included sequential acid stripping, X-ray diffractometry, scanning electron microscopy and energy dispersive X-ray spectrometry. The corrosion products on the Monel-400 were mainly nickel, copper, nickel oxide and nickel-deficient nickel ferrite and varied in composition and quantity as a function of both distance from the boiler inlet, and depth in the corrosion layer. The radioactive cobalt ( 60 Co) content was localized in 'streaks' deposited in the straight sections of the boiler tube, but distributed uniformly over the whole surface in the downstream bend section. The material covering the carbon steel surface comprised three phases: magnetite, aluminosilicate particles at the outermost surface, and a mixed cation spinel phase uniformly distributed over the surface at the corrosion film-water interface. The formation, composition and structure of the corrosion products are discussed. (author)

  10. 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

  11. 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)

  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. 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.)

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. 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.)

  19. 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

  20. 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

  1. 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)

  2. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. 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)

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

    International Nuclear Information System (INIS)

    2001-02-01

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

  9. 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

  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. 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)

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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.

  6. 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)

  7. 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)

  8. Dynamic behaviour of CANDU reactor

    International Nuclear Information System (INIS)

    Subramanian, M.G.; Srikantiah, G.; Pai, M.A.

    1976-01-01

    Understanding of the dynamic behaviour of a reactor system in a power station is essential for evolving control stragies as well as design modifications. The dynamic behaviour of Rajasthan Atomic Power Station is studied. Mathematical models for the reactor, the steam generator and the steam drum with the natural circulation loop are developed from physical principles like conservation of mass, momentum and energy. Each of these models is then simulated on a digital computer to obtain the characteristics during transients. The models are then combined to yield a dynamic mathematical model of the system comprising the reactor, the steam generator and the steam drum and this results in a nonlinear model. Using this model, responses of the system for various disturbances like step change in the area of the steam valve, step change in the temperature of feed water are obtained and are discussed. These models could be used to devise new control laws using optimal control theory or to evaluate the performance of existing control schemes. (author)

  9. Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

    International Nuclear Information System (INIS)

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

    2014-01-01

    The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

  10. Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent 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

    The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

  11. 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)

  12. 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

  13. 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

  14. Peak power and heavy water production from electrolytic H2 and O2 using CANDU reactors

    International Nuclear Information System (INIS)

    Hammerli, M.; Stevens, W.H.; Bradley, W.J.; Butler, J.P.

    1976-04-01

    A combined energy storage - heavy water production system is presented. Off-peak nuclear energy is stored in the form of electrolytic H 2 (and O 2 ) from which a large fraction of the deuterium has been transferred to water in an H 2 /H 2 O deuterium exchange catalytic column. The main features and advantages of the combined electrolysis -catalytic exchange D 2 O process are discussed. Significant quantities of D 2 O could be produced economically at reasonable peak to base power cost ratios. Thirty to forty percent of the primary electric energy should be available for peak energy via either gas-steam turbines or fuel cells. (author)

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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.

  1. 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)

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

    International Nuclear Information System (INIS)

    Lawrence, C.B.; Hall, D.S.

    1982-05-01

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

  3. 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

  4. Moderator circulation in CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Hussein, M.A.

    1989-01-01

    A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

  5. 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)

  6. 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)

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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)

  15. 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)

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. 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

  2. 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.

  3. 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)

  4. 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.

  5. 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

  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. 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

  8. 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

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. 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)

  14. 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.)

  15. 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)

  16. 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)

  17. 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.

  18. Nuclear power reactor safety

    International Nuclear Information System (INIS)

    Pon, G.A.

    1976-10-01

    This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. 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)

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. 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)

  13. 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

  14. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.

    1997-01-01

    CANDU calandria tubes are made from annealed Zircaloy-2 sheet formed into a cylinder and welded along its length to make the tube. The current calandria tubes have given exemplary service for many years. With more stringent regulations and the need to accommodate warm cooling water in tropical countries, we started a development program to increase the margins for failure during postulated accidents. These improvements involve increasing the tube strength and optimising the heat-transfer from an excessively hot fuel channel to the cool moderator. If the postulated accident involves a pressure tube break, it would be desirable if the calandria tube withstood the full pressure of the heat-transport system. The weakest link in current calandria tubes is the weld. Thickening the weld can increase the strength by 20% while seamless tubes can be 45% stronger than current tubes. The latter tubes can hold full system pressure for many hours without failure. If during the postulated accident the fuel and pressure tube become excessively hot but do not touch the calandria tube, the radiant heat loss must be maximised. Current calandria tubes have an absorptivity (emissivity) of about 0.2. To protect the fuel and the fuel channel we have devised a finish to the inside surface of the calandria tube that increases the emissivity to 0.7. If during the postulated accident the hot pressure tube touches the cool calandria tube, the contact conductance and the critical heat flux must be optimised to ensure nucleate boiling of the moderator at the outside surface of the calandria tube and therefore efficient exploitation of the moderator as a heat sink. In laboratory tests small ridges on the inside surface and roughening of the outside surface have been shown to increase the margins against failure and increase the possible moderator temperatures thus providing the opportunity to decrease the cost of the moderator heat-exchange system and remove restrictions on reactor operation in

  15. 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

  16. 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

  17. 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)

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

    International Nuclear Information System (INIS)

    Coutsiers, Eduardo E.; Pomerantz, Marcelo E.; Moreno, Carlos A.

    2000-01-01

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

  19. 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

  20. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. Dimensional response of CANDU fuel to power changes

    Energy Technology Data Exchange (ETDEWEB)

    Fehrenbach, P J [Fuel Engineering Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Hastings, I J; Morel, P A; Sage, R D; Smith, A D [Fuel Materials Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    1983-06-01

    The introduction of CANLUB-coated fuel cladding, modified fuel management schemes, and revisions to the sequence of control rod movements, have eliminated power ramping fuel failures in CANDU power reactors. However, an irradiation program continues at Chalk River Nuclear Laboratories to determine the effect of various design and operating parameters on the dimensional response of UO{sub 2} fuel elements to power changes, over a range of conditions outside those normally experienced by CANDU power reactor fuel. We have investigated the effect of power changes on element diameter for UO{sub 2} fuel with starting densities of 10.6 and 10.8 Mg/m{sup 3} clad in 0.4 mm thick Zircaloy, at burnups from 0 to 100 MW.h/kg U. Element diameter measurements were obtained at power using an In-Reactor Diameter Measuring Rig (IRDMR). Rates of power change over the range 0.0005 to 0.03 kW.m{sup -1}.s{sup -1} were achieved by a combination of reactor power control and use of a Helium-3 power cycling facility. Total diameter increases in unirradiated elements were about 1% at pellet interface locations for both fuel densities during the initial power increase to 60 kW/m. Diameter changes during subsequent power cycles of these elements from 55 to 100% maximum power were significantly larger for the higher density fuel, ranging from 0.3 to 0.5% compared to less than 0.1% for the standard density (10.6 Mg/m{sup 3}) fuel. In elements pre-irradiated at 27 kW/m to burnups of about 100 MW.h/kg U prior to power ramping, the diameter increases measured after ramping to 55 kW/m also varied with starting fuel density. Diameter changes at pellet interface locations were about 0.9% and 0.6% for higher density and standard density fuel respectively. (author)

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

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

  12. Qinshan Phase III (CANDU) nuclear power project quality assurance

    International Nuclear Information System (INIS)

    Wang Lingen; Du Jinxiang

    2001-01-01

    The completion and implementation of quality assurance system of Qinshan Phase III (CANDU) nuclear power project are presented. Some comments and understanding with consideration of the project characteristics are put forward

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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)

  18. 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

  19. 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

  20. Advanced CANDU Design With Negative Power Feedback

    International Nuclear Information System (INIS)

    Andang-Widi-Harto; Muslim

    2004-01-01

    The problem of positive power feedback in the recent PHWR-CANDU design, especially related to coolant void increase, will be overcame by the use of dual moderator concept, in which two moderator systems are used, i.e. a main moderator outside the calandria tube and an annular moderator inside the annular space. Annular moderator is allowed to boil in the case of overheating. The numerical calculations have been performed for two core design namely HWR-DM-ST and HWR-DM-XI which can reach burn up of 16,000 and 17,500 MWd/ ton U respectively. The results for the two designs is that the values of k at fully annular moderator filling condition are 1.0054 (HWR-DM-ST) and 1.0019 (HWR-DM-XI), while at completely empty annular moderator condition are 0.9634 (HWR-DM-ST) and 0.9143 (HWR-DM-XI). The decrease of coolant flow rate from 3,043 kg/s to 853 kg/s decrease k values of 0.0109 (HWR-DM-ST) and 0.0232 (HWR-DM-XI). While increasing inlet coolant enthalpy from 2,950 kJ/kg to 3,175 kJ/kg decreases of k values of 0.0074 (HWR-DM-ST) and 0.0239 (HWR-DM-XI). Thus, it can be summarized that the HWR-DM design has negative power reactivity feedback.(author)

  1. An analysis of cobalt irradiation in CANDU 6 reactor core

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Dumitrache, I.

    2003-01-01

    In CANDU reactors, one has the ability to replace the stainless steel adjuster rods with neutronically equivalent Co assemblies with a minimum impact on the power plant safety and efficiency. The 60 Co produced by 59 Co irradiation is used extensively in medicine and industry. The paper mainly describes some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronically equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and the heating of the irradiated cobalt rods are performed using the Monte Carlo codes MCNP5 and MONTEBURNS2.1. The 60 Co activity and heating evaluations are closely related to the neutronics computations and to the density evolution of cobalt isotopes during assumed in-core irradiation period. Unfortunately, the activities of these isotopes could not be evaluated directly using the burn-up capabilities of the MONTEBURNS code because of the lack of their neutron cross-section from the MCNP5 code library. Additional MCNP5 runs for all the cobalt assemblies have been done in order to compute the flux-spectrum, the 59 Co and the 60 Co radiative capture reaction rates in the adjusters. The 60m Co cross-section was estimated using the flux-spectrum and the ORIGEN2.1 code capabilities THERM and RES. These computational steps allowed the evaluation of the one-group cross-section for the radiative capture reactions of cobalt isotopes. The values obtained replaced the corresponding ones from the ORIGEN library, which have been estimated using the flux-spectrum specific to the fuel. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

  2. 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

  3. 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)

  4. 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.

  5. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    Green, R.E.; Boczar, P.G.

    1990-04-01

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

  6. Stochastic maintenance optimization at Candu power plants

    International Nuclear Information System (INIS)

    Doyle, E.K.; Duchesne, T.; Lee, C.G.; Cho, D.I.

    2004-01-01

    The use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems as previously reported at ICONE 6 in New Orleans (1996). Further refinement of the station maintenance strategy was evaluated via the applicability of statistical analysis of historical failure data. The viability of stochastic methods in Candu maintenance was illustrated at ICONE 10 in Washington DC (2002). The next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principal impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring 'baby boom generation'. Initial indications have been quite positive. The current reality of global competition necessitates the pursuit of all financial optimizers. The next construction phase in the power generation industry will soon begin on a worldwide basis. With the relatively high initial capital cost of new nuclear generation all possible avenues of financial optimization must be evaluated and implemented. (authors)

  7. 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

  8. 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

  9. 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)

  10. 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)

  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. 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

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. Occupational radiation exposures at Canadian CANDU nuclear power stations

    International Nuclear Information System (INIS)

    LeSurf, J.E.; Taylor, G.F.

    1982-09-01

    In Canada, methods to reduce the radiation exposure to workers at nuclear power reactors have been studied and implemented since the early days of the CANDU reactor program. Close collaboration between the designers, the operators, and the manufacturers has reduced the total exposure at each station, the dose requirement to operate and maintain each successive station compared with earlier stations, and the average annual exposure per worker. Specific methods developed to achieve dose reduction include water chemistry; corrosion resistant materials; low cobalt materials; decontamination; hot filtration, improved equipment reliability, maintainability, and accessibility; improved shielding design and location; planning of work for low exposure; improved operating and maintenance procedures; removal of tritium from D 2 O systems and work environments; improved protective clothing; on-power refuelling; worker awareness and training; and many other small improvements. The 1981 occupational dose productivity factors for Pickering A and Bruce A nuclear generating stations were respectively 0.43 and 0.2 rem/MW(e).a

  18. Reactor physics assessment of modified 37-element CANDU fuel bundles

    International Nuclear Information System (INIS)

    Pristavu, R.; Rizoiu, A.

    2016-01-01

    Reducing the central element diameter in order to improve the total flow area of CANDU fuel bundle and redistribute the power density of all remaining elements was studied in Canada and Korea when considering the effect of aging pressure tube diametral creep. The aim of this paper is to study the modified bundle behavior using the transport codes WIMS and DRAGON. In calculations, a WIMS nuclear data library on 172 energy groups was used. 2-D transport calculations were performed with WIMS and DRAGON, leading to similar results in estimated cell parameters. Additionally, 3-D DRAGON calculations were carried on in order to evaluate the local flux distribution shift, as well as the incremental cross sections for supercells containing modified CANDU bundles and reactivity devices. The overall effect of using modified fuel bundles was meaningless for both cell and supercell parameters, thus ensuring this possibility of fuel improvement for thermal-hydraulic purposes only. (authors)

  19. 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)

  20. 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

  1. 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.

  2. 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)

  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. Conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    Beamer, N.V.; Bourns, W.T.; Buckley, L.P.; Speranzini, R.A.

    1981-12-01

    A Waste Treatment Centre (WTC) is being constructed at the Chalk River Nuclear Laboratories to develop and demonstrate processes for converting reactor wastes to a form suitable for disposal. The WTC contains a starved air incinerator for reducing the volume of combustible solid wastes, a reverse osmosis section for reducing the volume of liquid wastes and an immobilization section for incorporating the conditioned wastes in bitumen. The incinerator is commissioned on inactive waste: approximately 16.5 Mg of waste packaged in polyethylene bags has been incinerated in 17 burns. Average weight and volume reductions of 8.4:1 and 32:1, respectively, have been achieved. Construction of the reverse osmosis section of WTC is complete and inactive commissioning will begin in 1982 January. The reverse osmosis section was designed to process 30,000 m 3 /a of dilute radioactive waste. The incinerator ash and concentrated aqueous waste will be immobiblized in bitumen using a horizontal mixer and wiped-film evaporator. Results obtained during inactive commissioning of the incinerator are described along with recent results of laboratory programs directed at demonstrating the reverse osmosis and bituminization processes

  5. 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)

  6. 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

  7. 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

  8. 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.

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

    International Nuclear Information System (INIS)

    Daverio, Hernando J.

    2003-01-01

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

  10. Transmutation of actinides in power reactors.

    Science.gov (United States)

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

    2005-01-01

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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  16. A short history of the CANDU nuclear power system

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, G L

    1993-04-01

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

  17. A short history of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1993-04-01

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

  18. 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)

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. Types and properties of elastomer materials used in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of)

    1996-05-01

    Properties and kinds of elastomer materials used in a CANDU power plant have been described. The elastomer materials have been used as a sealing material in the components f nuclear power plant since they have many excellent properties that can not be seen in other materials. It is very important to select proper elastomer materials used in the nuclear power plant are required to have resistance to temperature as well as radiation. According to the experimental results performed at some laboratories including the Chalk River Laboratory of AECL, elastomer materials with high resistance to temperature and radiation are Nitrile, Ethylene, Propylene and Butyl. These materials have been used in a lot of components of Wolsong unit 1 and Wolsong 2, 3 and 4 which are under elastomer material. Therefore, the studies on the standardization are currently under way to limit about 10 different kinds of elastomer materials to be used in the plant. 16 tabs., 1 fig., 12 refs. (Author) .new.

  5. 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

  6. 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.

  7. 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

  8. Some novel on-power refuelling features of CANDU stations

    International Nuclear Information System (INIS)

    Erwin, D.; Pendlebury, B.; Watson, J.F.; Welch, A.C.

    1976-01-01

    Part A of the paper describes the reasons for, and advantages resulting from, the use of flow assisted refuelling in the CANDU type nuclear reactors at the Pickering Generating Station. A separate fuel handling system is used for each reactor unit, as distinct from the system employed at the Bruce Generating station, where the fuel handling system is shared among several units. Part B of the paper describes some of the advantages of the shared concept with particular emphasis on the availability of the fuel handling system. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  10. 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

  11. 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

  12. Safety of CANDU nuclear power stations

    International Nuclear Information System (INIS)

    Snell, V.G.

    1978-11-01

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

  13. Assessment of the thorium fuel cycle in power reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  14. Analysis of power variation in a CANDU-6 with a loss of moderator

    International Nuclear Information System (INIS)

    Fan, Y.

    2008-01-01

    A loss of heavy water in a postulated small failure in the horizontal unpressurized calandria vessel of a CANDU-6 reactor will lead to a drop in the moderator level in the reactor core. The STEPBACK and SETBACK functions at the initial moment of the drop in moderator level ensure a reactor shutdown and a reduction in total reactor power during this 900 seconds postulated transient. If the STEPBACK and SETBACK functions are unavailable, the reactor's regulating system will try to compensate for the negative reactivity resulting from the loss of the moderator. This kind of compensation will lead to power distortions from top to bottom in the reactor core. .Comparisons of different moderator leakage rates were used in the analysis to determine the relationships between the power and the moderator leakage rates. Maximum bundle and channel powers obtained were insensitive to the moderator leakage rate. .In a complete analysis for a moderator leakage rate of 40 1/s, it was found that, without the STEPBACK and SETBACK functions, serious power distortions would occur during the 900 seconds transient. The maximization of bundle and channel power during this transient happened in the bottom part of the reactor , and the regulating system worsened this power distortion. .From the above analysis, it was concluded that the maximum bundle power attained during the loss of the moderator was 1.18% of its initial value. The risk of bundle dryout was, therefore, quite small. (author)

  15. 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)

  16. 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

  17. 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)

  18. 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)

  19. Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

    International Nuclear Information System (INIS)

    Kim, Yonghee

    2015-01-01

    LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

  20. Prediction of power-ramp defects in CANDU fuel

    International Nuclear Information System (INIS)

    Gillespie, P.; Wadsworth, S.; Daniels, T.

    2010-01-01

    Power ramps result in fuel pellet expansion and can lead to fuel sheath failures by fission product induced stress corrosion cracking (SCC). Historically, empirical models fit to experimental test data were used to predict the onset of power-ramp failures in CANDU fuel. In 1988, a power-ramped fuel defect event at PNGS-1 led to the refinement of these empirical models. This defect event has recently been re-analyzed and the empirical model updated. The empirical model is supported by a physically based model which can be used to extrapolate to fuel conditions (density, burnup) outside of the 1988 data set. (author)

  1. 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)

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. 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.

  7. Candu

    International Nuclear Information System (INIS)

    Melvin, J.G.

    1983-01-01

    Electricity consumption in Canada over the past twenty years correlates well with the real gross national product (GNP). No effect of electricity conservation measures has yet been discernible because, unlike oil, electricity has never been used very wastefully. A 4% annual expansion of real GNP implies a 5% extrapolated annual growth in electricity demand. To sell reactors, the nuclear industry must promote the use of electricity. Canadian industry generally, even the nuclear industry, has been slow to use electricity for process heating. The advantages include stability of price, good control, and absence of pollution

  8. Strategic provisioning of replacement parts for CANDU power plants

    International Nuclear Information System (INIS)

    Mizuno, G.; Tume, P.; Prentice, J.

    2000-01-01

    Provisioning of replacement parts and management of critical spares are key factors in optimizing maintenance programs for CANDU power plants. With a view to supply assurance, Atomic Energy Canada Limited (AECL) has created a Spare Parts Branch (SPB) to provide a clear pipeline from the client to the delivered replacement part(s). SPB provides the client with assured access to a qualified supplier database, computer aided design, engineering and manufacturing services and material upgrades and design registration through the authorized inspection agency. The AECL spare parts strategic provisioning service plan that has four thrusts: 1) the efficient delivery of cost-effective replacement parts; 2) obsolete parts resolution; 3) a website that will provide our clients with real-time access to replacement part data; and 4) inventory recovery opportunities. Thrusts one and two are actively ensuring plant maintenance for on-shore and off-shore CANDU clients. Thrusts three and four are longer-term commitments. This paper will explore these thrusts in the context of our CANDU business practices. (author)

  9. 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.

  10. 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

  11. 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%

  12. 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

  13. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  14. 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)

  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. Low Power Shutdown PSA for CANDU Type Plants

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yeon Kyoung; Kim, Myung Su [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    KHNP also have concentrated on full power PSA. Some recently constructed OPR1000 type plants and APR1400 type plants have performed the low power and shutdown (LPSD) PSA. The purpose of LPSD PSA is to identify the main contributors on the accident sequences of core damage and to find the measure of safety improvement. After the Fukushima accident, Korean regulatory agency required the shutdown severe accident management guidelines (SSAMG) development for safety enhancement. For the reliability of SSAMG, KHNP should develop the LPSD PSA. Especially, the LPSD PSA for CANDU type plant had developed for the first time in Korea. This paper illustrates how the LPSD PSA for CANDU type developed and the core damage frequency (CDF) is different with that of full power PSA. KHNP performed LPSD PSA to develop the SSAMG after the Fukushima accidents. The results show that risk at the specific operation mode during outage is higher than that of full power operation. Also, the results indicated that recovery failure of class 4 power at the POS 5A, 5B contribute dominantly to the total CDF from importances analysis. LPSD PSA results such as CDF with initiating events and POSs, risk results with plant damage state, and containment failure probability and frequency with POSs can be used by inputs for developing the SSAMG.

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

    International Nuclear Information System (INIS)

    Singh, N.; Duff, C.G.

    1979-10-01

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

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

    International Nuclear Information System (INIS)

    Rouben, Benjamin

    2008-01-01

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

  19. An intelligent safety system concept for future CANDU reactors

    International Nuclear Information System (INIS)

    Hinds, H.W.

    1980-01-01

    A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

  20. Evaluation of Required Water Sources during Extended Loss of All AC Power for CANDU NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Woo Jae; Lee, Kyung Jin; Kim, Min Ki; Kim, Keon Yeop; Park, Da Hee; Oh, Seo Bin [FNC Technology Co., Yongin (Korea, Republic of); Chang, Young Jin; Byun, Choong Seop [KHNP, Daejeon (Korea, Republic of)

    2016-10-15

    Fukushima accident was caused by lasting long hours of Station Black-Out (SBO) triggered from natural disaster. This accident had resulted in the reactor core damage. The purpose of this study is to evaluate the required water sources to maintain hot standby conditions until 72 hours during ELAP situation. The analysis was performed with CATHENA code. CATHENA code has been developed for the best-estimated transient simulation of CANDU plants. This study was carried out to evaluate the strategy to maintain hot standby conditions during ELAP situation in CANDU reactors. In this analysis, water was supplied to SG by MSSV open and by the gravity feed. It can cool the core without damage until the dousing tank depletion. Before dousing tank depletion, the emergency water supply pump was available by emergency power restoration. The pump continuously fed water to SG. So it is expected that the reactor core can be cooled down without damage for 72 hours if water source is enough to feed. This result is useful to make a strategy against SBO including ELAP situation.

  1. Experience of oil in CANDU moderator during A831 planned outage at Bruce Power

    International Nuclear Information System (INIS)

    Ma, G.; Nashiem, R.; Matheson, S.; Stuart, C.; Roberts, J.G.

    2011-01-01

    In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

  2. Experience of oil in CANDU moderator during A831 planned outage at Bruce Power

    Energy Technology Data Exchange (ETDEWEB)

    Ma, G.; Nashiem, R.; Matheson, S. [Bruce Power, Tiverton, Ontario (Canada); Stuart, C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Roberts, J.G. [CANTECH Associates Ltd., Burlington, Ontario (Canada)

    2011-03-15

    In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

  3. 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)

  4. 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)

  5. 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

  6. Development of techniques for radwaste systems in CANDU power stations

    International Nuclear Information System (INIS)

    Bourns, W.T.; Buckley, L.P.; Burrill, K.A.

    1979-04-01

    Techniques to reduce the volume of CANDU reactor wastes and to bitumenize them are being developed at Chalk River Nuclear Laboratories. Reverse osmosis is suitable for initial purification of dilute radioactive aqueous wastes. Tubular membranes are used to concentrate wastes to 5 weight percent solids, and while the membranes do foul, they may be cleaned mechanically, chemically, or with fresh feed. A wiped-film evaporator then concentrates the retentate to a 20 weight-percent slurry. A twin-screw extruder-evaporator has been used to bitumenize this slurry, and it will also handle ion exchange resin and dry incinerator ash. Work on a wiped-film evaporator as a bitumenizer for various feeds is in progress. More experience in handling solid feeds is needed before work can proceed to the demonstraton phase. (auth)

  7. 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.

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

    International Nuclear Information System (INIS)

    Khartabil, H.F.

    2000-01-01

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

  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. 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)

  11. 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

  12. 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)

  13. 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)

  14. 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)

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

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; S. A. Usmani

    1996-03-01

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

  16. Value addition initiatives for CANDU reactor operation performance

    International Nuclear Information System (INIS)

    Chugh, V.; Parmar, R.; Schut, J.; Sherin, J.; Xie, H.; Zobin, D.

    2013-01-01

    Recently, AMEC NSS initiated projects for CANDU® station performance engineering with potentially high returns for the utilities. This paper discusses three initiatives. Firstly, optimization of instrument calibration interval from 1 to 3 years will reduce time commitments on the maintenance resources on top of financial savings ~$3,500 per instrument. Secondly, reactor thermal power uncertainty assessment shows the level of operation which is believed to have an over-conservative margin that can be used to increase power by up to 0.75%. Finally, as an alternative means for controlling Reactor Inlet Header Temperature (RIHT), physical modifications to the High Pressure (HP) feedwater heaters can be useful for partially recovering RIHT resulting in increased production by 10-12 MWe. (author)

  17. 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

  18. Integrated control centre concepts for CANDU power plants

    International Nuclear Information System (INIS)

    Lupton, L.R.; Davey, E.C.; Lapointe, P.A.; Shah, R.R.

    1990-01-01

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

  19. Integrated control centre concepts for CANDU power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lupton, L. R.; Davey, E. C.; Lapointe, P. A.; Shah, R. R.

    1990-01-15

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

  20. 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).

  1. 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

  2. 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

  3. 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)

  4. 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

  5. Calculation of the negative reactivity inserted by the shutdown system number two (SDS2) of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Arsenault, B [Ecole Polytechnique, Montreal, PQ (Canada)

    1994-12-31

    The secondary shutdown system (SDS2) of a CANDU reactor consists of liquid poison injection through nozzles disposed horizontally across the core. The nominal concentration of gadolinium nitrate poison is 8000 ppm. With the methods available to the nuclear industry for calculating the negative reactivity inserted by the SDS2, some approximations are needed, and a simplified model of poison propagation has to be used to calculate the differential cross sections. The objective of this paper is to evaluate the errors introduced by the approximations in the supercell and core calculations. The MULTICELL and EXCELL codes gave different power distributions, and further work was recommended. 9 refs., 2 tabs., 4 figs.

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

    International Nuclear Information System (INIS)

    Wood, J.C.

    1979-01-01

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

  7. 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)

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

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

    2005-03-01

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

  9. 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

  10. Improvement of Candu-1000 MW(e) power cycle by moderator heat recovery

    International Nuclear Information System (INIS)

    Fath, H.E.S.

    1988-01-01

    Four different moderator heat recovery circuits are proposed for CANDU-1000 MW(e) reactors. The proposed circuits utilize all, or part, of the 155 MW(th) moderator heat load (at 70 0 C moderator outlet temperature from calandria) to the first stage of the feed water heating system. An economics study was carried out and indicated that the direct circulation of feed water through the moderator heat exchanger (with full heat recovery) is the most economical scheme. For this scheme the saved steam from the turbine extraction was found to produce additional electric power of 8 MW(e). This additional power represents a 0.7% increase in the plants nominal electric output. The outstanding features and advantages of the selected scheme are also presented. (author)

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Experience with digital instrumentation and control systems for CANDU power plant modifications

    International Nuclear Information System (INIS)

    Basu, S.

    1997-01-01

    Over the last years, Ontario Hydro CANDU power plants have gone through many modifications. This includes modification from analog hardwired controls to digital and solid state controls and replacement of the existing digital controls with the latest hardware and software technology. Examples of digital modifications at Bruce A and other CANDU power plants are briefly described and categorized. Most of the I and C technology development has been supported by the CANDU Owners Group (COG) a consortium of Canadian nuclear utilities and the Atomic Energy Canada Limited (AECL). (author)

  18. Experience with digital instrumentation and control systems for CANDU power plant modifications

    Energy Technology Data Exchange (ETDEWEB)

    Basu, S [Ontario Hydro, Toronto, ON (Canada)

    1997-07-01

    Over the last years, Ontario Hydro CANDU power plants have gone through many modifications. This includes modification from analog hardwired controls to digital and solid state controls and replacement of the existing digital controls with the latest hardware and software technology. Examples of digital modifications at Bruce A and other CANDU power plants are briefly described and categorized. Most of the I and C technology development has been supported by the CANDU Owners Group (COG) a consortium of Canadian nuclear utilities and the Atomic Energy Canada Limited (AECL). (author).

  19. 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.

  20. General requirements for pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1991-11-01

    This standard specifies the general requirements for the design, fabrication and installation of pressure-retaining systems, components, and their supports in CANDU nuclear power plants. (16 figs., 2 tabs., 25 refs.)

  1. Heat transport inventory monitoring for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Hussein, E.; Luxat, J.C.

    1984-01-01

    A computer-based D 2 O coolant inventory monitoring system proposed for implementation on the digital computer controllers at Ontario Hydro's CANDU generating units is discussed. By monitoring process parameters and utilizing probabilistically-based decision algorithms, timely indication of any significant loss of D 2 O inventory will be provided to the operator. The monitoring is performed in a co-ordinated manner such that D 2 O losses from either the heat transport system or the inventory control system can be detected. (orig.)

  2. 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)

  3. 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.

  4. Econometric modelling of certain nuclear power systems based on thermal and fast breeder reactors

    International Nuclear Information System (INIS)

    Pavelescu, M.; Pioaru, C.; Ursu, I.

    1988-01-01

    Certain known economic analysis models for a LMFBR fast breeder and CANDU thermal solitary reactors are presented, based on the concepts of discounting and levelization. These models are subsequently utilized as a basis for establishing an original model for the econometric analysis of certain thermal reactor systems or/and fast breeder reactors. Case studies are subsequently conducted with the systems: 1-CANDU, 2-LMFBR, 3-CANDU + LMFBR which enables us to draw certain interesting conclusions for a long range nuclear power policy. (author)

  5. 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)

  6. The next generation of CANDU: reactor design to meet future energy markets

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Love, J.W.; Wren, D.J.

    2001-01-01

    Nuclear power plant designs for the future must respond to increasingly demanding market requirements. This means that value can be gained from substantial product development directed at these requirements. For the CANDU system, AECL has adopted the revolutionary approach, accommodating significant changes to design while retaining traditional CANDU strengths. The focus of the new design is to achieve a 40% reduction in capital cost, quicken construction time and higher efficiency. Key aspects of the new design include: light water coolant, smaller core, slightly enriched fuel, higher temperature and pressure coolant. Work is well advanced on the preliminary design

  7. 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.

  8. 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

  9. The control of emissions from nuclear power reactors in Canada

    International Nuclear Information System (INIS)

    Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

    1988-01-01

    Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

  10. 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)

  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. 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)

  13. 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)

  14. 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.

  15. 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

  16. Long-term performance of the CANDU-type of vanadium self-powered neutron detectors in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: leungt@aecl.ca

    2007-07-01

    The CANDU-type of in-core vanadium self-powered neutron flux detectors have been installed in NRU to monitor the axial neutron flux distributions adjacent to the loop fuel test sites since 1996. This paper describes how the thermal neutron fluxes were measured at two monitoring sites, and presents a method of correcting the vanadium burn-up effect, which can be up to 2 to 3% per year, depending on the detector locations in the reactor. It also presents the results of measurements from neutron flux detectors that have operated for over eight-years in NRU. There is good agreement between the measured and simulated neutron fluxes, to within {+-} 6.5%, and the long-term performance of the CANDU-type of vanadium neutron flux detectors in NRU is satisfactory. (author)

  17. 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)

  18. 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

  19. 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

  20. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  1. 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)

  2. 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)

  3. 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)

  4. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  5. 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

  6. Management of radioactive wastes at power reactor sites in India

    International Nuclear Information System (INIS)

    Amalraj, R.V.; Balu, K.

    Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

  7. Study on models for gap conductance between fuel and sheath for CANDU reactors

    International Nuclear Information System (INIS)

    Lee, K.M.; Ohn, M.Y.; Lim, H.S.; Choi, J.H.; Hwang, S.T.

    1995-01-01

    The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the two recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope. (author)

  8. 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)

  9. 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)

  10. 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)

  11. Reconstruction of intra-bundle fission density profile during a postulated LOCA in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, D. [Oak Ridge National Laboratory (United States); Rahnema, F. [Georgia Inst. of Technology (United States); Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Inst. of Technology, Atlanta, GA 30332-0405 (United States); Serghiuta, D. [Canadian Nuclear Safety Commission (Canada); Sarsour, H.; Turinsky, P. J. [North Carolina State Univ. (United States); Stamm' ler, R. [Studsvik Scandpower AS (Norway)

    2006-07-01

    In this paper, results related to the reconstruction of intra-bundle fission density profile for a 37-pin CANDU-6 bundle with the highest enthalpy deposition during a postulated large LOCA stagnation break in a Bruce B core are presented. Bruce B is a nuclear power plant in Kincardine, Ontario (Canada)), on the shores of Lake Huron with 4 CANDU reactors that are rated at about 750 MWe. The reconstruction of the fuel pin fission densities is based on steady-state, three-dimensional simulations with the Monte Carlo code MCNP for a subset of 27 out of 69 time steps during the first two seconds of the power pulse predicted for the fuel bundle at core location V13/8. Two-group cross section data libraries are generated for MCNP at each time step by the lattice depletion neutron transport code HELIOS-1.7. To include the effect of the surrounding core environment, the calculations are performed with time-dependent albedo boundary conditions inferred from a full core simulation of the transient by the nodal diffusion code NESTLE with HELIOS homogenized cross-sections. It is found that the local peaking factor (LPF) in the outer ring varies during the transient, but never exceeds its value before the transient. Inclusion of the core environment increases the LPF in the outer ring. For the analyzed case, the increase is 0.72% with a relative error of 0.01% for the LPF before the transient and 0.55% (with a relative error of 0.01%) for the maximum average LPF during the transient. The latter is based on only four selected transient time points. Note that the immediate environment of the 'hot bundle' does not contain any reactivity devices or other perturbing factors. As a result, the increases observed in the LPF in the outer ring may not be representative of the situations in which 'other' core environment perturbing factors are present. To determine the effect of these factors on the LPF, further analyses of a bundle in the proximity of control devices

  12. 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

  13. 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

  14. 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.

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

    International Nuclear Information System (INIS)

    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

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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

  3. 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

  4. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  5. 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)

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

    International Nuclear Information System (INIS)

    1990-08-01

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

  7. 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)

  8. 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.

  9. 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)

  10. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

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

    1996-01-01

    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/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-line fuelling provide flexibility to respond to changing fuel-cycle requirements in the short term and in the indefinite future

  11. Aspects regarding the lifetime of a fuel channel in a CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Calinescu, A.

    1998-01-01

    The paper presents the analysis of factors influencing upon the time life of a fuel channel of CANDU reactors built in Romania. Fuel channels are made of Zr-2.5%Nb alloy. Means and methodology to detect cracking of fuel channels are described, as well as improvements to increase life time of Cernavoda NPP fuel channels and national programme in this area. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    International Nuclear Information System (INIS)

    Im, Sunghyuk; Sung, Hyung Jin; Seo, Han; Bang, In Cheol; Kim, Hyoung Tae

    2015-01-01

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

  14. 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.)

  15. 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

  16. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  17. 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)

  18. 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)

  19. 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

  20. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

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

  1. 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

  2. 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)

  3. The basic concepts of a fuel-power detector for nuclear power reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.

    1979-01-01

    Fuel power is proposed as an alternative to neutron or gamma-ray flux for control and safety functions in CANDU power reactors. To satisfy in-core power monitoring requirements, a detector whose dynamic response corresponds to the heat production rate in the fuel is needed. This report explores the concept of tailoring the response characteristics of a mixed-response self-powered flux detector to match the requirements of an ideal fuel-power detector. (author)

  4. 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)

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  7. 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

  8. 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)

  9. 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)

  10. Comparison of Ontario Hydro's performance with world power reactors - 1981

    International Nuclear Information System (INIS)

    Dumka, B.R.

    1982-04-01

    The performance of Ontario Hydro's CANDU reactors in 1981 is compared with that of 123 world nuclear power reactors rated at 500 MW(e) or greater. The report is based on data extracted from publications, as well as correspondence with a number of utilities. The basis used is the gross capacity factor, which is defined as gross unit generation divided by the perfect gross output for the period of interest. The lowest of the published turbine and generator design ratings is used to determine the perfect gross output, unless the unit has been proven capable of consistently exceeding this value. The first six reactors in the rankings were CANDU reactors operated by Ontario Hydro

  11. 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

  12. Maintenance of ageing CANDU reactors. A regulatory perspective

    International Nuclear Information System (INIS)

    Dunstan, T.

    1996-01-01

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

  13. 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)

  14. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  15. 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

  16. Tricon hardware controller implementation of CANDU nuclear power plant shutdown system

    International Nuclear Information System (INIS)

    Zahedi, P.

    2007-01-01

    This paper introduces the implementation of logic functions associated with the shutdown systems of CANDU nuclear power plants. The experimental aspects of this work include development of control program embedded in shutdown systems of CANDU based NPPs. A physical test environment is designed to simulate the measurements of in-core flux detector (ICFD) and ion chamber (I/C) signals. The programmable logic used in this experimentation provides Triple Modular Redundant (TMR) architecture as well as a voting mechanism used upon execution of control program on each independent channel. (author)

  17. 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)

  18. Instrumentation and control systems for CANDU-PHW nuclear power plants

    International Nuclear Information System (INIS)

    Lepp, R.M.; Watkins, L.M.

    1982-02-01

    The instrumentation and control of CANDU nuclear power plants takes advantage of modern electronics technology in the extensive computerization of important control and man-machine functions. A description of these functions as well as those of the four Special Safety Systems is provided

  19. General requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This standard provides the general requirements used in the design, construction, testing, and commissioning of concrete containment structures for CANDU nuclear power plants designated as class containment and is directed to the owners, designers, manufacturers, fabricators, and constructors of the concrete components and parts

  20. 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)

  1. 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

  2. Evolution of on-power fuelling machines on Canadian natural uranium power reactors

    International Nuclear Information System (INIS)

    Isaac, P.

    1984-10-01

    The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

  3. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

  4. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

  5. Reactor power control device

    International Nuclear Information System (INIS)

    Doi, Kazuyori.

    1981-01-01

    Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

  6. 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

  7. 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

  8. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Hoizumi, Atsushi.

    1986-01-01

    Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

  9. 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

  10. 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

  11. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  12. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  13. 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

  14. 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

  15. 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

  16. 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)

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

    International Nuclear Information System (INIS)

    Rouben, B.; Chow, H.C.; Leung, L.K.H.; Inch, W.; Fink, J.; Moreno, C.

    2004-01-01

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

  18. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  19. 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.)

  20. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  1. Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Jae Yong; Na, Man Gyun

    2011-01-01

    Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the epistemic

  2. Cernavoda nuclear power plant: Modifications in the fire protection measures of the CANDU 6 standard design

    International Nuclear Information System (INIS)

    Covalschi, V.

    1998-01-01

    Having as purpose the improvement of fire safety at the Cernavoda NPP - both in the prevention and the protection aspects in the case of fire - we implemented some modifications in the CANDU 6 standard design. These improvements are inspired, mainly, from two sources: the world-wide achievements in the field of fire protection techniques, introduced in nuclear power plants since the middle of 70's, when the CANDU 6 design was completed; the national practice and experience in fire protection, usually applied in industrial objectives (conventional power plants, in particular). The absence of any incident may be considered as a proof of the efficiency of the implemented fire preventing and protection measures. (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. Initiating stochastic maintenance optimization at Candu Power Plants

    International Nuclear Information System (INIS)

    Doyle, E.K.

    2003-01-01

    As previously reported at ICONE 6 in New Orleans (1996), the use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems. Further cost refinement of the station maintenance strategy is being evaluated via the applicability of statistical analysis of historical failure data. Since the statistical evaluation was initiated in 1999 significant progress has been made in demonstrating the viability of stochastic methods in Candu maintenance. Some of the relevant results were presented at ICONE 10 in Washington DC (2002). Success with the graphical displays and the relatively easy to implement stochastic computer programs was sufficient to move the program along to the next significant phase. This next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principle impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring 'baby boom generation'. Initial indications have been quite positive. (author)

  5. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

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

    International Nuclear Information System (INIS)

    Kastanya, Doddy

    2012-01-01

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

  7. 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)

  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. 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

  10. Compact power reactor

    International Nuclear Information System (INIS)

    Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

  11. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  12. 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)

  13. Power reactor design trends

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1985-01-01

    Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  17. 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

  18. Expert systems use in present and future CANDU nuclear power supply systems

    International Nuclear Information System (INIS)

    Lupton, L.R.; Basso, R.A.J.; Anderson, L.L.; Anderson, J.W.D.

    1989-11-01

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

  19. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

    Directory of Open Access Journals (Sweden)

    Simon Younan

    2018-01-01

    Full Text Available The objective of this study was to evaluate accident-tolerant fuel (ATF concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo and fully ceramic microencapsulated (FCM fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

  20. 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)

  1. 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)

  2. Fuel performance in aging CANDU reactors - a quick overview of the CNSC regulatory oversight activities of the past 15 years and of the lessons learned

    International Nuclear Information System (INIS)

    Couture, M.

    2013-01-01

    'Full text:' The operating conditions (coolant flows, temperatures, pressures) of the Heat Transport System (HTS) of a CANDU reactor are affected by the aging of it's components. For a given fuel bundle design, those changing conditions result in the lower dryout powers, and in the absence of any corrective actions addressing the root cause of those changes, the decrease will continue as the aging of the HTS components progresses. As a result of this situation, safety margins for several relatively high frequency Design Based Accidents (DBAs) in Deterministic Safety Analysis will also be decreasing as a function of time. Eventually, defence-in-depth will be compromised if no corrective actions are taken, and ultimately reactor deratings (reactor operating less than 100% full power) will be required in order to ensure, for those postulated DBAs, that shutdown system effectiveness at protecting the integrity of physical barriers to the release of radioactive materials is maintained at all times. Depending on its size and duration, the economic impact of deratings on licensees could be significant. The situation described above, as well as means to address it, has been the heart of numerous discussions and licensing activities between the CNSC and the industry for more than 15 years now. During that period, licensees developed HTS aging management strategies aimed at delaying as long as possible the need to derate strategies which led to many developments including new fuel designs with better heat transfer properties, new methodologies to calculate safety margins in deterministic safety analysis including the use of less conservative CHF correlations, and to the proposal by an expert panel, after a review of the experimental data on CANDU behaviour in post dryout conditions, of a new set of less conservative derived, acceptance criteria that could be in principle be used to assess, for certain DBAs, safety margins in aging CANDU reactors. All this

  3. Fuel performance in aging CANDU reactors - a quick overview of the CNSC regulatory oversight activities of the past 15 years and of the lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Couture, M. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2013-07-01

    'Full text:' The operating conditions (coolant flows, temperatures, pressures) of the Heat Transport System (HTS) of a CANDU reactor are affected by the aging of it's components. For a given fuel bundle design, those changing conditions result in the lower dryout powers, and in the absence of any corrective actions addressing the root cause of those changes, the decrease will continue as the aging of the HTS components progresses. As a result of this situation, safety margins for several relatively high frequency Design Based Accidents (DBAs) in Deterministic Safety Analysis will also be decreasing as a function of time. Eventually, defence-in-depth will be compromised if no corrective actions are taken, and ultimately reactor deratings (reactor operating less than 100% full power) will be required in order to ensure, for those postulated DBAs, that shutdown system effectiveness at protecting the integrity of physical barriers to the release of radioactive materials is maintained at all times. Depending on its size and duration, the economic impact of deratings on licensees could be significant. The situation described above, as well as means to address it, has been the heart of numerous discussions and licensing activities between the CNSC and the industry for more than 15 years now. During that period, licensees developed HTS aging management strategies aimed at delaying as long as possible the need to derate strategies which led to many developments including new fuel designs with better heat transfer properties, new methodologies to calculate safety margins in deterministic safety analysis including the use of less conservative CHF correlations, and to the proposal by an expert panel, after a review of the experimental data on CANDU behaviour in post dryout conditions, of a new set of less conservative derived, acceptance criteria that could be in principle be used to assess, for certain DBAs, safety margins in aging CANDU reactors. All this

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

    International Nuclear Information System (INIS)

    Rapeanu, S.N.; Bujor, A.; Comsa, O.

    1998-01-01

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

  5. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  6. 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.)

  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. 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)

  9. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

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

    International Nuclear Information System (INIS)

    Krasznai, J.P.; Vaughan, B.R.; Williamson, A.S.

    1990-01-01

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

  11. 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

  12. Technical description of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    Husseini, S.D.

    1977-01-01

    The lecture consists of: 1.) KANUPP description: general - Containment- Reactor Cooling System - Moderator Helium System. 2.) Operating experience: Plant Performances (as applicable to Kanupp) - Major failures of primary system - Performance of primary circulating pumps. - Heavy Water Leakage Control. - Radiation Dose Control - Inadvertant Operations. (orig.) [de

  13. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

  14. 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)

  15. Change in CANDU-6 reactivity following a power reduction at low PHT purity

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Soulard, M.R.; Baudouin, A.

    1995-01-01

    The reactivity effect of a power reduction in CANDU-6 is examined using a three-dimensional, steady-state, coupled neutronics/thermalhydraulics methodology, starting from a global irradiation distribution matched to site data. The power reduction is sufficient to suppress coolant boiling in the fuel channels, and thus the significant parameters affecting reactivity are an increase in coolant density and a decrease in fuel temperature. These individual components are estimated using infinite-lattice-cell methodology. The effect of using newer methodology, particularly for the thermalhydraulic analysis, is examined by comparison with previous simulations. (author). 10 refs., 7 tabs., 1 fig

  16. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  17. 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)

  18. CANDU fuel deposits and chemistry optimizations. Recent regulatory experience in Canadian Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kameswaran, Ram

    2014-01-01

    Water chemistry of the Primary Heat Transport System (PHT) of CANDU – Pressurised Heavy Water Reactors profoundly influences the transport of corrosion products around the Heat Transport System (HTS), where they can be deposited as crud on steam generators, feeder pipes and on the fuel. Fuel cladding can be covered with deposits which have precipitated from the coolant as a result of temperature changes or non-optimal coolant pH. Precipitation of deposits in-core must be avoided as far as possible, as it leads to fouling of the fuel, loss of heat transfer efficiency, and increased radiation fields. In the recent years a Canadian NPP experienced increased instances of black deposits being observed on fuel bundles discharged from one of the units. The black deposits were initially observed in 2008 during in-bay fuel inspections. Since then it has been determined that all the discharged fuel bundles have black deposits on them and that observed deposits have been increasing in size (thickness and surface area). This negative trend has persisted through to 2012, when one of fuel bundles was observed with significantly larger deposit than previously seen. Initial analysis of the deposit indicated it to be iron oxide (magnetite). Flow Accelerated Corrosion (FAC) of carbon steel feeder pipes is the primary source of iron, which deposits as magnetite on HTS surfaces. The black deposits have predominantly been located immediately downstream of the bearing pads of the fuel bundle. Deposits have also tended to form on the bottom-downstream quadrant of the fuel bundles. The deposits were most prevalent in low power channels, but some deposits have been observed on high power channels. It was reported by the utility that the PHT system chemistry has been maintained in specification for most of the time during normal operation but the chemistry control during outages was inadequate. Due to design constraints, purification circuit was not available during outages and ion

  19. A report of the overall working group of the AEC Committee on Development of Advanced Power Reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The AEC Committee on Development of Advanced Power Reactors was set up in April, 1978, following on the previous AEC Special Committee on Development of Advanced Power Reactors, in order to study on the complementary power reactors between current LWRs and future FBRs. The subjects of study by the overall working group are the status of advanced power reactors in views of the nuclear fuel cycle, the impacts on industries, the selection of reactor types under present international circumstances, and the evaluation of advanced power reactors in their technology and economy. The following matters are described: evaluations in view of the nuclear fuel cycle, i.e. the features of the ATR of Japan and CANDU reactors of Canada; international problems concerning nuclear nonproliferation and securing of uranium; problems in the diversification of power reactor types concerning the expenditure and technology; problems of technology in the ATR of Japan, CANDU reactors of Canada and Pu utilization for LWRs; and the economy of D 2 O power reactors, i.e. the ATR of Japan and CANDU reactors of Canada. (J.P.N.)

  20. 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

  1. Pressure drop variation as a function of axial and radial power distribution in CANDU fuel channel with standard and CANFLEX 43 bundles

    International Nuclear Information System (INIS)

    Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift

  2. A review of critical heat flux prediction technique and its application in CANDU reactor

    International Nuclear Information System (INIS)

    Park, Jee Won; Roh, Gyu Hong

    1997-09-01

    The CHF prediction method being used for CANDU reactor have been critically reviewed. The AECL's CHF prediction totally depends on the look-up table which has been developed from many CHF databank. These databanks include not only the water-cooled bundle-CHF data but also the freon-cooled bundle-CHF data. The CHF look-up tables have been developed by smoothing and interpolating (with some extrapolations) the experimental data to construct a practically useful CHF table. Therefore, the table look-up method has advantages of accuracy, consistency in a wide range of thermal-hydraulic parameters. It seems, however, that since the existing look-up table is constructed by many steps of modification of the original experimental data (e.g., the look-up table is constructed not only using the horizontal flow data but also the vertical flow data), one should be very careful when one try to generate a look-up table for other fuel geometries. In other words, a reliable look-up table can be constructed by performing experiments for new fuel geometry. Finally, it should be noted that the modifications to the original experimental data has simple form with many modification parameters for taking into account of different geometrical effects. This report presents the backbone and the validity of AECL CHF look-up table. (author). 22 refs., 2 tabs., 2 figs

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

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Osborne, R.V.

    1979-07-01

    Tritium is produced in CANDU heavy water reactors mainly by neutron activation of deuterium. The typical production rate is 2.4 kCi per megawatt-year (89 TBq. per megawatt-year. In Pickering Generating Station the average concentration of tritium in the moderators has reached 16 Ci.kg -1 (0.6 TBq.kg -1 ) and in coolants, 0.5 Ci.kg -1 (0.02 TBq.kg -1 ). Concentrations will continue to increase towards an equilibrium determined by the production rate, the tritium decay rate and heavy water replacement. Tritium removal methods that are being considered for a pilot plant design are catalytic exchange of DTO with D 2 and electrolysis of D 2 O/DTO to provide feed for cryogenic distillation of D 2 /DT/T 2 . Storage methods for the removed tritium - as elemental gas, as metal hydrides and in cements - are also being investigated. Transport of tritiated wastes should not be a particularly difficult problem in light of extensive experience in transporting tritiated heavy water. Methods for determining the presence of tritium in the environment of any tritium handling facility are well established and have the capability of measuring concentrations of tritium down to current ambient values. (author)

  4. 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

  5. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Kono, Shigehiro.

    1990-01-01

    Among a plurality of power monitoring programs in a reactor power monitoring device, rapid response is required for a scram judging program for the power judging processing of scram signals. Therefore, the scram judging program is stored independently from other power monitoring programs, applied with a priority order, and executed in parallel with other programs, to output scram signals when the detected data exceeds a predetermined value. As a result, the capacity required for the scram judging program is reduced and the processing can be conducted in a short period of time. In addition, since high priority is applied to the scram judging program which is divided into a small capacity, it is executed at higher frequency than other programs when they are executed in parallel. That is, since the entire processings for the power monitoring program are repeated in a short cycle, the response speed of the scram signals required for high responsivity can be increased. (N.H.)

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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

  14. Analysis of Liquid Zone Control Valve Oscillation Problem in CANDU Reactors

    Directory of Open Access Journals (Sweden)

    Elnara Nasimi

    2013-01-01

    Full Text Available This paper looks at the existing challenges with steady-state Liquid Zone control at some CANDU (CANada Deuterium Uranium stations, where—contrary to expectations for equilibrium flow—Liquid Zone Control Valve oscillations have proven to be a chronic, unanticipated challenge. Currently, the exact causes of this behaviour are not fully understood, although it is confirmed that the Control Valve oscillations are not due to automatic power adjustment requests or zone level changes due to process leaks. This phenomenon was analysed based on a case study of one domestic nuclear power station to determine whether it could be attributed to inherent controller properties. Next, a proposal is made in an attempt to improve current performance with minimal changes to the existing system hardware and logic using conventional technologies. Finally, a proposal was made to consider Model Predictive Control-based technology to minimize the undesirable Control Valve oscillations at steady state based on the obtained simulation results and discussion of other available alternatives.

  15. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude [fr

  16. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

  17. New approach to derive linear power/burnup history input for CANDU fuel codes

    International Nuclear Information System (INIS)

    Lac Tang, T.; Richards, M.; Parent, G.

    2003-01-01

    The fuel element linear power / burnup history is a required input for the ELESTRES code in order to simulate CANDU fuel behavior during normal operating conditions and also to provide input for the accident analysis codes ELOCA and SOURCE. The purpose of this paper is to present a new approach to derive 'true', or at least more realistic linear power / burnup histories. Such an approach can be used to recreate any typical bundle power history if only a single pair of instantaneous values of bundle power and burnup, together with the position in the channel, are known. The histories obtained could be useful to perform more realistic simulations for safety analyses for cases where the reference (overpower) history is not appropriate. (author)

  18. 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)

  19. Multivariable controller for a 600 MWe CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Mensah, S.

    1982-11-01

    The problems of designing a multivariable regulator for a nuclear power station of the Gentilly-2 type are studied. A reduced model, G2LDM, linearized around steady state operating conditions, is derived from the non-linear model G2SIM. The resulting linear model is described by state-space equations. Good agreement is demonstrated between the transient responses of both models. Properties of G2LDM are assessed by performing controllability and observability tests, cyclicity and rank tests, and eigenanalysis. A comprehensive set of application-orinented algorithms which allow multivariable controller design with closed-loop pole-assignment techniques are implemented in a computer-aided design package via several modules. A general scheme for the implementation of a multivariable controller in G2SIM is designed, and simulation tests show satisfactory performance of the controller [fr

  20. 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)

  1. 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)

  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. 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. 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

  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. Use of 'tail' as spent fuel dilution factor of Angra-1 (PWR) for use in the Embalse (Candu) reactor

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Maiorino, Jose Rubens

    1995-01-01

    This work purposes a process to use the tail of isotopic enrichment as a factor of dilution (blending) for the burned fuel of Angra-I reactor (PWR) for final utilization in the Embalse (Candu). It was made use of the same technic in previous works that used natural uranium. For this purpose, it was made a tail parametrization inside of the traditional limits of enrichment (between 0.2 and 0.3%). The study showed that the tail utilization represents great savings for the uranium supplies and environment and economic advantages. (author). 8 refs, 4 figs, 11 tabs

  7. Availability increase evaluation of a CANDU-600 reactor based on the 2 out of 4 shutdown logic

    International Nuclear Information System (INIS)

    Stefanescu, P.; Stancu-Ciolac, O.

    1995-01-01

    Quality, reliability and maintenance are the three directly decisive factors for the availability of nuclear process and safety systems of a Nuclear Power Plant. Since the reliability of nuclear equipment and components based on the efforts performed for perfecting them can rapidly reach a 'saturation' point, the only way to improve a system availability is to find those possibilities to optimize its structure so that to strongly minimize its unavailability to work. Reliability analysis prove that very good results have been obtained by replacing the simple reduplicate schemes (1 out of 2; 2 out of 3) with more sophisticate once (2 out of 4; 2 x 2 out of 3). The paper reveals the advantages gained by a CANDU-600 reactor if the Shut Down System Number 1 is based on the 2 out of 4 logic instead of 2 out of 3. The investigation's framework is a new 2 out of 4 shutdown scheme, entailing only relay changes and aiming for identical design requirements and purposes as the initial one. The calculation use the classical logical block diagrams, reliability factors and equations and demonstrate the advantage of the proposed logic by computing and comparing the availability factors for 2 out of 4 and 2 out of 3 logic. The efficiency of the method is established by estimating in comparison with the initial 2 out of 3 logic the implicit investment owed to the additional 4 release and 1 measurement channel. The determined increase of the availability factor (5.86·10 -6 year/years) and the subsequent rise of investment (8.6 millions lei) sustain the proposed method. (Author) 2 Tabs

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-15

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

  10. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  11. 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)

  12. Specific aspects in the manufacturing and operating of CANDU reactor pressure tubes (P/T)

    International Nuclear Information System (INIS)

    Muscaloiu, C.

    1997-01-01

    The CANDU reactor design is based on a number of individual P/T in which nuclear fuel bundles are located. P/T are required to be operated in an environment of elevated temperature (300 o C), internal pressure (10 Mpa), fast neutron flux (E>1 MeV) and heavy water. The most suitable material which can provide the desired neutron economy and still maintain its mechanical properties along with corrosion resistance is zirconium alloys Zr+ 2.5 % Nb with the following composition: niobium, 2.5 to 2.8 weight percent; oxygen, 1,000 to 1,300 ppm; zirconium + allowed impurities - balance. A total of 380 pressure tubes are installed into reactor. Each pressure tube is attached at each end to a stainless steel end fitting by means of a grooved, expanded joint. The installation works were performed by ANM Bucuresti, under the technical support of General Electric Canada. The integrity of P/T after installation was examined as follows: - the surface of the rolled area on unrolled internal surface extending 25 mm beyond rolled area was inspected for irregularities by means of a boroscope; - all pressure tubes were subjected to the helium leak test after F/C installation. During P/T operating life periodical inspections according to Canadian Standard CSA N285.4 are performed. The selection of the P/T for inspection is based either on particular properties or on the operating conditions of the fuel channel. The inspection consists in: a) Base Line Inspection within 2 years period commencing after 7,000 EFPH of operation which will include a volumetric inspection over P/T full length and measurements of P/T sag, ID, wall thickness and F/C bearing positions; b) Periodic Inspection in the same conditions plus material surveillance (on the four most significant indication P/T detected during the Base Line Inspection). The inspection will be performed on 14 selected P/T. (author)

  13. Nuclear reactor power supply

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector is interposed between the protection system and the control system. This selector prevents a parameter signal of a set of signals, which differs from the other parameters signals of the set by more than twice the allowable variation of the sensors which produce the set, from passing to the control system. The selectors include a pair of signal selection units, one unit sending selected process signals to primary control channels and the other sending selected process signals to back-up control channels. Test signals are periodically impressed by a test unit on a selected pair of a selected unit and control channels. When test signals are so impressed the selected control channel is disabled from transmitting control signals to the reactor and/or its associated components. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test

  14. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    Notley, M.J.F.

    1983-07-01

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  15. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

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

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

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

  17. Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

    International Nuclear Information System (INIS)

    Daverio, H; Juanico, L

    2000-01-01

    Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

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

    International Nuclear Information System (INIS)

    1998-08-01

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

  19. 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.

  20. 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

  1. Power reactors in member states

    International Nuclear Information System (INIS)

    1975-01-01

    This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

  2. 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

  3. 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

  4. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  5. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  6. Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core

    International Nuclear Information System (INIS)

    Gugiu, D.; Dumitrache, I.

    2005-01-01

    The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60 Co produced by 59 Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

  7. Transient signal analysis in power reactors by means of the wavelet technique

    International Nuclear Information System (INIS)

    Wentzeis, Luis

    1999-01-01

    The application of the wavelet technique, had enabled to study the time evolution of the properties (amplitude and frequency content) of a signals set, measured in the Embalse nuclear power plant (CANDU 600 M we), in the low frequency range and for different operating conditions. Particularly, by means of this technique, we studied the time evolution of the signals in the non-stationary state of the reactor (during a raise in power), where the Fourier analysis results inadequate. (author)

  8. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  9. Tendencies in operating power reactors

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1987-01-01

    A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

  10. 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)

  11. 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)

  12. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  13. 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

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

    International Nuclear Information System (INIS)

    Walker, J.R.

    1990-02-01

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

  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. 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

  17. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

  18. Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

    International Nuclear Information System (INIS)

    Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

    2003-01-01

    One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

  19. 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

  20. 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 sub