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Sample records for aecl

  1. AECL passive autocatalytic recombiners

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, L.B.; Marcinkowska, K. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-03-15

    Atomic Energy of Canada Limited's (AECL) Passive Autocatalytic Recombiner (PAR) is a passive device used for hydrogen mitigation under post-accident conditions in nuclear reactor containment. The PAR employs a proprietary AECL catalyst which promotes the exothermal reaction between hydrogen and oxygen to form water vapour. The heat of reaction combined with the PAR geometry establishes a convective flow through the recombiner, where ambient hydrogen-rich gas enters the PAR inlet and hot, humid, hydrogen-depleted gas exits the outlet. AECL's PAR has been extensively qualified for CANDU and light water reactors (LWRs), and has been supplied to France, Finland, Ukraine, South Korea and is currently being deployed in Canadian nuclear power plants. (author)

  2. AECL passive autocatalytic recombiners

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, L.B.; Marcinkowska, K. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    Atomic Energy of Canada Limited's (AECL) Passive Autocatalytic Recombiner (PAR) is a passive device used for hydrogen mitigation under post-accident conditions in nuclear reactor containment. The PAR employs a proprietary AECL catalyst which promotes the exothermal reaction between hydrogen and oxygen to form water vapour. The heat of reaction combined with the PAR geometry establishes a convective flow through the recombiner, where ambient hydrogen-rich gas enters the PAR inlet and hot, humid, hydrogen-depleted gas exits the outlet. AECL's PAR has been extensively qualified for CANDU and light water reactors (LWRs), and has been supplied to France, Finland, Ukraine, South Korea and is currently being deployed in Canadian nuclear power plants. (author)

  3. The AECL operator companion

    International Nuclear Information System (INIS)

    As CANDU plants become more complex, and are operated under tighter constraints and for longer periods between outages, plant operations staff will have to absorb more information to correctly and rapidly respond to upsets. A development program is underway at AECL 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

  4. AECL's reliability and maintainability program

    International Nuclear Information System (INIS)

    AECL's reliability and maintainability program for nuclear generating stations is described. How the various resources of the company are organized to design and construct stations that operate reliably and safely is shown. Reliability and maintainability includes not only special mathematically oriented techniques, but also the technical skills and organizational abilities of the company. (author)

  5. Annual report 1997--1998. AECL research number AECL-11964

    International Nuclear Information System (INIS)

    This is the Annual report of AECL, the legal name of Atomic Energy of Canada Limited. Its mandate is to undertake research into nuclear energy and to develop commercial applications for its developments. This annual report presents information on marketing and commercial operations, product development, CANDU research, waste management and nuclear sciences, environmental management and site refurbishment. A financial review is included, along with management responsibility, an Auditor's report, financial statements, a five-year financial summary, and a list of directors and locations

  6. AECL: Changing to meet the challenge

    International Nuclear Information System (INIS)

    In this paper, the president of AECL (Atomic Energy of Canada Ltd.) shares some thoughts on reorganization in general, and the on-going reorganization of AECL in particular. He explains that downsizing and the drive for efficiency are not enough: the organization must be customer-oriented, which means meeting with potential customers and listening to them, as well as thinking about their needs, and planning accordingly. Not only AECL, but the whole Canadian nuclear industry needs to be market-driven and to improve its marketing skills

  7. Final report of the AECL/SKB Cigar Lake analog study. AECL research No. AECL-10851

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, J.J.; Smellie, J.A.T. (eds.)

    1994-07-15

    AECL has conducted natural analog studies on the Cigar Lake uranium deposit in northern Saskatchewan since 1984 as part of the Canadian Nuclear Fuel Waste Management Program. This report provides background information and summarizes the results of the study, emphasizing the analog aspects and the implications of modelling activities related to the performance assessment of disposal concepts for nuclear fuel wastes developed in both Canada and Sweden. The study was undertaken to obtain an understanding of the process involved in, and the effects of, steady-state water-rock interaction and trace-element migration in and around the deposit, including paleo-migration processes since the deposit was formed. To achieve these objectives, databases and models were produced to evaluate the equilibrium thermodynamic codes and databases; the role of colloids, organics, and microbes in transport processes for radionuclides; and the stability of UO2 and the influence of radiolysis on UO2 dissolution and radionuclide migration.

  8. AECL programs in advanced systems research

    International Nuclear Information System (INIS)

    The AECL program in advanced systems research is directed in the long term to securing the option of obtaining fissile fuel by electronuclear breeding (accelerator breeder or fusion breeder) and to providing a basis from which AECL might move into stand alone fusion energy if warranted. In the short term the program is directed to reaping benefits from electronuclear technology. This report outlines the main activities and research facilities in both the long-term and short-term subprograms

  9. The AECL reactor development programme

    International Nuclear Information System (INIS)

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

  10. AECL annual review 1991-1992

    International Nuclear Information System (INIS)

    Formed as a Crown Corporation in 1952, AECL consists of two main divisions: AECL CANDU, based in Missisauga and Montreal, responsible for the development, design, marketing and project management of CANDU nuclear power projects; and AECL Research, with its head office in Ottawa and laboratories in Chalk River, Ontario and Pinawa, Manitoba, which supports CANDU and performs the research, development, demonstration and marketing required to apply nuclear sciences and their associated technologies. A strategic plan is under development, which will address the issues of market identification, key partnerships, securing the CANDU technology base, export financing and optimum business structure. In 1991/92 operating income was $16.4 million, up from $7.8 million in 1990/91. Good progress was made on goals to revitalize and upgrade AECL employee's skills and productivity. Key goals for AECL CANDU were: launching the Wolsung 2 reactor project in south Korea; closing the timing and product options for Wolsong 3 and 4; securing new business for Cernavoda 1; and attaining an agreement with either Saskatchewan Power Corp. or the New Brunswick Electric Power Commission regarding the timing of their CANDU 3 projects. Some success was achieved in the first three goals; Saskatchewan has chosen not to proceed with its CANDU 3 plant, but negotiations are continuing in New Brunswick. Key goals for AECL Research were: securing an advanced CANDU research and development program outside the CANDU Owners Group; Disposing of remaining non-nuclear technologies by spin-off, licensing or close-out; rationalizing commercial operations to generate increased revenues; and obtaining the Atomic Energy Control Board's approval of the NRU reactor assessment basis document. Progress was made on all goals

  11. AECL annual report 1996-1997

    International Nuclear Information System (INIS)

    The 1996/1997 Annual Report of Atomic Energy of Canada Ltd. (AECL) is published and submitted to the Honourable member of parliament, Minister of Natural Resources. Included in this report are messages from marketing, commercial operations, product development, CANDU research, waste management, environmental management, financial review and copies of financial statements

  12. AECL's business prospects with China improve

    International Nuclear Information System (INIS)

    In November 1994, Atomic Energy of Canada Ltd. (AECL) and the China National Nuclear Corp. signed a memorandum of understanding which opens the door for the eventual sale of two 685 MW Candu reactors worth a total of C$3.5-billion

  13. AECL annual report 1996-1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The 1996/1997 Annual Report of Atomic Energy of Canada Ltd. (AECL) is published and submitted to the Honourable member of parliament, Minister of Natural Resources. Included in this report are messages from marketing, commercial operations, product development, CANDU research, waste management, environmental management, financial review and copies of financial statements.

  14. AECL's support to operating plants world wide

    International Nuclear Information System (INIS)

    Through their operating records, CANDU reactors have established themselves as a successful and cost-effective source of electricity in Canada and abroad. They have proven to be safe, reliable and economical. A variety of factors have contributed to the enviable CANDU record, such as a sound design based on proven principles supported by effective development programs, along with dedicated plant owners committed to excellence in safely maintaining and operating their plants. Atomic Energy of Canada Limited (AECL), the CANDU designer, has continuously maintained a close relationship with owners/operators of the plants in Canada, Argentina, Romania and South Korea. AECL and the plant operators have all benefited from this strengthening relationship by sharing experience and information. CANDU plant operators have been required to respond decisively to the economic realities of downward cost pressures and deregulation. Operating, Maintenance and Administration (OM and A) costs are being given a new focus as plant owners review each cost element to improve the economic returns from their investments. Amongst the three main OM and A constituents, plant maintenance costs are the most variable and have the largest influence on effective plant operations. The correlation between effective plant maintenance and high capacity factors shows clearly the importance of proactive maintenance planning to reduce the frequency and duration of forced plant outages and their negative impacts on plant economics. This paper describes the management processes and organizational structures m AECL that support plant operations and maintenance in operating CANDU plants with cost effective products and services. (author)

  15. AECL experience in fuel channel inspection

    International Nuclear Information System (INIS)

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

  16. Overview of research reactor operation within AECL

    International Nuclear Information System (INIS)

    This paper presents information on reactor operations within the Research Company of Atomic Energy of Canada (AECL) today relative to a few years ago, and speculates on future operations. In recent years, the need for Research Company reactors has diminished. This, combined with economic pressures, has led to the shutdown of some of the company's major reactors. However, compliance with the government agenda to privatize government companies in Canada, and a Research Company policy of business development, has led to some offsetting activities. The building of a pool-type 10 MWt MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor for isotope production will assist in the sale of the AECL isotopes marketing company. A Low Enriched Uranium (LEU) fuel fabrication facility and a Tritium Extraction Plant (TEP), both currently under construction, are needed in support of the NRU (National Research Universal) reactor and are in line with business development strategies. The research program demands on NRU stretch many years into the future and the strategies for achieving effective operation of this aging reactor, now 32 years old, are discussed. The repair of the leaking light-water reflector of the NRU reactor is highlighted. The isotope business requires that a second reactor be available for back-up production and the operation of the 42 year old NRX (National Research Experimental) reactor in its present 'hot standby' mode is believed to be unique in the world

  17. Some AECL facilities to relocate in Saskatoon

    International Nuclear Information System (INIS)

    'Full-text': Under the terms of memorandum of understanding (MOU) signed by the federal and Saskatchewan governments, Atomic Energy of Canada Limited (AECL) will relocate its design, engineering and marketing offices for CANDU 3 reactors to Saskatoon. This will mean 115 new high-technology jobs for the city in the first year, which might increase to 140 jobs in the second year. As well, the MOU calls for feasibility studies on the establishment of a nuclear accelerator technology centre with accelerator development and marketing components, a nuclear simulator and training facility, a Slowpoke Energy System business, and other related technology in the areas of medicine, agriculture and industry. The provincial government and AECL will cost-share the new arrangement to a maximum of $20 million each over the four year term of the agreement. The MOU is significantly different from the one signed in September, 1991 in that there is no pre-commitment, or any commitment, on the part of the province to purchase or build a CANDU reactor for nuclear generation, nor will there be any study or discussion of development of a nuclear waste site in the province. (author)

  18. Validation of the AECL response time tester

    International Nuclear Information System (INIS)

    The response time of a nuclear safety (trip) channel is an important safety parameter, and an ISA standard requires nuclear operators to measure the response times of their trip instrumentation. As a major aid to facilitate this measurement, AECL (Chalk River) has designed and built a Response Time Tester (RTT) for pressure and differential-pressure transmitters. The RTT is mostly automated for ease of use, is self-checking, and complies with the requirements of ISA Standard, S67.06. The RTT was first checked for repeatability and self-consistency. Secondly, it was successfully validated against an independent measurement, namely the transfer function as measured using the natural in-service noise. This validation was done using two Bailey transmitters, which had the unfortunate property of having their response times as functions of the testing conditions. In all instances, after correcting for this Bailey nonlinearity, the RTT performance met its accuracy specification of ±(5% + 5 ms). (author)

  19. A bibliography of AECL publications on reactor safety

    International Nuclear Information System (INIS)

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

  20. The AECL study for an intense neutron - generator (technical details)

    International Nuclear Information System (INIS)

    The AECL study for an intense neutron-generator has been in progress for two years. Recently the scientific and technical details and the conceptual designs were compiled in a report supporting proposals addressed to AECL's Board of Directors for further work. The compilation is being issued in this form to permit further discussion of the technical aspects. However readers are asked to appreciate that it was written primarily for an AECL audience, and specifically that those chapters giving tentative information about costs, the rate of investment and similar items have been omitted or modified, many references have been made to interim internal reports in order to complete the local documentation, but these references do not imply that the reports themselves can be made generally available. (author)

  1. The AECL study for an intense neutron - generator (technical details)

    Energy Technology Data Exchange (ETDEWEB)

    Bartholomew, G.A.; Tunnicliffe, P.R

    1966-07-01

    The AECL study for an intense neutron-generator has been in progress for two years. Recently the scientific and technical details and the conceptual designs were compiled in a report supporting proposals addressed to AECL's Board of Directors for further work. The compilation is being issued in this form to permit further discussion of the technical aspects. However readers are asked to appreciate that it was written primarily for an AECL audience, and specifically that those chapters giving tentative information about costs, the rate of investment and similar items have been omitted or modified, many references have been made to interim internal reports in order to complete the local documentation, but these references do not imply that the reports themselves can be made generally available. (author)

  2. Modeling the critical hydrogen concentration in the AECL test reactor

    International Nuclear Information System (INIS)

    Hydrogen is added to a pressurized water reactor (PWR) to suppress radiolysis and maintain reducing conditions. The minimum hydrogen concentration needed to prevent radiolysis is referred to as the critical hydrogen concentration (CHC). The CHC was measured experimentally in the mid-1990s by Elliot and Stuart in a reactor loop at Atomic Energy of Canada (AECL), and was found to be approximately 0.5 scc/kg for typical PWR conditions. This value is well below industry-normal PWR operating levels near 40 scc/kg. Radiation chemistry models have also predicted a low CHC, even below the AECL experimental result. In the last few years some of the radiation chemical kinetic rate constants have been re-measured and G-values have been reassessed by Elliot and Bartels. These new data have been used in this work to revise the models and compare them with AECL experimental data. It is quite clear that the scavenging yields tabulated for high-LET radiolysis by Elliot and Bartels are not appropriate to use in the present context, where track-escape yields are needed to describe the homogeneous recombination kinetics in the mixed radiation field. In the absence of such data for high temperature PWR conditions, we have used the neutron G-values as fitting parameters. Even with this expedient, the model predicts at least a factor of two smaller CHC than was observed. We demonstrate that to recover the reported CHC result, the chemistry of ammonia impurity must be included. - Highlights: ► Hydrogen is added to nuclear reactor cooling loops to prevent radiolysis. ► Tests at AECL were carried out to determine the critical hydrogen concentration. ► Neutron radiolysis G-values need to be modified to understand the results. ► Ammonia impurity needs to be included for quantitative modeling.

  3. The Atomic Energy of Canada Limited (AECL) employee health study

    International Nuclear Information System (INIS)

    A preliminary examination of records relating to past Chalk River employees provides some reassurance that large numbers of cancer deaths that might be related to occupational radiation exposure do not exist in the groups of employees studied to the end of 1982. The lack of reliable information on deaths of ex-employees who left AECL for other employment prevented the inclusion of this group in this preliminary study. This information will presumably be obtained during the course of the more comprehensive Atomic Energy of Canada Ltd. employee health study. 6 refs

  4. Compendium of the data used with the SYVAC3-CC3 system model. AECL research No. AECL-11013

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    AECL is evaluating a concept for disposing of nuclear fuel waste from CANDU reactors deep in plutonic rock of the Canadian Shield. As part of this evaluation, models of the physical, chemical, geological, and biological processes that could occur in a sealed disposal vault designed to limit transport of contaminants to the accessible environment were developed. The mathematical models of the transport of radionuclides and toxic chemicals from nuclear fuel waste are incorporated into a computer model named the Systems Variability Analysis Code, Generation 3, and Canadian Concept Model, Generation 3 (SYVAC3-CC3). The report presents the data in the master database used by SYVAC3-CC3 for the postclosure assessment of deep geological disposal, derived from a major program of laboratory and field studies conducted by AECL Research over the past 15 years. The data represents characteristics of a hypothetical vault, certain geologic characteristics of the Whiteshell Research Area, and a general surface environment with a human population living a rural lifestyle on a portion of the Canadian Shield in central Canada.

  5. A study of the mortality of AECL employees. V

    International Nuclear Information System (INIS)

    A study has been underway since 1980 on the mortality of past and present AECL employees. The study population consists of 13,491 persons, 9997 males and 3494 females, for a total of 262,403.5 person-years at risk. During the period 1950-1985, 1299 deaths occurred in this population. The number of female deaths (121) is too few for detailed analysis, but the 1178 deaths in the male population represent a useful basis for this study. The present report examines mortality patterns in the AECL cohort between 1950 and 1985 by comparing the observed mortality with that expected in the general population for three groups of workers: those with no exposure, those with up to 50 mSv, and those with more than 50 mSv. Comparisons among the three groups of employees are discussed. The number of deaths is fewer than would be expected on the basis of general population statistics for both males who were exposed to ionizing radiation and those who were not exposed. The findings were similar for the 'all cancer' and 'all other deaths' groupings. In the group of exposed males, elevated Standardized Mortality Ratios (SMRs) are seen for non-Hodgkin's lymphoma and for buccal cavity, rectum and rectosigmoid junction, and prostate cancers. There are elevated SMRs for lymphatic and myeloid leukemias and for large intestine, prostate, brain and biliary system cancers in the 'unexposed' male group. The number of cases identified in all of these cancers is small and the confidence intervals are wide, such that none of the elevated SMRs is statistically significant. The report compares the findings of this study with those of similar studies published in the past decade. (Author) (28 tabs., 33 refs., 2 figs.)

  6. AECL experience with low-level radioactive waste technologies

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL), as the Canadian government agency responsible for research and development of peaceful uses of nuclear energy, has had experience in handling a wide variety of radioactive wastes for over 40 years. Low-level radioactive waste (LLRW) is generated in Canada from nuclear fuel manufacturers and nuclear power facilities, from medical and industrial uses of radioisotopes and from research facilities. The technologies with which AECL has strength lie in the areas of processing, storage, disposal and safety assessment of LLRW. While compaction and incineration are the predominant methods practised for solid wastes, purification techniques and volume reduction methods are used for liquid wastes. The methods for processing continue to be developed to improve and increase the efficiency of operation and to accommodate the transition from storage of the waste to disposal. Site-specific studies and planning for a LLRW disposal repository to replace current storage facilities are well underway with in-service operation to begin in 1991. The waste will be disposed of in an intrusion-resistant underground structure designed to have a service life of over 500 years. Beyond this period of time the radioactivity in the waste will have decayed to innocuous levels. Safety assessments of LLRW disposal are performed with the aid of a series of interconnected mathematical models developed at Chalk River specifically to predict the movement of radionuclides through and away from the repository after its closure and the subsequent health effects of the released radionuclides on the public. The various technologies for dealing with radioactive wastes from their creation to disposal will be discussed. 14 refs

  7. Impact of ENDF/B-VII.0 for AECL applications

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, Ken S.; Altiparmakov, Dimitar V. [AECL - Chalk River Laboratories, Chalk River (Canada)

    2008-07-01

    This paper examines the impact of the new evaluated nuclear data library ENDF/B-VII.0 on selected reactor physics applications at AECL. The twin objectives are to provide feedback to the nuclear data community concerning the practical impact of their work and preliminary guidance to end-users. This work is based on comparison of the results of MCNP simulations with critical measurements involving both the ZED-2 zero power reactor and the MAPLE dedicated isotope production reactors at the Chalk River Laboratories. Significant improvement in the reactivity agreement with the measurements is obtained with ENDF/B-VII.0 for the specific ZED-2 measurements analysed; however, improvements associated with the thermal scattering law data for UO{sub 2} that had been observed initially were subsequently determined to be fortuitous, due to the inadvertent omission of the elastic neutron scattering component. Additionally, the net reactivity impact of major changes to the {sup 90}Zr and {sup 91}Zr capture cross sections with ENDF/B-VII.0 is examined in the MAPLE reactor context and found to be modest due primarily to the offsetting effects of the specific nuclides involved. (authors)

  8. Coupling of Wims-AECL and Origen-S for depletion calculations - 357

    International Nuclear Information System (INIS)

    One of the more powerful tools for isotope depletion calculations in neutron-irradiated material is the SCALE (Standardized Computer Analyses for Licensing Evaluation) module ORIGEN-S, maintained and developed by Oak Ridge National Laboratory. ORIGEN-S takes as input, in addition to a material description, a problem-dependent cross section library in which relative reaction rates for each nuclear process have been pre-evaluated. Creating different libraries for different stages of burnup, and for different materials, allows the 'point' code phenomenology of ORIGEN-S to be extended to more complicated geometries. To this end, AECL (Atomic Energy of Canada Limited) has coupled its successful 2-D neutron transport solver WIMS-AECL 2.5d to ORIGEN-S to create the coupled code 'WOBI' (WIMS-ORIGEN Burnup Integration). This code has been validated against PIE (post irradiation examination) results for CANDUTM reactors and for light-water reactors, and is extensively used at AECL to calculate exit compositions and decay heats for high and low enriched uranium fuels at the NRU (National Research Universal) research reactor located at the Chalk River Laboratories. In addition, because of the significantly expanded list of reactions available in ORIGEN-S, WOBI is more useful for advanced fuel cycle studies than WIMS-AECL alone. This paper discusses the validation results, and verification of WOBI against simple WIMS-AECL and ORIGEN-S stand-alone models. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  10. AECL's concept for the disposal of nuclear fuel waste and the importance of its implementation

    International Nuclear Information System (INIS)

    Since 1978, Canada has been investigating a concept for permanently dealing with the nuclear fuel waste from Canadian CANDU (Canada Deuterium Uranium) nuclear generating stations. The concept is based on disposing of the waste in a vault excavated 500 to 1000 m deep in intrusive igneous rock of the Canadian Shield. AECL Research will soon be submitting an environmental impact statement (EIS) on the concept for review by a Panel through the federal environmental assessment and review process (EARP). In accordance with AECL Research's mandate and in keeping with the detailed requirements of the review Panel, AECL Research has conducted extensive studies on a wide variety of technical and socio-economic issues associated with the concept. If the concept is accepted, we can and should continue our responsible approach and take the next steps towards constructing a disposal facility for Canada's used nuclear fuel waste

  11. AECL's concept for the disposal of nuclear fuel waste and the importance of its implementation

    International Nuclear Information System (INIS)

    Since 1978, Canada has been investigating a concept for permanently dealing with the nuclear fuel waste from Canadian CANDU nuclear generating stations. The concept is based on disposing of the waste in a vault excavated 500 to 1000 m deep in intrusive igneous rock of the Canadian Shield. AECL will soon be submitting an environmental impact statement on the concept to a federal environmental assessment review panel. In accordance with AECL's mandate, and in keeping with the detailed requirements of the panel, AECL has conducted extensive studies on a wide variety of technical and socio-economic issues associated with the concept. If the concept is accepted, we can and should continue our responsible approach, and take the next steps towards constructing a disposal facility for Canada's used fuel wastes. 16 refs

  12. AECL's participation in the commissioning of Point Lepreau generating station unit 1

    International Nuclear Information System (INIS)

    Support from Atomic Energy of Canada Ltd. (AECL) to Point Lepreau during the commissioning program has been in the form of: seconded staff for commissioning program management, preparation of commissioning procedures, and hands-on commissioning of several systems; analysis of test results; engineering service for problem solving and modifications; design engineering for changes and additions; procurement of urgently-needed parts and materials; technological advice; review of operational limits; interpretation of design manuals and assistance with and preparation of submissions to regulatory authorities; and development of equipment and procedures for inspection and repairs. This, together with AECL's experience in the commissioning of other 600 MWe stations, Douglas Point and Ontario Hydro stations, provides AECL with a wide range of expertise for providing operating station support services for CANDU stations

  13. AECL hot-cell facilities and post-irradiation examination services

    International Nuclear Information System (INIS)

    This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analyses are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

  14. Co-operative projects with AECL in the fields of hydrogeology and geochemistry

    International Nuclear Information System (INIS)

    The report covers collaborative study with Atomic Energy of Canada Limited on geological aspects of waste disposal in crystalline rocks. A field test of the sinusoidal hydraulic pressure pulse method was carried out at the URL site to try to define hydraulic properties of major horizontal fractures. The trials were generally successful and observable sine and square wave signals were transmitted. Owing to the limited scale of the programme, and some equipment problems, the results proved difficult to interpret, although the speed and flexibility of the method was demonstrated. A second aspect of collaboration was to be the field comparison of the AECL and NERC/BGS borehole geochemical probes. In the event, the AECL probe development programme was curtailed and a Swedish design selected for purchase. Effort thus switched to technical comparison of the SGAB probe with the NERC/BGS design. Since both are still at various development points the collaboration was limited to technical exchange. The results are presented. (author)

  15. Microbial analysis of the buffer/container experiment at AECL's underground research laboratory

    International Nuclear Information System (INIS)

    The Buffer/Container Experiment (BCE) was carried out at AECL's Underground Research Laboratory (URL) for 2.5 years to examine the in situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived the conditions (i.e., compaction, heat and desiccation) in the BCE and to determine which group(s) of microorganisms would be dominant in such a simulated vault environment. Such knowledge will be very useful in assessing the potential effects of microbial activity on the concept for deep disposal of Canada's nuclear fuel waste, proposed by AECL. 46 refs., 31 tabs., 35 figs

  16. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  17. AECL R and D's role in promoting nuclear research and education

    International Nuclear Information System (INIS)

    Nuclear renaissance has created new opportunities for new technology development and has also brought along the challenge of meeting the growing demand of trained personnel in the nuclear science and engineering. Towards meeting this challenge, AECL R and D organization is actively promoting and supporting the creation of nuclear research capabilities at the universities and also effectively leveraging the R and D at the universities. It has also put in place several new initiatives to attract and develop the talented young people for careers in nuclear science and engineering. This paper describes various interactions and collaborations with the universities that supports the nuclear R and D at the universities and develop highly qualified personnel for the future nuclear R and D needs. (author)

  18. Final report of the AECL/SKB Cigar Lake analog study

    International Nuclear Information System (INIS)

    The Cigar Lake uranium deposit is located in northern Saskatchewan, Canada. The 1.3-billion-year-old deposit is located at a depth of about 450 m below surface in a water-saturated sandstone at the unconformity contact with the high-grade metamorphic rocks of the Canadian Shield. The Cigar Lake deposit has many features that parallel those being considered within the Canadian concept for disposal of nuclear fuel waste. The study of these natural structures and processes provides valuable insight toward the eventual design and site selection of a nuclear fuel waste repository. The main feature of this analog is the absence of any indication on the surface of the rich uranium ore 450 m below. This indicates that the combination of natural barriers has been effective in isolating the uranium ore from the surface environment. More specifically, the deposit provides analog information relevant to the stability of UO2 fuel waste, the performance of clay-based barriers, radionuclide migration, colloid formation, radiolysis, fission-product geochemistry and general aspects of water-rock interaction. The main geochemical studies on this deposit focus on the evolution of groundwater compositions in the deposit and on their redox chemistry with respect to the uranium, iron and sulphide systems. Since 1984, through cooperation from the owners of the Cigar Lake deposit, analog studies have been conducted. AECL, with support from Ontario Hydro under the auspices of the CANDU Owners Group, initiated international participation in 1989 through collaboration with the Swedish Nuclear Fuel and Waste Management Company (SKB) and, more recently, with the Los Alamos National Laboratory (LANL). This report gives the results of the various studies carried out during the 3-year collaboration between AECL and SKB, as well as a summery of the LANL study. It provides detailed information on the generated databases and models, and integrates this information into conclusions for use in safety

  19. Final report of the AECL/SKB Cigar Lake analog study

    International Nuclear Information System (INIS)

    The Cigar Lake uranium deposit is located in northern Saskatchewan, Canada. The 1.3-billion-year-old deposit is located at a depth of about 450 m below surface in a water-saturated sandstone at the unconformity contact with the high-grade metamorphic rocks of the Canadian Shield. The uranium mineralization, consisting primarily of uraninite (UO2), is surrounded by a clay-rich halo in both sandstone and basement rocks, and remains extremely well preserved and intact. The average grade of the mineralization is ∼ 8 wt.% U; locally grades are as high as ∼ 55 wt.%U. The Cigar lake deposit has many features that parallel those being considered within the Canadian concept for disposal of nuclear fuel waste. Specifically, the deposit provides analog information relevant to the stability of UO2 fuel waste, the performance of clay-based barriers, radionuclide migration, colloid formation, radiolysis, fission-product geochemistry and general aspects of water-rock interaction. The main geochemical studies on this deposit focus on the evolution of groundwater compositions in the deposit and on their redox chemistry with respect to the uranium, iron and sulphide systems. Since 1984, through cooperation from the owners of the Cigar lake deposit, analog studies have been conducted. AECL, with support from Ontario Hydro under the auspices of the CANDU Owners Group, initiated international participation in 1989 through collaboration with the Swedish Nuclear Fuel and Waste Management Company (SKB) and, more recently, with the Los Alamos National Laboratory (LANL). This report gives the results of the various studies carried out during the 3-year collaboration between AECL and SKB, as well as a summary of the LANL study. It provides detailed information on the generated databases and models, and integrates this information into conclusions for use in safety assessment of the Canadian, Swedish and United States disposal concepts. 15 refs., 25 figs., 55 tabs

  20. Validation of WIMS-AECL/(MULTICELL)/RFSP system by the results of phase-B test at Wolsung-II unit

    Energy Technology Data Exchange (ETDEWEB)

    Hong, In Seob; Min, Byung Joo; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The object of this study is the validation of WIMS-AECL lattice code which has been proposed for the substitution of POWDERPUFS-V(PPV) code. For the validation of this code, WIMS-AECL/(MULTICELL)/RFSP (lattice calculation/(incremental cross section calculation)/core calculation) code system has been used for the Post-Simulation of Phase-B physics Test at Wolsung-II unit. This code system had been used for the Wolsong-I and Point Lepraeu reactors, but after a few modifications of WIMS-AECL input values for Wolsong-II, the results of WIMS-AECL/RFSP code calculations are much improved to those of the old ones. Most of the results show good estimation except moderator temperature coefficient test. And the verification of this result must be done, which is one of the further work. 6 figs., 15 tabs. (Author)

  1. Current status of the waste identification program at AECL's Chalk River Laboratories

    International Nuclear Information System (INIS)

    The management of routine operating waste by Waste Management and Decommissioning (WM and D) at Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) is supported by the Waste Identification (WI) Program. The principal purpose of the WI Program is to minimize the cost and the effort associated with waste characterization and waste tracking, which are needed to optimize waste handling, storage and disposal. The major steps in the WI Program are: (1) identify and characterize the processes that generate the routine radioactive wastes accepted by WM and D - radioisotope production, radioisotope use, reactor operation, fuel fabrication, et cetera (2) identify and characterize the routine blocks of waste generated by each process or activity - the initial characterization is based on inference (process knowledge) (3) prepare customized, template data sheets for each routine waste block - templates contain information such as package type, waste material, waste type, solidifying agent, the average non-radiological contaminant inventory, the average radiological contaminant inventory, and the waste class (4) ensure generators 'use the right piece of paper with the right waste' when they transfer waste to WM and D - that is they use the correct template data sheets to transfer routine wastes, by: identifying and marking waste collection points in the generator's facility; ensuring that generators implement effective waste collection/segregation procedures; implementing standard procedures to transfer waste to WM and D; and, auditing waste collection and segregation within a generator's facility (5) determine any additional waste block characterization requirements (is anything needed beyond the original characterization by process knowledge?) This paper describes the WI Program, it provides an example of its implementation, and it summarizes the current status of its implementation for both CRL and non-CRL waste generators. (author)

  2. AECL strategy for surface-based investigations of potential disposal sites and the development of a geosphere model for a site

    International Nuclear Information System (INIS)

    The objective of this report is to summarize AECL's strategy for surface-based geotechnical site investigations used in screening and evaluating candidate areas and candidate sites for a nuclear fuel waste repository and for the development of geosphere models of sites. The report is one of several prepared by national nuclear fuel waste management programs for the Swedish Nuclear Fuel and Waste Management Co. (SKB) to provide international background on site investigations for SKB's R and D programme on siting.The scope of the report is limited to surface-based investigations of the geosphere, those done at surface or in boreholes drilled from surface. The report discusses AECL's investigation strategy and the methods proposed for use in surface-based reconnaissance and detailed site investigations at potential repository sites. Site investigations done for AECL's Underground Research Laboratory are used to illustrate the approach. The report also discusses AECL's strategy for developing conceptual and mathematical models of geological conditions at sites and the use of these models in developing a model (Geosphere Model) for use in assessing the performance of the disposal system after a repository is closed. Models based on the site data obtained at the URL are used to illustrate the approach. Finally, the report summarizes the lessons learned from AECL's R and D program on site investigations and mentions some recent developments in the R and D program. 120 refs, 33 figs, 7 tabs

  3. Analysis of the results for the AECL cohort in the IARC study on the radiogenic cancer risk among nuclear industry workers in fifteen countries

    International Nuclear Information System (INIS)

    Over the last two decades there have been attempts to estimate the risks from occupational exposure in the nuclear industry by epidemiological assessments on cohorts of workers. However, generally low doses and relatively small worker populations have limited the precision of such studies. In 1995 the International Agency for Research on Cancer (IARC) completed a study that involved workers from facilities in the USA, UK and AECL. In 2005, IARC completed a further study involving nuclear workers from 15 countries including Canada. Surprisingly, the risk ascribed to the Canadian cohort for all cancers excluding leukaemia, driven by the AECL component, was significantly higher than the cohort as a whole. The work described in this report is an attempt to unravel what might have accounted for the divergence between the results for the AECL cohort and the others

  4. Analysis of the results for the AECL cohort in the IARC study on the radiogenic cancer risk among nuclear industry workers in fifteen countries

    Energy Technology Data Exchange (ETDEWEB)

    Ashmore, J.P. [Ponsonby and Associates, Manotick, Ontario (Canada); Gentner, N.E. [Consultant, Petawawa, Ontario (Canada); Osborne, R.V. [Ranasara Consultants Inc., Deep River, Ontario (Canada)

    2007-03-31

    Over the last two decades there have been attempts to estimate the risks from occupational exposure in the nuclear industry by epidemiological assessments on cohorts of workers. However, generally low doses and relatively small worker populations have limited the precision of such studies. In 1995 the International Agency for Research on Cancer (IARC) completed a study that involved workers from facilities in the USA, UK and AECL. In 2005, IARC completed a further study involving nuclear workers from 15 countries including Canada. Surprisingly, the risk ascribed to the Canadian cohort for all cancers excluding leukaemia, driven by the AECL component, was significantly higher than the cohort as a whole. The work described in this report is an attempt to unravel what might have accounted for the divergence between the results for the AECL cohort and the others.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

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

  7. Research on radionuclide migration under subsurface geochemical conditions. JAERI/AECL Phase II Collaborative Program Year 1 (joint research)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-11-01

    A radionuclide migration experiment program for fractured rocks was performed under the JAERI/AECL Phase-II Collaborative Program on research and development in radioactive waste management. The program started in the fiscal year 1993, as a five-year program consists of Quarried block radionuclide migration program, Speciation of long-lived radionuclides in groundwater, Isotopic hydrogeology and Groundwater flow model development. During the first year of the program (Program Year 1: March 18, 1994 - September 30, 1994), a plan was developed to take out granite blocks containing part of natural water-bearing fracture from the wall of the experimental gallery at the depth of 240 m, and literature reviews were done in the area of the speciation of long-lived radionuclides in groundwater, isotopic hydrogeology and the groundwater flow model development to proceed further work for the Program Year 2. (author)

  8. An analysis of the AECL/CEC field experiment on the transport of 82Br through a single fracture

    International Nuclear Information System (INIS)

    An analysis of the joint AECL/CEC field experiment performed at the Chalk River test site in Canada in 1983 is presented. A pulse of 82Br tracer was injected into a steady dipole flow field set up in a single fracture between two boreholes 10.6 m apart. A model is presented accounting for dispersal by the dipole flow field and for hydrodynamic dispersion within the fracture. The model is fitted to the experimental data of the breakthrough curve by varying a dispersion length and the water travel time along the line joining the boreholes. In addition, the predicted recovery is compared with an estimate of the actual recovery. Recommendations are made for future experiments. (author)

  9. Research on radionuclide migration under subsurface geochemical conditions. JAERI/AECL Phase II Collaborative Program Year 1 (joint research)

    International Nuclear Information System (INIS)

    A radionuclide migration experiment program for fractured rocks was performed under the JAERI/AECL Phase-II Collaborative Program on research and development in radioactive waste management. The program started in the fiscal year 1993, as a five-year program consists of Quarried block radionuclide migration program, Speciation of long-lived radionuclides in groundwater, Isotopic hydrogeology and Groundwater flow model development. During the first year of the program (Program Year 1: March 18, 1994 - September 30, 1994), a plan was developed to take out granite blocks containing part of natural water-bearing fracture from the wall of the experimental gallery at the depth of 240 m, and literature reviews were done in the area of the speciation of long-lived radionuclides in groundwater, isotopic hydrogeology and the groundwater flow model development to proceed further work for the Program Year 2. (author)

  10. Factors controlling the population size of microbes in groundwater from AECL's Underground Research Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Hamon, C. [Atomic Energy of Canada Limited, Whiteshell Labs., Pinawa, Manitoba (Canada); Mills, K. [University of Saskatoon, Saskatoon, SK (Canada); Rana, S.; Vaidyanathan, S. [Deep River Science Academy, Whiteshell Campus Summer 1997, Pinawa, Manitoba (Canada)

    2001-01-01

    Microbial populations in groundwaters from AECL's Underground Research Laboratory (URL) range from 10{sup 3} to 10{sup 5} cells/mL. Based on the total dissolved organic carbon (DOC), nitrate and phosphate content of these waters, populations of about 10{sup 5} to 10{sup 7} cells/mL should be possible. Upon storage of groundwater samples, total cell counts generally increase and viable cell counts always increase. A study was undertaken to determine what controls the in situ microbial population size in groundwater and what causes this population to grow upon sampling. Fresh URL groundwater was filter-sterilized, inoculated with small quantities of the unaltered water and incubated in the absence and presence of added nutrients (nitrate, phosphate and glucose). Unfiltered groundwater and R2A growth medium inoculated with unaltered groundwater, were also incubated. Microbial changes over time were followed by total and viable (on R2A medium) cell counts. Results showed that in the absence of any nutrient addition, populations grew to between 5 x 10{sup 5} to 4 x 10{sup 6} cells/mL, regardless of the initial size of the population ({approx}10{sup 1} to 10{sup 4} cells/mL), suggesting that nutrients for growth were available in the unamended groundwater. It was hypothesized that the original groundwater population was in 'equilibrium' with the underground environment, which likely included a large population of sessile cells in biofilms on fracture surfaces. Sampling of the groundwater removed the large demand on nutrient supplies by the sessile population which subsequently allowed the planktonic population to grow to a new 'equilibrium' with the available nutrients in the sample bottles. Addition of single nutrients (C, N or P) did not increase cell numbers, suggesting that more than one nutrient is limiting growth. Glucose was used very efficiently aerobically in the presence of both added N and P, but somewhat less under anaerobic

  11. The use of HANDIDET reg-sign non-electric detonator assemblies to reduce blast-induced overpressure at AECL's Underground Research Laboratory

    International Nuclear Information System (INIS)

    A number of aspects of the Canadian concept for nuclear fuel waste disposal are being assessed by Atomic Energy of Canada Limited (AECL) in a series of experiments at its Underground Research Laboratory (URL) near Lac du Bonnet, Manitoba, Canada. One of the major objectives of the work being carried out at the URL is to develop and evaluate the methods and technology to ensure safe, permanent disposal of Canada's nuclear fuel waste. In 1994, AECL excavated access tunnels and a laboratory room for the Quarried Block Fracture Migration Experiment (QBFME) at the 240 Level of the URL. This facility will be used to study the transport of radionuclides in natural fractures in quarried blocks of granite under in-situ groundwater conditions. The experiment is being carried out under a cooperative agreement with the Japan Atomic Energy Research Institute. The excavation of the QBFME access tunnels and laboratory was carried out using controlled blasting techniques that minimized blast-induced overpressure which could have damaged or interrupted other ongoing experiments in the vicinity. The majority of the blasts used conventional long delay non-electric detonators but a number of blasts were carried out using HANDIDET 250/6000 non-electric long delay detonator assemblies and HTD reg-sign non-electric short delay trunkline detonator assemblies. The tunnel and laboratory excavation was monitored to determine the levels of blast-induced overpressure. This paper describes the blasting and monitoring results of the blasts using HANDIDET non-electric detonator assemblies and the effectiveness of these detonators in reducing blast-induced overpressure

  12. Past and future fracturing in AECL Research areas in the superior province of the Canadian Precambrian Shield, with emphasis on the Lac du Bonnet Batholith

    International Nuclear Information System (INIS)

    The likelihood that future fracturing, arising from geologic causes, could occur in the vicinity of a nuclear fuel waste repository in plutonic rock of the Canadian Precambrian Shield, is examined. The report discusses the possible causes of fracturing (both past and future) in Shield rocks. The report then examines case histories of fracture formation in Precambrian plutonic rocks in AECL's Research Areas, especially the history of the Lac du Bonnet Batholith, in the Whiteshell Area, Manitoba. Initially, fractures can be introduced into intrusive plutonic rocks during crystallization and cooling of an intrusive magma. These fractures are found at all size scales; as late residual magma dyking, hydraulic fracturing by retrograde boiling off of hydrothermal fluids, and, in some cases, through local differential cooling. Subsequent fracturing is largely caused by changes in environmental temperature and stress field, rather than by alteration of the material behaviour of the rock. Pluton emplacement during orogeny is commonly accompanied by uplift and erosional exhumation, altering both the tectonic and the lithostatic stresses, the rock temperature gradient and the pore fluid characteristics

  13. Rock stability considerations for siting and constructing a KBS-3 repository. Based on experiences from Aespoe HRL, AECL's URL, tunnelling and mining

    Energy Technology Data Exchange (ETDEWEB)

    Martin, C.D. [Univ. of Alberta, Edmonton (Canada); Christiansson, Rolf [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Soederhaell, J. [VBB VIAK AB, Stockholm (Sweden)

    2001-12-01

    Over the past 25 years the international nuclear community has carried out extensive research into the deep geological disposal of nuclear waste in hard rocks. In two cases this research has resulted in the construction of dedicated underground research facilities: SKB's Aespoe Hard Rock Laboratory, Sweden and AECL's Underground Research Laboratory, Canada. Both laboratories are located in hard rocks considered representative of the Fennoscandian and Canadian Shields, respectively. This report is intended to synthesize the important rock mechanics findings from these research programs. In particular the application of these finding to assessing the stability of underground openings. As such the report draws heavily on the published results from the SKB's ZEDEX Experiment in Sweden and AECL's Mine- by Experiment in Canada. The objectives of this report are to: 1. Describe, using the current state of knowledge, the role rock engineering can play in siting and constructing a KBS-3 repository. 2. Define the key rock mechanics parameters that should be determined in order to facilitate repository siting and construction. 3. Discuss possible construction issues, linked to rock stability, that may arise during the excavation of the underground openings of a KBS-3 repository. 4. Form a reference document for the rock stability analysis that has to be carried out as a part of the design works parallel to the site investigations. While there is no unique or single rock mechanics property or condition that would render the performance of a nuclear waste repository unacceptable, certain conditions can be treated as negative factors. Outlined below are major rock mechanics issues that should be addressed during the siting, construction and closure of a nuclear waste repository in Sweden in hard crystalline rock. During the site investigations phase, rock mechanics information will be predominately gathered from examination and testing of the rock core and

  14. Rock stability considerations for siting and constructing a KBS-3 repository. Based on experiences from Aespoe HRL, AECL's URL, tunnelling and mining

    International Nuclear Information System (INIS)

    Over the past 25 years the international nuclear community has carried out extensive research into the deep geological disposal of nuclear waste in hard rocks. In two cases this research has resulted in the construction of dedicated underground research facilities: SKB's Aespoe Hard Rock Laboratory, Sweden and AECL's Underground Research Laboratory, Canada. Both laboratories are located in hard rocks considered representative of the Fennoscandian and Canadian Shields, respectively. This report is intended to synthesize the important rock mechanics findings from these research programs. In particular the application of these finding to assessing the stability of underground openings. As such the report draws heavily on the published results from the SKB's ZEDEX Experiment in Sweden and AECL's Mine- by Experiment in Canada. The objectives of this report are to: 1. Describe, using the current state of knowledge, the role rock engineering can play in siting and constructing a KBS-3 repository. 2. Define the key rock mechanics parameters that should be determined in order to facilitate repository siting and construction. 3. Discuss possible construction issues, linked to rock stability, that may arise during the excavation of the underground openings of a KBS-3 repository. 4. Form a reference document for the rock stability analysis that has to be carried out as a part of the design works parallel to the site investigations. While there is no unique or single rock mechanics property or condition that would render the performance of a nuclear waste repository unacceptable, certain conditions can be treated as negative factors. Outlined below are major rock mechanics issues that should be addressed during the siting, construction and closure of a nuclear waste repository in Sweden in hard crystalline rock. During the site investigations phase, rock mechanics information will be predominately gathered from examination and testing of the rock core and mapping of the

  15. AECL/US INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors -- Fuel Requirements and Down-Select Report

    Energy Technology Data Exchange (ETDEWEB)

    William Carmack; Randy D. Lee; Pavel Medvedev; Mitch Meyer; Michael Todosow; Holly B. Hamilton; Juan Nino; Simon Philpot; James Tulenko

    2005-06-01

    The U.S. Advanced Fuel Cycle Program and the Atomic Energy Canada Ltd (AECL) seek to develop and demonstrate the technologies needed to minimize the overall Pu and minor actinides present in the light water reactor (LWR) nuclear fuel cycles. It is proposed to reuse the Pu from LWR spent fuel both for the energy it contains and to decrease the hazard and proliferation impact resulting from storage of the Pu and minor actinides. The use of fuel compositions with a combination of U and Pu oxide (MOX) has been proposed as a way to recycle Pu and/or minor actinides in LWRs. It has also been proposed to replace the fertile U{sup 238} matrix of MOX with a fertile-free matrix (IMF) to reduce the production of Pu{sup 239} in the fuel system. It is important to demonstrate the performance of these fuels with the appropriate mixture of isotopes and determine what impact there might be from trace elements or contaminants. Previous work has already been done to look at weapons-grade (WG) Pu in the MOX configuration [1][2] and the reactor-grade (RG) Pu in a MOX configuration including small (4000 ppm additions of Neptunium). This program will add to the existing database by developing a wide variety of MOX fuel compositions along with new fuel compositions called inert-matrix fuel (IMF). The goal of this program is to determine the general fabrication and irradiation behavior of the proposed IMF fuel compositions. Successful performance of these compositions will lead to further selection and development of IMF for use in LWRs. This experiment will also test various inert matrix material compositions with and without quantities of the minor actinides Americium and Neptunium to determine feasibility of incorporation into the fuel matrices for destruction. There is interest in the U.S. and world-wide in the investigation of IMF (inert matrix fuels) for scenarios involving stabilization or burn down of plutonium in the fleet of existing commercial power reactors. IMF offer the

  16. AECL research programs in life sciences

    International Nuclear Information System (INIS)

    The present report summarizes the current research activities in life sciences in the Atomic Energy of Canada Limited-Research Company. The research is carried out at its two main research sites: the Chalk River Nuclear Laboratories and the Whiteshell Nuclear Research Establishment. The summaries cover the following areas of research: radiation biology, medical biophysics, epidemiology, environmental research and dosimetry. (author)

  17. AECL devises new nuclear welding system

    International Nuclear Information System (INIS)

    Automatic autogenous TIG pipe butt welding equipment has been developed for producing joints in reactor coolant monitoring systems for tubes of between 6 and 25 mm diameter and up to 3 mm wall thickness in stainless steel. The equipment is designed to work on site with power requirements of up to 2.2 KW maximum. A major feature of the design, therefore, was a welding system of sufficiently small size, portability and ruggedness to be able to withstand on-site conditions. Quality control is carried out automatically by a comparison of welding parameters with those of a standard acceptable weld. Details of power source characteristics and welding procedure are given. (author)

  18. AECL programs in basic physics research

    International Nuclear Information System (INIS)

    This report describes the CRNL program of research into the basic properties of atomic nuclei and condensed matter (liquids and solids). Brief descriptions are given of some of the current experimental programs done principally at the NRU reactor and MP tandem accelerator, the associated theoretical studies, and some highlights of past achievements

  19. Canada, Atomic Energy of Canada Limited (AECL), Chalk River Labs: Reuse and Licence Termination of a Number of Facilities at the Chalk River Labs to Allow for Refurbishment of the Site. Annex A. I-1

    International Nuclear Information System (INIS)

    Chalk River Labs is located along the Ottawa River in Ontario, Canada, approximately 200 km north-west of Ottawa. The site began construction in 1944 following the expropriation of approximately 1 500 ha of land. A number of research reactors were constructed at the site along with numerous nuclear labs, hot cells and administrative facilities in support of the research and development work planned for the site. The principal occupants of the Chalk River site are AECL employees with a strong presence from National Resources Canada (NRC) and other small research groups. The site is undergoing substantial changes with an emphasis on minimizing the impact of increasing the builtup area footprint in conjunction with site upgrades and new build projects. To accomplish this task, a number of refurbishment and decommissioning projects were planned. Decommissioning projects were initiated to make room for new development through a number of initiatives. The decommissioning mandate includes the removal of a select group of original deteriorating facilities to make room for new construction and to decommission other facilities to facilitate redevelopment and reuse of the available space. In Canada, the Canadian Nuclear Safety Commission (CNSC) issues nuclear licences. The licensees must demonstrate that it is safe to continue operations of the nuclear site and request a renewal of their licence. CNSC will issue a new operating licence for a specific period of time at which the licensee must demonstrate that it is safe to proceed with a licence renewal. A request to terminate a licensable activity must be submitted to the CNSC. Upon approval to proceed, it must be demonstrated that the licensable activities have ceased and the facility has been appropriately decommissioned. Licence termination requires a demonstration that the land or previous activities presents a low risk and that the process can be used to support redevelopment because it results in a scrutinized

  20. Microbial analysis of the buffer/container experiment at AECL`s Underground Research Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Hamon, C.J.; Haveman, S.A.; Delaney, T.L. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs; Pedersen, K.; Ekendahl, S.; Jahromi, N.; Arlinger, J.; Hallbeck, L. [Univ. of Goeteborg, (Sweden). Dept. of General and Marine Microbiology; Daumas, S.; Dekeyser, K. [Guiges Recherche Appliquee en Microbiologie, Aix-en-Provence, (France)

    1996-05-01

    The Buffer/Container experiment was carried out for 2.5 years to examine the in-situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived to conditions and to determine which groups of microorganisms would be dominant in such a simulated vault environment. Microbial analyses were initiated within 24 hour of sampling for all types of samples taken. The culture results showed an almost universal disappearance of viable microorganisms in the samples taken from near the heater surface. The microbial activity measurements confirmed the lack of viable organisms with very weak or no activity measured in most of these samples. Generally, aerobic heterotrophic culture conditions gave the highest mean colony-forming units (CFU) values at both 25 and 50 C. Under anaerobic conditions, and especially at 50 C, lower mean CFU values were obtained. In all samples analyzed, numbers of sulfate reducing bacteria were less than 1000 CFU/g dry material. Methanogens were either not present or were found in very low numbers. Anaerobic sulfur oxidizing bacteria were found in higher numbers in most sample types with sufficient moisture. The statistical evaluation of the culture data demonstrated clearly that the water content was the variable limiting the viability of the bacteria present, and not the temperature. 68 refs, 35 figs, 37 tabs.

  1. Processing of LLRW arising from AECL nuclear research centres

    International Nuclear Information System (INIS)

    Operation of nuclear research reactors and laboratories results in the generation of a wide variety of solid and liquid radioactive wastes. This paper describes practical experience with processing of low-level radioactive wastes at two major nuclear research centres in Canada

  2. AECL programs for new applications for nuclear energy

    International Nuclear Information System (INIS)

    This document reports the activities of the New Applications Steering Committee (NASC) of Atomic Energy of Canada Ltd. The NASC is intended to develop future RβD programs, and more specifically to promote certain existing ideas that have not yet become part of established programs, stimulate new idaas, identify needs and opportunities for RβD, evaluate proposals for RβD programs, initiate action on new ideas, and provide feedback to a staff who may be expected to generate ideas. Major areas and technologies that have been studied by the NASC and are covered in this report include oil substitution by nuclear heat and by electricity, energy storage and the role of hydrogen, nuclear energy in liquid fuel production, assessment of Canadian energy resources, and computer modelling of energy systems

  3. Reorganization of AECL and the future marketing program

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. Engineering Co. has been reorganized to support the new emphasis on foreign sales of CANDU reactors. Much has been learned from reactor sales to Argentina, Korea, and Romania, but Canada needs to sell one 600 MWe reactor a year in order to avoid a decline in its nuclear industry. (LL)

  4. Radiation applications research and facilities in AECL research company

    Science.gov (United States)

    Iverson, S. L.

    In the 60's and 70's Atomic Energy of Canada had a very active R&D program to discover and develop applications of ionizing radiation. Out of this grew the technology underlying the company's current product line of industrial irradiators. With the commercial success of that product line the company turned its R&D attention to other activities. Presently, widespread interest in the use of radiation for food processing and the possibility of developing reliable and competitive machine sources of radiation hold out the promise of a major increase in industrial use of radiation. While many of the applications being considered are straightforward applications of existing knowledge, others depend on more subtle effects including combined effects of two or more agents. Further research is required in these areas. In March 1985 a new branch, Radiation Applications Research, began operations with the objective of working closely with industry to develop and assist the introduction of new uses of ionizing radiation. The Branch is equipped with appropriate analytical equipment including HPLC (high performance liquid chromatograph) and GC/MS (gas chromatograph/mass spectrometer) as well as a Gammacell 220 and an I-10/1, one kilowatt 10 MeV electron accelerator. The accelerator is located in a specially designed facility equipped for experimental irradiation of test quantities of packaged products as well as solids, liquids and gases in various configurations. A conveyor system moves the packaged products from the receiving area, through a maze, past the electron beam at a controlled rate and finally to the shipping area. Other necessary capabilities, such as gamma and electron dosimetry and a microbiology laboratory, have also been developed. Initial projects in areas ranging from food through environmental and industrial applications have been assessed and the most promising have been selected for further work. As an example, the use of charcoal adsorbent beds to concentrate the components of gas or liquid waste streams requiring treatment is showing promise as a method of significantly reducing the cost of radiation treatment for some effluents. A number of other projects are described.

  5. Predicting the effects of microbial activity on the corrosion of copper nuclear fuel waste disposal containers. AECL research No. AECL-11598

    Energy Technology Data Exchange (ETDEWEB)

    King, F.; Stroes-Gascoyne, S.

    1996-12-31

    Microbially influenced corrosion of copper nuclear fuel waste containers may occur in a disposal vault buried in granitic rock. The extent and diversity of microbial activity in the vault is expected to be limited initially because of the aggressive conditions produced by gamma radiation, elevated temperatures, and desiccation of the clay-based buffer in which the containers will be emplaced. This paper presents new evidence regarding the claim that a virtually sterile zone will be created around the container, and describes experiments studying the effects of remote sulphate-reducing bacteria activity on the long-term corrosion of the container. A method for predicting the consequences for the container lifetime is also presented.

  6. Microbial analysis of the buffer/container experiment at AECL's Underground Research Laboratory

    International Nuclear Information System (INIS)

    The Buffer/Container experiment was carried out for 2.5 years to examine the in-situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived to conditions and to determine which groups of microorganisms would be dominant in such a simulated vault environment. Microbial analyses were initiated within 24 hour of sampling for all types of samples taken. The culture results showed an almost universal disappearance of viable microorganisms in the samples taken from near the heater surface. The microbial activity measurements confirmed the lack of viable organisms with very weak or no activity measured in most of these samples. Generally, aerobic heterotrophic culture conditions gave the highest mean colony-forming units (CFU) values at both 25 and 50 C. Under anaerobic conditions, and especially at 50 C, lower mean CFU values were obtained. In all samples analyzed, numbers of sulfate reducing bacteria were less than 1000 CFU/g dry material. Methanogens were either not present or were found in very low numbers. Anaerobic sulfur oxidizing bacteria were found in higher numbers in most sample types with sufficient moisture. The statistical evaluation of the culture data demonstrated clearly that the water content was the variable limiting the viability of the bacteria present, and not the temperature. 68 refs, 35 figs, 37 tabs

  7. Post-irradiation examination of the 37M fuel bundle at Chalk River Laboratories (AECL)

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    The modified (-element (37M) fuel bundle was designed by Ontario Power Generation (OPG) to improve Critical Heat Flux (CHF) performance in ageing pressure tubes. A modification of the conventional 37-element fuel bundle design, the 37M fuel bundle allows more coolant flow through the interior sub-channels by way of a smaller central element. A demonstration irradiation (DI) of thirty-two fuel bundles was completed in 2011 at OPG's Darlington Nuclear Generating Station to confirm the suitability of the 37M fuel bundles for full core implementation. In support of the DI, fuel elements were examined in the Chalk River Laboratories Hot Cells. Inspection activities included: Bundle and element visual examination; Bundle and element dimensional measurements; Verification of bundle and element integrity; and Internal Gas Volume Measurements. The inspection results for 37M were comparable to that of conventional 37-element CANDU fuel. Fuel performance parameters of the 37M DI fuel bundle and fuel elements were within the range observed for similarly operated conventional 37-element CANDU fuel. Based on these Post Irradiation Examination (PIE) results, 37M fuel performed satisfactorily. (author)

  8. Load following testing by AECL in collaboration with the Institute for Nuclear Research in Romania

    International Nuclear Information System (INIS)

    Tests are planned to confirm and demonstrate that the load following (LF) operation of CANDU reactors would have no deleterious effect on fuel performance. Current operating experience with LF has not identified any new limiting criteria for LF operation. Thus far, fission-gas release and sheath strains have been consistent with those of baseline operation. As part of the collaboration under the Romania-Canada Memorandum for Cooperation in research and development of nuclear energy and technology, one of the areas of focus is LF experiments at the Institute for Nuclear Research (SCN) in Pitesti, Romania, where both in-reactor and out-reactor testing will be performed. This paper describes the irradiation and post-irradiation examination facilities at SCN in Pitesti, the operational experience with power-cycling testing performed in-reactor, and a description of the ongoing in-reactor testing in the SCN TRIGA reactor. This paper also describes the out-reactor test methodology and test matrix that will be used in the SCF tests at SCN. (author)

  9. Verification and characterization of continuum behavior of fractured rock at AECL Underground Research Laboratory

    International Nuclear Information System (INIS)

    The purposes of this study are to determine when a fracture system behaves as a porous medium and what the corresponding permeability tensor is. A two-dimensional fracture system model is developed with density, size, orientation, and location of fractures in an impermeable matrix as random variables. Simulated flow tests through the models measure directional permeability, K/sub g/. Polar coordinate plots of 1/√K/sub g/, which are ellipses for equivalent anistropic homogeneous porous media, are graphed and best fit ellipses are calculated. Fracture length and areal density were varied such that fracture frequency was held constant. The examples showed the permeability increased with fracture length. The modeling techniques were applied to data from the Atomic Energy of Canada Ltd.'s Underground Research Laboratory facility in Manitoba, Canada by assuming the fracture pattern at the surface persists at depth. Well test data were used to estimate the aperture distribution by both correlating and not correlating the aperture with fracture length. The permeability of models with uncorrelated length and aperture were smaller than those for correlated models. A Monte Carlo type study showed that analysis of steady state packer tests consistently underestimate the mean aperture. Finally, a three-dimensional model in which fractures are discs randomly located in space, interactions between the fractures are line segments, and the solution of the steady state flow equations is based on image theory was discussed

  10. Remote robotic inspection of irregular surfaces on the inner diameter of the AECL NRU reactor

    International Nuclear Information System (INIS)

    In May of 2009, the NRU (National Research Universal) reactor was forced to shut down after a small heavy water leak. In 2009-2010 repairs were performed in order to restart medical isotope production mid-August 2010. Since the NRU vessel's return to service, a series of periodic inspections is required to ensure the safe operation of the reactor. Eclipse Scientific in collaboration with Utex Scientific Instruments and Liburdi Automation developed the NDE inspection system for the In-Service Inspection program of the NRU vessel. In addition to the difficult environmental, delivery and inspection circumstances the inspection team was faced with the problem of doing an immersion inspection of the inside surface of the reactor vessel through a small 120 mm access port at a distance of more than 10 m to the inspection area at the bottom of the reactor. The vessel was built over 50 years ago and as the inner surface was modified by the repair program during the forced outage, there were no accurate drawings of the inner surface of the vessel that an automated system could rely upon. Eclipse Scientific in collaboration with Liburdi Automation developed a robotic arm designed to enter from the remote access port to deploy the Phased Array and Eddy Current Array inspection heads into the reactor vessel. The motion control and data acquisition system was developed in collaboration with Utex Scientific Instruments using their Inspection Ware software. This paper will highlight the challenges faced in the development of an inspection system capable of using ultrasonic signals to learn a surface and, using this acquired surface topography, effectively and safely deploy and articulate the different inspection heads required to perform the In-Service Inspection of the NRU vessel. (author)

  11. Information Exchange among COG Member Stations, Utility/AECL Design and External Nuclear Organizations

    International Nuclear Information System (INIS)

    The paper presents the COG Information Exchange Program the mandate of which reads: 'To promote the safety reliability and excellence of CANDU plants worldwide by facilitating the sharing of operating experience amongst the members of COG'. To fulfill its mandate the COG operates Information Exchange Program which: 1. Provides a user-friendly facility, COGNET, for staff of COG member organizations to communicate with each other and with external stations, utilities and organizations on topics applicable to CANDU operation, safety, maintenance, design and performance; 2. Offers one-stop shopping for information applicable to the design, operation, maintenance, safety and performance of CANDU's; 3. Reports and compares the performance of all CANDU stations; 4. Organizes opportunities for individuals involved with the operation of CANDU's to meet with their peers and with CANDU industry experts to share operating experience; 5. Facilitates the identification of generic CANDU problems which leads to the addressing of these problems by others through co-operative projects, designer feedback and R and D programs. The paper has the following content: 1. COGNET; 1.1. COGNET Message Forums; 1.2. COGNET Operations Forums; 1.3. COGNET Private Messages; 2. Report Databases and Library; 2.1. REPEX (Technical Reports); 2.2. PCN (CANDU Plant Modifications); 2.3. SEREX (CANDU Station Events); 2.4. INPO (International Events); 3. CANDU Performance; 3.1. COG NEWSLETTERS; 3.2. Performance Indicators; 4. Workshops; 4.1. COG Workshops

  12. Summary of geoscience work at the AECL research site near Atikokan, Ontario

    International Nuclear Information System (INIS)

    Since 1979 June, geolgical, geophysical and hydro-geological investigations have been conducted at Research Area 4 north of Atikokan, Ontario as part of the Canadian Nuclear Fuel Waste Management Program. Composition, shape and internal structure of the Eye-Dashwa pluton were the subjects of regional field studies. Detailed research concentrated on the detection and characterization of surface and subsurface fractures within a 400-m x 800-m grid area, where five boreholes were drilled to depths of between 200 m and 1100 m. Fracture zones in the area were readily detected by surface mapping, ground very low frequency electromagnetic (VLF-EM) surveys and borehole logging. Borehole logs, downhole tube-wave seismic surveys, and thermal and television logging were successful in detecting open fractures in boreholes

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

    International Nuclear Information System (INIS)

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

  14. The five year report of the Tunnel Sealing Experiment: an international project of AECL, JNC, ANDRA and WIPP

    International Nuclear Information System (INIS)

    The Tunnel Sealing Experiment (TSX) was conducted to address construction and performance issues of full-scale seals for potential application to deep geological repositories for radioactive waste. The TSX was performed by an international partnership representing Japan, France, the United States and Canada. The experiment was installed at the 420-m depth of Atomic Energy of Canada Limited's Underground Research Laboratory in the granite rock of the Precambrian Canadian Shield. The experiment involved the construction of two full-scale tunnel seals at either end of a single excavation. One seal was an assembly of pre-compacted sand-bentonite blocks and the second seal was a single cast of Low-Heat High-Performance concrete. The objective of the TSX was to assess the applicability of technologies for construction of practicable concrete and bentonite bulkheads; to evaluate the performance of each bulkhead; and to identify and document the parameters that affect that performance. This report documents the construction and operation of the experiment over its first five years. During this period, the experiment was designed, tunnels were excavated, and the seals were constructed. The sand-filled region between the two bulkhead seals was filled and pressurized with water to 800 and 2000 kPa. A tracer test was conducted at a tunnel pressure of 800 kPa to assess the solute transport characteristics of full-scale tunnel seals. The most important outcome from the TSX is that functional full-scale repository seals can be constructed using currently available technology. Factors identified as potentially affecting seal performance included: excavation method and minimizing the excavation damaged zone (EDZ); keying bulkheads into the rock to interrupt the EDZ; compacted sand-bentonite placement method; treatment of clay bulkhead-rock interface; rate of clay saturation compared with the rate of water pressurization; clay bulkhead volume expansion; the resealing properties of bentonite; concrete heat of hydration; concrete shrinkage; grouting of the both EDZ and the concrete-rock interface; cracking of the concrete and debonding of the concrete-rock interface. The conclusions arising from the TSX are directly applicable to either tunnel or shaft seals constructed in potential host rock environments under consideration for radioactive waste disposal, and the results have direct application to the repository sealing programs of the participating countries. (author)

  15. Some rock mechanics laboratory testing related to the construction and operation of AECL's Underground Research Laboratory (URL)

    International Nuclear Information System (INIS)

    In preparation for and in support of the geomechanical investigations during the Construction and Operating Phases of the URL, considerable rock mechanics laboratory testing work has been conducted over a range of conditions. The main objectives of the laboratory testing programs are twofold: (1) to provide a rock properties database for the URL rock mass for input into numerical models and the design of in situ experiments; (ii) to study the behaviours of the rocks under the repository conditions when they are subjected to changes in stress, temperature, humidity and other time-dependent factors. This paper discusses some testing programs undertaken in the Mining, Laboratories - Nepean (formerly known as Mining Research Laboratories, CANMET) of Natural Resources Canada. Particular emphasis is placed on relating the laboratory-scale behaviour of the rock samples to the in situ behaviour of the rock mass. (author)

  16. The five year report of the Tunnel Sealing Experiment: an international project of AECL, JNC, ANDRA and WIPP

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, N.A.; Cournut, A.; Dixon, D. (and others)

    2002-07-01

    The Tunnel Sealing Experiment (TSX) was conducted to address construction and performance issues of full-scale seals for potential application to deep geological repositories for radioactive waste. The TSX was performed by an international partnership representing Japan, France, the United States and Canada. The experiment was installed at the 420-m depth of Atomic Energy of Canada Limited's Underground Research Laboratory in the granite rock of the Precambrian Canadian Shield. The experiment involved the construction of two full-scale tunnel seals at either end of a single excavation. One seal was an assembly of pre-compacted sand-bentonite blocks and the second seal was a single cast of Low-Heat High-Performance concrete. The objective of the TSX was to assess the applicability of technologies for construction of practicable concrete and bentonite bulkheads; to evaluate the performance of each bulkhead; and to identify and document the parameters that affect that performance. This report documents the construction and operation of the experiment over its first five years. During this period, the experiment was designed, tunnels were excavated, and the seals were constructed. The sand-filled region between the two bulkhead seals was filled and pressurized with water to 800 and 2000 kPa. A tracer test was conducted at a tunnel pressure of 800 kPa to assess the solute transport characteristics of full-scale tunnel seals. The most important outcome from the TSX is that functional full-scale repository seals can be constructed using currently available technology. Factors identified as potentially affecting seal performance included: excavation method and minimizing the excavation damaged zone (EDZ); keying bulkheads into the rock to interrupt the EDZ; compacted sand-bentonite placement method; treatment of clay bulkhead-rock interface; rate of clay saturation compared with the rate of water pressurization; clay bulkhead volume expansion; the resealing properties of bentonite; concrete heat of hydration; concrete shrinkage; grouting of the both EDZ and the concrete-rock interface; cracking of the concrete and debonding of the concrete-rock interface. The conclusions arising from the TSX are directly applicable to either tunnel or shaft seals constructed in potential host rock environments under consideration for radioactive waste disposal, and the results have direct application to the repository sealing programs of the participating countries. (author)

  17. List of publications 1986-1987

    International Nuclear Information System (INIS)

    This list includes all the scientific and technical publications of Atomic Energy of Canada Limited - reports, reprints of journal articles, and translations - issued from 1986 April to 1987 December. Together with the earlier cumulative lists (AECL-5000, AECL-5001, AECL-5002, AECL--5003, AECL--5004, AECL--5005), it provides a complete catalogue of publications in the AECL-series. In the future, lists will be produced at twelve month intervals. The titles and other bibliographic information are arranged in several categories, each devoted to a broad subject area. In addition, each document is identified with an AECL number (for example, AECL-12345) which should be used in ordering reports and making enquiries

  18. AECL/U.S. INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors Fuel Requirements and Down-Select Report

    Energy Technology Data Exchange (ETDEWEB)

    William Carmack; Randy Fielding; Pavel Medvedev; Mitch Meyer

    2005-08-01

    This report documents the first milestone of the International Nuclear Energy Research Initiative (INERI) U.S./Euratom Joint Proposal 1.8 entitled “Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Light-Water Reactors.” The milestone represents the assessment and preliminary study of a variety of fuels that hold promise as transmutation and minor actinide burning fuel compositions for light-water reactors. The most promising fuels of interest to the participants on this INERI program have been selected for further study. These fuel compositions are discussed in this report.

  19. 1993 Annual progress report for subsidiary agreement No. 2 (1991--1996) between AECL and US/DOE for a radioactive waste management technical co-operative program

    International Nuclear Information System (INIS)

    A coordinated research program on radioactive waste disposal is being carried out by the Atomic Energy of Canada Limited and the US Department of Energy. This annual report describes progress in the following eight studies: Fundamental materials investigations; In-situ stress determination; Development of a spent fuel dissolution model; Large block tracer test--Experimental testing of retardation models; Laboratory and field tests of in-situ hydrochemical tools; Cigar Lake--Analogue study, actinide and fission product geochemistry; Performance assessment technology exchange; and Development of multiple-well hydraulic test and field tracer test methods

  20. Annual report 1993-1994

    International Nuclear Information System (INIS)

    Established in 1952 as a Crown corporation, AECL reports to Parliament through the Ministry of Natural Resources. Its mandate is to undertake research into nuclear energy and prepare and develop its commercial applications. AECL's mission is to secure the maximum economic benefit for Canada from CANDU technology and associated research and development - the CANDU business. AECL's accomplishments include the development of products and services which, through diligent marketing efforts, are now in use worldwide. The corporation's world-renowned flagship product, the CANDU reactor, currently satisfies 16% of Canada's electricity requirements and is a key component of the energy programs in five other countries. AECL's vision over the next 20 years is to: be a world-leading supplier of full-scope nuclear power capability; be a long-term business with at least a quarter share of the emerging global market for the next generation of nuclear power plants; have a comprehensive ongoing research program to: maintain at the highest levels the performance and safety of operating CANDU plants; advance the CANDU technology and the science that underlies it; develop knowledge on the health effects of radiation and on the safety and environmental impacts of nuclear reactor operation. AECL currently employees 4000 people. With headquarters in Ottawa, AECL operates two research laboratories, one in Ontario and one in Manitoba, and engineering and design offices in Ontario, Quebec, Saskatchewan, and New Brunswick, as well as offices abroad. AECL has a subsidiary, AECL Technologies Inc., situated in Washington, DC

  1. Qualification plan for the Genmod-PC computer program

    Energy Technology Data Exchange (ETDEWEB)

    Richardson, R.B.; Wright, G.M.; Dunford, D.W.; Linauskas, S.H

    2002-07-01

    Genmod-PC is an internal dosimetry code that uses Microsoft Windows operating system, and that currently calculates radionuclide doses and intakes for an adult male. This report provides a plan for specifying the quality assurance measures that conform to the recommendations of the Canadian Standards Association, as well as AECL procedural requirements for a legacy computer program developed at AECL. (author)

  2. Atomic Energy of Canada Limited annual report 1989-1990

    International Nuclear Information System (INIS)

    In 1990, after a comprehensive industry review, the Canadian government announced that steps would be taken to revitalize the nuclear industry. Canada's nuclear utilities made a commitment to bear a large share of the cost of nuclear research and development. Atomic Energy of Canada Limited (AECL) reported its first financial loss in twelve years, as anticipated at the start of the year. Four of the 20 CANDU reactors operating worldwide were in the top ten based on lifetime performance. By year-end one foreign and two domestic utilities had announced their intention to build more CANDU units. The federal government has agreed to stabilize AECL's research funding at 1989-90 levels ($31.5 million above levels planned in 1985), has authorized AECL to negotiate with New Brunswick to build Point Lepreau-2 as the prototype for the CANDU-3 reactor, and has allowed the restructuring of AECL so utility and private sector investors can become equity partners in AECL CANDU

  3. Restructuring of the Canadian nuclear industry

    International Nuclear Information System (INIS)

    Issues of structural change pertaining to Atomic Energy of Canada Limited (AECL) are discussed. AECL is responsible for the CANDU design, construction and engineering program as well as Canada's nuclear research and development programs, along with Ontario Hydro, a provincially owned electric power utility. Restructuring of these two organizations will have significant impacts on the entire nuclear industry because of the major role they play in the industry. The roles and structures of AECL and Ontario Hydro are described, the trends forcing restructuring of these two organizations and efforts underway to adapt them to the 'new realities'. (R.P.)

  4. Nuclear industry perspectives: change and challenge

    International Nuclear Information System (INIS)

    The President and Chief Executive Officer of AECL discusses the prospects for new reactor sales and other business. Future sales of CANDU reactors to a number of countries, especially, Korea, China and Turkey, are distinctly possible. Another area of possible opportunity is the use of CANDU to burn plutonium from the curtailed weapons programs of the US and the former USSR. AECL is looking to replace its single operating research reactor, namely NRU. Recently, AECL has submitted an environmental impact statement for its concept for permanent disposal of used fuel

  5. Radioactive waste management in Canada

    International Nuclear Information System (INIS)

    This bibliography is a review of the Canadian literature on radioactive waste management from 1953 to the present. It incorporates the references from the previous AECL--6186 revisions, and adds the current data and some of the references that had been omitted. Publications from outside organizations of concern to the Canadian Nuclear Fuel Waste Program are included in addition to AECL Research reports and papers. This report is intended as an aid in the preparation of the Concept Assessment Document and is complementary to AECL Research's internal document-ready references on the MASS-11 word processing systems

  6. Integrated plant information technology design support functionality

    International Nuclear Information System (INIS)

    This technical report was written as a result of Integrated Plant Information System (IPIS) feasibility study on CANDU 9 project which had been carried out from January, 1994 to March, 1994 at AECL (Atomic Energy Canada Limited) in Canada. From 1987, AECL had done endeavour to change engineering work process from paper based work process to computer based work process through CANDU 3 project. Even though AECL had a lot of good results form computerizing the Process Engineering, Instrumentation Control and Electrical Engineering, Mechanical Engineering, Computer Aided Design and Drafting, and Document Management System, but there remains the problem of information isolation and integration. On this feasibility study, IPIS design support functionality guideline was suggested by evaluating current AECL CAE tools, analyzing computer aided engineering task and work flow, investigating request for implementing integrated computer aided engineering and describing Korean request for future CANDU design including CANDU 9. 6 figs. (Author)

  7. Fuel cycles

    International Nuclear Information System (INIS)

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  8. A compilation of subsurface hydrogeologic data

    International Nuclear Information System (INIS)

    This volume contains the storage coefficient, porosity, compressibility and fracture data for the research sites discussed in Volume 1 which have been studied in sufficient detail to allow for analysis. These sites are the following: Stripa Mine, Sweden; Finnsjon, Kamlunge, Fjallveden, Gidea, Svartboberget, Sweden; Olkiluoto, Loviisa, Lavia, Finland; Climax Granite Nevada Test Site; OCRD Room, Colorado School of Mines; Savannah River Plant, Aiken, South Carolina; Oracle, Arizona; Basalt Waste Isolation Project (BWIP), Hanford, Washington; Underground Research Laboratory, AECL, Canada; Atikokan Research Area, AECL; Chalk River Research Area, AECL; Whiteshell Research Area, AECL. Other sources of information have been included where sufficient site specific geologic and hydrogeologic information is provided. The fracture data for the first three of the sites listed above are contained in this volume. The fracture data for the remaining research research sites are discussed in Volume 4

  9. Atomic Energy of Canada Limited annual report 1987-88

    International Nuclear Information System (INIS)

    The annual report of Atomic Energy of Canada Limited for the fiscal year ended March 31, 1988 covers: Research Company; CANDU Operations; Radiochemical Company; Medical Products Division; The Future; Financial Sections; Board of Directors and Officers; and AECL locations

  10. Korea signs for 2nd CANDU at Wolsong

    International Nuclear Information System (INIS)

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

  11. New applications of radioisotopes

    International Nuclear Information System (INIS)

    The Radiochemical Company of Atomic Energy of Canada Ltd. is developing new uses for radioisotpes. This paper discusses three of them. The first is positron emission tomography. AECL, together with the Montreal Neurological Institute, has developed a new PET scanner, the Therascan 3128. A second area of interest is radiopharmaceuticals, which AECL is beginning to produce in patient-ready form. Finally, investigations are being carried out into the use of cobalt 60 gamma sources as food and waste irradiators

  12. Annual report 1994-1995

    International Nuclear Information System (INIS)

    Established in 1952 as a Crown corporation, AECL reports to Parliament through the Ministry of Natural Resources. Its mandate is to undertake research into nuclear energy and prepare and develop its commercial applications. AECL's objective is to secure the maximum economic benefit for Canada from CANDU technology and the associated research and development. AECL's accomplishments include the development of products and services which are now in use worldwide. The corporation's world-renowned flagship product, the CANDU reactor, supplies almost one-fifth of Canada's electricity requirements and is an important component of the energy programs in five other countries. Building upon these achievements, AECL continues to consolidate its position as a world-leading supplies of full-scope nuclear power capabilities with expectations to capture a substantial share of the emerging global nuclear power market. The immediate goals are to meet the customers' requirements in the delivery of current projects, to ensure that operating CANDU stations continue to maintain a high level of performance, to secure further CANDU sales, to garner increased revenues from further commercialization of CANDU technology, and to serve the Government of Canada's nuclear policy initiatives. AECL currently employs 3900 people in a variety of locations, including two major research laboratories located in Ontario and Manitoba, as well as business, engineering and design offices in Ottawa, Toronto, Montreal, Saskatoon and Fredericton. AECL has two subsidiaries, AECL Technologies Inc., in Washington, DC, and AECL Technologies B.V., in the Hague, Netherlands, and maintains a significant presence in South Korea, as well as several smaller offices in other countries

  13. Nuclear energy. Unmasking the mystery

    International Nuclear Information System (INIS)

    The Standing Committee on Energy, Mines and Resources of the House of Commons of Canada undertook a study of the economics of nuclear power in Canada. This is its report on the evidence it heard. It found that maintaining the nuclear power option is vital to Canada's interests. The Committee recommended that: the schedule for establishing a commercial high-level radioactive waste repository be advanced; the basic insurance coverage on nuclear facilities be raised; the federal government increase its financial support of Atomic Energy of Canada Ltd. (AECL); AECL expand its research and development activities, including non-nuclear R and D; AECL be allowed to hold a minority interest in any component of AECL that is privatized; any new entity created by privatization from AECL be required to remain under Canadian control; the Atomic Energy Control Act be altered to allow the Atomic Energy Control Board (AECB) to recover costs through licensing fees and user charges, while the AECB's parliamentary appropriation is increased to offset remaining costs of operations; membership on the AECB be increased from one to five full-time members, retaining the present four part-time members; the AECB hold its hearings in public; the name of the AECB be changed so it is more readily distinguishable from AECL; the AECB establish an office of public information; and that federal and provincial governments cooperate more closely to identify opportunities where more efficient use of electricity could be achieved and to promote those measures that can attain the greatest economic efficiency

  14. Annual report 1997-1998

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) was established in 1952 as a Crown Corporation and reports to parliament through the Minister of Natural Resources. As an annual report, financial statements are an integral element, financial analysis and review are also ongoing. AECL is very active in marketing the science culture which is key to public understanding and acceptance of the nuclear industry. In commercial operations, the CANDU is still the flagship to be marketed in many countries. AECL is the main producer of medical isotopes for the global market. AECL and MDS Nordion signed agreements to secure the ongoing supply of isotopes and to build and operate two MAPLE reactors at the Chalk River site. Activities at AECL are focused on improved economics, further enhanced safety systems and fuel cycle flexibility in the research and product development programs. Waste management and nuclear sciences i e. health and environmental sciences are ongoing studies. Site refurbishment focuses on replacing and refurbishing major facilities to meet business needs

  15. Annual report 1997-1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    Atomic Energy of Canada Limited (AECL) was established in 1952 as a Crown Corporation and reports to parliament through the Minister of Natural Resources. As an annual report, financial statements are an integral element, financial analysis and review are also ongoing. AECL is very active in marketing the science culture which is key to public understanding and acceptance of the nuclear industry. In commercial operations, the CANDU is still the flagship to be marketed in many countries. AECL is the main producer of medical isotopes for the global market. AECL and MDS Nordion signed agreements to secure the ongoing supply of isotopes and to build and operate two MAPLE reactors at the Chalk River site. Activities at AECL are focused on improved economics, further enhanced safety systems and fuel cycle flexibility in the research and product development programs. Waste management and nuclear sciences i e. health and environmental sciences are ongoing studies. Site refurbishment focuses on replacing and refurbishing major facilities to meet business needs.

  16. Progress report, physics and health sciences, physics section, 1986 January 01 - June 30

    International Nuclear Information System (INIS)

    The two progress reports PR-PHS-P-1 (AECL-9262) and PR-PHS-HS-1 (AECL-9263) are continuations of the former series in Physics, PR-P-142, (AECL-9103) and in Health Sciences, PH-HS-20 (AECL-9102). The new series have been initiated to take into account the reorganization of the Research Company effective 1986 February 1. It is intended to issue the reports semi-annually on June 30 and December 31 covering the previous six months. The new series cover the same areas as before except that the Accelerator Physics Branch and the Mathematics and Computation Branch activities are no longer included in Physics, and the activities of the Medical Biophysics Branch at Whiteshell are now included in Health Sciences. The latest progress report on the Medical Biophysics work appeared in the WNRE report PR-WHS-73. This report (AECL-9262) covers the research, business and commercial activities of Nuclear Physics, TASCC Operations, Neutron and Solid State Physics, Theoretical Physics and the Fusion Office

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  18. Nuclear medicine tomorrow

    International Nuclear Information System (INIS)

    The purpose of this Workshop was to discuss and promote future nuclear medicine applications. Atomic Energy of Canada Limited (AECL) is determined to assist in this role. A major aim of this gathering was to form an interface that was meaningful, representative of the two entities, and above all, on-going. In the opening address, given by Mr. J. Donnelly, President of AECL, this strong commitment was emphasized. In the individual sessions, AECL participants outlined R and D programs and unique expertise that promised to be of interest to members of the nuclear medicine community. The latter group, in turn, described what they saw as some problems and needs of nuclear medicine, especially in the near future. These Proceedings comprise the record of the formal presentations. Additionally, a system of reporting by rapporteurs insured a summary of informal discussions at the sessions and brought to focus pertinent conclusions of the workshop attendees

  19. Epp names new interim execs to head Atomic Energy Canada

    International Nuclear Information System (INIS)

    Federal Energy Minister Jake Epp has named Mrs. Marnie Paiken as acting chairman and Bruce Howe as acting president of AECL (formerly Atomic Energy Canada Ltd.), the federal Crown corporation charged with the development and utilization of nuclear energy. Both appointments were made necessary by the resignations of Robert Ferchat as chairman and Stanley Hatcher as president, each citing deep differences in their respective approaches to the management of the corporation. Mrs. Paiken has been a member of AECL's board since 1985, and previously served as acting chairman from March 1989 to July 1990. Howe has been deputy minister of the federal energy department since 1988, a position he will retain while carrying out his duties as president of AECL. A search has begun to find permanent replacements

  20. New opportunities from nuclear R and D

    International Nuclear Information System (INIS)

    The author presents a new initiative within Atomic Energy of Canada Ltd. (AECL), the intention to look for spin-off business opportunities from main-line research and development. In 1982 AECL began encouraging ideas for spin-off applications. Some problems were encountered: the reluctance of staff to divert attention from the CANDU program; resource allocation; difficulties in getting market input; and difficulties in deciding what to license and what to retain as an in-house business opportunity. Successes have come in the areas of using CANDU technology in LWRs, SLOWPOKE reactors, industrial accelerators, stable isotope production, intelligent sensing systems, and deuterated lucite for fibre optics. (L.L.)

  1. Value added services to CANDU plants

    International Nuclear Information System (INIS)

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

  2. Radioactive waste management in Canada: a bibliography of published literature

    International Nuclear Information System (INIS)

    In view of the increased interest in the management of radioactive wastes, a need has been felt for a listing of Canadian publications in this field. Over one hundred AECL reports and other Canadian papers are included as well as a list of selected international conferences on the topic. (author)

  3. Proceedings of the international conference on nuclear structure at high angular momentum and the workshop on large gamma-ray detector arrays. Conference summaries

    International Nuclear Information System (INIS)

    The proceedings of the conference/workshop is being published in two volumes. Abstracts of all contributed papers are contained in volume 1. The final proceedings will be published as volume 2 under the same report number (AECL--10613), and will contain expanded versions of abstracts submitted by registered participants together with the invited and contributed talks

  4. List of publications 1994 January - 1995 December

    International Nuclear Information System (INIS)

    AECL's mandate is to undertake research into nuclear energy and from that develop commercial applications. Its objective is to secure the maximum economic benefit for Canada from CANDU technology and the associated research and development. Among our most important products are scientific reports, publications and conference presentations. This document fists our publications for 1994-95. (author)

  5. List of publications, 1989 January - December

    International Nuclear Information System (INIS)

    This list includes all the scientific and technical publications of Atomic Energy of Canada Limited. This includes both technical reports and reprints of journal articles and conference proceedings issued from 1989 January to 1989 December. The titles and other bibliographic information are arranged in several categories, each devoted to a broad subject area. In addition, each document is identified with an AECL number

  6. Progress report. Physics and Health Sciences, Health Sciences Section (1988 July 01 - December 31)

    International Nuclear Information System (INIS)

    The screening assay for inherited variations in radiosensitivity has been tested. The object is to determine whether those individuals whose cells are abnormally radiosensitive are in fact prone to cancer. Follow-up of the health of radiation workers at AECL continues. As noted in the Hare report (Ontario Nuclear Safety Review), 'epidemiological analysis of the exposed workers of AECL ... shows cancer mortality to be below that in the general public'. These studies are being extended in order to ensure that the initial conclusions remain valid with up-to-date information. A new, very sensitive thermoluminescent material has been adapted for use in AECL dosimetry. The new material results in a much improved performance for measuring small doses and in addition, for accurate dose estimates of low energy beta rays. Much of the work of the Environmental Research Branch concerns modelling. In the atmosphere, our work on atmospheric plume dispersion and metabolic modelling has led naturally to AECL staff contributing to the high profile international study, BIOMOVS. Similarly, the release of a small quantity of tritiated heavy water provided an excellent opportunity to test our model of surface water flow in the Ottawa River. This rather simple model provided a surprisingly accurate prediction, and gave the best estimate of the total release. Finally, continuing analysis of Twin Lakes tracer data is making significant contributions to our very sophisticated model of groundwater flow in porous, heterogeneous media. Conversion of this model to run under NOS/VE on the new Cyber 990 computer is essentially complete

  7. Progress report - Physical and Environmental Sciences - Physics Division. 1994 January 1 to December 31

    International Nuclear Information System (INIS)

    This report marks the change from biannual to annual reports recording technical developments in Physics Division. During this period, AECL has continued with its restructuring program, with Physics Division now included in an expanded Physical and Environmental Sciences Unit. The Division itself remains unchanged, with major activities on neutron scattering, the Sudbury Neutrino Observatory and developments and applications of accelerator technology. (author)

  8. Third international conference on CANDU fuel

    International Nuclear Information System (INIS)

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

  9. Atomic Energy of Canada Limited, annual report, 1995-1996

    International Nuclear Information System (INIS)

    The 1996 Annual Report of Atomic Energy of Canada Ltd. (AECL) is published and submitted to the Honourable member of Parliament, Minister of Natural Resources. Included in this report are messages from Marketing and Commercial Operation, Product Development, i e.CANDU and Research Reactors, CANDU research, Waste Management, Environmental Management, Financial Review and also included are copies of the financial statements

  10. List of publications 1993 January - December

    International Nuclear Information System (INIS)

    AECL research is engaged in research and development related to the peaceful applications of nuclear energy. Specifically, the company's mission is to perform the research, development, demonstration and marketing required to apply nuclear sciences and their related technologies for the maximum benefit of Canada. Among our most important products are scientific reports, publications and conference presentations. This document lists our publications for 1993. (author)

  11. MARS - a multidetector array for reaction studies

    International Nuclear Information System (INIS)

    The proposal for MARS, a Multidetector Array for Reaction Studies is presented. MARS consists of a large, high-vacuum vessel enclosing an array of 128 scintillation detectors for use in studies of heavy-ion collisions at TASCC. The instrument will be funded and owned jointly by AECL and NSERC

  12. List of publications 1990

    International Nuclear Information System (INIS)

    AECL Research is engaged in research and development related to the peaceful applications of nuclear energy. Specifically, the company's mission is to perform the research, development, demonstration and marketing required to apply nuclear sciences and their related technologies for the maximum benefit of Canada. Among our most important products are scientific reports, publications and conference presentations. This document lists our publications for 1990

  13. Disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    In 1978, the governments of Canada and Ontario established the Nuclear Fuel Waste Management program. As of the time of the conference, the research performed by AECL was jointly funded by AECL and Ontario Hydro through the CANDU owners' group. Ontario Hydro have also done some of the research on disposal containers and vault seals. From 1978 to 1992, AECL's research and development on disposal cost about C$413 million, of which C$305 was from funds provided to AECL by the federal government, and C$77 million was from Ontario Hydro. The concept involves the construction of a waste vault 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield. Used fuel (or possibly solidified reprocessing waste) would be sealed into containers (of copper, titanium or special steel) and emplaced (probably in boreholes) in the vault floor, surrounded by sealing material (buffer). Disposal rooms might be excavated on more than one level. Eventually all excavated openings in the rock would be backfilled and sealed. Research is organized under the following headings: disposal container, waste form, vault seals, geosphere, surface environment, total system, assessment of environmental effects. A federal Environmental Assessment Panel is assessing the concept (holding public hearings for the purpose) and will eventually make recommendations to assist the governments of Canada and Ontario in deciding whether to accept the concept, and how to manage nuclear fuel waste. 16 refs., 1 tab., 3 figs

  14. Annual report, 1978-79

    International Nuclear Information System (INIS)

    Activities of AECL for the year ending March 31, 1979 are described. Progress in research, production and marketing of radioisotopes, heavy water manufacture, and construction of CANDU reactors is discussed. The company was restructured in 1978. Hearings on nuclear issues meant that more attention was paid to public affairs. A financial statement is included. (L.L.)

  15. A short history of the CANDU nuclear power system

    International Nuclear Information System (INIS)

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

  16. Eddy current manual, volume 2

    International Nuclear Information System (INIS)

    This report on eddy current testing is divided into three sections: (a) Demonstration of Basic Principles, (b) Practical (Laboratory) Tests and, (c) Typical Certification Questions. It is intended to be used as a supplement to ΣEddy Current Manual, Volume 1Σ (AECL-7523) during CSNDT Foundation Level II and III courses

  17. Detection of defects in logs using computer assisted tomography (CAT) scanning

    International Nuclear Information System (INIS)

    The Chalk River Nuclear Laboratories of AECL have performed a preliminary feasibility study on the applicability of computer assisted tomographic techniques to detect the internal structure of logs. Cross sections of three logs have been obtained using a medical CAT scanner. The results show that knots, rot and growth rings are easily recognized in both dry and wet logs

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

    International Nuclear Information System (INIS)

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

  19. Twenty plus years of underground research at Canada's URL

    International Nuclear Information System (INIS)

    AECL's Underground Research Laboratory (URL) addressed the needs of Canada's Nuclear Fuel Waste Management Program with a comprehensive research and development program of geologic characterization and large-scale geotechnical experiments in granite. This program contributed to the technical acceptance in 1998 of an Environmental Impact Statement to move forward in the development of a deep geologic repository for used CANDU fuel in the crystalline rock of the Canadian Shield. The URL is presently in its final R and D phase, closure. The URL is now 25-years old and is closing. The URL R and D program has met most of its objectives and will continue to do so by capturing valuable information pertaining to its sealing and closure. AECL has 30 years of R and D experience in used-fuel characterization, metallurgy, geosciences and geo-engineering. These comprehensive, multidisciplinary and integrated R and D programs have led to a technically robust concept for a used CANDU fuel deep geologic repository. The NWMO is moving forward based on the technologies pioneered by AECL and the international community. AECL continues to service the NWMO's R and D programs by means of surface-based laboratory and office studies and by participating, under the NWMO's auspices, in international underground research facility programs. AECL's staff have a wealth of experience in solute transport, rock-mass characterization, excavation design and engineered-barrier design within a wide variety of scientific and engineering disciplines. Separate to the service provided to Canada's NWMO, the expert geotechnical R and D capabilities of AECL are being provided via service contracts to a number of international customers. The key messages from this paper are not just that a great deal of important scientific work has been performed at the URL to date, but that AECL's experienced staff are still going strong and continue to play an important role in Canada and internationally. The URL provided

  20. Integrated assessment approach to ensure removal of decay heat with RCW out of service

    International Nuclear Information System (INIS)

    In the unusual event that Recirculated Cooling Water System of a CANDU NPP is drained for maintenance or generally incapacitated and the reactor is in the Guaranteed Shutdown State, an alternative source of cooling water is required to remove decay heat. For the Cernavoda Unit 1 NPP, CNE-PROD assessed and proposed a novel arrangement to address this issue, prepared a design, and requested AECL to assess the proposed approach. As part of the assessment, AECL developed and performed an integrated approach that brought together a variety of design, operational, maintenance and safety considerations. First, a simplified hydraulic analysis was performed, the heat removal requirements of the system were reviewed and alternative heat sinks were evaluated. Then, the maintenance and operational implications were considered in detail, using AECL's Systematic Assessment of Maintenance (SAM) process. This included a detailed assessment of failure modes of interest and associated effects (Failure Modes and Effects Analysis). This assessment included consideration of the impact on operational chemistry. In addition, the safety implications (potential for human error and safety consequences of failure) were reviewed. Based upon the system requirements, hydraulic assessment, and maintenance review, several alternatives or improvements to the CNE-PROD proposal were developed. In addition, a number of operational suggestions, as well as additional concerns and considerations were identified. Finally, potential alternative solutions were looked at based upon previous AECL experience. This paper will describe the novel CNE-PROD arrangement and provide an overview of the integrated assessment approach taken by AECL (using Plant Life Management technologies) to evaluate the proposed arrangement and to identify additional considerations. (author)

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

    International Nuclear Information System (INIS)

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

  2. Defining criteria for cemented waste produced from legacy liquids

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has several hundred cubic metres of legacy radioactive waste stored in underground tanks at the Chalk River Laboratories (CRL) site in Chalk River, Ontario. As part of a larger campaign to reduce its legacy liabilities, AECL intends to remove and immobilize this waste using a cementation system. AECL plans to hire an external contractor to design and operate a cementation skid to remove and condition the liquid wastes. Clear and measurable waste form criteria must be determined and provided to the contractor in order for the contractor to demonstrate that a safe and stable waste form has been produced. AECL has reviewed industry-standard test methods and best practices related to cementation of liquid nuclear wastes. Where suitable, these test methods and practices have been incorporated into Product Performance Criteria. An extensive test program has been performed to obtain cement formulations for the legacy wastes; the resulting sample cemented wastes have been tested and the results compared to the Product Performance Criteria. Modifications to the criteria have been made as required based on knowledge gained during this process. In addition, since no industry standards had previously been identified to measure homogeneity, 3 potential test methods have been identified. Regardless of the amount of testing performed and the stringency of the performance criteria, some risk remains that the waste will deteriorate over time. However, by performing a rigorous review of industry practice and an extensive series of tests under various conditions, AECL believes that it has addressed the risks in a reasonable and prudent manner and has selected the appropriate Product Performance Criteria to achieve a safe and stable waste product

  3. Power and the future generation

    International Nuclear Information System (INIS)

    In this keynote address, the author, who was acting president of AECL at the time of the conference, emphasizes the importance of nuclear energy to Canada, and its future importance to the developing countries. In 1992, nuclear energy supplied 15% of Canada's electricity, employed 30,000 people in Canada, created at least 10,000 jobs in other sectors, generated federal tax revenues of C$700 million, and by supplanting coal and gas imports saved about C$1 billion. Export sales prospects in China, Korea, Turkey, the Philippines, Indonesia and Thailand are indicated. AECL is presently undergoing reorganization for greater efficiency. A public opinion poll indicated about 70% Canadian public support for nuclear energy

  4. Annual report 1998-1999

    International Nuclear Information System (INIS)

    This is the Annual Report of the Atomic Energy of Canada Limited for the year ending March 31, 1999 and summarizes the activities of AECL during the period 1998-1999. The Activities covered in this Report include the CANDU Reactor Business, with excellent progress reported on the construction of two 700 MWe-class CANDU reactors in Qinshan, China. In the Republic of Korea, Wolsong Unit entered into commercial operation and Wolsong Unit 4 achieved sustained nuclear reaction. The Report also covers AECL's R and D and Waste Management programs. In the R and D section, the report outlines the development of the CANFLEX fuel bundle, Fuel Channels, Reactor Safety, Code Validation, Fuels and Fuel Cycles as well as Heavy Water production. Progress in the Waste Management program is also discussed

  5. Medical isotope shortage 2009-2010 and future options NRU, SLOWPOKE and MAPLE

    Energy Technology Data Exchange (ETDEWEB)

    Hilborn, J. [Deep River, Ontario (Canada)

    2013-07-01

    The 15 month shutdown of NRU and the unexpected termination of the AECL/Nordion MAPLE project caused a world-wide shortage of medical isotopes. After the recent repair of NRU, AECL is confident that it could continue operating safely and reliably as a multi-purpose reactor until 2021 or longer. There is convincing evidence that the restoration of the MAPLE reactors is technically feasible, but it is highly improbable that a 10 MW MAPLE production reactor can ever be cost-effective. However, conversion of the present 10 MW reactors to 3 MW, without major changes to the structural hardware, warrants serious consideration. Finally, even the 20 kW SLOWPOKE reactor could produce useful quantities of Mo-99. If the present fuel rods were replaced with a small tank containing a solution of low-enriched uranyl sulphate in water, three of these liquid core reactors could supply all of Canada. (author)

  6. Whiteshell labs closure: crisis or opportunity?

    International Nuclear Information System (INIS)

    L. Simpson, Mayor, Local Government District of Pinawa, Manitoba, described the impacts and public concerns produced by a hastily planned and executed withdrawal of the primary employer from a dependent company town. The Whiteshell Laboratories of the Crown corporation Atomic Energy of Canada Limited (AECL) were established in Eastern Manitoba in 1963, and Pinawa was created 15 kilometres away. Located in a provincial park region, Pinawa has also become a popular holiday cottage area with 20 000 residents inside a 30-minute radius. In 1995, the AECL Reactor Safety Research Program was moved to Chalk River, and the Nuclear Waste Management Program (NWMP) was left in limbo. Commercial negotiations to go on operating business on the site broke down. The town of Pinawa, the major stakeholder, was kept at arm's length from all discussions. (author)

  7. Microbial issues pertaining to the Canadian concept for the disposal of nuclear fuel waste

    International Nuclear Information System (INIS)

    AECL Research is developing a concept for the permanent disposal of nuclear fuel waste in plutonic rock of the Canadian Shield. The Federal Environmental Assessment Review Panel has issued a set of guidelines to be used by AECL Research in preparing an Environmental Impact Statement (EIS) for this concept. These guidelines require that the EIS address a number of microbiological factors and their potential to affect the integrity of the multiple barrier system on which the disposal concept is based. This report formulates a number of views and positions on microbiological factors that could influence the performance of a disposal vault in plutonic rock. Microbiological factors discussed include the presence and survival of microbes, biofilms, corrosion, biodegradation (of emplaced materials), gas production, geochemical changes, radionuclides migration, colloid formation, mutation, pathogens and methylation. Not all issues can be fully resolved with the current state of knowledge. Studies being performed to underscore and strengthen current knowledge are briefly discussed. (author). 92 refs., 1 tab

  8. Microbial studies in the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed a concept for permanent geological disposal of nuclear fuel waste in Canada. An accelerated program was initiated in 1991 to address and quantify the potential effects of microbial action on the integrity of the disposal concept's multiple barrier system. This microbial program focuses on answering specific questions in areas such as the survival of bacteria in compacted clay-based buffer materials under relevant radiation and desiccation conditions; mobility of microbes in compacted buffer materials; the potential for microbially-influenced corrosion of containers; microbial gas production in backfill material; introduction of nutrients as a result of vault excavation and operation; the presence and activity of microbes in deep granitic groundwaters; and the effects of biofilms on radionuclide migration in the geosphere. This paper summarizes the current research activities at AECL in these areas. (author)

  9. A study of the health of the employees of Atomic Energy of Canada Limited. 1

    International Nuclear Information System (INIS)

    This report summarizes the status of Atomic Energy of Canada Ltd.'s health study of its present and past employees, and is a description of the steps which have been taken up to the time of writing. During the design phase there was a shift in the emphasis of the study. What was originally proposed as a study of mortality in a population of radiation workers, related spacifically to radiation exposure, has become a study of mortality data for all AECL employees. The interest in mortality as a function of occupational radiation expksure remains, but it is recognized that the data available to the study will probably be inadequate for the definition of a dose-effect relationship, although it will be useful in conjuction with other similar studies. The importance of cancer incidence is recognized, and the possibility of linking the AECL data to that contained in the National Cancer Incidence Reporting System is being pursued

  10. The disposal of Canada's nuclear fuel waste: public involvement and social aspects

    International Nuclear Information System (INIS)

    This report describes the activities undertaken to provide information to the public about the Canadian Nuclear Fuel Waste Management Program as well as the opportunities for public involvement in the direction and development of the disposal concept through government inquiries and commissions and specific initiatives undertaken by AECL. Public viewpoints and the major issues identified by the public to be of particular concern and importance in evaluating the acceptability of the concept are described. In addition, how the issues have been addressed during the development of the disposal concept or how they could be addressed during implementation of the disposal concept are presented. There is also discussion of public perspectives of risk, the ethical aspects of nuclear fuel waste disposal, and public involvement in siting a nuclear fuel waste disposal facility. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

  11. Collaborative approach in developing a small supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A joint Research and Development (R and D) project between University of Saskatchewan and Atomic Energy of Canada (AECL) is being established to develop a concept of the small Canadian supercritical water-cooled reactor (SCWR) for power generation and process heat in remote areas. This project will be led by professors at the university and supported by technology experts from AECL. It integrates student training with a significant contribution to the reactor concept development. Students from various disciplines will combine results from physics, fuel, thermalhydraulic, control, material, and chemistry analyses to develop the core and fuel channel configurations and fuel design. This project would enhance the R and D expertise and capability of University of Saskatchewan and facilitate training of highly qualified persons (HQPs) for nuclear and non-nuclear industries at Saskatchewan and in Canada. (author)

  12. Annual report 1998-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    This is the Annual Report of the Atomic Energy of Canada Limited for the year ending March 31, 1999 and summarizes the activities of AECL during the period 1998-1999. The Activities covered in this Report include the CANDU Reactor Business, with excellent progress reported on the construction of two 700 MWe-class CANDU reactors in Qinshan, China. In the Republic of Korea, Wolsong Unit entered into commercial operation and Wolsong Unit 4 achieved sustained nuclear reaction. The Report also covers AECL's R and D and Waste Management programs. In the R and D section, the report outlines the development of the CANFLEX fuel bundle, Fuel Channels, Reactor Safety, Code Validation, Fuels and Fuel Cycles as well as Heavy Water production. Progress in the Waste Management program is also discussed.

  13. WLUP burnable absorber isotopic influence on coolant void reactivity in an ACR lattice

    International Nuclear Information System (INIS)

    ACRTM-1000 is the topmost nuclear power reactor promoted by AECL during the next years as a response to increasing competitiveness in the nuclear energy market. Recent AECL innovations allowed overriding for the first time the main CANDU drawback - the positive Coolant Void Reactivity (CVR). The solution was using of burnable absorbers in the central element (CE) whose radius was significantly increased. The paper's goal is to evaluate the isotopic influence on CVR and, as result, on nuclear safety when the central element is filled one by one with the most common oxide of burnable isotopes from the IAEA updated WIMS library (WLUP). The isotopes taken into account are: Dysprosium, Hafnium, Gadolinium, Erbium and Holmium. A comparison between CVRs given at the using of above lanthanides and their suitability to be used in the central element design is illustrated in the paper. (authors)

  14. Update on Canada's fuel waste management program: preparing for the environmental review of the concept

    International Nuclear Information System (INIS)

    The Canadian Nuclear Fuel Waste Program was established in 1978 as a joint initiative by the governments of Canada and Ontario. Under the program, Atomic Energy of Canada Ltd. (AECL) is responsible for developing and assessing a concept to dispose of nuclear fuel wastes in plutonic rocks of the Canadian Shield. A series of engineered and natural barriers will isolate the nuclear fuel waste from the biosphere. The ultimate choice of methods, materials, and designs will depend on the site chosen, availability, cost, and practicality. AECL has submitted the Environmental Impact Statement (EIS) on the concept for review by the Environmental Assessment Panel appointed by the federal Minister of the Environment. Three reasons for implementing the concept if accepted are as follows: to show environmental leadership by reducing the burden on future generation, to foster public confidence in nuclear energy, and to forestall 'inaction by default'. 19 refs

  15. Human health considerations in the assessment of Canadian concept for the disposal of nuclear fuel wastes

    International Nuclear Information System (INIS)

    In 1978, AECL was mandated by the government of Ontario and the federal government to find a permanent disposal solution for spent nuclear fuels. Canada opted for disposal in plutonic rocks of the Canadian shield. The Canadian concept calls for disposal in crystalline rocks at a depth of 500 to 1000 m below the surface. The spent fuel would be contained in a canister, the canister would be emplaced in a vault containing clay-based buffer materials, and the cavity would be backfilled and sealed with natural materials. A Federal Environmental Assessment Review Panel was formed in 1992 to assess the concept for disposal of the spent fuel. In this paper a brief discussion of the human health impacts of the proposed concept is presented. Our assessment is based on the information provided by AECL, namely, the main EIS document, a summary and nine other supporting documents

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

    International Nuclear Information System (INIS)

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

  17. Nuclear platform research and development - 2008-09 highlights

    International Nuclear Information System (INIS)

    The Nuclear Platform R and D Program has lead responsibility for the maintenance and further development of the CANDU intellectual property covering the safety, licensing and design basis for nuclear facilities. The Nuclear Platform R and D Program is part of the Research and Technology Operation (RTO) unit of AECL and is managed through the Research and Development division, which has responsibility for maintaining and enhancing the knowledge and technology base. The RTO is also responsible for managing AECL's nuclear facilities and infrastructure (including laboratories and R and D facilities), the nuclear waste management program and other legacy liabilities (e.g., decommissioning) to demonstrate and grow shareholder value. The Nuclear Platform also provides the technology base from which new products and services can be developed to meet customer needs (including ACR and commercial products and services). (author)

  18. Chernobyl - a Canadian technical perspective

    International Nuclear Information System (INIS)

    In this report we present the design review done to date in Canada by AECL. From the Canadian point of view it covers: 1) relevant information on the Chernobyl design and the accident, both as presented by the Soviets at the Post-Accident Review Meeting (PARM) held in Vienna from August 25-29, 1986, and as deduced from publicly available Soviet documentation; and 2) details of AECL's technical review of the CANDU PHWR (Pressurized Heavy Water Reactor) against the background of the Chernobyl accident, and implications of the Chernobyl accident. Reviews of operational aspects are underway by the Canadian electrical utilities and a review by the Canadian regulatory agency (the Atomic Energy Control Board) is near completion

  19. Electron-processing technology: a promising application for the viscose industry

    International Nuclear Information System (INIS)

    In marketing its IMPELA[reg] line of high-power, high-throughput industrial accelerators, Atomic Energy of Canada Limited (AECL) is working with viscose (rayon) companies world-wide to integrate electron-processing technology as part of the viscose manufacturing process. The viscose industry converts cellulose wood pulp into products such as staple fiber, filament, cord, film, packaging, and non-edible sausage casings. This multibillion dollar industry is currently suffering from high production costs, and is facing increasingly stringent environmental regulations. The use of electron-treated pulp can significantly lower production costs and can provide equally significant environmental benefits. This paper describes our current understanding of the benefits of using electron-treated pulp in this process, and AECL's efforts in developing this technology

  20. Electron-processing technology: a promising application for the viscose industry

    International Nuclear Information System (INIS)

    In marketing its IMPELA line of high-power, high-throughput industrial accelerators, Atomic Energy of Canada Limited (AECL) is working with viscose (rayon) companies world-wide to integrate electron-processing technology as part of the viscose manufacturing process. The viscose industry converts cellulose wood pulp into products such as staple fiber, filament, cord, film, packaging, and non-edible sausage casings. This multibillion dollar industry is currently suffering from high production costs, and is facing increasingly stringent environmental regulations. The use of electron-treated pulp can significantly lower production costs and can provide equally significant environmental benefits. This paper describes our current understanding of the benefits of using electron-treated pulp in this process, and AECL's efforts in developing this technology. (author)

  1. Safety and licensing program for the proposed irradiation research facility

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) proposes to replace NRU with a dual-purpose irradiation-research facility (IRF) to test Canada deuterium uranium (CANDU) fuels and materials and to perform materials research using neutrons. The reference IRF concept was estimated to cost $500 million and would require 87 months to complete. Approval of the IRF project is not expected to occur before 1997, and a favorable decision will be influenced by the estimated cost and confidence in the estimate. Accordingly, AECL has initiated a preproject program that includes code validation, analysis, development and testing, safety and licensing, and concept design activities to reduce uncertainties in the reference IRF project cost and schedule, and to develop cost and schedule reductions

  2. Electron processing of fibre-reinforced advanced composites

    Science.gov (United States)

    Singh, Ajit; Saunders, Chris B.; Barnard, John W.; Lopata, Vince J.; Kremers, Walter; McDougall, Tom E.; Chung, Minda; Tateishi, Miyoko

    1996-08-01

    Advanced composites, such as carbon-fibre-reinforced epoxies, are used in the aircraft, aerospace, sporting goods, and transportation industries. Though thermal curing is the dominant industrial process for advanced composites, electron curing of similar composites containing acrylated epoxy matrices has been demonstrated by our work. The main attraction of electron processing technology over thermal technology is the advantages it offers which include ambient temperature curing, reduced curing times, reduced volatile emissions, better material handling, and reduced costs. Electron curing technology allows for the curing of many types of products, such as complex shaped, those containing different types of fibres, and up to 15 cm thick. Our work has been done principally with the AECL's 10 MeV, 1 kW electron accelerator; we have also done some comparative work with an AECL Gammacell 220. In this paper we briefly review our work on the various aspects of electron curing of advanced composites and their properties.

  3. Transport modeling of sorbing tracers in artificial fractures

    Energy Technology Data Exchange (ETDEWEB)

    Keum, Dong Kwon; Baik, Min Hoon; Park, Chung Kyun; Cho, Young Hwan; Hahn, Phil Soo

    1998-02-01

    This study was performed as part of a fifty-man year attachment program between AECL (Atomic Energy Canada Limited) and KAERI. Three kinds of computer code, HDD, POMKAP and VAMKAP, were developed to predict transport of contaminants in fractured rock. MDDM was to calculate the mass transport of contaminants in a single fracture using a simple hydrodynamic dispersion diffusion model. POMKAP was to predict the mass transport of contaminants by a two-dimensional variable aperture model. In parallel with modeling, the validation of models was also performed through the analysis of the migration experimental data obtained in acrylic plastic and granite artificial fracture system at the Whiteshell laboratories, AECL, Canada. (author). 34 refs., 11 tabs., 76 figs.

  4. Program of experiments for the operating phase of the Underground Research Laboratory

    International Nuclear Information System (INIS)

    The Underground Research Laboratory (URL) is one of the major research and development facilities that AECL Research has constructed in support of the Canadian Nuclear Fuel Waste Management Program. The URL is a unique geotechnical research facility constructed in previously undisturbed plutonic rock, which was well characterized before construction. The site evaluation and construction phases of the URL project have been completed and the operating phase is beginning. A program of operating phase experiments that address AECL's objectives for in situ testing has been selected. These experiments were subjected to an external peer review and a subsequent review by the URL Experiment Committee in 1989. The comments from the external peer review were incorporated into the experiment plans, and the revised experiments were accepted by the URL Experiment Committee. Summaries of both reviews are presented. The schedule for implementing the experiments and the quality assurance to be applied during implementation are also summarized. (Author) (9 refs., 11 figs.)

  5. Environmental scan

    International Nuclear Information System (INIS)

    Trends in current affairs and public policy that emerged in 1988 are analyzed to provide a tool for Atomic Energy of Canada's decision makers. This issue provides a general overview of the economic and energy environments; and international review of nuclear policies, equipment orders, and trends that will position Canada and AECL within the global nuclear community; a precis of Canadian public opinion polls on economic, environmental and energy issues; a summary of major trends identifies by Canadian opinion leaders from the public and private sectors; a review of opinions and recommendations of influential Canadian policy institutes; a report on government policies and actions that affect AECL directly or indirectly; an analysis of new coverage by print and electronic media; a review of anti-nuclear organizations; and conclusions and recommendations

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

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

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

  7. A registry for the study of the health of radiation workers employed by Atomic Energy of Canada Limited

    International Nuclear Information System (INIS)

    Factors to be considered in formulating a study of the health of radiation workers are discussed, and a proposal is made for the establishment of such a study in relation to the employees of Atomic Energy of Canada Limited. By setting up a registry of AECL radiation workers, data could be accumulated suitable for the long-term followup of their health, and for preparing periodic interim reports on mortality and morbidity. (author)

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

    International Nuclear Information System (INIS)

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

  9. Field test of wireless sensor network in the nuclear environment

    International Nuclear Information System (INIS)

    Wireless sensor networks (WSNs) are appealing options for the health monitoring of nuclear power plants due to their low cost and flexibility. Before they can be used in highly regulated nuclear environments, their reliability in the nuclear environment and compatibility with existing devices have to be assessed. In situ electromagnetic interference tests, wireless signal propagation tests, and nuclear radiation hardness tests conducted on candidate WSN systems at AECL Chalk River Labs are presented. The results are favourable to WSN in nuclear applications. (author)

  10. Proposed Atomic Energy of Canada Ltd. 99Mo waste calcination process

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL), at its Chalk River Laboratory, generates from 3000 to 5000 L/year of high-level fissile waste solution from the production of 99Mo. In this Mo process, highly enriched uranium (93 wt % 235U, total uranium basis) contained in uranium-aluminum alloy target rods is irradiated to produce the 99Mo product. The targets are removed from the reactor and dissolved in a mercury nitrate-catalyzed reaction with nitric acid. The 99Mo product is then recovered by passing the solution through an alumina (Al2O3) column. During discussions with personnel from the Oak Ridge National Laboratory (ORNL) on September 10, 1992, the ORNL-developed technology formerly applied to the solidification of aqueous uranium waste (Consolidated Edison Uranium Solidification Program or CEUSP) was judged potentially applicable to the AECL 99Mo waste. Under a Work-for-Others contract (no. ERD-92-1132), which began May 24, 1993, ORNL was tasked to determine the feasibility of applying the CEUSP (or a similar) calcination process to solidify AECL's 99Mo waste for > 30 years of safe dry storage. This study was to provide sufficient detailed information on the applicability of a CEUSP-type waste solidification process to allow AECL to select the process which best suited its needs. As with the CEUSP process, evaporation of the waste and a simultaneously partial destruction of acid by reaction with formaldehyde followed by in situ waste can thermal denitration waste was chosen as the best means of solidification. Unlike the CEUSP material, the 99Mo waste has a considerable number of problem volatile and semivolatile constituents which must be recovered in the off-gas treatment system. Mercury removal before calcination was seen as the best option

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

    International Nuclear Information System (INIS)

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

  12. 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.

  13. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  14. Canadian experience with spin-offs from nuclear technology

    International Nuclear Information System (INIS)

    The innovation process introduced into AECL's research laboratories is described, with its achievements in increased commercial and spin-off businesses. In particular, the role of the champion or entrepreneur is emphasized in the manner in which he/she interacts within a dedicated team to pursue each opportunity. Examples are provided of several commercial and business development opportunities resulting from the background research programs

  15. The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Under the auspices of the IAEA, a consultants' meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena

  16. Human Reliability Analysis in Third Qinshan Nuclear Power Plant%秦山第三核电厂人因可靠性分析

    Institute of Scientific and Technical Information of China (English)

    张力; 戴立操; 赵明; 曾春; 宋明海; 彭晓春

    2012-01-01

    Human reliability analysis (HRA) is an important component of probabilistic safety assessment (PSA). The design HRA was conducted by AECL and the technique was oversimplified. In order to make HRA represent the operational state of Third Qinshan Nuclear Power Plant more realistically, HRA was re-analyzed, to which dependence between events was added. On the basis of a comparison on internationally prevailing HRA techniques, a standardized THERP+HCR technique was adopted. Compared with AECL results, the updated is basically consistent with AECL analysis, while rationality and accuracy are obviously improved and results are more truthful%人因可靠性分析(HRA)是概率安全评价(PSA)的重要组成部分.秦山第三核电厂(简称秦山三核)初版HRA由加拿大原子能公司(AECL)完成,其采用的HRA方法为简化的ASEP-HRA.为获得更符合秦山三核运行状态实际的HRA结论,本工作对秦山三核重新进行了HRA分析,并增加了事件间的相关性分析.在对国际HRA方法比较研究的基础上,秦山三核HRA采用了规范化的THERP+HCR分析方法.新分析所得数据与AECL数据比较分析结果表明,新分析与AECL的分析判断基本一致,但在合理性和准确性方面较原分析有明显提高,分析结论更符合秦山三核实际.

  17. A review of the prospects for fusion breeding of fissile material

    International Nuclear Information System (INIS)

    This report is the result of an eight month study by the AECL Fusion Status Study Group. The objectives of this study were to review the current status of fusion research, to evaluate the neutronic performance of various fusion-breeder systems, and to assess the economic and technological outlook for the fusion breeder as a source of fissile material to support CANDU reactors operating on the thorium fuel cycle

  18. Update on seamless calandria tube development and qualification

    International Nuclear Information System (INIS)

    AECL is undertaking the qualification of the production of seamless calandria tubes as replacement components for installation in reactors during retubing. Seamless tube prototypes made from Zircaloy-2 possessing a suitable crystallographic texture have been shown to be significantly stronger than seam-welded tubes under both rising pressure and sustained pressure conditions in a simulated reactor loading. This paper describes the seamless calandria tube development program and current status. (author)

  19. Implementation and ongoing development of a comprehensive program to deal with Canada's nuclear legacy liabilities - 16039

    International Nuclear Information System (INIS)

    Nuclear legacy liabilities have resulted from 60 years of nuclear research and development carried out on behalf of Canada by the National Research Council (1944 to 1952) and Atomic Energy of Canada Limited (AECL, 1952 to present). These liabilities are located at AECL research and prototype reactor sites, and consist of shutdown reactors, research facilities and associated infrastructure, a wide variety of buried and stored waste, and contaminated lands. In 2006, the Government of Canada adopted a new long-term strategy to deal with the nuclear legacy liabilities and initiated a five-year, $520 million (Canadian dollars) start-up phase, thereby creating the Nuclear Legacy Liabilities Program (NLLP). The objective of the long-term strategy is to safely and cost-effectively reduce risks and liabilities based on sound waste management and environmental principles in the best interests of Canadians. The five-year plan is directed at addressing health, safety and environmental priorities, accelerating the decontamination and demolition of shutdown buildings, and laying the groundwork for future phases of the strategy. It also includes public consultation to inform the further development of the strategy and provides for continued care and maintenance activities at the sites. The NLLP is being implemented through a Memorandum of Understanding between Natural Resources Canada (NRCan) and AECL whereby NRCan is responsible for policy direction and oversight, including control of funding, and AECL is responsible for carrying out the work and holding and administering all licences, facilities and lands. The paper summarizes achievements during the first three years of program implementation in the areas of decommissioning and dismantling; waste recovery and environmental restoration; the construction of enabling facilities to analyze, handle and store the legacy waste; and, planning for the long-term management of the radioactive waste. (authors)

  20. Marketing CANDU internationally

    International Nuclear Information System (INIS)

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

  1. Field test of wireless sensor network in the nuclear environment

    Energy Technology Data Exchange (ETDEWEB)

    Li, L., E-mail: lil@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Wang, Q.; Bari, A. [Univ. of Western Ontario, London, Ontario (Canada); Deng, C.; Chen, D. [Univ. of Electronic Science and Technology of China, Chengdu, Sichuan (China); Jiang, J. [Univ. of Western Ontario, London, Ontario (Canada); Alexander, Q.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-06-15

    Wireless sensor networks (WSNs) are appealing options for the health monitoring of nuclear power plants due to their low cost and flexibility. Before they can be used in highly regulated nuclear environments, their reliability in the nuclear environment and compatibility with existing devices have to be assessed. In situ electromagnetic interference tests, wireless signal propagation tests, and nuclear radiation hardness tests conducted on candidate WSN systems at AECL Chalk River Labs are presented. The results are favourable to WSN in nuclear applications. (author)

  2. The WRAPUP project: recovering information from the operation of WR-1

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S.; Mills, P.J.; Gibb, R.A. [ACSION (Canada)

    2013-07-01

    The WRAPUP (Whiteshell Reactor Applied Physics data Utilization and Preservation) Project was established in response to an inquiry received in 2011 May from staff at the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) who are involved with the International Reactor Physics Benchmark Experiments (IRPhE) Project. The IRPhE Project collects, archives and evaluates integral reactor physics experimental data from measurements performed at various research laboratories, worldwide, and manages a handbook of evaluated experimental data and benchmark simulations pertaining to them. The IRPhE Project wanted to know if AECL (Atomic Energy of Canada Limited) would be interested in contributing information from WR-1 (Whiteshell Reactor No. 1) physics experiments to its database. AECL - Chalk River Laboratories (CRL) is an active participant in the IRPhE Project, having contributed experimental data from the ZED-2 (Zero Energy Deuterium) reactor at the CRL, including CANDU (Canada Deuterium Uranium) related experimental data with the support of the CANDU Owners Group (COG), and having participated in the review of the contributions from other national laboratories (currently representing fourteen countries). AECL recognizes the value of this work to the global reactor physics community for testing the computer codes and nuclear data used in reactor simulations of every reactor type and thereby improving their reliability. (author)

  3. Annual report, 1990-1991

    International Nuclear Information System (INIS)

    At the beginning of the 1990/91 fiscal year the government of Canada announced that it would maintain the CANDU nuclear option, increase R and D funding for AECL Research, and authorize the start of negotiations to build a prototype CANDU 3 reactor. Later in the year AECL signed contracts with the Korea Electric Power Corporation to supply a second CANDU reactor for the Wolsung site. Consolidated net income was $7.8 million, after a $10.2 million loss in 1990. Revenue from nuclear power operations increased 11 percent to $187 million, with a 23 percent increase in the contribution from nuclear supply and services. Research and development expenditures rose to $293 million in 1991 from $259 million in 1990. The increase was mainly in cost-shared work on waste management, safety, health and environmental programs. Cost recovery revenue, principally from Ontario Hydro, increased to $87 million, reducing the federal government's share to 53 percent compared to 87 percent in 1985. Federal funding of R and D has been maintained at the 1990 level. The net expense of R and D operations was reduced to $11.0 million compared to $22.6 million in the previous year. Cash flow from all sources amounted to $47.9 million, leaving AECL with adequate working capital for the next year. In the future higher capital investment than previously anticipated will be required for waste management facilities associated with commercial isotope production. All figures are given in Canadian dollars

  4. A compendium of the data used with the SYVAC3-CC3 system model

    International Nuclear Information System (INIS)

    AECL is evaluating a concept for disposing of nuclear fuel waste from Canada's CANDU reactors deep in plutonic rock of the Canadian Shield. As part of this evaluation, AECL has developed models of the physicals, chemical, geological and biological processes that could occur in a sealed accessible environment over thousands of years. The mathematical models of the transport of radionuclides and toxic chemicals from nuclear fuel waste to the environment are incorporated into a computer model named the SYstems Variability Analysis Code, generation 3, and Canadian Concept model, generation 3 (SYVAC3-CC3). This report reproduces the data in the master database used by SYVAC3-CC3 for the postclosure assessment of deep laboratory and field studies conducted by AECL Research over the past fifteen years, including the investigations at an Underground Research Laboratory excavated to a depth of 450 meters in a large granitic batholith within the Whiteshell Research area near Lac du Bonnet, Manitoba; conceptual engineering studies; detailed analyses of specific features, events and processes; and published literature. The data represent characteristics of a hypothetical vault, certain geological characteristics of the Whiteshell Research area, and a general surface environment with a human population living a rural lifestyle on a portion of the Canadian Shield in central Canada. The data are stored in a master database, which is used with a suite of computer programs to create the input data files used by SYVAC3-CC3. (author). 19 refs., 11 tabs., 2 figs

  5. Tritium handling experience at Atomic Energy of Canada Limited

    Energy Technology Data Exchange (ETDEWEB)

    Suppiah, S.; McCrimmon, K.; Lalonde, S.; Ryland, D.; Boniface, H.; Muirhead, C.; Castillo, I. [Atomic Energy of Canad Limited - AECL, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-03-15

    Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritium powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.

  6. Progress report - physical sciences - physics division 1990 July 01 - December 31

    International Nuclear Information System (INIS)

    A completely new administrative structure of AECL Research was implemented on 1990 July 1. All of the basic physics programs, together with accelerator physics, radiation applications and most of the chemistry programs of AECL, have been placed in a new organizational unit called Physical Sciences. This unit also includes the management of the National Fusion Program. The research programs of Physical Sciences are grouped into three divisions: Chemistry, Physics and TASCC. Progress in each division will henceforth be reported on a twice-yearly basis. This report is the first of the new series to be issued by the Physics Division. Of special note within the period covered by this report was the successful acceleration of over 75 mA of protons to 600 keV in RFQ1 making it the highest current RFQ in the world. Our electron accelerator expertise has been recognized by the award of one of the R and D 100 awards for the IMPELA (10 MeV 50 kW) machine. Considerable activity was associated with bringing the new dual beam neutron spectrometer DUALSPEC to completion. This instrument has been jointly funded by AECL and NSERC through McMaster University and will be a central component of the national neutron scattering facility at NRU in the 1990's. A major effort was made with the writing of a Project Definition Document for installation of a cold neutron source at the most opportune time

  7. Control of blast overpressure and vibrations at the Underground Research Laboratory

    International Nuclear Information System (INIS)

    AECL Research (AECL) has constructed an Underground Research Laboratory (URL) as a facility for research and development in the Canadian Nuclear Fuel Waste Management Program. The objectives of the program are to develop and evaluate the technology to ensure safe, permanent disposal of Canada's nuclear fuel waste. Several multidisciplinary experiments and engineering demonstrations are planned for the URL over the next ten years. In 1989, AECL excavated a test room for the Buffer/Container Experiment at the 240 Level. The blasts were designed to limit vibration and overpressure damage because the excavation was located close to existing furnishings and services that were very susceptible to blast-induced vibration and overpressure. An experimental room, which contained sensitive instrumentation, was located within 30 m of the initial blasts. A concrete floor slab, timber curtains and a bulkhead were installed to protect furnishings and services from fly-rock and overpressure. Five of the initial blasts were monitored. This paper describes the results of the monitoring program and the effectiveness of the blast design, floor slab and timber curtains and bulkhead in reducing blast overpressure and vibrations at the blast site. It is shown that greater than a 20-fold reduction in both blast vibrations and air overpressures can be achieved with specific combinations of blast design, installation of timber curtains and construction of a concrete floor slab

  8. CANDU spent fuel dry storage interim technique

    International Nuclear Information System (INIS)

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

  9. Update on Canada's nuclear fuel waste management program

    International Nuclear Information System (INIS)

    The Canadian Nuclear Fuel Waste Management Program (CNFWMP) was launched in 1978 as a joint initiative by the governments of Canada and Ontario. Under the program, AECL has been developing and assessing a generic concept to dispose of nuclear fuel waste in plutonic rock of the Canadian Shield. The disposal concept has been referred for review under the Environmental Assessment and Review Process. AECL will submit an Environmental Impact Statement (EIS) to an Environmental Assessment Panel, which was appointed in late 1989. Hearings will be held in areas that have a particular interest in the concept and its application. At the end of the review, the Panel will make recommendations as to the acceptability of the concept and the course of future action. The federal government will decide on the next steps to be taken. In the spring of 1990 public open houses were held to tell prospective participants how to enter the process. Sessions designed to assist the Panel in determining the scope of the EIS took place in the autumn of 1990. In June 1991 the Panel issued for comment a set of draft guidelines for the EIS. More than 30 groups and individuals submitted comments. The final guidelines were issued in March 1992, and AECL expects to submit its EIS to the Panel in 1993. If the concept review is completed by 1995 and if the concept is approved, disposal could begin some time after 2025. (L.L.) (12 refs.)

  10. Update on Canada's fuel waste management program. Preparing for the environmental review of the concept

    International Nuclear Information System (INIS)

    The Canadian Nuclear Fuel Waste Management Program (CNFWMP) was established in 1978 as a joint initiative by the governments of Canada and Ontario. Under the program, AECL is responsible for developing and assessing a concept to dispose of nuclear fuel wastes in plutonic rock of the Canadian Shield. Ontario Hydro has advanced the technologies for interim storage and transportation of used fuel. The aim of the concept is to isolate the used fuel waste from the biosphere by a series of engineered and natural barriers. During the past fourteen years, AECL has carried out detailed studies on each component of this barrier system. A robust concept has been developed, with options for the choice of materials and designs for the different components. The disposal concept is being reviewed under the Environmental Assessment and Review Process (EARP). AECL is the 'Proponent' for this review, and will submit an Environmental Impact Statement (EIS) describing the disposal concept. The EIS has been written to respond to guidelines issued by the Environmental Assessment Panel responsible for carrying out the review. The future direction of the CNFWMP will depend on the recommendations of the Panel and on the resulting governmental decisions on the appropriate next steps. If the concept review is completed by 1996, as currently expected, and the concept is approved, the many steps that would be involved with siting and construction of a disposal facility, mean that disposal would not begin before about 2025. (J.P.N.)

  11. Vendor provision for operations and maintenance : a route to nuclear excellence

    International Nuclear Information System (INIS)

    The CANDU pressurized heavy water design has been established over three decades of operation. Historically CANDU has been developed by a successful partnership between AECL, the designer and developer, and the operating utilities. Maintaining good performance of a nuclear power plant is primarily the responsibility of the owner operator; but there is a common interest between the design organization and the operator in achieving good performance. AECL is emphasizing activities and partnerships to provide engineering and Research and Development (R and D) to reduce operating costs for new plants and support existing plants. Among these are: a comprehensive, structured process to deal with project, operations and other feedback; teams to study issues with impact on capacity factor and OM and A cost and identify initiatives in response; an integrated plant life management program in conjunction with CANDU 6 utilities; and strategic information technology products installed at the plants in support of improved station operation, sponsored by the CANDU Owner's Group. The breadth of these programs gives AECL a unique ability to link between support to operating plants, and the development of improved Nuclear Power Plant (NPP) designs for the future. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  13. Progress on Developing an Interface Program between WIMSD-5B and RFSP

    Energy Technology Data Exchange (ETDEWEB)

    You, Guk Jong; Kim, Won Young; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    WIMS (Winfrith Improved Multigroup Scheme) code is a multi-group transport code for the reactor lattice calculations which includes a fuel depletion or burn-up routine. The code, created at the United Kingdom Atomic Energy Authority Establishment, Winfrith (AEEW), was intended to perform the lattice calculations with an acceptable accuracy for the analysis of the experiments in a wide range of geometries. As one of its branches, WIMSD-5B is a code which was released from OECD/NEA Data Bank in 1998 and now has been used widely for thermal research and power reactor calculation. Also one of WIMS codes, WIMS-AECL, has been developed by AECL in Canada as an independent version of the original AEEW code. While WIMS-AECL produces a data file which can generate the information required by other code such as RFSP, WIMSD-5B does not. The data file is used for the reactor analysis by WIMSAECL in connection with RFSP. This study is to develop an interface data file (Tape 16) of WIMSD-5B with RFSP and to develop a process utility to provide the group collapsing and cell average cross-section generation for a CANDU-6 core analysis on the WINDOW system. With this utility, the physics analysis of a CANDU-6 reactor will be performed by RFSP code using the lattice parameters generated by WIMSD-5B.

  14. A human reliability assessment screening method for the NRU upgrade project

    International Nuclear Information System (INIS)

    The National Research Universal (NRU) reactor is a 130MW, low pressure, heavy water cooled and moderated research reactor. The reactor is used for research, both in support of Canada's CANDU development program, and for a wide variety of other research applications. In addition, NRU plays an important part in the production of medical isotopes, e.g., generating 80% of worldwide supplies of Molybdenum-99. NRU is owned and operated by Atomic Energy of Canada Ltd. (AECL), and is currently undergoing upgrading as part of AECL's continuing commitment to operate their facilities in a safe manner. As part of these upgrades both deterministic and probabilistic safety assessments are being carried out. It was recognized that the assignment of Human Error Probabilities (HEPs) is an important part of the Probabilistic Safety Assessment (PSA) studies, particularly for a facility whose design predates modern ergonomic practices, and which will undergo a series of backfitted modifications whilst continuing to operate. A simple Human Reliability Assessment (HRA) screening method, looking at both pre- and post-accident errors, was used in the initial safety studies. However, following review of this method within AECL and externally by the regulator, it was judged that benefits could be gained for future error reduction by including additional features, as later described in this document. The HRA development project consisted of several stages; needs analysis, literature review, development of method (including testing and evaluation), and implementation. This paper discusses each of these stages in further detail. (author)

  15. International standard problem (ISP) no. 41 follow up exercise: Containment iodine computer code exercise: parametric studies

    Energy Technology Data Exchange (ETDEWEB)

    Ball, J.; Glowa, G.; Wren, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ewig, F. [GRS Koln (Germany); Dickenson, S. [AEAT, (United Kingdom); Billarand, Y.; Cantrel, L. [IPSN (France); Rydl, A. [NRIR (Czech Republic); Royen, J. [OECD/NEA (France)

    2001-11-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I{sup -} concentration. The codes used in this exercise were IODE(IPSN), IODE(NRIR), IMPAIR(GRS), INSPECT(AEAT), IMOD(AECL) and LIRIC(AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained front intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility, (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (author)

  16. 1979-80 annual report

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. added a fifth semi-autonomous unit, Atomic Energy of Canda International Company, in June 1979, to assist in the company's pursuit of world wide acceptance of the CANDU system. Evaluations of reactor safety following the Three Mile Island accident, public hearings in Ontario, and the report of the Porter Commission all reaffirm the safety of Canadian nuclear power plants. Continuing efforts were made to bring information on nuclear power to the public. AECL continued to participate in the International Fuel Cycle Evaluation; INFCE findings confirm the competitiveness of the CANDU reactor. Emphasis in the AECL research program was on safety, safeguards, health effects of radiation, waste management and new applications for nuclear power. The Radiochemical Company had sales of $49 million, with 97% of its business being in the export market. The Engineering Company is working on eight major projects totalling 12000 MW(e) as well as providing consulting service for the CANDU stations already in operation. The Chemical Company produced over 5000000 kg of heavy water. AECL revenues were $497.1 million in 1979, an increase of 41.6 percent over the previous year. Research and development expenditures increased 2 percent to $127.2 million. Net income for the year increased to $11.2 million from $5.2 million for the previous year. (LL)

  17. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

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

  18. Incomplete data on the Canadian cohort may have affected the results of the study by the International Agency for Research on Cancer on the radiogenic cancer risk among nuclear industry workers in 15 countries

    Energy Technology Data Exchange (ETDEWEB)

    Ashmore, J Patrick [Ponsonby and Associates, Manotick, ON (Canada); Gentner, Norman E; Osborne, Richard V, E-mail: osborner@magma.c [Ranasara Consultants Inc., PO Box 1116, Deep River, ON (Canada)

    2010-06-15

    In 1995 the International Agency for Research on Cancer (IARC) completed a study that involved nuclear workers from facilities in the USA, UK and Canada. The only significant, though weak, dose-related associations found were for leukaemia and multiple myeloma. The results for the Canadian cohort, which comprised workers from the facilities of Atomic Energy of Canada Limited (AECL), were compatible with those for the other national cohorts. In 2005, IARC completed a further study, involving nuclear workers from 15 countries, including Canada. In these results, the dose-related risk for leukaemia was not significant but the prominent finding was a statistically significant excess relative risk per sievert (ERR Sv{sup -1}) for 'all cancers excluding leukaemia'. Surprisingly, the risk ascribed to the Canadian cohort for all cancers excluding leukaemia, driven by the AECL sub-cohort, was significantly higher than the risk estimate for the 15-country cohort as a whole. We have attempted to identify why the results for the AECL cohort were so discrepant and had such a remarkable influence on the 15-country risk estimate. When considering the issues associated with data on the AECL cohorts and their handling, we noted a striking feature: a major change in outcome of studies that involved Canadian nuclear workers occurred concomitantly with the shift to when data from the National Dose Registry (NDR) of Canada were used directly rather than data from records at AECL. We concluded that an important contributor to the considerable upward shift in apparent risk in the 15-country and other Canadian studies that have been based on the NDR probably relates to pre-1971 data and, in particular, the absence from the NDR of the person-years of workers who had zero doses in the calendar years 1956 to 1970. Our recommendation was for there to be a comprehensive evaluation of the risks from radiation in nuclear industry workers in Canada, organisation by organisation, in

  19. Safe operation of the NRU research reactor now and beyond 2021

    International Nuclear Information System (INIS)

    This paper will describe the approach that has been taken by Atomic Energy of Canada Limited (AECL) to ensure that the National Research Universal (NRU) reactor designed in the 1940's continues to remain safe and reliable to operate now and for the near future (2021 and beyond). This paper focuses on two major projects, the NRU Upgrades Project undertaken in the 1990's and the Integrated Safety Review (ISR) resulting in the Integrated Implementation Plan (IIP) that is currently underway. Through the NRU Upgrades Project, AECL was able to identify areas for safety improvement and implement changes in the field. Following the NRU Upgrades Project, AECL was able to demonstrate that for design basis accidents that the reactor was able to meet the four basic safety requirements namely:- · It shall be possible to shut down the reactor and maintain it in that state indefinitely; · The capability of removing decay heat from the fuel during this shut down period shall be maintained; · The confinement structure shall continue to be capable of limiting radioactivity release; and · Continuous monitoring of reactor safety functions shall remain available. The NRU Upgrades Project enabled AECL to continue to operate the NRU reactor beyond the year 2000 but it was recognised in 2008 that if operations were to continue up to and beyond 2021 then another assessment was warranted. This assessment resulted in the ISR project. The ISR project consisted of reviewing the NRU design against current codes and standards and, where applicable, addressing gaps identified. This project identified not only gaps in the analysis basis for NRU, it also identified the need to replace ageing equipment that was reaching the end of its design life. The findings of the ISR project have been captured in the IIP; IIP has enabled AECL to prioritise equipment replacement to enable continued safe and reliable operation of the NRU reactor beyond 2021. The paper demonstrates that, in order to safely

  20. Summary of the Environmental Impact Statement on the concept for disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    This is the Summary of the Environmental Impact Statement (EIS) prepared by Atomic Energy of Canada Limited (AECL) on the concept for disposal of Canada's nuclear fuel waste. The proposed concept is a method for geological disposal, based on a system of natural and engineered barriers. The EIS provides information requested by the Environmental Assessment Panel reviewing the disposal concept and presents AECL's case for the acceptability of the concept. The introductory chapter of this Summary provides background information on several topics related to nuclear fuel waste, including current storage practices for used fuel, the need for eventual disposal of nuclear fuel waste, the options for disposal, and the reasons for Canada's focus on geological disposal. Chapter 2 describes the concept for disposal of nuclear fuel waste. Because the purpose of implementing the concept would he to protect human health and the natural environment far into the future, we discuss the long-term performance of a disposal system and present a case study of potential effects on human health and the natural environment after the closure of a disposal facility. The effects and social acceptability of disposal would depend greatly on how the concept was implemented. Chapter 3 describes AECL's proposed approach to concept implementation. We discuss how the public would be involved in implementation; activities that would be undertaken to protect human health, the natural environment, and the socio-economic environment; and a case study of the potential effects of disposal before the closure of a disposal facility. The last chapter presents AECL's Conclusion, based on more than 15 years of research and development, that implementation of the disposal concept represents a means by which Canada can safely dispose of its nuclear fuel waste. This chapter also presents AECL's recommendation that Canada progress toward disposal of its nuclear fuel waste by undertaking the first stage of concept

  1. Computer based plant display and digital control system of Wolsong NPP Tritium Removal Facility

    International Nuclear Information System (INIS)

    The Wolsong Tritium Removal Facility (WTRF) is an AECL-designed, first-of-a-kind facility that removes tritium from the heavy water that is used in systems of the CANDUM reactors in operation at the Wolsong Nuclear Power Plant in South Korea. The Plant Display and Control System (PDCS) provides digital plant monitoring and control for the WTRF and offers the advantages of state-of-the-art digital control system technologies for operations and maintenance. The overall features of the PDCS will be described and some of the specific approaches taken on the project to save construction time and costs, to reduce in-service life-cycle costs and to improve quality will be presented. The PDCS consists of two separate computer sub-systems: the Digital Control System (DCS) and the Plant Display System (PDS). The PDS provides the computer-based Human Machine Interface (HMI) for operators, and permits efficient supervisory or device level monitoring and control. A System Maintenance Console (SMC) is included in the PDS for the purpose of software and hardware configuration and on-line maintenance. A Historical Data System (HDS) is also included in the PDS as a data-server that continuously captures and logs process data and events for long-term storage and on-demand selective retrieval. The PDCS of WTRF has been designed and implemented based on an off-the-self PDS/DCS product combination, the Delta-V System from Emerson. The design includes fully redundant Ethernet network communications, controllers, power supplies and redundancy on selected I/O modules. The DCS provides field bus communications to interface with 3rd party controllers supplied on specialized skids, and supports HART communication with field transmitters. The DCS control logic was configured using a modular and graphical approach. The control strategies are primarily device control modules implemented as autonomous control loops, and implemented using IEC 61131-3 Function Block Diagram (FBD) and Structured

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

    International Nuclear Information System (INIS)

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

  3. Maintaining Quality in a Decommissioning Environment

    International Nuclear Information System (INIS)

    The decommissioning of AECL's Whiteshell Laboratories is Canada's largest nuclear decommissioning project to date. This research laboratory has operated for forty years since it was set up in 1963 in eastern Manitoba as the Whiteshell Nuclear Research Establishment, complete with 60 MW(Th) test reactor, hot cells, particle accelerators, and multiple large-scale research programs. Returning the site to almost complete green state will require several decades of steady work in combination with periods of storage-with-surveillance. In this paper our approach to maintaining quality during the long decommissioning period is explained. In this context, 'quality' includes both regulatory aspects (compliance with required standards) and business aspects (meeting the customers' needs and exceeding their expectations). Both aspects are discussed, including examples and lessons learned. The five years of development and implementation of a quality assurance program for decommissioning the WL site have led to a number of lessons learned. Many of these are also relevant to other decommissioning projects, in Canada and elsewhere: - Early discussions with the regulator can save time and effort later in the process; - An iterative process in developing documentation allows for steady improvements and input throughout the process; - Consistent 2-way communication with staff regarding the benefits of a quality program assists greatly in adoption of the philosophy and procedures; - Top-level management must lead in promoting quality; - Field trials of procedures ('beta testing') ensures they are easy to use as well as useful. Success in decommissioning the Whiteshell Laboratories depends on the successful implementation of a rigorous quality program. This will help to ensure both safety and efficiency of all activities on site, from planning through execution and reporting. The many aspects of maintaining this program will continue to occupy quality practitioners in AECL, reaping

  4. Evaluation of spiral wound reverse osmosis for four radioactive waste processing applications

    International Nuclear Information System (INIS)

    A pilot-scale spiral wound reverse osmosis rig was used to treat four significantly different radioactive waste streams, three of which were generated at the Chalk River Laboratories at AECL. These streams included: 1. A chemical decontamination (CD/DC) waste stream which is routinely treated by the plant-scale membrane system at CRL; 2. Reactor waste which is a dilute radioactive waste stream (containing primarily tritium and organic acids), and it an effluent from the operating reactors at AECL; 3. An ion exchange regenerant waste stream which contains a mixture of stream (1) (CD/DC), blended with secondary waste from ion exchange regeneration; 4. Boric acid simulated waste which is a by-product waste of the PWR reactors. This was the only stream treated that was not generated as a waste liquid at AECL. For the first three streams specified above, reverse osmosis was used to remove chemical and radiochemical impurities from the water with efficiencies usually exceeding 99%. In these three cases the 'permeate' or clean water was the product of the process. In the case of stream 4, reverse osmosis was used in a recovery application for the purpose of recycling boric acid back to the reactor, with the concentrate being the 'product'. Reverse osmosis technology was successfully demonstrated for the treatment of all four streams. Prefiltration and oxidation (with photocatalytic continuous oxidation technology) were evaluated as pretreatment alternatives for streams 1, 2, and 3. The results indicated that the effective crossflow velocity through and membrane vessel was more important in determining the extent of membrane fouling than the specific pretreatment strategy employed. (author)

  5. Atomic solution? The nuclear option is again touted for Alberta's oilsands

    International Nuclear Information System (INIS)

    As early as the 1950s, nuclear blasts were considered as a potential way to unlock the huge potential contained in northern Alberta's tarsands deposits. A plan, which came close to receiving government sanction, was devised to detonate an atomic device underground to melt the bitumen. Today, public reaction would not permit any revival of such a plan, but a less explosive, modern-day solution is being considered. Atomic Energy of Canada Ltd (AECL) is promoting a new medium-sized nuclear reactor called the Advanced CANDU Reactor (ACR) for use in the oilsands industry. AECL recently commissioned a$35,000 study into the economics of atomic energy compared to natural gas to produce large amounts of steam that is needed to separate oil from the sand. Preliminary results suggest that nuclear power may be a viable option for oilsand extraction, requiring much less energy than the currently used steam assisted gravity drainage process (SAGD). In addition, nuclear power could solve the problem of projected greenhouse gas (GHG) emissions. Unlike natural gas-fired cogeneration facilities , nuclear energy does not emit GHGs. The Canadian Energy Research Institute (CERI) is examining the economics of nuclear energy for oilsand extraction. AECL claims the ACR-700 is competitive with the best-advanced gas-fired technology based on projections for 2010 and beyond. It will also move to light water from heavy water cooling by using slightly enriched uranium, thereby extending fuel life and reducing operating costs. Public perception, however, may be the biggest challenge. Opponents argue that storage and disposal of spent fuel rods still needs to be addressed. (author)

  6. Atomic solution? The nuclear option is again touted for Alberta's oilsands

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M.

    2003-03-01

    As early as the 1950s, nuclear blasts were considered as a potential way to unlock the huge potential contained in northern Alberta's tarsands deposits. A plan, which came close to receiving government sanction, was devised to detonate an atomic device underground to melt the bitumen. Today, public reaction would not permit any revival of such a plan, but a less explosive, modern-day solution is being considered. Atomic Energy of Canada Ltd. (AECL) is promoting a new medium-sized nuclear reactor called the Advanced Candu Reactor (ACR) for use in the oilsands industry. AECL recently commissioned a $35,000 study into the economics of atomic energy compared to natural gas to produce large amounts of steam that is needed to separate oil from the sand. Preliminary results suggest that nuclear power may be a viable option for oilsand extraction, requiring much less energy than the currently used steam assisted gravity drainage process (SAGD). In addition, nuclear power could solve the problem of projected greenhouse gas (GHG) emissions. Unlike natural gas-fired cogeneration facilities, nuclear energy does not emit GHGs. The Canadian Energy Research Institute (CERI) is examining the economics of nuclear energy for oilsand extraction. AECL claims the ACR-700 is competitive with the best-advanced gas-fired technology based on projections for 2010 and beyond. It will also move to light water from heavy water cooling by using slightly enriched uranium, thereby extending fuel life and reducing operating costs. Public perception, however, may be the biggest challenge. Opponents argue that storage and disposal of spent fuel rods still needs to be addressed. 1 fig.

  7. CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

    International Nuclear Information System (INIS)

    Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y+ < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

  8. Revised delayed neutron data for Pickering NGS B

    International Nuclear Information System (INIS)

    Revised delayed neutron fractions and constants were calculated specifically for Pickering NGS B using the latest available delayed neutron data for fissionable isotopes, the currently recommended CANDU delayed photoneutron data, and the formulae provided by Laughton. Burnup-dependent number densities of the fissionable isotopes were computed with WIMS-AECL-IST. Validation and assessment of the new fractions and constants was performed by comparison of historical operating data with point kinetics simulation using the new values. Recommendations are made regarding possible improvements to the delayed neutron data. (author)

  9. ACR-1000: Operator - based development

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU* reactors to establish Generation III+ Advanced CANDU ReactorTM (ACRTM) technology. The ACR-1000TM nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability. The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution. Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDUTM technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants. As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of main program

  10. Semi-annual status report of the Canadian Nuclear Fuel Waste Management Program, April 1--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Wright, E.D. [comp.

    1992-02-01

    This report is the eleventh in a series of semi-annual status reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. it describes progress achieved in the three major subprograms, engineered systems, natural systems and performance assessment, from 1991 April 1 to September 30. It also gives a brief description of the activities being carried out in preparation for the public and governmental review of the disposal concept. Since 1987, this program has been jointly funded by AECL and Ontario Hydro under the auspices of the CANDU Owners Group (COG).

  11. Hardware replacements and software tools for digital control computers

    International Nuclear Information System (INIS)

    Technological obsolescence is an on-going challenge for all computer use. By design, and to some extent good fortune, AECL has had a good track record with respect to the march of obsolescence in CANDU digital control computer technology. Recognizing obsolescence as a fact of life, AECL has undertaken a program of supporting the digital control technology of existing CANDU plants. Other AECL groups are developing complete replacement systems for the digital control computers, and more advanced systems for the digital control computers of the future CANDU reactors. This paper presents the results of the efforts of AECL's DCC service support group to replace obsolete digital control computer and related components and to provide friendlier software technology related to the maintenance and use of digital control computers in CANDU. These efforts are expected to extend the current lifespan of existing digital control computers through their mandated life. This group applied two simple rules; the product, whether new or replacement should have a generic basis, and the products should be applicable to both existing CANDU plants and to 'repeat' plant designs built using current design guidelines. While some exceptions do apply, the rules have been met. The generic requirement dictates that the product should not be dependent on any brand technology, and should back-fit to and interface with any such technology which remains in the control design. The application requirement dictates that the product should have universal use and be user friendly to the greatest extent possible. Furthermore, both requirements were designed to anticipate user involvement, modifications and alternate user defined applications. The replacements for hardware components such as paper tape reader/punch, moving arm disk, contact scanner and Ramtek are discussed. The development of these hardware replacements coincide with the development of a gateway system for selected CANDU digital control

  12. Management of the installation of a 10 MeV, 50 kW electron-beam irradiator

    International Nuclear Information System (INIS)

    An IMPELA-10/50 electron-beam irradiator has been installed by AECL Accelerators in Iotron Industries' service centre near Vancouver. Construction of the facility, installation of the accelerator and conveyor, and commissioning to the full rated power of 50 kW were completed in 12 months. Iotron began commercial irradiation immediately and the first continuous operation achieved 250 hours of production in 12 days. The engineering, production and project management organizations and activities to complete the on-schedule installation and commissioning are reviewed. (author). 3 refs., 2 tabs., 1 fig

  13. Analysis of the impact of coolant density variations in the high efficiency channel of a pressure tube super critical water reactor

    International Nuclear Information System (INIS)

    The Pressure Tube (PT) Supercritical Water Reactor (SCWR) is based on a light water coolant operating at pressures above the thermodynamic critical pressure; a separate low temperature and low pressure moderator. The coolant density changes by an order of magnitude depending on its local enthalpy in the porous ceramic insulator tube. This causes significant changes in the neutron transport characteristics, axially and radially, in the fuel channel. This work performs lattice physics calculations for a 78-element Pu-Th fuel at zero burnup and examines the effect of assumptions related to coolant density in the radial direction of a HEC, using the neutron transport code WIMS-AECL. (author)

  14. Advances in radiation processing of polymeric materials

    International Nuclear Information System (INIS)

    In this paper we review recent advances in industrial applications of electron-beam irradiation in the field of polymer processing at the Takasaki Radiation Chemistry Research Establishment (TRCRE) of JAERI (Japan Atomic Energy Research Institute), and the Whiteshell Laboratories of AECL Research, Canada. Irradiation of a substrate with ionizing radiation produces free radicals through ionization and excitation events. The subsequent chemistry of these radicals is used in radiation processing as a substitute for conventional processing techniques based on heating and/or the addition of chemicals. The advantages of radiation processing include the formation of novel products with desirable material properties, favourable overall process economics and, often, environmental benefits

  15. Validation study on reliability analysis of main safety system in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Cho, Chang Keun; Kim, Yong Hui; Kim, Tae Hyeong; Hong, Seo Kee; Park, Keon Woo; Park, Chang Jea [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Cheong, Woo Sik [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Moon Kyu [KEPRI, Taejon (Korea, Republic of)

    1993-12-15

    The scope and contents of this validation study are to review the design changes of the four main safety systems in Wolsong 2/3/4 Nuclear Power Plants, to review the consideration of the above design changes in the AECL reports, the structure of fault trees, and the data base used in the quantification of the fault trees, to quantify the unavailabilities of main safety systems and check them if they meet the requirements, and to recommend desirable design changes in the emergency core cooling system to reduce the unavailability.

  16. New nuclear power plants for Ontario

    International Nuclear Information System (INIS)

    Towards the end of this year the Ontario government will select the technology for its future nuclear power plants. To clarify the differences between the contending reactors I have put together the following quick overview. Ontario's requirement is for a stand-alone two-unit nuclear power plant to provide around 2,000 to 3,500 MWe of baseload generating capacity at a site to he specified with an option for one or two additional units. It is likely that the first units will be located at either the Darlington site near Bowmanville or the Bruce site near Kincardine. However the output from the Bruce site is presently transmission constrained. All nuclear-electric generation in Ontario comes from Atomic Energy of Canada Limited's (AECL) CANDU reactors at Pickering, Darlington and Bruce. The contenders are, AECL's 1085 MWe (net) ACR-1000 (Advanced CANDU Reactor), Westinghouse Electric Company's 1117 MWe (net) AP1000 (Advanced Passive), AREVA NP's 1600 MWe (net) U.S. EPR (United States Evolutionary Pressurized Reactor) and the 1550 MWe (net) GE Hitachi Nuclear Energy's ESBWR (Economic and Simplified Boiling Water Reactor). Westinghouse has Toshiba as a majority shareholder, AREVA has the government of France as a majority shareholder and GE-Hitachi has GE as the major shareholder. AECL is a federal crown corporation and is part of Team CANDU consisting of Babcock and Wilcox Canada, GE-Hitachi Nuclear Energy Canada Inc., Hitachi Canada Limited and SNC-Lavalin Nuclear Inc. Generally the engineering split in Team CANDU would be, AECL, Mississauga, Ontario, responsible for the design of the nuclear steam plant including reactor and safety systems; Babcock and Wilcox Canada, Cambridge, Ontario, responsible for supply of the steam generators and other pressure retaining components; GE-Hitachi Nuclear Energy Canada Inc., Peterborough, Ontario for the fuel handling equipment; Hitachi Canada Limited, Mississauga, for the balance of plant steam to electricity conversion

  17. Build your own Candu reactor

    International Nuclear Information System (INIS)

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

  18. Review of release models used in source-term codes

    International Nuclear Information System (INIS)

    Throughout this reviews, the limitations of current release models are identified and ways of improving them suggested, By incorporation recent experimental results, recommendations for future release modeling activities can be made. All release under review were compared with respect to the following six items: scenario, assumptions, mathematical formulations, solution method, radioactive decay chain considered, and geometry. The following nine models are considered for review: SOTEC and SCCEX (CNWRA), DOE/INTERA, TSPA (SNL), Vault Model (AECL), CCALIBRE (SKI), AREST (PNL), Risk Assessment (EPRI), TOSPAC (SNL). (author)

  19. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  20. Configuration management for CANDU feeder refurbishment

    International Nuclear Information System (INIS)

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

  1. Review of the nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Progress over the previous year in the nuclear fuel waste management program is reviewed. Universities, industry and consultants have become increasingly involved, and the work is being overseen by a Technical Advisory Committee. The program has also been investigated by Ontario's Porter Commission and Select Committe on Ontario Hydro Affairs. A public information program has been extended to cover most of the Canadian Shield region of Ontario. Ontario Hydro is studying spent fuel storage and transportation, while AECL is covering immobilization of spent fuel or processing wastes, geotechnical and geochemical research in the laboratory and in the field, design of disposal facilities, and environmental and safety assessments. (L.L.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  4. Using Fortran modules to design and develop modular reactor analysis software components

    International Nuclear Information System (INIS)

    This paper presents a design for heavy and light water property calculation routines using object oriented design techniques. The designed routines are part of a new thermalhydraulics code being developed by Atomic Energy of Canada Limited (AECL). We demonstrate how application of object-oriented methodology leads to Fortran modules that use new features of Fortran 95 effectively. We also present performance metrics. This paper contributes in two ways. Firstly, it provides a methodology that can be used to systematically identify objects, assign responsibilities to the objects and establish the interaction among objects. Secondly, it shows how the designs can be communicated using the Unified Modeling Language (UML). (author)

  5. A general description of the NRX reactor

    International Nuclear Information System (INIS)

    The NRX Reactor structure, equipment and experimental facilities are described. The purpose of the various components is explained using photographs and diagrams as much as possible. Dimensions are given so that the reader can visualize the relative sizes of the components. The report is meant to be an introduction to the NRX Design and Operating Manuals, from which detailed information can be obtained. It is expected that the report will be of value to trainee NRX Reactor Operations personnel and to those persons who require only a general knowledge of the reactor. A bibliography of AECL reports pertaining to NRX is given. (author)

  6. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

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

  7. Proceedings of DUPIC fuel workshop 97

    International Nuclear Information System (INIS)

    The researchers discuss the technical aspects of DUPIC fuel fabrication in the workshop as follows; 1) The DUPIC fuel development program in KAERI 2) AECL's progress in developing the DUPIC fuel fabrication process 3) Mechanical decladding 4) Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept 5) Assessment of DUPIC fuel compatibility with CANDU-6 6) The development of combination software for spent PWR fuel to fabricate the homogeneous DUPIC fuel 7) Thermodynamic properties of the DUPIC fuel and its performance 8) Captural properties of cesium and ruthenium 9) A secondary fuel removal process : Plasma processing 10) Technology development for DUPIC process safeguards

  8. Biological effects of ionizing radiation

    International Nuclear Information System (INIS)

    In this review radiation produced by the nuclear industry is placed into context with other sources of radiation in our world. Human health effects of radiation, derivation of standards and risk estimates are reviewed in this document. The implications of exposing the worker and the general population to radiation generated by nuclear power are assessed. Effects of radiation are also reviewed. Finally, gaps in our knowledge concerning radiation are identified and current research on biological effects, on environmental aspects, and on dosimetry of radiation within AECL and Canada is documented in this report. (author)

  9. Management of legacy spent nuclear fuel wastes at the Chalk River Laboratories: the challenges and innovative solutions implemented

    International Nuclear Information System (INIS)

    AECL has operated research reactors at the Chalk River Laboratories (CRL) site since 1947, for the purpose of nuclear energy and scientific research and for the production of radioisotopes. During the 1950s and 60s, a variety of spent nuclear fuel wastes were produced by irradiating metallic uranium and other prototype fuels. These legacy waste fuels were initially stored in water-filled fuel storage bays for a period of several years before being placed in storage containers and transferred to the CRL Waste Management Areas (WMAs), where they have been stored in below-grade, vertical cylindrical steel and concrete structures called 'tile holes'. (author)

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

    International Nuclear Information System (INIS)

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

  11. Analysis specifications for the CC3 biosphere model biotrac

    Energy Technology Data Exchange (ETDEWEB)

    Szekely, J.G.; Wojciechowski, L.C.; Stephens, M.E.; Halliday, H.A.

    1994-12-01

    The CC3 (Canadian Concept, generation 3) model BIOTRAC (Biosphere Transport and Consequences) describes the movement in the biosphere of releases from an underground disposal vault, and the consequent radiological dose to a reference individual. Concentrations of toxic substances in different parts of the biosphere are also calculated. BIOTRAC was created specifically for the postclosure analyses of the Environmental Impact Statement that AECL is preparing on the concept for disposal of Canada`s nuclear fuel waste. The model relies on certain assumptions and constraints on the system, which are described by Davis et al. Accordingly, great care must be exercised if BIOTRAC is used for any other purpose.

  12. CRNL library serials list

    International Nuclear Information System (INIS)

    A list of 1900 serial publications (periodicals, society transactions and proceedings, annuals and directories, indexes, newspapers, etc.) is presented with volumes and years held by the Main Library. This library is the largest in AECL as well as one of the largest scientific and technical libraries in North America, and functions as a Canadian resource for nuclear information. A main alphabetical list is followed by broad subject field lists representing research interests, and lists of abstract and index serials, general bibliographic serials, conference indexes, press releases, English translations, and original language journals

  13. Discussing spent nuclear fuel in high school classrooms: addressing public fears through early education

    Energy Technology Data Exchange (ETDEWEB)

    Winkel, S. [Deep River Science Academy, 20 Forest Ave. P.O. Box 600, Deep River, Ontario K0J 1P0 (Canada); Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Sullivan, J.; Jones, S.; Sullivan, K. [Deep River Science Academy, 20 Forest Ave. P.O. Box 600, Deep River, Ontario K0J 1P0 (Canada); Hyland, B.; Pencer, J.; Colton, A. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    The Inreach program combines the Deep River Science Academy (DRSA) 'learning through research' approach with state of the art communication technology to bring scientific research to high school classrooms. The Inreach program follows the DRSA teaching model where a university student tutor works on a research project with scientific staff at AECL's Chalk River Laboratories. Participating high school classes are located across Canada. The high school students learn about the ongoing research activities via weekly web conferences. In order to engage the students and encourage participation in the conferences, themed exercises linked to the research project are provided to the students. The DRSA's Inreach program uses a cost-effective internet technology to reach a wide audience, in an interactive setting, without anyone leaving their desks or offices. An example Inreach research project is presented here: an investigation of the potential of the Canadian supercritical water cooled reactor (SCWR) concept to burn transuranic elements (Np, Pu, Am, Cm) to reduce the impact of used nuclear fuel. During this project a university student worked with AECL (Atomic Energy of Canada Limited) researchers on technical aspects of the project, and high school students followed their progress and learned about the composition, hazards, and disposition options for used nuclear fuel. Previous projects included the effects of tritium on cellular viability and neutron diffraction measurement of residual stresses in automobile engines.

  14. Multi-purpose hydrogen isotopes separation plant design

    Energy Technology Data Exchange (ETDEWEB)

    Boniface, H.A.; Gnanapragasam, N.V.; Ryland, D.K.; Suppiah, S.; Castillo, I. [Atomic Energy of Canada Limited - AECL, Chalk River, ON (Canada)

    2015-03-15

    There is a potential interest at AECL's Chalk River Laboratories to remove tritium from moderately tritiated light water and to reclaim tritiated, downgraded heavy water. With only a few limitations, a single CECE (Combined Electrolysis and Catalytic Exchange) process configuration can be designed to remove tritium from heavy water or light water and upgrade heavy water. Such a design would have some restrictions on the nature of the feed-stock and tritium product, but could produce essentially tritium-free light or heavy water that is chemically pure. The extracted tritium is produced as a small quantity of tritiated heavy water. The overall plant capacity is fixed by the total amount of electrolysis and volume of catalyst. In this proposal, with 60 kA of electrolysis a throughput of 15 kg*h{sup -1} light water for detritiation, about 4 kg*h{sup -1} of heavy water for detritiation and about 27 kg*h{sup -1} of 98% heavy water for upgrading can be processed. Such a plant requires about 1,000 liters of AECL isotope exchange catalyst. The general design features and details of this multi-purpose CECE process are described in this paper, based on some practical choices of design criteria. In addition, we outline the small differences that must be accommodated and some compromises that must be made to make the plant capable of such flexible operation. (authors)

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

    Directory of Open Access Journals (Sweden)

    Seung Yeol Yoo

    2015-01-01

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

  16. Features, events, processes, and safety factor analysis applied to a near-surface low-level radioactive waste disposal facility

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, M.E.; Dolinar, G.M.; Lange, B.A. [Atomic Energy of Canada Limited, Ontario (Canada)] [and others

    1995-12-31

    An analysis of features, events, processes (FEPs) and other safety factors was applied to AECL`s proposed IRUS (Intrusion Resistant Underground Structure) near-surface LLRW disposal facility. The FEP analysis process which had been developed for and applied to high-level and transuranic disposal concepts was adapted for application to a low-level facility for which significant efforts in developing a safety case had already been made. The starting point for this process was a series of meetings of the project team to identify and briefly describe FEPs or safety factors which they thought should be considered. At this early stage participants were specifically asked not to screen ideas. This initial list was supplemented by selecting FEPs documented in other programs and comments received from an initial regulatory review. The entire list was then sorted by topic and common issues were grouped, and issues were classified in three priority categories and assigned to individuals for resolution. In this paper, the issue identification and resolution process will be described, from the initial description of an issue to its resolution and inclusion in the various levels of the safety case documentation.

  17. Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

    1997-08-01

    Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

  18. Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

    International Nuclear Information System (INIS)

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% 235U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% 235U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO2 powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report

  19. Advanced in fuel channel gauging tool - instrumenting a SLAR tool for dual purpose

    International Nuclear Information System (INIS)

    This paper describes the latest inspection technology to be implemented on a SLARette tool. In 2002, a gauging module was developed and qualified to replace the SLARette tool's blister module. This gauging module had five ultrasonic transducers for diameter creeping and wall thickness measurements. The results of the 2002 SLARette campaign were excellent; the data obtained came within ten microns of that collected with a CANDE tool in 2003. Gentilly-2 decided to continue developing gauging techniques and apparatus to be mounted on a SLARette tool. The new front-end module incorporates both sag measurement and revolutionary pressure tube (PT)/calandria tube (CT) gap modules. Advancements were made along several lines: (1) selection of a radiation-resistant sag module, (2) development of a sag simulator, (3) improvement of AECL gap measurement technology and finally, (4) design of a front-end encompassing module with motorized lift-off capability. This front-end module is only 19 centimeters long and is capable of performing all of the gauging measurements required for fuel channel life-cycle management. This paper will detail the development efforts of Hydro-Quebec, IREQ and AECL in improving fuel channel gauging technology, as well as the implementation and field results of the 2005 Gentilly-2 inspection campaign. (author)

  20. Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

    International Nuclear Information System (INIS)

    This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results

  1. Hydrogen isotope enrichment by hydrophobic Pt-catalyst in Japan and Western countries

    International Nuclear Information System (INIS)

    The activities of the studies on hydrogen isotope separation by hydrophobic Pt-catalyst in Japan and the Western countries between 1970 and 1990 were reviewed. The R and D of tritium separation from heavy water or light water by the H2/H2O-isotopic exchange system with the aid of hydrophobic Pt-catalyst at the CRNL in Canada, the Mound Facility in the USA, the University of Karlsruhe in Germany and the Nuclear Center of Mol in Belgium were technically successful, but the construction of a commercial H2/H2O-isotopic exchange plant was abandoned or suspended because of political or budgetary problems. On the other hand, the Fugen heavy water upgrader using hydrogen isotope exchange by hydrophobic Pt-catalyst was constructed commercially and is treating on the average of about 10 m3 degraded heavy water (about 30%D) a year since 1986. In Canada, the major thrust of the development program of a new heavy water production process is directed at the CIRCE process. The catalyst used in this process is based on AECL's structured-type hydrophobic screen Pt-catalysts which are about three times more active than the corresponding random-type. Recent advances have allowed a fourfold reduction in Pt-metal loading without affecting catalyst performance. The objective of AECL is to develop the catalyst technology to allow commitment of a prototype CIRCE plant in 1994. (author)

  2. Analysis of common cause failure in Wolsong 2/3/4 NPPs PSA

    International Nuclear Information System (INIS)

    Wolsong 2/3/4 Nuclear Power Plants (WS 2/3/4 NPPs) are CANDU 6 type Pressurized Heavy Water Reactors (PHWRs) being built in Wolsong site of Korea by Korea Power Electric Corporation (KEPCO). WS 2/3/4 NPPs are designed by Atomic Energy of Canada Ltd. (AECL). AECL performs Probabilistic Safety Assessment (PSA) for these NPPs. In this PSA, however, the effect of Common Cause Failure (CCF) is not analyzed due to the limitation of CANDU reliability data. Since the CCF is regarded as one of the most dominant contributors to the total core damage frequency (CDF) in Pressurized Water Reactor (PWR) PSA. KEPCO and Korea Atomic Energy Research Institute (KAERI) initiated WS 2/3/4 NPPs Level 2 PSA which includes the CCF analysis and detailed HRA. In WS 2/3/4 NPPs Level 2 PSA, the authors reviewed the CCF data used in PWR PSA, and applied these data for WS 2/3/4 NPPs Level 2 PSA. The MGL method is implemented for the CCF analysis of WS 2/3/4 NPPs. The analyzed results show that the effect of CCF is not negligible in CANDU PSA as well as in PWR PSA. So the CCF data of CANDU plants should be collected and incorporated into CANDU PSA

  3. Geological disposal concept hearings

    International Nuclear Information System (INIS)

    The article outlines the progress to date on AECL spent-nuclear fuel geological disposal concept. Hearings for discussion, organised by the federal Environmental Assessment Review Panel, of issues related to this type of disposal method occur in three phases, phase I focuses on broad societal issues related to long term management of nuclear fuel waste; phase II will focus on the technical aspects of this method of disposal; and phase III will consist of community visits in New Brunswick, Quebec, Ontario, Manitoba and Saskatchewan. This article provides the events surrounding the first two weeks of phase I hearings (extracted from UNECAN NEWS). In the first week of hearings, where submissions on general societal issues was the focus, there were 50 presentations including those by Natural Resources Canada, Energy Probe, Ontario Hydro, AECL, Canadian Nuclear Society, Aboriginal groups, environmental activist organizations (Northwatch, Saskatchewan Environmental Society, the Inter-Church Uranium Committee, and the Canadian Coalition for Nuclear responsibility). In the second week of hearings there was 33 presentations in which issues related to siting and implementation of a disposal facility was the focus. Phase II hearings dates are June 10-14, 17-21 and 27-28 in Toronto

  4. Repair of the NRU Reactor Vessel: Technical Challenges and Lessons Learned

    International Nuclear Information System (INIS)

    Full text: In May 2009, following a Class 4 power outage that affected most of Eastern Ontario, including the Chalk River Laboratories site, Atomic Energy of Canada Limited (AECL) announced to its various stakeholders that a small heavy-water leak in the NRU reactor had been detected during routine monitoring while the reactor was being readied for return to service. Over the next 15 months AECL located, inspected, repaired and returned the NRU reactor to service. This presentation will focus on the extensive efforts required to support the unique activities associated with reactor vessel inspection and repair including initial assessment, repair site challenges, repair preparation and finally repair execution. The presentation will summarize: - Initial leak search and assessment of the vessel condition through the use of specialized tooling and non-destructive evaluation which resulted in one of the largest single NDE inspection campaigns ever carried out in the nuclear industry; - Challenges of executing a repair through 12 cm access ports at a distance of nine meters including the development of the specialized tooling; - The importance of development of repair techniques through mock up testing to perform welding repairs on a thin wall aluminium vessel and the measures taken and engineering challenges overcome to achieve a successful repair; - The final repair process, including site preparation, weld execution and final NDE inspection techniques; - Challenges encountered and lesson learned during the execution of weld repair, NDE inspections, and return-to-service of the reactor. (author)

  5. An Assessment of Resonance Treatment in WIMSD-5B

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

    WIMSD-5B is a lattice code with a depletion capability for the analysis of reactor physics problems related to a design and safety. It is released from the OECD/NEA Data Bank in 1998 and is now being used widely for thermal research and power reactor calculations. The purpose of this study is to assess and improve the resonance treatment method in WIMSD- 5B, through the introduction of a new method with a high accuracy in treating the resonance, as one of the development works for WIMS/CANDU, which is being developed for replacing WIMS-AECL, for the physics analysis of CANDU reactors. In this article, we specifically describe the recent improvements in the resonance integral method using the Carlvik's approximation. As a result, a comparison for the resonance calculation on the CANDU-6 fuel lattice was performed between the WIMSD-5B code and the WIMS/CANDU code with the 69-energy groups of the ENDF/B-VI nuclear data library and the WIMS-AECL code with the 89-energy group of the ENDF/B-VI nuclear data library.

  6. Upgrading from the Dicon Wiring Management system to IntEC at the Gentilly 2 station

    International Nuclear Information System (INIS)

    The General Electric DICON Wiring Management system supplied to HQ during the construction of G2 is currently being replaced by the stand-alone version of the IntEC software developed by AECL. The reasons for replacing DICON and choosing lntEC are discussed. The different aspects of the two year DICON data conversion project are presented with the problems encountered and the means that were taken to resolve the problems. lntEC has shown our DICON data to be considerably more deficient than we had thought. This has increased the cost and the duration of the conversion process. However, correcting the errors during the conversion process provides us with much more accurate data. This should be viewed as an investment in configuration management. Many potential causes of future errors and potentially critical path delays have been removed. We have chosen to document the detailed procedures for the use of lntEC in our plant using a Windows Help File compiler. This also has been found to be extremely useful as a training tool as well as providing on-line help. The DICON data conversion into lntEC will not be completed until 1996. lntEC is not perfect. However, from what we have up to now, we are satisfied with the conviviality and efficiency of lntEC and with AECL's diligence in constantly aspiring in making it a better product. (author)

  7. Annual report, 1981-82

    International Nuclear Information System (INIS)

    Recent operational restructuring implemented grouped the Engineering, Chemical, and International Companies under CANDU Operations. The Research Company was charged with finding products and markets to bridge the gap in new orders for reactors apparent for the next few years. Net income rose 46 percent to $19.7 million. Economic slowdown in Canada and elsewhere had little effect as AECL continued to fufill obligations on previously negotiated multi-year contracts. Over 60 percent of commercial revenue came from outside Canada, and at $234 million was marginally higher than 1980-81. Development of the superconducting cyclotron continued at Chalk River, with successful testing of magnetic field and radiofrequency systems. The nuclear fuel waste management program continued, with selection of a site for an underground research laboratory near Pinawa, Manitoba. The Therac-25 high energy accelerator for cancer therapy neared completion of its development and manufacturing program. There are more than 10 orders already booked. A record 15.2 million curies of cobalt 60 were shipped, an increase of 25 percent in orders for gamma irradiation processing. The prototype Douglas Point generating station was returned to full power and reached its highest annual capacity factor since 1975. Conceptual design of the new standardized two 950MW-unit CANDU PHWR generating station was completed. AECL responded to a request for quotations from the Mexican government for its nuclear power program

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

    International Nuclear Information System (INIS)

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

  9. Investigating a Link Between Knowledge Management and NPP Organizational Performance

    International Nuclear Information System (INIS)

    Effective knowledge management (KM) enhances a firm's capability to assimilate, create and exploit knowledge. KM is widely recognized in both the management literature and in practice as an important enabler of long term organizational performance. Its role is particularly important in technology intensive industries such as the nuclear power industry. Various nuclear power plants (NPPs) around the globe have begun to recognize the strategic importance of KM initiatives in achieving sustained high levels of operational performance. Although these organizations have been involved in KM-related activities for some time, they typically have not viewed and managed these activities from a KM perspective. Several NPPs have been early adopters of KM practices in the nuclear industry, and have been proactive in implementing company-wide KM programmes. However, at most NPPs, the concepts and benefits of KM are only beginning to be understood and have yet to be applied. Atomic Energy Canada Ltd. (AECL) undertook a research project entitled Knowledge Management (KM) for Nuclear Power Plants (NPPs), which formed part of the IAEA's Coordinated Research Project (CRP) on Comparative Analysis of Methods and Tools for Nuclear Knowledge Preservation. The programme scope of work included the following activities: conducting a literature review, participation in IAEA KM meetings and Assist Visits to NPPs, development of an industry survey, peer review and trial of the survey, and preparation of a summary report. This paper summarizes findings of the research done at AECL under this CRP.

  10. The geochemistry, age, and origin of groundwater in a mafic pluton, East Bull Lake, Ontario, Canada

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. (AECL) is conducting geoscience investigations of several plutons in Canada's Precambrian Shield as part of the Canadian Nuclear Fuel Waste Management Program, to determine whether such rock masses are suitable for the safe disposal of nuclear fuel waste. The East Bull Lake (EBL) gabbro-anorthosite layered complex is unique in this program as it is the only mafic pluton in which hydrogeological and hydrogeochemical studies have been conducted. These results can be compared with those of similar studies of granitic rocks which have been investigated more extensively. During the period, 1983-85, hydrogeological testing and hydrochemical sampling were conducted by the National Hydrology Research Institute of Environment Canada and AECL in boreholes drilled to depths of up to 850 m into the EBL pluton (Raven et al., 1987). This paper discusses the hydrogeochemistry of the pluton and identifies the major rock-water interactions controlling the chemistry. The spatial variability in chemistry will be shown to be related to the nature of the groundwater flow systems present at this site. The ages and origins of the groundwaters and their solutes are inferred from isotopic analyses

  11. Progress report - physical sciences TASCC division 1990 July 01 - December 31

    International Nuclear Information System (INIS)

    A completely new administrative structure of AECL Research was implemented on 1990 July 1. All of the basic physics programs, together with accelerator physics, radiation applications and most of the chemistry programs of AECL, have been placed in a new organizational unit called Physical Sciences. This unit also includes the management of the National Fusion Program. The research programs of Physical Sciences are grouped into three divisions: Chemistry, Physics and TASCC. Progress in each division will henceforth be reported on a twice-yearly basis. This report is the first of the new series to be issued by the TASCC Division. During the period covered by this report, the operation of the superconducting cyclotron has matured considerably, with over 30 accelerated ion beams more-or-less routinely available for a wide variety of nuclear physics experiments. The TASCC team, together with all the engineers, trades-people and other staff members who contributed to the design, constructed and commissioning of the Tandem Accelerator Superconducting Cyclotron facility, are to be heartily congratulated on bringing it to its present highly successful state in an unusually short period of time. In conjunction with our many outside collaborators, we are now engaged on exciting experiments in several areas of nuclear physics research, as reported in the following pages. We are well on the way to the establishment of a truly National Centre for Nuclear Physics research in Canada

  12. Technologies for tritium control in fission reactors moderated with heavy water

    International Nuclear Information System (INIS)

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  13. Waste Management Improvement Initiatives at Atomic Energy of Canada Limited - 13091

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) has been in operation for over 60 years. Radioactive, mixed, hazardous and non-hazardous wastes have been and continue to be generated at CRL as a result of research and development, radioisotope production, reactor operation and facility decommissioning activities. AECL has implemented several improvement initiatives at CRL to simplify the interface between waste generators and waste receivers: - Introduction of trained Waste Officers representing their facilities or activities at CRL; - Establishment of a Waste Management Customer Support Service as a Single-Point of Contact to provide guidance to waste generators for all waste management processes; and - Implementation of a streamlined approach for waste identification with emphasis on early identification of waste types and potential disposition paths. As a result of implementing these improvement initiatives, improvements in waste management and waste transfer efficiencies have been realized at CRL. These included: 1) waste generators contacting the Customer Support Service for information or guidance instead of various waste receivers; 2) more clear and consistent guidance provided to waste generators for waste management through the Customer Support Service; 3) more consistent and correct waste information provided to waste receivers through Waste Officers, resulting in reduced time and resources required for waste management (i.e., overall cost); 4) improved waste minimization and segregation approaches, as identified by in-house Waste Officers; and 5) enhanced communication between waste generators and waste management groups. (authors)

  14. Nuclear knowledge management strategies in Canada

    International Nuclear Information System (INIS)

    Full text: The Canadian Nuclear Industry recognizes the importance of nuclear knowledge management and has already implemented a number of initiatives to maintain competency, capture and preserve existing knowledge, advance the nuclear technology, develop future nuclear workers, and maintain a critical R and D capability. Although this paper addresses the Canadian scene in general, it will focus on knowledge management from a technology development point of view. Therefore, special emphasis will be placed on activities underway at present at Atomic Energy of Canada Limited (AECL). Maintaining competency is a high priority issue. With the on-going retirement of nuclear workers, resource management, succession planning and technical training programs are all in place at AECL. For example, a comprehensive assessment was recently completed to identify critical core competencies and the potential and timing of future retirements. Using a risk-based approach, the technology disciplines were prioritized and a plan was developed to address the requirements. The plan is now being implemented to hire, train, mentor and develop a new core of technical experts. Collaboration and knowledge sharing are important success factors in that regard. This is being achieved through cross-functional teamwork, consolidation of expertise, on-going work on nuclear power plant projects (e.g., the just completed units in China and ongoing work on unit 2 at the Romanian Cernavoda site), developing and designing new products (Advanced CANDU Reactor, ACR-700), adopting and improving Quality Management Systems (e.g., ISO 9001:2000 Global Certification and pursuing business excellence through the adoption of the Canadian Framework for Business Excellence). Capturing and preserving existing knowledge as well as advancing nuclear technology have also received significant attention. Fully computerized engineering tools have been developed and used to document the complete design of CANDU plants, and

  15. Laboratory and modeling studies in search of the critical hydrogen concentration

    International Nuclear Information System (INIS)

    The great success of hydrogen water chemistry (HWC) for primary coolant in nuclear power plants is due to the prevention of net radiolysis and to maintenance of the corrosion potential below -230 mV (SHE) where the rate of stress corrosion cracking is minimized. The critical hydrogen concentration or CHC has been defined as that concentration of excess H2 in primary coolant water, which prevents net water radiolysis via the chain reaction OH + H2 ↔H2O + H (1, -1) H + H2O2 → H2O + OH (2) The principle oxidizing free radical (OH) is thus converted into a reducing radical (H), oxidation products are reduced back to water, and the net result is no chemical change. A set of benchmark experiments at the U2 reactor in Chalk River have been reported in an extensive AECL report, which indicate that the CHC in this reactor is ca. 25 micro-molar. Using the review of yields and reaction rates set forth in another recent AECL report, the Chalk River experiments have been modelled in work at NNL, Harwell. The model was not able to successfully reproduce the experimental CHC, or the steady-state H2 concentrations (SSH2) in the absence of excess hydrogen. A sensitivity analysis of the entire model was carried out. Essentially three important variables have been found to dominate the result. Reaction rate (1) is overwhelmingly important in determining how much H2 is needed to accomplish the chain back-reaction. Almost with equal importance, the back reaction (-1) needs to be considered at 300 deg. C, but there is some uncertainty of its magnitude. Finally, the relative yields of radicals and molecular products (i.e. H2, H2O2 ) in particular H2:OH from the radiolysis are critical. Laboratory studies of hydrogenated water radiation chemistry have been carried out with a van de Graaff electron accelerator at Notre Dame Radiation Laboratory. Modelling of the hydrogen produced as a function of the hydrogen input, suggests that the reaction rate (-1) is ca. two times larger at 300

  16. Replacement of steam generators for Embalse NGS - the steam generator cartridge design and manufacturing issues, localization and site assembly challenges

    International Nuclear Information System (INIS)

    Embalse Nuclear Generating Station (Central Nuclear Embalse) was placed in service in 1983 and the outage for refurbishment is foreseen for 2011/2012. Embalse is equipped with four vertical inverted 'U' tube-type Steam Generators (SG) with integral preheater, I-800 tubes and carbon steel internals. Between 2002-2006, the owner assessed the potential for SG life extension; Nucleoelectrica Argentina S.A. (NA-SA) and AECL and a number of actions were completed towards meeting this objective (i.e.: primary divider plate replacement, additional U Bend support and inspection port installation). However, degradation of the tube supports (carbon steel broached plate) and U-bend supports due to Flow-accelerated corrosion (FAC) compromised the possibility for life extension of these Steam Generators. This issue, coupled with the plan to increase the plant power output during the life extension of the station, resulted in the strategic decision by NA-SA, to replace the Steam Generators. Several options were considered for SG replacement: In-situ replacement of the SG tube bundle, the original steam drum to be re-used; Removal and replacement of the entire SG (including the steam drum); and, Replacement of the bottom portion of the SG, i.e. the shell, the tube bundle, the tube sheet, the primary head and its internals and the primary nozzles with a factory assembled cartridge (collectively called the 'SG cartridge'). In this option, the original steam drum would be retained for the extended life. The final decision, based on the recommendations from the Life Assessment Study performed during the Pre-project Condition Assessment Process, is to replace only the Steam Generator cartridges. NA-SA requested AECL's support for the preparation of the Technical Specification for the replacement cartridges, allowing for the higher plant output. This paper presents the design basis for the technical requirements covered in the Technical Specification. The specified requirements include

  17. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Darlington NGS fuel damage investigation

    International Nuclear Information System (INIS)

    Darlington Unit 2 operated successfully from July to November 1990, but then a fuelling machine jammed, and the problem was soon traced to fuel debris. Initial inspections done with the CIGAR video camera, and subsequent metallurgical inspections at AECL, showed that vibration had produced fatigue cracking of some fuel end plates, and the damaged fuel had fretted one or two pressure tubes. The inspection and modelling programs showed that the cause of the trouble was resonant amplification of pressure pulsations produced by the primary heat transport system pumps. The trouble, which also affected the operation of Unit 1, has been cured by changing the number of vanes on each impeller from five to seven. Other possible solutions were considered

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

    International Nuclear Information System (INIS)

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

  1. Accelerating the Whiteshell Laboratories Decommissioning Through the Implementation of a Projectized and Delivery-Focused Organization - 13074

    International Nuclear Information System (INIS)

    Whiteshell Laboratories (WL) is a nuclear research site in Canada that was commissioned in 1964 by Atomic Energy of Canada Limited. It covers a total area of approximately 4,375 hectares (10,800 acres) and includes the main campus site, the Waste Management Area (WMA) and outer areas of land identified as not used for or impacted by nuclear development or operations. The WL site employed up to 1100 staff. Site activities included the successful operation of a 60 MW organic liquid-cooled research reactor from 1965 to 1985, and various research programs including reactor safety research, small reactor development, fuel development, biophysics and radiation applications, as well as work under the Canadian Nuclear Fuel Waste Management Program. In 1997, AECL made a business decision to discontinue research programs and operations at WL, and obtained government concurrence in 1998. The Nuclear Legacy Liabilities Program (NLLP) was established in 2006 by the Canadian Government to remediate nuclear legacy liabilities in a safe and cost effective manner, including the WL site. The NLLP is being implemented by AECL under the governance of a Natural Resources Canada (NRCan)/AECL Joint Oversight Committee (JOC). Significant progress has since been made, and the WL site currently holds the only Canadian Nuclear Safety Commission (CNSC) nuclear research site decommissioning license in Canada. The current decommissioning license is in place until the end of 2018. The present schedule planned for main campus decommissioning is 30 years (to 2037), followed by institutional control of the WMA until a National plan is implemented for the long-term management of nuclear waste. There is an impetus to advance work and complete decommissioning sooner. To accomplish this, AECL has added significant resources, reorganized and moved to a projectized environment. This presentation outlines changes made to the organization, the tools implemented to foster projectization, and the benefits

  2. The smallest SMRs

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K. [ACSION (Canada)

    2013-07-01

    An overview is presented on the subject of Small Modular Reactors (SMRs) for the generation of electricity and/or process heat in the Canadian context, with a particular focus on very small systems, up to about 30 MWe (less than about 100 MWt) output capacity. The potential Canadian market for such systems is examined, especially as a substitute for electricity generation by diesel engines in remote locations. Past experience with SMR systems in Canada and elsewhere is briefly reviewed, including AECL's earlier SLOWPOKE Energy System and Nuclear Battery development projects. Technology options and some recently proposed systems in this size range are discussed along with some of the requirements of an ideal SMR system for remote Canadian applications. (author)

  3. Learning from history: A case study in nuclear fuel

    International Nuclear Information System (INIS)

    The award of the 1993 W.J. Kroll Zirconium Medal recognized the value of cooperative, multidisciplinary, applied research in tackling practical problems. This paper suggests that several other lessons relevant to the current debate on science-and-technology (S and T) policy can be drawn from experience a quarter of a century ago. It outlines how close cooperation among those involved with the fuel for the Canadian CANDU heavy-water reactors identified a problem, then proceeded to solve it expeditiously. This capability for a rapid response to an unforeseen problem was no accident, but arose out of the conditions that existed at the Chalk River Laboratory of Atomic Energy of Canada Limited (AECL) and a deliberate policy to maintain this capability even when the utility's power reactors were demonstrating excellent performance

  4. The disposal of Canada's nuclear fuel waste: the geosphere model for postclosure assessment

    International Nuclear Information System (INIS)

    AECL is preparing an Environmental Impact Statement (EIS) of a concept for disposing of Canada's nuclear fuel waste. The disposal concept is that of a sealed vault constructed at a depth of 500 to 1 000 m in plutonic rock of the Canadian Shield. This report is one of nine primary references for the EIS. A probabilistic system variability analysis code (SYVAC3) has been used to perform a case study assessment of the long-term safety and environmental impacts for the EIS. This report describes the methodology for developing the SYVAC3-CC3 Geosphere Model (GEONET) which simulates the transport of contaminants from the vault through the geosphere to the biosphere. It also discusses the data used to construct the model, as well as assumptions and justifications for the data and model. The geosphere consists of the rock mass surrounding the vault, including the groundwater in the pores and cracks in the rock, the materials used to seal the shafts and exploratory boreholes at the site, and a domestic water well that is assumed to intersect the pathway of most rapid transport from the vault to the biosphere. GEONET simulates the movement of groundwater from the vault through the geosphere to discharge locations at the biosphere; the movement of contaminants in the groundwater by advection, hydrodynamic dispersion, and molecular diffusion; chemical sorption of contaminants onto minerals in the rock during transport; radioactive decay; and the rate of discharge of vault contaminants to the biosphere. Development of the Geosphere Model involves several steps. The initial step is to construct a conceptual model of the subsurface geological structure and ground water flow conditions using data from site investigations and laboratory tests. Once a conceptual model has been constructed, the coupled equations describing 3-D groundwater flow and heat transport are solved using the MOTIF finite-element code to calculate hydraulic head and groundwater velocity distributions. Next

  5. Phase doppler anemometer - commissioning tests for measurement of water aerosol sizes and velocities in flashing jets

    International Nuclear Information System (INIS)

    A state-of-the-art phase Doppler Anemometer (PDA) has been commissioned at AECL Research, Whiteshell Laboratories to undertake the measurement of size and velocity of water droplets generated in flashing jets. Experimental data on size and velocity distribution of water aerosols in flashing jets are required to support licensing of current multi-unit and single-unit CANDU (CANada Deuterium Uranium) stations. This paper presents the methodology involved in choosing the magnitudes of the various operating parameters of the PDA such as laser power and sensitivity of photomultiplier tubes in obtaining the experimental data. The various calibration and validation procedures used are also discussed. Size and velocity distributions in a typical flashing jet are presented. (author)

  6. The potential for vault-induced seismicity in nuclear fuel waste disposal: experience from Canadian mines

    International Nuclear Information System (INIS)

    A seismic event which causes damage to an underground opening is called a rockburst. Practical experience indicates that these damaging seismic events are associated with deep mines where extraction ratios are greater than 0.6. For the arrangement being considered by AECL for nuclear fuel waste disposal vaults, extraction ratios, for the room and pillar design, will be less than 0.3. At this extraction ratio the stress magnitudes will not be sufficient to induce seismic events that can damage the underground openings. Documented world-wide experience shows that unless the underground opening is very close to the source of a naturally occurring seismic event, such as an earthquake, the opening will also not experience any significant damage. Backfilling a disposal vault will improve its resistance to earthquake damage. Backfilling a disposal vault will also reduce the total convergence of the openings caused by thermal loads and hence minimize the potential for thermally-induced seismic events. (author)

  7. Modelling iodine behaviour using LIRIC 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wren, J.C.; Glowa, G.A.; Ball, J.M. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-01

    The overall objective of the iodine chemistry research program at the Whiteshell Laboratories of AECL is to develop and validate the LIRIC (Library of Iodine Reactions In Containment) model. The model, once validated, is intended as either a stand-alone analytical tool or for incorporation into a code for licensing analyses of fission-product behaviour in containment. LIRIC is currently being used to assess the role and importance of individual phenomena on iodine volatility under reactor accident conditions and, thus, help to establish priorities within the iodine research program. The LIRIC model has undergone significant alterations since it was last reported (LIRIC 2.0), mainly as a result of considerable development in understanding of iodine behaviour over the last few years. The new version, LIRIC 3.0, has been used to simulate various results from the Radioiodine Test Facility (RTF) with reasonable success, although under somewhat limited conditions.

  8. User's manual for the CC3 computer models of the concept for disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) is assessing a concept for disposing of CANDU reactor fuel waste in a vault deep in plutonic rock of the Canadian Shield. A computer program called the Systems Variability Analysis Code (SYVAC) has been developed as an analytical tool for the postclosure (long-term) assessment of the concept, and for environmental assessments of other systems. SYVAC3, the third generation of the code, is an executive program that directs repeated simulation of the disposal system, which is represented by the CC3 (Canadian Concept, generation 3) models comprising a design-specific vault, a site-specific geosphere and a biosphere typical of the Canadian Shield. (author). 23 refs., 7 tabs., 21 figs

  9. Dosimetry practice for irradiation of the Mediterranean fruit fly Ceratitis capitata (Wied.)

    International Nuclear Information System (INIS)

    In a sterile insect technique (SIT) programme the sterility of mass-reared insects, in our case Mediterranean fruit flies, is of primary importance. Mediterranean fruit fly pupae are irradiated in an AECL-CP-JS-7400 irradiator. Originally the capacity was 31,300 Ci, but because of the natural decay of cobalt, the actual source strength is 14,836 Ci. Thus, the dose with which the pupae are irradiated is 14.5 +- 1 krad (145 +- 10 Gy). A great risk in the daily release of sterile flies is that some batches of fertile flies may also be released. To ensure that this does not occur, continuous dosimetric check-ups have to routinely be carried out. Fricke dosimetry is ideal for this purpose because it has a range of response to doses of 4 to 40 krad (40 to 400 Gy) and because it is an economic and simple dosimetric system. (author)

  10. Creep and shrinkage analysis for concrete spent fuel dry storage module

    International Nuclear Information System (INIS)

    CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

  11. Implementation of the electronic DDA workflow for NSSS system design

    International Nuclear Information System (INIS)

    For improving NSSS design quality, and productivity several cases of the nuclear developed nation's integrated management system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Powes's (USA) were investigated, and it was studied in this report that the system implementation of NSSS design document computerization and the major workflow process of the DDA (Document Distribution for Agreement). On the basis of the requirements of design document computerization which covered preparation, review, approval and distribution of the engineering documents, KAERI Engineering Information Management System (KEIMS) was implemented. Major effects of this report are to implement GUI panel for input and retrieval of the document index information, to setup electronic document workflow, and to provide quality assurance verification by tracing the workflow history. Major effects of NSSS design document computerization are the improvement of efficiency and reliability and the engineering cost reduction by means of the fast documents verification capability and electronic document transferring system. 2 tabs., 16 figs., 9 refs. (Author)

  12. Deep geological disposal of nuclear fuel waste: background information and regulatory requirements regarding the concept assessment phase

    International Nuclear Information System (INIS)

    In their Joint Statement of August 1981, the governments of Canada and Ontario noted that the Nuclear Fuel Waste Program had been established to assure the safe and permanent disposal of radioactive waste from nuclear power reactors. The statement addressed the scope and schedule of the 'Concept Assessment Phase' of the Program, and identified the participating organizations and their responsibilities. The scope of this initial phase includes the development and assessment by Ontario Hydro and Atomic Energy of Canada Limited (AECL) of a disposal concept and its subsequent review by the regulatory agencies and government. The Atomic Energy Control Board (AECB), as lead regulatory agency is issuing this statement to outline its position with respect to evaluation of the concept

  13. Development of an FPGA-based controller for safety critical application

    International Nuclear Information System (INIS)

    In implementing safety functions, Field Programmable Gate Arrays (FPGA) technology offers a distinct combination of benefits and advantages over microprocessor-based systems. FPGAs can be designed such that the final product is purely hardware, without any overhead runtime software, bringing the design closer to a conventional hardware-based solution. On the other hand, FPGAs can implement more complex safety logic that would generally require microprocessor-based safety systems. There are now qualified FPGA-based platforms available on the market with a credible use history in safety applications in nuclear power plants. Atomic Energy of Canada (AECL), in collaboration with RPC Radiy, has initiated a development program to define a vigorous FPGA engineering process suitable for implementing safety critical functions at the application development level. This paper provides an update on the FPGA development program along with the proposed design model using function block diagrams for the development of safety controllers in CANDU applications. (author)

  14. Atomic Energy of Canada Limited annual report 1999-2000

    International Nuclear Information System (INIS)

    This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2000, and summarizes the activities of AECL during the period 1999-2000. The activities covered in this report include the CANDU reactor business, with the completion of the Wolsong unit 4 in the Republic of Korea, progress in the construction of two CANDU reactors for the Qinshan CANDU project in China, as well as the service business with Ontario Power Generation in the rehabilitation and life extension of operating CANDU reactors. In the R and D programs there is on-going effort towards the next generation of reactor technologies for CANDU nuclear power plants, discussions continue on the funding for the Canadian Neutron Facility for materials research (CNF) and progress being made on the Maple medical isotope reactor

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

    International Nuclear Information System (INIS)

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

  16. Atomic Energy of Canada Limited annual report 2000-2001

    International Nuclear Information System (INIS)

    This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2001 and summarizes the activities of AECL during the period 2000-2001. The activities covered in this report include the CANDU reactor business, with progress being reported in the construction of two CANDU 6 reactors for the Qinshan CANDU project in China, the anticipated completion of Cernavoda unit 2, the completion of spent fuel storage at Cernavoda unit 1 in Romania, as well as the service business with New Brunswick Power, Ontario Power Generation, Bruce Power and Hydro Quebec in the refurbishment of operating, CANDU reactors. In the R and D programs discussions continue on funding for the Canadian Neutron Facility for Materials Research (CNF) and progress on the Maple medical isotope reactor

  17. Canada

    International Nuclear Information System (INIS)

    Nuclear research and development in Canada started in the 1940s as a responsibility of the federal government. An engineering design team was established at Chalk River, Ontario, to carry out research on heavy water moderated lattices. A zero-energy heavy water moderated research reactor, ZEEP, was built and achieved criticality in September 1945; it was in fact the first human-made operating reactor outside the USA. In 1947, the 20 MW heavy water moderated national research experimental reactor (NRX) started up. It served as one of the most valuable research reactors in the world, and provided the basis for Canada's development of the very successful CANDU series of pressurised heavy water reactors (PHWR) for power generation. Atomic Energy of Canada Limited (AECL) was established in 1952 as a federal Crown Corporation. It has both a public and a commercial mandate. AECL has overall responsibility for Canada's nuclear research and development programme (its public mandate) as well as for the Canadian reactor design (CANDU), engineering and marketing programme (its commercial mandate). Nuclear energy in Canada is a $5 billion per-year industry, representing about 150 firms, 21 000 direct jobs and 10 000 indirect jobs, and ∼$1.2 billion in exports - the value to the country's economy is much higher than the research and development funding provided by the federal government. The CANDU nuclear reactor system was developed by AECL in close collaboration with the Canadian nuclear industry, and in particular with Ontario Hydro (now Ontario Power Generation). Currently, Canada operates 17 CANDU reactors, which contribute 16% of the country's current electricity consumption. There are also 12 CANDU reactors operating abroad (in Argentina, China, India, the Republic of Korea, Pakistan and Romania). AECL is now developing the 'third generation plus' Advanced CANDU Reactor (ACR-1000), and also has the leading role internationally in developing the Generation IV

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

    International Nuclear Information System (INIS)

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

  19. Canada's national policy on the long term management of nuclear fuel waste

    International Nuclear Information System (INIS)

    Nuclear energy is an important part of Canada's diversified energy mix. There are 22 CANDU reactors in Canada located in Ontario, New Brunswick, and Quebec. Like any other industry, nuclear fuel cycle operations produce some waste, and for this paper, we will focus on nuclear fuel waste, i.e., the irradiated fuel taken out of nuclear reactors at the end of their useful life. Canada has no plans to reprocess and recycle this fuel, so current plans are based on direct long term management of the waste fuel. Although nuclear fuel waste is currently in safe storage, steps are now underway to develop and proceed effectively with the implementation of more long term management solutions. A deep geological disposal concept was developed by the federal crown corporation Atomic Energy of Canada Limited (AECL) and Ontario Hydro, and, in October 1988, it was referred for review by a federal independent environmental assessment panel. AECL submitted the Environmental Impact Statement to the Panel in 1994. The Panel released its report with conclusions and recommendations on the acceptability of the concept in March 1998. It found that 'from a technical perspective, safety of the AECL concept has been on balance adequately demonstrated for a conceptual stage of development, but from a social perspective, it is not. As it stands, the AECL concept for deep geological disposal has not been demonstrated to have broad public support. The concept in its current form does not have the required level of acceptability to be adopted as Canada's approach for managing nuclear fuel waste'. Thus it was clear that Canada should increase public confidence before proceeding with any general approach on the long term management. With the Panel's recommendations in mind, and with further consultations with stakeholders, including the public, the Government of Canada developed the Nuclear Fuel Waste Act (NFW), which came into force on 15 November 2002. The NFW Act is a stand-alone piece of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  1. Simulated maintenance a virtual reality

    International Nuclear Information System (INIS)

    The article describes potential applications of personal computer-based virtual reality software. The applications are being investigated by Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories for the Canadian deuterium-uranium (Candu) reactor. Objectives include: (1) reduction of outage duration and improved safety, (2) cost-effective and safe maintenance of equipment, (3) reduction of exposure times and identification of overexposure situations, (4) cost-effective training in a virtual control room simulator, (5) human factors evaluation of design interface, and (6) visualization of conceptual and detailed designs of critical nuclear field environments. A demonstration model of a typical reactor control room, the use of virtual reality in outage planning, and safety issues are outlined

  2. General fire protection guidelines for Egyptian nuclear installations

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish the regulatory requirements that will provide and ensure fire protection of Egyptian nuclear installations. Two or more classes of occupancy are considered to occur in the same building or structure. Fire protection measures and systems were reviewed for four of the Egyptian nuclear installations. These are Egypt's first research reactor (ET-RR-1) building and systems, hot laboratories buildings and facilities, the building including the AECL type JS-6500 industrial cobalt-60 gamma irradiator ''Egypt's Mega Gamma I'' and Egypt's second research multi-purpose reactor (MPR). A brief review is given about fire incidents in Egypt, and descriptions of the only fire reported at one of the Egyptian nuclear installations over more than 35 years of operating these installations. The study outlines the various aspects of fire protection with a view to define the relevant highlights and scope of an Egyptian guidelines. (author)

  3. CIGAR - a remote inspection system

    International Nuclear Information System (INIS)

    CIGAR (Channel Inspection and Gauging Apparatus for Reactors) is an automated system that has been developed by Ontario Hydro, Canadian General Electric and AECL for the inspection of pressure tubes in CANDU reactors. It performs volumetric flaw detection, diameter and wall thickness measurement, sag profile measurement and spacer location determination. The inspection systems employed may have application in other situations and they well demonstrate the capabilities that exist in Canada for the development of sophisticated technology to solve complex problems. Part 1 of this paper discusses the requirements for the system and describes the CIGAR inspection, data acquisition and processing components that were developed by Ontario Hydro's Research Division. Part 2 describes the CIGAR delivery system that was developed by Canadian General Electric

  4. Nuclear power plant accident handbook: a CNSC emergency operations centre tool

    International Nuclear Information System (INIS)

    In response to the Fukushima Nuclear Emergency and the subsequent Emergency Operations Centre (EOC) response, the Canadian Nuclear Safety Commission (CNSC) Fukushima Task Force recommended that hardcopy and electronic version reference packages for all Canadian nuclear reactor sites are readily available to the Technical Support Team. CNSC staff, in a cooperative agreement with Atomic Energy of Canada Limited at Chalk River Laboratories (AECL-CRL), has begun implementing this recommendation through the development of the Nuclear Power Plant (NPP) Accident Handbook. The NPP Accident Handbook will provide readily available reference material for technical staff involved in EOC operations. The NPP Accident Handbook will assist technical staff in finding site-specific and accident-specific details that will help them provide expert advice to the EOC team during a nuclear power plant accident. (author)

  5. Economic impact of Hydro-Quebec's nuclear activities

    International Nuclear Information System (INIS)

    Gentilly 2 nuclear power plant has benefited the regions of Becancour and Trois Rivieres, with spin-off at the provincial level. Gentilly 2 is Hydro Quebec's only nuclear plant. Its 675 MW provide nearly 3% of Hydro Quebec's production. Over 664 permanent jobs were created, 70% of them highly specialized and multi-skilled. In 1993, out of C$99 spent, 57.3 were for wages, 16.3 for equipment and supplies (including fuel and heavy water), 18 for professional services provided by AECL and others, and the remainder included fees, permits, contract work, and miscellaneous. Gentilly 2 has fostered technological development and inventions which are used at other CANDU stations. 7 ills

  6. Atomic Energy of Canada Limited annual report 2000-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2001 and summarizes the activities of AECL during the period 2000-2001. The activities covered in this report include the CANDU reactor business, with progress being reported in the construction of two CANDU 6 reactors for the Qinshan CANDU project in China, the anticipated completion of Cernavoda unit 2, the completion of spent fuel storage at Cernavoda unit 1 in Romania, as well as the service business with New Brunswick Power, Ontario Power Generation, Bruce Power and Hydro Quebec in the refurbishment of operating, CANDU reactors. In the R and D programs discussions continue on funding for the Canadian Neutron Facility for Materials Research (CNF) and progress on the Maple medical isotope reactor.

  7. Evaluation of BICRON NE MCP DXT-RAD passive extremity dosemeter

    CERN Document Server

    Yuen, P S; Frketich, G; Rotunda, J

    1999-01-01

    Passive extremity dosemeters currently used in dosimetry communities worldwide have shortcomings. In general, an extremity dosemeter has too thick a detector element, and the dosemeter response is highly energy dependent for beta rays with energies ranging from 200 keV to 2 MeV. It often does not have dosemeter identification, causing problems in the chain of custody. It is often read manually, rendering reading/packing operations very labour intensive. As a result of collaboration between AECL and BICRON NE, a new extremity dosemeter, incorporating a highly sensitive LiF:Mg,Cu,P TLD and tentatively code named MCP DXT-RAD, was developed. It has been evaluated for radiological performance against an ISO draft standard for extremity dosemeters in twelve categories: homogeneity, detection threshold, beta ray energy response, beta angular response, photon energy response, photon angular response, reproducibility, stability under various climatic conditions, linearity, residue, self irradiation, and effect of ligh...

  8. Accelerating the Whiteshell Laboratories Decommissioning Through the Implementation of a Projectized and Delivery-Focused Organization - 13074

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, Brian; Mellor, Russ; Michaluk, Craig [Atomic Energy of Canada Limited, Whiteshell Laboratories, Pinawa, Manitoba (Canada)

    2013-07-01

    Whiteshell Laboratories (WL) is a nuclear research site in Canada that was commissioned in 1964 by Atomic Energy of Canada Limited. It covers a total area of approximately 4,375 hectares (10,800 acres) and includes the main campus site, the Waste Management Area (WMA) and outer areas of land identified as not used for or impacted by nuclear development or operations. The WL site employed up to 1100 staff. Site activities included the successful operation of a 60 MW organic liquid-cooled research reactor from 1965 to 1985, and various research programs including reactor safety research, small reactor development, fuel development, biophysics and radiation applications, as well as work under the Canadian Nuclear Fuel Waste Management Program. In 1997, AECL made a business decision to discontinue research programs and operations at WL, and obtained government concurrence in 1998. The Nuclear Legacy Liabilities Program (NLLP) was established in 2006 by the Canadian Government to remediate nuclear legacy liabilities in a safe and cost effective manner, including the WL site. The NLLP is being implemented by AECL under the governance of a Natural Resources Canada (NRCan)/AECL Joint Oversight Committee (JOC). Significant progress has since been made, and the WL site currently holds the only Canadian Nuclear Safety Commission (CNSC) nuclear research site decommissioning license in Canada. The current decommissioning license is in place until the end of 2018. The present schedule planned for main campus decommissioning is 30 years (to 2037), followed by institutional control of the WMA until a National plan is implemented for the long-term management of nuclear waste. There is an impetus to advance work and complete decommissioning sooner. To accomplish this, AECL has added significant resources, reorganized and moved to a projectized environment. This presentation outlines changes made to the organization, the tools implemented to foster projectization, and the benefits

  9. The predictive analysis of wear work-rates in wear test rigs

    Energy Technology Data Exchange (ETDEWEB)

    Phalippou, C.; Delaune, X.

    1996-12-31

    Impact and sliding wear in components is classically studied, as far as the wear laws are concerned, in specific wear test rigs that simulate the vibratory motion induced by the flow. In this paper, an experimental and numerical study on the impact forces and wear work-rates of a typical AECL rig is presented. The mode shapes and frequencies are measured and compared with finite element computations. Impact and sliding motions between the wear specimens are calculated and compared to the experimental results. Impact forces, mean values of wear work-rates as well as the specimen relative motions are found to be close to the experimental data. (authors). 14 refs., 9 figs., 5 tabs.

  10. Nuclear fuel waste disposal. Canada's consultative approach

    International Nuclear Information System (INIS)

    Over the past two decades, society has increasingly demanded more public participation and public input into decision-making by governments. Development of the Canadian concept for deep geological disposal of used nuclear fuel has proceeded in a manner that has taken account of the requirements for social acceptability as well as technical excellence. As the agency responsible for development of the disposal concept, Atomic Energy of Canada Limited (AECL) has devoted considerable effort to consultation with the various publics that have an interest in the concept. This evolutionary interactive and consultative process, which has been underway for some 14 years, has attempted to keep the public informed of the technical development of the concept and to invite feedback. This paper describes the major elements of this evolutionary process, which will continue throughout the concept assessment and review process currently in progress. (author)

  11. Considerations in managing the assessment of the Canadian nuclear fuel waste disposal concept

    International Nuclear Information System (INIS)

    This paper reports that in developing a concept for disposal of Canada's nuclear fuel waste, AECL has faced challenges because the acceptability of the concept must be established before a site is selected, no agency has been made responsible for implementing the concept if it is selected, and many stakeholders in the review must be satisfied if the concept is to be accepted. The challenges have thus far been met by a program that is well-integrated technically and administratively. However, interactions with stakeholders reviewing the concept present a problem in communication. The authors believe the nature of the nuclear fuel waste disposal issue calls for a cooperative rather than an adversarial approach to problem solving to efficiently deal with the requirements of all the stakeholders

  12. Nuclear fuel waste disposal: The Canadian consultative approach

    International Nuclear Information System (INIS)

    Over the past two decades society has demanded more public participation and public input into decision-making governments. Accordingly, development of the Canadian concept for deep geological disposal of used nuclear fuel has taken into account the requirements for social acceptability as well as technical excellence. As the agency responsible for developing the disposal concept, AECL Research, the research and development arm of Atomic Energy of Canada Ltd., has included, in its program, consultation with the various publics that have an interest in the concept. This interactive and consultative process, which has been underway for some 14 years, has attempted to ensure that the public has had the opportunity to become familiar with the technical development of the concept and to provide input into it. This process will continue throughout the concept assessment and review currently in progress

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-15

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

  14. Contribution of international cooperation in achieving the Romanian nuclear power objectives

    International Nuclear Information System (INIS)

    The international cooperation implied by the Romanian nuclear power program has on the short term the goal of supporting the Romanian effort of obtaining new nuclear fuels, improving the radioactive waste management, developing the technology and software transfer, training of Romanian researchers and participations in international projects. On long term the international cooperation is aiming at rising the scientific standards and the degree of global integration of Romanian research and technology. This policy is supported also by the agreements convened with different international organizations as for instance 'The Agreement for Nuclear Safety' or 'Common agreement for safe management of spent fuel and radioactive waste'. The INR at Pitesti and CITON - Bucharest are involved in works for cooperation technical projects supported by IAEA Vienna aiming at the nuclear safe and safeguard and power programs. At present underway are research contracts concerning the CANDU pressure tube, the corrosion in the primary cooling circuit of the CANDU reactors, as well as, the behavior under irradiation and intermediate disposal of oxide fuels. In the frame of IAEA assisted technical programs the TRIGA reactor of INR Pitesti is transferred from HEU to LEU fuels, new technologies for using recovered uranium and slightly enriched uranium fuels are developed, as well as a data acquisition system with advanced on-line characteristics. The cooperation between INR Pitesti and AECL Canada comprised in the frame of the 1998 agreement is oriented towards three fields of common interest: nuclear safety, nuclear fuel and management of in-service life of NPPs. Already transferred were computer codes from AECL devoted to accident analysis (WIMS, CATHENA and ELOCA) which were implemented and reproduced accurately all the test cases provided by AECL. For installing these codes specialists from our institute participated in training courses, on-the-job training, as well as, in validation

  15. Study on applicability of clay-based grout injection in the excavated damaged zone around the plug (TSX project)

    International Nuclear Information System (INIS)

    JNC has joined the international joint project, the TSX project, with AECL at the Underground Research Laboratory (URL) in Canada. Full-scale sealing technologies are applied to an underground tunnel in the TSX project. Regarding clay grouting, which supports the performance of the clay plug, a grouting experiment in the Excavated Damaged Zone around the tunnel was performed in the TSX project. A pre-injection test was the trial for the development of the grouting procedure, and the injection test was to evaluate the grouting effectiveness of the grouting in the EDZ around the tunnel. The results of the experiments showed the efficiency injection concentration of the grout slurry was between 4.0 and 6.0wt%. Grouted EDZ had lower hydraulic conductivity than that before grouting. (author)

  16. Report on a visit to Canada to discuss tritium instrumentation and radiological protection

    International Nuclear Information System (INIS)

    A report is presented of a visit to Canada on behalf of the CEC DG II/Fusion between the 8th to 13th April 1984. Discussions were arranged by the Canadian Fusion-Fuels Technology Project near Toronto and covered all aspects of tritium technology but especially radiological protection. Visits included the CFFTP Centre, Pickering Nuclear Generating Section, Ontario Hydro's Head Office, Safety Services Department and Research Division, Scintrex Ltd (tritium instrument manufacturers) and the Atomic Energy of Canada (AECL) Chalk River Nuclear laboratories (CRNL). There are clearly many areas for the use of Canadian Technology in Europe, particularly with CRNL and Scintrex on the development of 3H2/3H2O discriminating monitors. There is some doubt whether these development will be in time for applications at the JET laboratory and the JRC at ISPRA but this collaboration will be pursued. (author)

  17. Canada's high-level nuclear waste disposal concept: The evaluation process and a review of some aspects of the research work

    International Nuclear Information System (INIS)

    The concept of disposing of high-level nuclear waste in granitic rocks in the Canadian Shield, developed by Atomic Energy of Canada Limited (AECL), is anticipated to undergo a national public review within two years. A document which comprehensively describes the disposal concept is being prepared as a environmental impact statement (EIS). The process for EIS review and concept evaluation, including the role of the public, government and the scientific/engineering community, is summarized. A Technical Advisory Committee (TAC) has provided external peer review of the program since 1979 and its findings are published in annual reports which are publicly available. TAC's current views of certain geologic and geotechnical aspects of the program are presented along with a description of the safety and performance assessment of the disposal concept. (author). 35 refs., 2 figs

  18. Perceived risks of nuclear fuel waste disposal: trust, compensation, and public acceptance in Canada

    International Nuclear Information System (INIS)

    AECL's recommendation to place the high-level radioactive waste in corrosion resistant containers and bury it in underground vaults several hundred metres deep in the rock of the Canadian shield is presently under federal review. If and when the disposal concept is approved by the federal review panel, a search will begin for a suitable host community. Given that siting guidelines prevent the government from unilaterally imposing the waste on a reluctant community, identifying a suitable site may represent the single greatest obstacle to successfully implementing the disposal concept. Even if the concept is approved by the review panel, it may be very difficult to find a community that is willing to accept the waste. In the US, efforts to site an underground disposal facility for high-level nuclear waste at Yucca Mountain has run into strong opposition from local residents and politicians, resulting in long delays and major cost overruns

  19. Canada's national policy on the long-term management of nuclear fuel waste

    International Nuclear Information System (INIS)

    Full text: Nuclear energy is an important part of Canada's diversified energy mix. There are 22 CANDU reactors in Canada located in the provinces of Ontario, New Brunswick, and Quebec. Like any other industry, nuclear fuel cycle operations produce some waste, and for this paper, we will focus on nuclear fuel waste, i.e., the irradiated fuel taken out of nuclear reactors at the end of their useful life. Canada has no plans to reprocess and recycle this used nuclear fuel, so current plans are based on direct long-term management. Although nuclear fuel wastes is currently in safe storage, steps are now underway to develop and proceed effectively with the implementation of long-term management solutions. A cornerstone of Canada's approach to addressing radioactive waste management is the Government of Canada's 1996 Policy Framework for Radioactive Waste, which has set general policy for dealing with a all radioactive waste from the nuclear fuel cycle (nuclear fuel waste, low level radioactive waste, and uranium mine and mill waste). The Framework clearly indicates that the federal government will ensure safe, environmentally sound, comprehensive, cost-effective and integrated waste management, including disposal; that it will develop policy, regulate and oversee the waste owners to ensure compliance with legal and financial requirement in accordance with approved disposal plans; and that the waste owners are responsible for the funding, organization, management and operation of long term management, including disposal, facilities. With respect to the long-term management of nuclear fuel waste, a deep geological disposal concept was developed by the federal crown corporation Atomic Energy of Canada Limited (AECL) and Ontario Hydro, and, in October 1988, it was referred by the government for review by an independent Federal Environmental Assessment Panel. AECL submitted the Environmental Impact Statement to the Panel in 1994. The Panel reported its conclusions and

  20. Insights from the panel review process

    International Nuclear Information System (INIS)

    The environmental review process for nuclear waste management and disposal was unusual in that the Panel was asked to examine a concept rather than a specific project at a specific site. The Panel was charged with commenting on the safety and acceptability of the AECL concept, examining criteria for determining the safety and acceptability of any concept for managing nuclear fuel waste, and examining future steps which should be taken. In short, it was asked to provide policy advice to governments. The Panel concluded that safety is a key part, but only one part, of acceptability, and that safety must be viewed from both a technical and a social perspective. It judged that safety of the AECL concept had been adequately demonstrated from a technical perspective, but not from a social perspective. It also concluded that the AECL concept does not have the required level of public acceptability to be adopted as Canada's approach for managing nuclear fuel wastes. The paper examines in some detail the various aspects of the public concerns surrounding the nuclear cycle in general, and the safety of the proposals put forward by AECL for nuclear fuel waste management in particular. It notes the differences between those who look at safety from a technical perspective, and those who look at safety from a social perspective. And it lists the concerns related to acceptability in addition to the key factor of safety. After outlining the Panel's recommendations to governments on future steps to be taken, the paper discusses the extent to which the recommendations respond to the public's concerns. It stresses the importance of Aboriginal participation; of the creation of a new agency to deal with the full range of activities, technical and social, related to long-term management; of the public and decision-makers having more than one viable option to choose from; and of the essentiality of an inter-active process of public participation at all stages of decision-making. Finally

  1. Integrated plant for treatment of liquid radwaste

    International Nuclear Information System (INIS)

    In the early 1980's, AECL Research, at its Chalk River Laboratories (CRL) site, built a Waste Treatment Centre for managing low-level radioactive aqueous liquid wastes. At present, two industrial liquid waste streams are being routinely treated. One stream originates from the central Decontamination Centre (DC), where reactor components, protective plastic clothing, and respirators are cleaned. The other Active Drain (AD) stream is produced from a large and diverse number of research laboratories and radioisotope production facilities. The two waste streams, totalling about 2500 m per year (0.66 million US gallons), are volume reduced by a combination of continuous crossflow microfiltration (MF), spiral wound reverse osmosis (SWRO), and tubular reverse osmosis (TRO) membrane technologies; two thin-film evaporators (TFE) are employed for (i) the final volume reduction step, and (ii) the subsequent solidification of evaporator bottom with bitumen for containment of the radioactivity

  2. The gentle giants of healing

    International Nuclear Information System (INIS)

    Nuclear medicine, radiation therapy, and medical radioisotope production are explained at a popular level, for the non-specialist. Nuclear medicine in Canada uses either Positron emission tomography (PET), or single photon emission computerized tomography (SPECT). PET is used at the Montreal Neurological Institute to study epilepsy, brain tumours, stroke, or arterio-venous malformations. The much cheaper SPECT technique does many of the things that PET will do, and may eventually replace it to a considerable extent. This article features the manufacture of radioisotopes by Nordion Ltd., formerly known as AECL Radiochemical Co. Nordion supplies more than 20 isotopes, including about 80% of the world demand for 60Co, and 70% of all reactor isotopes, including the medically important 99Tc(m), 125I and 201Tl. Also featured is the intended acquisition (now cancelled) by Sherbrooke University of a 10-MW Slowpoke heating and isotope production reactor

  3. Test requirement for PIE of HANARO irradiated fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Lim, I. C.; Cho, Y. G

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U{sub 3}Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE.

  4. Field characterization and personal dosimetry at a high energy ion accelerator

    International Nuclear Information System (INIS)

    The response of a variety of dosimeters was evaluated in the radiation field outside the shielding of the Lawrence Berkeley Laboratory Bevalac Biomedical Facility. The primary beam was 580 MeV/center dot/A neon ions, incident upon a 30.5-cm polyethylene cube. The field was characterized by a neutron spectrometer consisting of Bonner spheres and other detectors and by estimates of charged particle fluences in NTA film and in the Berklet spectrometer. The responses of American Acrylics CR-39 track-etch plastic detectors and AECL (Canada) type BD-100 Bubble Detectors were compared to those of NTA film, Andersson-Braun remmeter and recombination-chamber results as well as to reference dose equivalents based upon the unfolded neutron spectrum. Evaluations of these dosimeters are discussed. 7 refs., 4 figs

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

    International Nuclear Information System (INIS)

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

  6. Single and two-phase flow pressure drop for CANFLEX bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  7. Effects of end-of-life power ramping on UO2 fuel

    International Nuclear Information System (INIS)

    A power-ramp test performed on fuel from the Nuclear Power Demonstration (NPD) reactor as part of the AECL Research extended-burnup program is described. The National Research Universal (NRU) reactor at Chalk River was used to ramp NPD fuel burned up to about 35 MW.d/kg U at declining power. The NRU ramp to about 36 kW/m proved sufficient to cause significant fission-gas release and diametral strain in the fuel elements. The ceramographic observations on both the ex-NPD and ex-NRU fuel are described and contrasted. Parallels are drawn with other recent observations on high-burnup fuel, and causative mechanisms for increased diametral strain and gas release are inferred

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Creep and shrinkage analysis for concrete spent fuel dry storage module

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: zhangd@aecl.ca

    2009-07-01

    CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

  11. A web-based resource for the nuclear science/technology high school curriculum - a summary

    International Nuclear Information System (INIS)

    On November 15, 2008, the CNA launched a new Nuclear Science Technology High School Curriculum Website. Located at www.cna.ca the site was developed over a decade, first with funding from AECL and finally by the CNA, as a tool to explain concepts and issues related to energy and in particular nuclear energy targeting the public, teachers and students in grades 9-12. It draws upon the expertise of leading nuclear scientists and science educators. Full lesson plans for the teacher, videos for discussion, animations, games, electronic publications, laboratory exercises and quick question and answer sheets will give the student greater knowledge, skills and attitudes necessary to solve problems and to critically examine issues in making decisions. Eight modules focus on key areas: Canada's Nuclear History, Atomic Theory, What is Radiation?, Biological Effects of Radiation, World Energy Sources, Nuclear Technology at Work, Safety (includes Waste Disposal) in the Nuclear Industry and Careers. (author)

  12. The temperature dependence of the rate constants and yields for the simulation of the radiolysis of heavy water

    International Nuclear Information System (INIS)

    At Chalk River Laboratories, a computer code is being developed to model the radiolysis of the heavy water in the moderator and the heat-transport system in CANDU reactors. This report collects together, for heavy water, the current knowledge regarding the rate constants, pKa's, yields and diffusion coefficients based on measurements in this laboratory and reports in the literature. The latest data available for the radiolysis of light water are generally included for comparison, which forms a partial update to the report on the radiolysis of light water (Elliot, AECL- 11073, COG-94-167, 1994). There are some reactions where little or no data are available at ambient or elevated temperatures; in these cases, an indication is given of the approach that will be taken to measure or estimate the required parameters. (author)

  13. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  16. Natural colloids in groundwater from granite and their potential impact on radionuclide transport

    International Nuclear Information System (INIS)

    AECL Research is assessing the concept of nuclear fuel waste disposal in an engineered vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield and submitted an Environmental Impact Statement to the Canadian Environmental Agency for review. Radionuclide transport in groundwater is the only likely path for radionuclide migration to the biosphere through the mass of rock surrounding a disposal vault. To evaluate the potential impact of natural particles on radionuclide migration it is necessary to determine the range of particle concentrations in ground water, which is a measure of their sorption capacity for radionuclides. An understanding of particle formation, stability and size distribution is important for predicting migration properties. This paper discusses the sampling and characterization of groundwater particles from the Whiteshell Research Area (WRA) and provides information on particle size, concentration, and composition. The significance of radiocolloid formation with colloids in groundwater from granite is also discussed

  17. A review of CANDU plant lifetime management

    International Nuclear Information System (INIS)

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

  18. BEATRIX-2 Program third annual report, January 1990--December 1990

    International Nuclear Information System (INIS)

    The BEATRIX-2 experiment is an International Energy Agency (IEA) sponsored collaborative experiment between Japan, Canada, and the United States. The purpose of the experiment is to evaluate the performance of ceramic solid breeder materials in a fast neutron environment. To do this, an in-situ tritium recovery experiment is being conducted in the Fast Flux Test Facility (FFTF), located on the Hanford site near Richland, Washington, and operated by Westinghouse Hanford Company (WHC). The Pacific Northwest Laboratory (PNL), Richland, Washington, together with the Japan Atomic Energy Research Institute (JAERI) and Atomic Energy of Canada Limited (AECL) are responsible for conducting the experiment. This work is divided into two phases: Phase 1 was irradiated from January 1990 until March 1991 in Cycle 11 of FFTF, while Phase 2 will be irradiated in Cycle 12, which began in June 1991 and is scheduled to continue until approximately October of 1991 for 300 effective full power days (EFPD)

  19. A reappraisal of some Cigar Lake issues of importance to performance assessment

    International Nuclear Information System (INIS)

    The AECL/SKB Cigar Lake Analogue Study was published in 1994. Data from this study, relevant for repository performance assessments, have been reappraised in the light of greater exposure to analogue studies and the development of more realistic models used in performance assessment. Several of the areas proved to have been adequately addressed in the original study, but one of the areas that particularly benefited from the renewed analysis concerned radiolysis. In this case a model for radiolysis was developed and tested, significantly narrowing the gap between calculated and predicted oxidant production. Considerable progress was also made in understanding and modelling the initial formation of the deposit under hydrothermal conditions, and using this conceptual model to evaluate the changes that have subsequently occurred under 'ambient' repository conditions over geological timescales. Moreover, the physical properties of clay as a potential buffer to groundwater flow and radionuclide migration were addressed with some success. 99 refs

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

    International Nuclear Information System (INIS)

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

  1. Research and development experience

    International Nuclear Information System (INIS)

    In the early 1950s, Atomic Energy of Canada Limited (AECL), in collaboration with Canadian industry and the power utilities, started on the task of developing and establishing the CANDU power reactor system and the necessary industrial infrastructure. While international activity provided a useful background and information to support Canadian activities, there were several unique features of the CANDU reactor which demanded specific programs of research and development work with a physics orientation. The four major areas were basic reactor physics, reactor control, heavy water and tritium monitoring and instruments and, finally, the potential of alternative fuel cycles. These four topics are discussed with the objective of providing an overview of what has been accomplished and what remains to be done. (auth)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  3. 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.

  4. Training of engineering personnel - Ad-hoc or science

    International Nuclear Information System (INIS)

    The training programs at AECL, with emphasis on training of technical employees is described and discussed. Implementation of Systematic Approach to Training (SAT) principles, establishments of individual and organizational unit career development and training plans, commitment for training budget, and assessment on training effectiveness are presented. There is also recognition of the importance of informal training for on-going continued development of staff, around key specialists/subject-matter experts, through mentoring and establishment of small centers of excellence, as part of the technical succession planning program. There is a discussion on what works well and what needs improvements. These approaches to training promise improvements in training and career development of individual employees and implementation of the Corporate training objectives. (author)

  5. Atomic Energy of Canada Limited annual report 1999-2000

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2000, and summarizes the activities of AECL during the period 1999-2000. The activities covered in this report include the CANDU reactor business, with the completion of the Wolsong unit 4 in the Republic of Korea, progress in the construction of two CANDU reactors for the Qinshan CANDU project in China, as well as the service business with Ontario Power Generation in the rehabilitation and life extension of operating CANDU reactors. In the R and D programs there is on-going effort towards the next generation of reactor technologies for CANDU nuclear power plants, discussions continue on the funding for the Canadian Neutron Facility for materials research (CNF) and progress being made on the Maple medical isotope reactor.

  6. Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

    2014-08-15

    A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

  7. Update on Cernavoda

    International Nuclear Information System (INIS)

    The Cernavoda project in Romania is for five 700 MWe CANDU units. Construction began in 1980, but because of problems occurring, many of them associated with political, social, and economic changes in Romania, in 1991 the management of construction of Unit 1 was transferred from RENEL (Regia Nationala de Electricitate) to a consortium formed by AECL and Ansaldo. This had been one of the chief recommendations of a pr-operational safety review conducted by the I.A.E.A. By May 1994, Unit 1 was 93.4% complete, but progress was hampered by the need to rework a number of components, mainly piping. A continuous quality assurance program was in place. Ninety-two Romanian operators were trained by New Brunswick Power at Point Lepreau. As of the date of the conference, the target date for grid connection was March 1995, nine months ahead of the contractual target

  8. Update from Cernavoda

    International Nuclear Information System (INIS)

    On April 16 1995 Cernavoda Unit 1 achieved criticality. This event marked the culmination of roughly 17 years of effort and one of the most troubled histories experienced by any nuclear construction project worldwide. It was also a major milestone in the program begun by AAC (AECL-ANSALDO CONSORTIUM) in 1991 and due to end in June 1997 with the hand-over to RENEL (The Romanian National Electric Utility) of a fully operational CANDU 600 Power Plant. This paper briefly outlines the history of the project, the organisational structure in place and the funding schemes used to ensure its completion. Most attention is given to the present status of the project and those areas and/or issues which have been or continue to be significant problem areas. The future program for the achievement of 100% power and hand-over of the station to RENEL is outlined and potential future problem areas are discussed. (author). 9 appendices

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

    International Nuclear Information System (INIS)

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

  10. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  11. Proceedings of the WIN-Global 2008 conference

    International Nuclear Information System (INIS)

    WiN-France hosted the 16. WIN-Global conference May 26-30, 2008, in Marseille, France. The conference was attended by over 150 delegates, representing 30 countries. Canadian participants, from many diverse backgrounds, attended the annual conference from AECL, Bruce Power, CNSC, NB Power and OPG. The theme: Maintaining Key Competencies, Arising Key Competencies for Nuclear Energy: A Challenge and Opportunity for Diversity Development, emphasized the challenges ahead in providing a skilled workforce for the nuclear renaissance, as new build projects and a vast number of retirements are expected around the world within the next 5 years. The conference addressed such questions as 'How will nuclear, attract, develop and retain staff?' A technical tour of Marcoule invited conference attendees to visit one of: Atalante, a high level nuclear chemistry laboratory; Phenix, a fast breeding research reactor; or AVM, a vitrification plant. A subsequent technical tour visited Cadarache providing the opportunity to view ITER, the international fusion research project

  12. Supplement to dose conversion factors for air, water, soil and building materials

    International Nuclear Information System (INIS)

    The calculations described in the original report, AECL-9825, have been repeated on a more recent database obtained from Oak Ridge National Laboratories (US). This database was apparently derived from the Evaluated Nuclear Structure Data File (ENSDF) and was used as the basis for the preparation of ICRP publication number 38. Data for 826 nuclides are given, including Argon-42 which was added to the database manually. Because the new database is more detailed than that used previously, and the beta-spectrum calculation sub-routine was improved, these new calculations are believed to be more accurate than those in the original report. This report does not entirely replace the original version, however, since there are a few nuclides that were included in that report but not in the current one

  13. A web-based resource for the nuclear science/technology high school curriculum - a summary

    Energy Technology Data Exchange (ETDEWEB)

    Ripley, C. [Atomic Energy of Canada Limited, Saint John, New Brunswick (Canada)], E-mail: ripleyc@aecl.ca

    2009-07-01

    On November 15, 2008, the CNA launched a new Nuclear Science Technology High School Curriculum Website. Located at www.cna.ca the site was developed over a decade, first with funding from AECL and finally by the CNA, as a tool to explain concepts and issues related to energy and in particular nuclear energy targeting the public, teachers and students in grades 9-12. It draws upon the expertise of leading nuclear scientists and science educators. Full lesson plans for the teacher, videos for discussion, animations, games, electronic publications, laboratory exercises and quick question and answer sheets will give the student greater knowledge, skills and attitudes necessary to solve problems and to critically examine issues in making decisions. Eight modules focus on key areas: Canada's Nuclear History, Atomic Theory, What is Radiation?, Biological Effects of Radiation, World Energy Sources, Nuclear Technology at Work, Safety (includes Waste Disposal) in the Nuclear Industry and Careers. (author)

  14. Development of DCC software dynamic test facility: past and future

    International Nuclear Information System (INIS)

    This paper describes a test facility for future dynamic testing of DCC software used in the control computers of CANDU nuclear power stations. It is a network of three computers: the DCC emulator, the dynamic CANDU plant simulator and the testing computer. Shared network files are used for input/output data exchange between computers. The DCC emulator runs directly on the binary image of the DCC software. The dynamic CANDU plant simulator accepts control signals from the DCC emulator and returns realistic plant behaviour. The testing computer accepts test scripts written in AECL Test Language. Both dynamic test and static tests may be performed on the DCC software to verify control program outputs and dynamic responses. (author)

  15. Development of a test rig and its application for validation and reliability testing of safety-critical software

    International Nuclear Information System (INIS)

    This paper describes a versatile test rig developed by AECL for functional testing of safety-critical software used in the process trip computers of the Wolsong CANDU stations. The description covers the hardware and software aspects of the test rig, the test language and its interpreter, and other major testing software utilities such as the test oracle, sampler and profiler. The paper also discusses the application of the rig in the final stages of testing of the process trip computer software, namely validation and reliability tests. It shows how random test cases are generated, test scripts prepared and automatically run on the test rig. The versatility of the rig is further demonstrated in other types of testing such as sub-system tests, verification of the test oracle, testing of newly-developed test script, self-test and calibration. (author). 5 tabs., 10 figs

  16. Recent developments in radiation equipment and radioisotopes

    International Nuclear Information System (INIS)

    A review is given of the technology of the uses of radiation equipment and radioisotopes, in which field Canada has long been a world leader. AECL Commercial Products has pioneered many of the most important applications. The development and sale of Co-60 radiation teletherapy units for cancer treatment is a familiar example of such an application and Commercial Products dominates the world market. Another such example is the marketing of Mo-99, which is produced in the NRX and NRU reactors at Chalk River, and from which the short-lived daughter Tc-99 is eluted as required for use in in-vivo diagnosis. New products coming into use for this purpose include Tl-201, I-123, Ga-67 and In-111, all produced in the TRIUMF cyclotron in Vancouver, while I-125 continues to be in demand for in-vitro radioimmunoassays. Radioisotopes continue to play an important part in manufacturing, where their well-known uses include controlling thickness, contents, etc., in production, and industrial radiography. The application of large industrial irradiators for the sterilization of medical products is now a major world industry for which Commercial Products is the main manufacturer. Isotopes are also used in products such as smoke detectors. Isotopes continue to find extensive use as tracers, both in industrial applications and in animal and plant biology studies. Some more recent uses include pest control by the Σsterile maleΣ technique and neutron activation and delayed neutron counting in uranium assay. The review concludes with an account of the development and prospects of AECL Commercial Products. (author)

  17. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

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

  18. Annual report, 1988-1989

    International Nuclear Information System (INIS)

    For the eleventh consecutive year Atomic Energy of Canada Limited realized a profit. However, for the second half of the year the profitable Radiochemical Company and the Medical Products Division were incorporated and operated separately as Nordion International Inc. and Theratronics International Limited. AECL now has two core operations, Nuclear Power, comprising the business carried out by the CANDU Operations division, and Research and Development, which is the responsibility of the Research Company division. Consolidated net income rose to $23.2 million from $10.4 million in 1988. Revenue from Nuclear Power operations declined 6 percent to $88.9 million, although the ratio of costs of sales and services to revenue is improving. Work on CANDU 3 pre-project engineering is continuing with the support of the federal government. AECL has made some progress towards its goal of covering the cost of its R and D operations from external sources. In 1989 the net expense of R and D operations was $19.3 million. The size of the fourth tranche of the five-year cutback in federal government funding was reduced to $10 million instead of the $25 million originally scheduled to be effective in 1989. Cash generated from operations in the year totalled $47.4 million compared to $6.9 million in 1988. A dividend of $16.4 million was paid to the shareholder, the federal government, at year end. It will be possible to continue research and CANDU 3 development activities for the first few months of the 1990 fiscal year at approximately 1989 levels. All figures are in Canadian dollars

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

    International Nuclear Information System (INIS)

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

  20. Dating fractures and fracture movement in the Lac du Bonnet Batholith

    International Nuclear Information System (INIS)

    This report examines and summarizes all work that has been done from 1980 to the present in determining the age of rock crystallization, fracture initiation, fracture reactivation and rates of fracture movement in the Lac du Bonnet Batholith to provide information for Atomic Energy of Canada Limited's (AECL) Canadian Nuclear Fuel Waste Management Program. Geological and petrographical indicators of relative age (e.g. cross-cutting relationships, sequences of fracture infilling minerals, P-T characteristics of primary and secondary minerals) are calibrated with radiometric age determinations on minerals and whole rock samples, using 87Rb-87Sr, 40K-39Ar, 40Ar-39Ar and fission track methods. Most fractures and fracture zones inclined at low angles are found to be ancient features, first formed in the Early Proterozoic under conditions of deuteric alteration. Following some movement on fractures in the Late Proterozoic and Early Paleozoic, reactivation of fractures during the Pleistocene is established from uranium-series dating methods and use of stable isotopic contents of fracture infilling minerals (mainly calcite). Some indication of movement on fracture zones during the Pleistocene is given by electron spin resonance dating techniques on fault gouge. The slow rate of propagation of fractures is indicated by mineral infillings, their P-T characteristics and U-series calcite ages in a fracture in sparsely fractured rock, accessible from AECL's Underground Research Laboratory. These results collectively indicate that deep fractures observed in the batholith are ancient features and the fracturing and jointing in the upper 200 m is relatively recent (< 1 Ma) and largely a result of stress release. (author)

  1. The bungling giant: Atomic Energy Canada Limited and next-generation nuclear technology, 1980--1994

    Science.gov (United States)

    Slater, Ian James

    From 1980--1994 Atomic Energy Canada Limited (AECL), the Crown Corporation responsible for the development of nuclear technology in Canada, ventured into the market for small-scale, decentralized power systems with the Slowpoke Energy System (SES), a 10MW nuclear reactor for space heating in urban and remote areas. The SES was designed to be "passively" or "inherently" safe, such that even the most catastrophic failure of the system would not result in a serious accident (e.g. a meltdown or an explosion). This Canadian initiative, a beneficiary of the National Energy Program, was the first and by far the most successful attempt at a passively safe, decentralized nuclear power system anywhere in the world. Part one uses archival documentation and interviews with project leaders to reconstruct the history of the SES. The standard explanations for the failure of the project, cheap oil, public resistance to the technology, and lack of commercial expertise, are rejected. Part two presents an alternative explanation for the failure of AECL to commercialize the SES. In short, technological momentum towards large-scale nuclear designs led to structural restrictions for the SES project. These restrictions manifested themselves internally to the company (e.g., marginalization of the SES) and externally to the company (e.g., licensing). In part three, the historical lessons of the SES are used to refine one of the central tenets of Popper's political philosophy, "piecemeal social engineering." Popper's presentation of the idea is lacking in detail; the analysis of the SES provides some empirical grounding for the concept. I argue that the institutions surrounding traditional nuclear power represent a form utopian social engineering, leading to consequences such as the suspension of civil liberties to guarantee security of the technology. The SES project was an example of a move from the utopian social engineering of large-scale centralized nuclear technology to the piecemeal

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  3. The disposal of Canada's nuclear fuel waste: comments on the postclosure assessment of a reference system

    International Nuclear Information System (INIS)

    Canada, like other countries, is developing technology for disposal of its nuclear fuel waste , based on the concept of geological disposal in stable plutonic rock of the Canadian Shield. The choice of methods, materials, and designs for a disposal system will ultimately be made on the basis of safety, taking into account the characteristics of the specific site on which the facility is to be developed, costs and practicality. As part of its work in developing the technology for the disposal of Canada's nuclear fuel waste, AECL analyzed the performance of a hypothetical disposal facility that incorporates specific design choices for the engineered barriers and that assumes a specific geological setting. This system, comprising the disposal facility and the geological setting, and the results of the performance analysis, is described in an Environmental Impact Statement that AECL submitted in 1994 and in a Primary Reference for the EIS 'The Disposal of Canada's Nuclear Fuel Waste: Postclosure Assessment of a Reference System.' The performance analysis was not intended to be a general proof of the safety of disposal, but rather it presents a safety analysis of one specific system to illustrate the postclosure assessment methodology and to demonstrate that safety could be achieved for the system in question. Although the design of the disposal facility analyzed and the geological setting have specific features, the results obtained from the safety analysis can, however, be used to provide considerable insight into the performance of the various components that comprise the multibarrier geological disposal system. Moreover, the results can show how changes in the performance of specific components can affect the overall performance of the system. This report discusses these aspects of the postclosure analysis. (author)

  4. The coolant void reactivity program in ZED-2

    International Nuclear Information System (INIS)

    The coolant void reactivity program at Chalk River produces reactor physics data for validating cell codes, in particular WIMS-AECL. One type of data provides detailed spectrum information within the various regions of a lattice cell. Most often however, the experiments are designed to provide the material buckling of a specific fuel type in a specific cell environment. Experiments are performed in the ZED-2 reactor. Fuel rod assemblies are positioned vertically, and the reactor is brought to criticality by reducing the neutron leakage from its top surface. This is done by raising the level of the heavy water moderator. The moderator height at the critical condition represents a key measurement in most experiments. The challenge is to covert this measurement, along with other supporting information such as foil activation data and moderator temperature, into the desired nuclear property of the test fuel - usually its material buckling. Since it's inception, the program has attempted to make measurements at conditions that are as close as possible to those in a power reactor. Most of the previous data available was for natural uranium at room temperature, the so called 'cold-clean' condition, and for the extreme ends of the coolant density range. Extending these conditions necessitated including effects of fuel burnup and the temperature of the fuel and coolant. A major component of the program has been to develop techniques for acquiring as much of that information as possible while operating within the constraints of a limited budget and the capabilities of a zero-energy critical facility. In the following sections, the progress made in developing some of the techniques necessary for generating data at power reactor conditions will be reviewed. A limited comparison with WIMS-AECL calculated values will also be made where appropriate. (author)

  5. Simulation of the multiple-fracture model. Phase 1, benchmark test 2 of the DECOVALEX project

    International Nuclear Information System (INIS)

    DECOVALEX is an international co-operative project for the development of coupled models and their validation against experiments in nuclear waste isolation. The emphasis of this project is on the coupled thermo-hydro-mechanical effects in jointed hard rock. In the first phase of DECOVALEX, two benchmark tests and one test case have been selected for modelling. This report describes the results of the second benchmark test, the Multiple-Fracture Model, as obtained by the AECL Research team. This problem relates to groundwater flow and coupled thermo-hydro-mechanical deformation in a simple system comprising several blocks of porous medium and several intersecting fractures. The simulation domain is defined to be a rectangular box that is made up of an assemblage of nine blocks separated by two sets of discontinuities (planar fractures). The rock mass is subjected to in situ stress and thermal loading as well as a hydraulic gradient. Both no-flow and adiabatic heat flux acting along a section of one of the lateral boundaries will induce expansion of the rock and cause shearing in the model. The MOTIF finite-element code, developed at AECL, has been employed to simulate this problem. The simulation results show that thermal expansion of the solid blocks reduced the aperture and, consequently, the permeability of the fractures. As a result, the fluid velocity along the horizontal fractures decreased with time, except in the vicinity close to the heat source, where the velocity initially increased and then decreased as a result of the decrease in permeability. (author). 9 refs., 7 tabs., 23 figs

  6. Retrievability - a matter of public acceptance? Reflections on the public review of the proposed nuclear fuel waste disposal concept in Canada

    International Nuclear Information System (INIS)

    Environmental assessment has been used as a planning tool in Canada for almost three decades. Public participation, one of its fundamental principles, is at the heart of environmental assessment in our country. To date, approximately 12 large projects related to nuclear energy have been the subject of public reviews by independent panels of experts appointed by the Government of Canada. These include: the development of uranium mines in Northern Saskatchewan; the construction and operation of two CANDU reactors in New-Brunswick, the second of which was never constructed; proposed uranium hexafluoride refineries in Ontario and Saskatchewan; expansion of a dry storage facility for nuclear spent fuel in Quebec; and decommissioning of uranium mine tailings areas in Ontario. All of the assessments mentioned above were conducted under the environmental assessment regimes of 1975 and 1984 that preceded the Canadian Environmental Assessment Act (1995). One of the public reviews of particular interest to this workshop is that of the proposed concept for deep geological disposal of nuclear fuel waste in Canada. This paper focuses exclusively on the public review of the Nuclear Fuel Waste Disposal Concept developed by Atomic Energy of Canada Limited (AECL), particularly as it relates to public acceptance of retrievability. The paper first describes the historical context in which AECL's concept was developed prior to the public review. It then briefly outlines the changes in the societal context that occurred between the time when decisions were made to proceed with the development of the concept in 1978 and the time when public hearings were held in 1996-1997 and the panel report was presented to the government in 1998. It also provides a short description of the concept itself. The paper then presents a discussion of the arguments used by the public in the panel review, arguments, which demonstrate a decrease in confidence in a concept lacking effective postclosure

  7. Nuclear energy for oil sands production: Providing security of energy and hydrogen supply at economic cost

    International Nuclear Information System (INIS)

    The development of Canada's 2000 EJ oil sands resource depends on a substantial energy input for extraction and upgrading. So far, this input has been supplied by natural gas, a resource that (a) is a premium fuel; (b) has limited availability; and (c) produces significant CO2 emissions. For the now preferred SAGD in-situ method of extracting oil-sands bitumen, nuclear heat can easily supply the steam at the ∼ 2.5 MPa requisite pressure. Studies by AECL and others show that steam from an Advanced CANDUTM Reactor (ACRTM) should produce steam for SAGD at lower cost than natural gas and also give far greater price stability. The large quantity of steam (2 to 2.5 volumes of condensate per volume of bitumen) for a typical project of 100 to 140 million barrels per day of bitumen provides a good match to the output of a 1900 MW(th) reactor, which would also produce about 200 MW of electricity. Electricity would be produced using a back-pressure turbine, yielding a very high overall energy efficiency. AECL work also shows economic competitiveness for electrolytic production of hydrogen, which is needed to upgrade the bitumen. Electrolysis would be interruptible, avoiding the short periods of high electricity prices experienced on the Alberta grid. Competitiveness with conventional steam-methane reforming is achieved by a combination of off-peak power and low-cost electrolytic cells. Using nuclear-generated steam and electricity produces negligible CO2, thus placing synthetic crude from the oil sands on a comparable basis to conventional crude with respect to greenhouse gas emissions. (author)

  8. Impact of recovered uranium cycle on the natural uranium production cycle and the environment

    International Nuclear Information System (INIS)

    The requirements by which future reactor and fuel cycle concepts must be judged are following: - properly utilize natural resources and national capabilities; - maximize the economic benefits; - effectively demonstrate the safety of fuel cycle facilities, and gain government and public approval for the enterprise; - satisfy national and international policies and goals; - contribute to sustainable energy supply. The ability to combine these five requirements ensures the success of the best options. Fuel utilization in thermal reactors can be improved in three ways: - lower the tails assay in the depleted stream of enrichment plants; - utilization of higher burnup fuel; - recycle plutonium. Recovered Uranium (RU) Cycle is a way to improve Slightly Enriched Uranium resulted from LWR spent fuel reprocessing which has 0.9-1.2% 235 U (dependent of the fuel history: reprocessing, burn up, reactor type) comparatively with 0.72% 235 U in natural Uranium. An international collaboration between Korea Atomic Energy Research Institute (KAERI), Atomic Energy of Canada Limited (AECL) and British Nuclear Fuel plc (BNFL) to use RU was developed. Since 1991, KAERI and AECL have introduced the Canadian Flexible (CANFLEX) fuel concept. A very attractive alternative to use RU in CANDU Reactors appears. Theoretically the quantity of 25,000 t (Europe and Japan) of RU would provide sufficient fuel for 500 CANDU reactor years of operation, knowing that the annual refueling requirement for a RU fuel burnup 13 MWd/KgU is around 50 t/y in comparison with 85 t/y for Natural Uranium (NU). Hereby, it is not necessary to mine about 42,500 t grade NU. In conclusion Recovered Uranium fuel cycle can be a very good option for the future of nuclear power in Romania. Moreover, waste resulted from uranium mining, waste resulted from uranium grade obtaining will disappear and financial costs, zones with nuclear activities and population exposed to irradiation will decrease. Also, the costs for fresh

  9. The DUPIC fuel cycle - Recycle without reprocessing

    International Nuclear Information System (INIS)

    Full text: The Generation IV International Forum, the IAEA's INPRO project and other international programs are pursuing enhanced proliferation resistance, in addition to enhancing economics, safety and radioactive waste management. Recent IAEA meetings have explored both technical and institutional aspects of this issue. Since 1991, KAERI (Korea Atomic Energy Research Institute), AECL (Atomic Energy of Canada Limited) and the USA (Department of State, Los Alamos National Laboratories), with the participation of IAEA, have been engaged in a practical exercise in developing a spent fuel recycle process to extend resources and reduce wastes, while enhancing proliferation resistance over typical recycle options. The concept of the DUPIC fuel cycle, DUPIC stands for Direct Use of PWR spent fuel In CANDU reactors, is to reuse spent pressurized water reactor fuel as a fuel for CANDU reactors without the reprocessing operations typical of recycling fuel cycles. The basic rationale of the DUPIC fuel cycle is that the typical fissile content of PWR spent fuel is approximately twice that of the natural uranium used in a CANDU reactor, and thus it can be used for fuel, even though it contains fission products and transuranic elements. This paper describes the basic requirements for the DUPIC fuel cycle development, the fuel fabrication process, the development and implementation of IAEA safeguards, the positive impact achieved on resource utilization and waste reduction and the factors resulting in enhanced proliferation resistance. DUPIC pellets and elements have been successfully manufactured at KAERI and AECL for irradiation tests at HANARO and NRU research reactors, respectively. The performance of DUPIC fuel is similar to that of conventional CANDU fuel, and more extensive work is under way to demonstrate DUPIC fuel performance under the power reactor condition. The technology and approach for safeguarding the DUPIC process has been developed and confirmed by the IAEA

  10. The bungling giant : Atomic Energy Canada Limited and next-generation nuclear technology, 1980-1994

    International Nuclear Information System (INIS)

    From 1980-1994 Atomic Energy Canada Limited (AECL), the Crown Corporation responsible for the development of nuclear technology in Canada, ventured into the market for small-scale, decentralized power systems with the Slowpoke Energy System (SES), a 10MW nuclear reactor for space heating in urban and remote areas. The SES was designed to be 'passively' or 'inherently' safe, such that even the most catastrophic failure of the system would not result in a serious accident (e.g. a meltdown or an explosion). This Canadian initiative, a beneficiary of the National Energy Program, was the first and by far the most successful attempt at a passively safe, decentralized nuclear power system anywhere in the world. Part one uses archival documentation and interviews with project leaders to reconstruct the history of the SES. The standard explanations for the failure of the project, cheap oil, public resistance to the technology, and lack of commercial expertise, are rejected. Part two presents an alternative explanation for the failure of AECL to commercialize the SES. In short, technological momentum towards large-scale nuclear designs led to structural restrictions for the SES project. These restrictions manifested themselves internally to the company (e.g., marginalization of the SES) and externally to the company (e.g., licensing). In part three, the historical lessons of the SES are used to refine one of the central tenets of Popper's political philosophy, 'piecemeal social engineering.' Popper's presentation of the idea is lacking in detail; the analysis of the SES provides some empirical grounding for the concept. I argue that the institutions surrounding traditional nuclear power represent a form utopian social engineering, leading to consequences such as the suspension of civil liberties to guarantee security of the technology. The SES project was an example of a move from the utopian social engineering of large-scale centralized nuclear technology to the piecemeal

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

    International Nuclear Information System (INIS)

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

  12. Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Isaenkova, Perlovich; Cheong, Y. M.; Kim, S. S.; Yim, K. S.; Kwon, Sang Chul

    2000-11-01

    For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in {beta}-phases of CANDU Zr-2.5%Nb pressure tubes was also established.

  13. ACR-1000TM - advanced Candu reactor design

    International Nuclear Information System (INIS)

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

  14. Measuring and reporting on decommissioning progress

    International Nuclear Information System (INIS)

    One of the challenges facing AECL, as well as other organizations charged with the responsibility of decommissioning nuclear facilities, is the means by which to measure and report on decommissioning progress to various audiences which, in some cases, may only have a peripheral knowledge or understanding of the complexities associated with the decommissioning process. The reporting and measurement of decommissioning progress is important for a number of reasons, i.e., It provides a vehicle by which to effectively communicate the nature of the decommissioning process; It ensures that stakeholders and shareholders are provided with a transparent and understandable means for assessing value for money; It provides a means by which to integrate the planning, measurement, and operational aspects of decommissioning One underlying reason behind the challenge of reporting decommissioning progress lies in the fact that decommissioning programs are generally executed over periods of time that far exceed those generally associated with typical design and build projects. For example, a decommissioning program could take decades to complete in which case progress on the order of a few percent in any one year might be typical. However, such progress may appear low compared to that seen with more typical projects that can be completed in a matter of years. As a consequence, AECL undertook to develop a system by which to measure decommissioning progress in a straightforward, meaningful, and understandable fashion. The system is not rigorously objective, and there are subjective aspects that are necessitated by the need to keep the system readily understandable. It is also important to note that while the system is simple in concept, there is, nonetheless, significant effort involved in generating and updating the parameters used as input, and in the actual calculations. (author)

  15. Accounting Systems for Heavy Water and Fissionable Materials

    International Nuclear Information System (INIS)

    Detailed accounting and reporting procedures used by Atomic Energy of Canada Limited (AECL) for maintaining adequate records and control of heavy water supplies and stocks of fissionable materials are described, along with the duties and responsibilities of those administering the system. An appraisal is made of these procedures with respect to their adaptability for use in rapidly expanding research and power programmes. In particular the use of electronic data processing equipment is evaluated. A senior management committee is responsible for ensuring that there is a proper system for recording, reporting and controlling fissionable materials. The Production Planning and Control Branch (Pp and C B) of the Operations Division at the Chalk River Nuclear Laboratories (CRNL) is responsible to the committee for keeping the over-all records and for the general administration of the system. The duties involved are detailed in the report. The system for fissionable materials is segregated into several accountability units 15 of which are allocated to AECL departments and the others to Canadian industries and research organizations. A control ledger is kept by PP and CB for each of the units; however, the units are responsible for preparing detailed accounts of all material under their jurisdiction. The basic recording procedures covering the movement Of materials between units, the changing of forms within units, the handling of gains and losses, and disposals, are outlined in the report. The transfer of this data to IBM cards, the ultimate processing through an IBM 1401 computer and the preparation of reports for management approval are described. The heavy-water accounting system based on the same principles as used for the fissionable materials is explained. In this case the control ledger lists the pounds of D2O allocated to each of the 15 accountability units. Again the basic recording methods and the use of a computer system are outlined. (author)

  16. Validation test for carbon-14 migration and accumulation in a Canadian shield lake

    International Nuclear Information System (INIS)

    This particular BIOMOVS II Technical Report is concerned with modelling the transfer of C-14 through the aquatic food chain following release to a Canadian shield lake. Model performance has been tested against field data supplied by Atomic Energy of Canada Limited. Carbon-14 was added in 1978 to the epilimnion of a small Canadian Shield lake to investigate primary production and carbon dynamics. Data from this experiment were used within BIOMOVS II to provide a validation test, which involved modelling the fate of the C-14 added to the lake. The nature of the spike and the subsequent monitoring allowed the investigation of both short-term processes relevant to evaluation of the impacts of accidental releases as well as longer-term processes relevant to routine release and to solid waste disposal. Four models participated in the scenario: 1) a simple mass balance model of a lake (AECL, Whiteshell Laboratories, Canada); 2) a relatively complex deterministic dynamic compartment model (QuantiSci Ltd.,UK); 3) a complex deterministic model (Studsvik Model A) and a more complex probabilistic model (Studsvik Model B; Studsvik Eco and Safety AB, Sweden). Endpoints were C-14 concentrations in water, sediment and whitefish over a thirteen year period. Each model produced reasonable predictions when compared to the observed data and when uncertainty is taken into consideration. About 0.2 to 0.4% of the initial C-14 inventory to the lakes remained in the water at the end of the study, because of internal recycling of C-14 from sediments. The simple AECL model did not account for this internal recycling of C-14 and, in this respect, its predictions were not as realistic as those of the QuantiSci and Studsvik models for concentrations in water. However, the AECL model predictions for the C-14 inventory remaining in lake sediment were closest to the observed values. Overall, Studsvik Model B was the most accurate in simulating C-14 concentrations in water and in whitefish, but

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

    International Nuclear Information System (INIS)

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

  18. Canada [National programme on the use of inert matrix fuel (IMF)

    International Nuclear Information System (INIS)

    Canadian effort on inert matrix fuels has been led by Atomic Energy of Canada Limited (AECL) with the participation of Queen's University in Ontario. This effort has focused on reactor physics analysis of the use of inert matrix fuels in a CANDU reactor, and on assessment of inert-matrix fuel candidates. The application has been either plutonium destruction or actinide burning. On-line refuelling, a simple bundle design, and high neutron economy provide the CANDU reactor with flexibility in these applications. A full core of inert matrix fuel, containing either plutonium, or a plutonium/actinide mix, could be used to fuel existing CANDU reactors, with high destruction rates. AECL has done preliminary evaluations of a number of inert-matrix fuel candidates, including ZrSiO4, MgAl2O4, CeO2, CePO4, ZrO2 doped with CaO, CeO2, Er2O3 or Y2O3, and SiC. Silicon carbide stood out early as a promising candidate material because of its high melting temperature and very high thermal conductivity, and because of its known resistance to attack by many corrosive agents, including oxygen, even at high temperatures. Good progress has been made in establishing methods of fabricating SiC containing a wide range of cerium concentrations. The methods are similar to those in current use for conventional UO2 fuel production. The thermal conductivity of such SiC-based inert-matrix fuels is high relative to UO2 and MOX. If high conductivity is maintained during irradiation, SiC-Based IMF will provide a clear benefit from the perspective of in-reactor fuel performance and safety. A critical parameter for candidate matrix materials is performance when exposed to highenergy fission fragments. AECL used its TANDEM accelerator to bombard candidate materials with a beam of 72 MeV iodine ions. This beam substitutes for fission fragments. And the bombarded areas were examined for damage, especially swelling. The results were also reasonably independent of temperature. The materials that showed the

  19. Canada country report

    International Nuclear Information System (INIS)

    1 - Nuclear 2007 highlights: New Build Applications and Environmental Assessments (Ontario Power Generation (OPG), Bruce Power, Bruce Power Alberta), Refurbishments (Bruce Power's Bruce A Units 1 and 2 Restart Project, NB Power's Refurbishment of Point Lepreau, New Brunswick, Atomic Energy of Canada Limited (AECL) NRU 50. Anniversary, expansion of the solid radioactive waste storage facilities at Gentilly-2 nuclear generating station, Ontario Power Generation (OPG) Deep Geologic Repository..); 2. Nuclear overview: a. Energy policy (Future of nuclear power, state of the projects, schedule, Refurbishment), b. Public acceptance, Statements from Government Officials in Canada; c. Nuclear equipment (number and type); d. Nuclear waste management, Deep Geologic Repository; e. Nuclear research at AECL; f. Other nuclear activities (Cameco Corporation, MDS Nordion); 3. Nuclear competencies; 4. WIN 2007 Main Achievements: GIRLS Science Club, Skills Canada, WiN-Canada Web site, Book Launch, WINFO, 2007 WiN-Canada conference 4 - Summary: - 14.6% of Canada's electricity is provided by Candu nuclear reactors; Nuclear equipment: 10 Research or isotope producing reactors - Pool-Type; Slowpoke 2; Sub-Critical assembly; NRU; and Maple; 22 Candu reactors providing electricity production - 18 of which are currently operating. Public acceptance: 41% feel nuclear should play more of a role, 67% support refurbishment, 48% support new build, 13% point gender gap in support, with men supporting more than women. Energy policy: Future of nuclear power - recognition that nuclear is part of the solution across Canada; New Build - 3 applications to regulator to prepare a site for new build, in Provinces of Ontario and Alberta, with one feasibility study underway in New Brunswick; Refurbishment - Provinces of Ontario (2010) and New Brunswick (2009). Nuclear waste management policy: Proposal submitted to regulator to prepare, construct and operate a deep geologic disposal facility in Ontario

  20. Case studies: Experience in Canada [Factors affecting public and political acceptance for the implementation of geological disposal

    International Nuclear Information System (INIS)

    Canada has 22 nuclear reactors that, at this time, supply approximately 15% of the country's electricity. Canada's spent nuclear fuel is now stored on an interim basis at licensed facilities at the reactor sites located in Ontario, Quebec and New Brunswick and at the Atomic Energy of Canada Limited (AECL) facility in Manitoba. The issue of long-term management of spent fuel has been the subject of considerable study in Canada. At the request of the federal and Ontario governments, and after a 20-year research program, a concept for the management of spent fuel was developed by AECL. In its 1998 report, the environmental assessment panel provided insight and direction on key issues that had to be addressed in order to move the decision-making forward. The Government of Canada considered and responded to the panel report, and in November 2002 brought into force the Nuclear Fuel Waste Act (NFWA) which assigned roles and responsibilities and established the legislative framework for decision-making framework for the long-term management of its spent nuclear fuel. As required by that legislation, the Nuclear Waste Management Organization (NWMO) was established by the nuclear energy corporations. The NWMO was tasked with conducting a study of storage and disposal options, and will be responsible for implementing the management approach selected by government. After the Government selects the approach, implementation will unfold under the ongoing oversight of Natural Resources Canada. The Canadian Nuclear Safety Commission is responsible for the licensing and ensuring NWMO's compliance with regulatory requirements. The immediate task of the NWMO was to conduct a three year examination of different management options. The NWMO believed that an approach that generates confidence for the long term would need to resonate with citizen values and objectives for the study, and bring to bear the best science and technical understanding. A dynamic and iterative study was conducted

  1. The disposal of Canada's nuclear fuel waste: site screening and site evaluation technology

    International Nuclear Information System (INIS)

    The concept for the disposal of Canada's nuclear fuel waste is to dispose of the waste in an underground vault, nominally at 500 m to 1000 m depth, at a suitable site in plutonic rock of the Canadian Shield. The feasibility of this concept and assessments of its impact on the environment and human health, will be documented by AECL in an Environmental Impact Statement (EIS). This report is one of nine primary references for the EIS. It describes the approach and methods that would be used during the siting stage of the disposal project to identify a preferred candidate disposal site and to confirm its suitability for constructing a disposal facility. The siting stage is divided into two distinct but closely related substages, site screening and site evaluation. Site screening would mainly involve reconnaissance investigations of siting regions of the Shield to identify potential candidate areas where suitable vault locations are likely to exist. Site screening would identify a small number of candidate areas where further detailed investigations were warranted. Site evaluation would involve progressively more detailed surface and subsurface investigations of the candidate areas to first identify potentially suitable vault locations within the candidate areas, and then characterize these potential disposal sites to identify the preferred candidate location for constructing the disposal vault. Site evaluation would conclude with the construction of exploratory shafts and tunnels at the preferred vault location, and underground characterization would be done to confirm the suitability of the preferred candidate site. An integrated program of geological, geophysical, hydrogeological, geochemical and geomechanical investigations would be implemented to obtain the geoscience information needed to assess the suitability of the candidate siting areas and candidate sites for locating a disposal vault. The candidate siting areas and candidate disposal vault sites would be

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

    International Nuclear Information System (INIS)

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

  3. Preserving the life extension option for Wolsong NPP Unit 1 through plant life management program

    International Nuclear Information System (INIS)

    The first CANDU 6 plants, including Wolsong Unit 1 nuclear power plant (NPP) (which entered service in the early 1980's) are now approaching two thirds of their thirty-year design life. Korea Electric Power Research Institute (KEPRI) and Atomic Energy of Canada Limited (AECL) have worked together since 2000 to develop and implement a comprehensive and integrated CANDU Plant Lifetime Management (PLiM) program for Wolsong Unit 1 NPP. PLiM will see this plant successfully and reliably through to the design life and preserve the option for life extension. The focus of the initial phase of the program is on the major critical components and structures and any potential aging phenomena that might affect plant safety and availability. In-depth life assessments of Wolsong Unit 1 critical systems, structures, and components (CSSCs) are under going. It is recognized that effective plant practices in inspection, monitoring, maintenance, and operations are the primary means of managing aging through the design life and necessary for preserving the life extension option. Hence, the PLiM assessments identify enhancements to current plant programs to mitigate aging effects and to ensure reliable life attainment and performance. The KEPRI/AECL co-operation for the PLiM program over the last few years is providing in-depth assessments and promising life prognosis for the key CSSCs of Wolsong Unit 1. The assessments are also identifying those areas for optimized plant inspection, monitoring and maintenance programs to achieve utility targets for safety, reliability and production capacity during extended life. These outcomes are important inputs to decision makings to embark upon a detailed Wolsong Unit 1 life management program. In this paper, the PLiM program assessment methods and techniques tailored to the components, like CANDU 6 steam generators, are described. A typical proactive aging management program for steam generators, aimed to continue current excellent service for

  4. Joint studies on large CANDU

    International Nuclear Information System (INIS)

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

  5. Liquid radwaste processing with spiral wound reverse osmosis

    International Nuclear Information System (INIS)

    Two different reverse osmosis systems were investigated. The first was a 50-element plant-scale system that is used to treat 2200 m3 of AECL liquid radwastes annually.It uses thin-film composite (TFC) membranes and operates at an applied pressure of 2760 kPa, with a fixed crossflow of about 40 L/min. The other system uses the same thin-film composite membranes for waste processing but is a two-element pilot-scale system. It is operated at pressures m ranging between 1500 and 7000 kPa, at a fixed crossflow of 55 L/min. The average lifetime of the thin-film composite membranes in the plant-scale processing application at AECL is about 3000 h. After this service life has expired the rejection efficiency declines rapidly from 99.5% to about 95% as the membranes become impaired from chemical cleaning procedures that are required after each 100 m3 of waste is treated. The permeation flux for the plant-scale system decreases from about 2.2 L/min/element to below 0.5 L/min/element at the end of the membrane's useful service life. The plant-scale membrane elements, fouled by an assortment of chemicals including calcium phosphate and various organics, were successfully regenerated by exposing them to a threestep chemical cleaning procedure, using detergent, HCI, and an alkaline-based cleaning with EDTA. The three-step procedure was successful in elevating the flux from 0.5 L/min for the spent membrane to 1.2 L/min after cleaning. The 1.2-L/min postcleaning flux could be maintained provided that the crossflow velocity remained high. The decontamination factor (DF) for cesium for the plant-scale system, decreased from about 100 when the membranes were new, to about 30 after they were replaced. The strontium DF was unaffected by the applied pressure. 9 refs., 1 tab., 6 figs

  6. Application of Ultrasonic Flow Measurement for Power Uprates Based on Measurement Uncertainty Recapture

    International Nuclear Information System (INIS)

    in the field of computer based instrumentation and control. An ultrasonic flow meter (UFM) is one of these technologies, which enables a significantly more accurate flow measurement compared to the traditional flow meters. In addition, the UFM has no fouling issues due to ageing, thereby retaining its accuracy through the life of the plant. Hitachi and AECL have initiated a feasibility study on UFM applications to enhance CANDU performance, focussing especially on the MUR reactor power uprate for CANDU. Hitachi has developed the UFM application over the last ten years in Japan and AECL has widespread CANDU technology and licensing experience. Thus, the collaboration of these two companies provides a good fit for introducing this technology into CANDU

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

    International Nuclear Information System (INIS)

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

  8. Inventory of radioactivity in Ottawa River-bed sediments near the Chalk River Laboratories

    International Nuclear Information System (INIS)

    AECL's Chalk River Laboratories (CRL) are situated on the Ontario side of the Ottawa River about 200 km NW of the City of Ottawa. Since 1947, water for cooling CRL's research reactors has been piped from and returned to the Ottawa River. From 1952 to the present time, cooling water has been discharged through the Process Sewer at a rate of 1.5 to 2 m3/s. The Outfall, which is the discharge from the Process Sewer, is in 18 m of water, 65 m offshore. Flow is directed toward the river surface through three 'diffuser vents,' creating a turbulent swirl at the surface and maintaining a patch of open water in winter. In addition to cooling water, the Outfall has, over the years, included small additional effluents from a heavy water recovery plant, a decontamination centre and a waste treatment centre. Although the effluent has been monitored and has met applicable regulatory requirements, investigations of the riverbed near the Outfall revealed radioactivity. In 2001, a riverbed reconnaissance and a detailed coring program were initiated for the purpose of determining the inventory of residual radioactivity. (author)

  9. Risk-based Prioritization of Facility Decommissioning and Environmental Restoration Projects in the National Nuclear Legacy Liabilities Program at the Chalk River Laboratory - 13564

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Jerel G.; Kruzic, Michael [WorleyParsons, Mississauga, ON, L4W 4H2 (United States); Castillo, Carlos [WorleyParsons, Las Vegas, NV 89128 (United States); Pavey, Todd [WorleyParsons, Idaho Falls, ID 83402 (United States); Alexan, Tamer [WorleyParsons, Burnaby, BC, V5C 6S7 (United States); Bainbridge, Ian [Atomic Energy Canada Limited, Chalk River Laboratories, Chalk River, ON, K0J1J0 (Canada)

    2013-07-01

    Chalk River Laboratory (CRL), located in Ontario Canada, has a large number of remediation projects currently in the Nuclear Legacy Liabilities Program (NLLP), including hundreds of facility decommissioning projects and over one hundred environmental remediation projects, all to be executed over the next 70 years. Atomic Energy of Canada Limited (AECL) utilized WorleyParsons to prioritize the NLLP projects at the CRL through a risk-based prioritization and ranking process, using the WorleyParsons Sequencing Unit Prioritization and Estimating Risk Model (SUPERmodel). The prioritization project made use of the SUPERmodel which has been previously used for other large-scale site prioritization and sequencing of facilities at nuclear laboratories in the United States. The process included development and vetting of risk parameter matrices as well as confirmation/validation of project risks. Detailed sensitivity studies were also conducted to understand the impacts that risk parameter weighting and scoring had on prioritization. The repeatable prioritization process yielded an objective, risk-based and technically defendable process for prioritization that gained concurrence from all stakeholders, including Natural Resources Canada (NRCan) who is responsible for the oversight of the NLLP. (authors)

  10. Risk-based Prioritization of Facility Decommissioning and Environmental Restoration Projects in the National Nuclear Legacy Liabilities Program at the Chalk River Laboratory - 13564

    International Nuclear Information System (INIS)

    Chalk River Laboratory (CRL), located in Ontario Canada, has a large number of remediation projects currently in the Nuclear Legacy Liabilities Program (NLLP), including hundreds of facility decommissioning projects and over one hundred environmental remediation projects, all to be executed over the next 70 years. Atomic Energy of Canada Limited (AECL) utilized WorleyParsons to prioritize the NLLP projects at the CRL through a risk-based prioritization and ranking process, using the WorleyParsons Sequencing Unit Prioritization and Estimating Risk Model (SUPERmodel). The prioritization project made use of the SUPERmodel which has been previously used for other large-scale site prioritization and sequencing of facilities at nuclear laboratories in the United States. The process included development and vetting of risk parameter matrices as well as confirmation/validation of project risks. Detailed sensitivity studies were also conducted to understand the impacts that risk parameter weighting and scoring had on prioritization. The repeatable prioritization process yielded an objective, risk-based and technically defendable process for prioritization that gained concurrence from all stakeholders, including Natural Resources Canada (NRCan) who is responsible for the oversight of the NLLP. (authors)

  11. Reviewing NPP Cernavoda site evaluation

    International Nuclear Information System (INIS)

    The Nuclear Power Plant Cernavoda site was selected before the IAEA Safety Guide issue, during NUSS program development. The Romanian codes issued in 1976, as a regulatory body requirements, establish general criteria regarding safety concept and concentration limits of different radionuclides in air and water body and limits of individual or collective dose. In 1979 the Romanian Authority signed the contract with AECL to improve the CANDU-600 concept in the nuclear development programme and erection of 4 units on the Cernavoda site. The construction work started in 1980. In 1983 the former Romanian Government decided to build up another unit (finally it will be 5 units) on Cernavoda site, so the total gross electrical power we have 3,500 MW. The Canadian safety and quality standards or requirements was harmonized with the Romanian rules and regulations. Many studies, investigations and research were done to qualify the site and have a good knowledge about its characteristics coupled with CANDU-600 performance. The new evolution of the site was performed by Romanian technical staff in CITON and the final conclusions were favourable for erection and operation of NPP. The first unit of Cernavoda NPP is on operation and now the efforts are concentrated to continue the works for the unit 2. The paper underlines how the Cernavoda site characteristics meet IAEA Code of Practice and Safety Guides issued until now. (author)

  12. Argentina: Nuclear power development and Atucha 2

    Energy Technology Data Exchange (ETDEWEB)

    Nogarin, Mauro

    2015-08-15

    In 2014, nuclear energy generated about 5,257 GWh of electricity or a total share of 4.05 % of the total electrical energy of about 129,747.63 GWh kWh produced in Argentina and there has been a trend for this production to increase. Argentina currently has a nuclear production capacity of 1,010 megawatts of electrical energy. However, when the Atucha 2 nuclear power plant is completed and starts commercial operation, it will add 745 megawatts to this electrical production capacity. There are two sites with nuclear power plants in Argentina: Atucha and Embalse. The Embalse nuclear power plant went into operation in 1984. At the Atucha site, the Atucha-1 nuclear power plant started operation in 1974. It was the first nuclear power plant in Latin America. Construction of Atucha-2 started in 1981 but advanced slowly due to funding and was suspended in 1994 when the plant was 81 % built. In 2003, new plans were approved to complete the Atucha 2. I summer 2014 the plant went critical for the first time. The construction was completed under a contract with AECL.

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

    International Nuclear Information System (INIS)

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

  14. Stability of flashing-driven natural circulation in a passive moderator cooling system for Canadian SCWR

    International Nuclear Information System (INIS)

    Highlights: • The stability in a passive moderator cooling system of a unique system in the Canadian SCWR. • Identify and analyze unstable oscillations using flashing-driven natural circulation test results. • The flashing-driven oscillations categorized as a flashing-driven Type-I density wave instability including a geysering-like feature. • A stability map on the dimensionless plane with the Subcooling number and Phase Change number. - Abstract: This paper presents an examination of the instability mechanisms in a Passive Moderator Cooling System for the Canadian SCWR (Supercritical Water-cooled Reactor). The passive system is being developed at AECL using a flashing-driven natural circulation loop. Unstable intermittent and sinusoidal oscillations were identified from experimental data of the flashing-driven natural circulation passive moderator cooling system. The oscillation periods were correlated with the boiling delay time. A stability map for a flashing-driven two-phase natural circulation loop was established on the dimensionless plane with Subcooling number and Phase Change number. It was observed that there is thermal non-equilibrium in the single-phase and two-phase oscillation stages of the flashing-driven natural circulation

  15. Radiation sensitivity of different citric pectins

    Energy Technology Data Exchange (ETDEWEB)

    Inamura, Patricia Y.; Mastro, Nelida L. del [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: patyoko@yahoo.com; nlmastro@ipen.br

    2007-07-01

    Pectic substances are important soluble polysaccharides of plant origin of considerable interest for food industry as gelling agent and stabilizer in jams, fruit jellies, yogurt drinks and lactic acid beverages. Polysaccharides can be degraded by ionizing radiation due to the free radical induced scission of the glycosidic bonds. Viscosity methods had been used to determine the efficiency of hydroxyl radical induced chain breaks generation in macromolecules. In the present work samples of pectin with different degree of methoxylation were employed in order to study their radiation sensitivity by means of viscosity measurements. Samples of citric pectin 1% solutions were irradiated with gamma rays at different doses, ranging from 0 to 15 kGy, using a {sup 60}Co Gammacell 220 (AECL), dose rate about 2 kGy/h. After irradiation the viscosity was measured on the viscometer Brookfield model LV-DVIII at 50, 60 and 70 deg C within a period of 48h. Pectin viscosity with high degree of methoxylation decreased sharply with the radiation dose remaining almost constant from 10 kGy. Pectin with low degree of methoxylation presented initially higher values of viscosity and the radiation induced decrease was also pronounced. Viscosity measurements decreased with the increase of the temperature applied for both kind of samples. The effect of radiation induced chain breaks generation in pectin molecules was evident through the viscosity reduction of irradiated pectin solutions although the viscosity presented diverse values depending of the degree of methoxylation of carboxyl groups in the backbone of polysaccharide macromolecules. (author)

  16. Demonstrating the sealing of a deep geologic repository: the RECAP project

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has operated an Underground Research Laboratory (URL) for twenty-three years (1982-2005). The URL was designed and constructed to carry out in situ geotechnical R and D needed for the Canadian Nuclear Fuel Waste Management program. The facility is now being closed, the final of several phases that have included siting, site evaluation, construction and operation. The closure phase presents a unique opportunity to develop and demonstrate the methodologies needed for closure and site restoration of a deep geologic repository for used nuclear fuel. A wealth of technical background and characterization data, dating back to before the first excavation work was carried out, are available to support closure activities. A number of closure-related activities are being proposed as part of a REpository Closure And Post-closure (RECAP) project. The RECAP project is proposed to include demonstrations of shaft and borehole sealing and monitoring as well as fracture sealing (grouting), room closure and monitoring system decommissioning, all activities that would occur when closing an actual repository. In addition to the closure-related activities, the RECAP project could provide a unique opportunity to conduct intrusion-monitoring demonstrations as part of a repository safeguards demonstration. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  19. The 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generating station-bundle manufacture and QA, fuel handling aspects, flasking and shipping and pie for the irradiated fuel, and follow-up documentation

    International Nuclear Information System (INIS)

    Korea Ministry of Science and Technology(MOST) has pushed and given a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel as one of Korea's National Nuclear Mid- and Long Term R and D Program. The Industrialization Program will be conducted for 3 years from 2000 November to efficiently utilize the CANFLEX fuel technology developed by KAERI and AECL jointly, where the KAERI's works have been conducted under the Korea's national program of the mid- and long-term nuclear R and D programs since 1992. This document is a report to guideline the following activities on the safety assessment for the 24 CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station: 'bundle manufacture and QA', 'Fuel handling aspects such as loading fuel, de-fuelling and segregation, and visual in-bay examinations', 'Flasking and shipping', 'Post-irradiation examination', and 'Follow-up documentation to be produced'

  20. Irradiation performance of (Th,U)02 fuel designed for advanced fuel cycle applications

    International Nuclear Information System (INIS)

    The reference fabrication route for Advanced Cycle thoria-based fuel is conventional in that it produces cold-pressed and sintered pellets. However we are also evaluating alternative fuels which offer the potential for simpler fabrication in a remote facility, and in some cases improved high burnup performance. These alternatives are impregnated, spherepac, and extruded thoria-based fuels. Spherepac fuel has been irradiated at a linear power of 50-60 kW/m to about 180 MW.h/kg H.E. There have been unexplained defects in fuel with both free-standing and collapsible cladding. Impregnated fuel has operated to 650 MW.h/kg H.E. at 50-60 KW/m. An experiment examining fuel from the sol-gel extrusion process has reached 450 MW.h/kg H.E. at a maximum linear power of 60 KW/m. The latter two experiments have operated without defects and with fission gas release less than that for U02 under identical conditions. The extruded fuel has a pellet geometry similar to that for conventional fuel and is AECL's first practical demonstration of thoria-based fuel with the fissile component distributed homogeneously on an atomic scale. We will continue monitoring the extruded fuel to a burnup approaching 1000 MW.h/kg H.E., as an indicator for the performance expected from co-precipitated (Th,U)02 or mechanically-mixed (Th,U)02 with good fissile homogeneity

  1. 2nd International technical meeting on small reactors

    International Nuclear Information System (INIS)

    The 2nd International Technical Meeting on Small Reactors was held on November 7-9, 2012 in Ottawa, Ontario. The meeting was hosted by Atomic Energy of Canada Limited (AECL) and Canadian Nuclear Society (CNS). There is growing international interest and activity in the development of small nuclear reactor technology. This meeting provided participants with an opportunity to share ideas and exchange information on new developments. This Technical Meeting covered topics of interest to designers, operators, researchers and analysts involved in the design, development and deployment of small reactors for power generation and research. A special session focussed on small modular reactors (SMR) for generating electricity and process heat, particularly in small grids and remote locations. Following the success of the first Technical Meeting in November 2010, which captured numerous accomplishments of low-power critical facilities and small reactors, the second Technical Meeting was dedicated to the achievements, capabilities, and future prospects of small reactors. This meeting also celebrated the 50th Anniversary of the Nuclear Power Demonstration (NPD) reactor which was the first small reactor (20 MWe) to generate electricity in Canada.

  2. Survey of in situ testing at underground laboratories with application to geologic disposal of spent fuel waste in crystalline rock

    International Nuclear Information System (INIS)

    This report is intended for use in designing testing programs, or as backup material for the review of 'R and D 92' which will be the next three-year plan for spent fuel repository siting and characterization activities in Sweden. There are eight major topics, each of which is addressed in a chapter of around 2000 to 10000 words. The major topics are defined to capture the reasons for testing, in a way that limits overlap between chapters. Other goals of this report are to provide current information on recent or ongoing tests in crystalline rock, and to describe insights which are important but not obvious from the literature. No data are presented, but the conclusions of testing programs are summarized. The principal sources were reports (in English) produced by the laboratory projects particularly the Stripa Project (SKB), the Underground Research Laboratory in Canada (AECL), and the Grimsel Test Site in Switzerland (Nagra). Articles from refereed journals have been used in lieu of project literature where possible and appropriate. (au)

  3. Addressing ethical considerations about nuclear fuel waste management

    International Nuclear Information System (INIS)

    Ethical considerations will be important in making decisions about the long-term management of nuclear fuel waste. Public discussions of nuclear fuel waste management are dominated by questions related to values, fairness, rights and responsibilities. To address public concerns, it is important to demonstrate that ethical responsibilities associated with the current management of the waste are being fulfilled. It is also important to show that our responsibilities to future generations can be met, and that ethical principles will be applied to the implementation of disposal. Canada's nuclear fuel waste disposal concept, as put forward in an Environmental Impact Statement by Atomic Energy of Canada Limited (AECL), is currently under public review by a Federal Environmental Assessment Panel. Following this review, recommendations will be made about the direction that Canada should take for the long-term management of this waste. This paper discusses the ethical principles that are seen to apply to geological disposal and illustrates how the Canadian approach to nuclear fuel waste management can meet the challenge of fulfilling these responsibilities. The author suggests that our ethical responsibilities require that adaptable technologies to site, design, construct, operate decommission and close disposal facilities should de developed. We cannot, and should not, present future generations from exercising control over what they inherit, nor control whether they modify or even reverse today's decisions if that is what they deem to be the right thing to do. (author)

  4. Identification of irradiated refrigerated pork with the DNA comet assay

    International Nuclear Information System (INIS)

    Food irradiation can contribute to a safer and more plentiful food supply by inactivating pathogens, eradicating pests and by extending shelf-life. Particularly in the case of pork meat, this process could be a useful way to inactivate harmful parasites such as Trichinella and Taenia solium. Ionizing radiation causes damage to the DNA of the cells (e.g. strand breaks), which can be used to detect irradiated food. Microelectrophoresis of single cells ('Comet Assay') is a simple and rapid test for DNA damage and can be used over a wide dose range and for a variety of products. Refrigerated pork meat was irradiated with a 60Co source, Gammacell 220 (A.E.C.L.) installed in IPEN (Sao Paulo, Brazil). The doses given were 0, 1.5, 3.0 and 4.5 kGy for refrigerated samples. Immediately after irradiation the samples were returned to the refrigerator (6 deg. C). Samples were kept in the refrigerator after irradiation. Pork meat was analyzed 1, 8 and 10 days after irradiation using the DNA 'Comet Assay'. This method showed to be an inexpensive and rapid technique for qualitative detection of irradiation treatment

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

    International Nuclear Information System (INIS)

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

  6. Effects of varying doses of gamma radiation on locally adapted Tradescantia clone 02 (BNL) (Brookhaven National Laboratory)

    International Nuclear Information System (INIS)

    This study determined the effects of gamma radiation on the meiotic cells of Tradescantia bracteata clone 02 (BNL). The flower buds collected were exposed through dosages ranging from 1 Gy to 5 Gy using gamma cell 220 machine (AECL) in a central axis position (c/a) and grown in Peralta's solution for three days. Out of the twenty buds designated for each dosages, ten buds were treated with 0.05% colchicine solution. The occurrence of micronuclei among the irradiated pollen mother cells suggested a linear relation with the quantity of radiation dose. The occurrence of MN among cells increased linearly from 1 Gy until it reached 3 Gy and 4 Gy. Beyond this maximum dose, cells were less responsive to the dose caused by inhibition of cell division, as demonstrated in the buds exposed to 5 Gy. This result was validated through the kruskal-Wallis test, where the computed h value was 3.44 (critical region of X20.05 = 9.49) Experimental results also showed chromosomal breaks, sticky chromosomes, and anaphase bridges in the pollen mother cells of irradiated buds. A significant numbers of cells were also found to have micronuclei, which may vary from 1 to 6 per pollen mother cell, and this showed no relationship with radiation dose. (Author)

  7. Materials chemical compatibility for the fabrication of small inherently safe nuclear reactors

    International Nuclear Information System (INIS)

    Aqueous nuclear fuels offer a unique set of characteristics for homogeneous reactor nuclear applications. Their advantages include high nuclear stability and inherent safety, high power density, high burn-up, simple preparation and reprocessing, easy fuel handling, high neutron economy, and simple control system leading to simple mechanical designs. The major disadvantages are corrosion, limited uranium concentration, and radiation decomposition of water. Likewise, organic coolants offer certain properties that are conducive for small reactor applications. These include reduced corrosion and activation, and low vapour pressures with good heat-transfer capabilities. Their major disadvantages are decomposition, fouling and flammability. A particular organic coolant, HB-40, has been extensively studied in Canada and was used for nineteen years in the 60-MWt organic-cooled WR-1 reactor at the Whiteshell Nuclear Research Establishment (WNRE) of Atomic Energy of Canada Limited (AECL). Proper attention to design and coolant chemistry in the nineteen years of operation in the WR-1 reactor kept the coolant aspects related to decomposition, fouling and flammability to acceptable levels. For small reactor applications, organic coolants are potentially superior to heavy water in terms of overall cost. The purpose of this thesis work was, through a literature review, to select the most suitable aqueous fuel and materials of construction for two proposed small inherently safe reactors, the QH-1 reactor and the homogeneous SLOWPOKE reactor under design at the Royal Military College of Canada.

  8. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

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

  9. An ESR study of the gamma radiolysis of aromatic polyesters containing isomeric naphthalene links

    International Nuclear Information System (INIS)

    Six polyesters were synthesised from 4,4'-oxy-bis(benzoyl chloride) and 1,4-, 1,5-, 1,6-, 2,3-, 2,6-, and 2,7-naphthalenediol isomers. The structures of the polyesters were characterised by means of IR, inherent viscosities in tetrachloroethane (TCE), solutions at 303 K and thermal analysis. The glass transition temperatures were in the range of 425-494 K by DSC thermal analysis. All of the polyesters were irradiated in an AECL Gammacell 220 unit at a dose rate of approximately 6.7 kGy/h to doses in the range of 0-15 kGy at 77 and 300 K. ESR spectroscopy was used to examine the radicals formed during radiolysis and to measure their yields. The G-values for radical formation in the polyesters were found to be in the range 0.18-1.41 at 77 K and 0.19-0.78 at 300 K. At 77 K, up to 15% of the radicals formed on radiolysis were found to be photo-bleachable anion radicals. Annealing experiments were carried out in order to identify the neutral radicals, which were assigned to naphthyl- or phenyl- and phenoxyl-type radicals

  10. Proceedings of the WIN-Global 2008 conference

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    WiN-France hosted the 16. WIN-Global conference May 26-30, 2008, in Marseille, France. The conference was attended by over 150 delegates, representing 30 countries. Canadian participants, from many diverse backgrounds, attended the annual conference from AECL, Bruce Power, CNSC, NB Power and OPG. The theme: Maintaining Key Competencies, Arising Key Competencies for Nuclear Energy: A Challenge and Opportunity for Diversity Development, emphasized the challenges ahead in providing a skilled workforce for the nuclear renaissance, as new build projects and a vast number of retirements are expected around the world within the next 5 years. The conference addressed such questions as 'How will nuclear, attract, develop and retain staff?' A technical tour of Marcoule invited conference attendees to visit one of: Atalante, a high level nuclear chemistry laboratory; Phenix, a fast breeding research reactor; or AVM, a vitrification plant. A subsequent technical tour visited Cadarache providing the opportunity to view ITER, the international fusion research project.

  11. Elements for evaluation of the potential of the thorium cycle in Argentina

    International Nuclear Information System (INIS)

    A comprehensive review of the most important elements to be taken into account for the evaluation and, eventually, the implementation of the introduction of thorium cycle strategies in argentinian heavy-water type power plants, and also of the associated development of the external fuel cycle, is presented. Particularly, the up-dated situations summarized here cover resources and prices of natural uranium and thorium, development of the various stages of the external fuel cycle, description of the most important strategies and their capabilities for the best use of mineral resources and, finally, the economic implications and the global comparison of those strategies. Various data and parameter values are added to those given in AECL's external reports. Some appendices are devoted to the definitions of a 'global fuel-cycle conversion factor' and to the analysis of the effective use of mineral resources, taking into account fissile and fertile material losses in the external fuel cycle., for the different cases of fuel conversion and breeding. (Author)

  12. Proceedings of the OECD/NEA/CSNI workshop on the implementation of hydrogen mitigation techniques

    Energy Technology Data Exchange (ETDEWEB)

    Koroll, G.W. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Rohde, J. [GRS, Koln (Germany); Royen, J. [OECD NEA, Issy-les-Moulineaux (France)

    1997-03-01

    The Workshop on the Implementation of Hydrogen Mitigation Techniques was held in Winnipeg, Manitoba,Canada from 1996 May 13 to 15. It was organized in collaboration with the Whiteshell Laboratories of Atomic Energy of Canada Limited (AECL), Ontario Hydro and the CANDU Owner's Group (COG). Sixty-five experts from twelve OECD Member countries and the Russian Federation attended the meeting. Papers presented in the sessions included topics: accident management and analysis, relevant aspects of hydrogen production, distribution and mixing, engineering, technology, possible side-effects consequences and new designs. The objectives of the Workshop were the following: to establish the state of the art of hydrogen mitigation techniques, with emphasis on igniters and catalytic recombiners; to exchange information on Member countries' strategies in managing hydrogen mitigation, and to establish dialogue as to differences in approach; to determine whether there is now an adequate technical basis for such strategies or whether more work is needed; to exchange information on future plans for implementation of hydrogen mitigation techniques.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  14. Evolution in performance assessment modeling as a result of regulatory review

    Energy Technology Data Exchange (ETDEWEB)

    Rowat, J.H.; Dolinar, G.M.; Stephens, M.E. [AECL Chalk River Labs., Ontario (Canada)] [and others

    1995-12-31

    AECL is planning to build the IRUS (Intrusion Resistant Underground Structure) facility for near-surface disposal of LLRW. The PSAR (preliminary safety assessment report) was subject to an initial regulatory review during mid-1992. The regulatory authority provided comments on many aspects of the safety assessment documentation including a number of questions on specific PA (Performance Assessment) modelling assumptions. As a result of these comments as well as a separate detailed review of the IRUS disposal concept, changes were made to the conceptual and mathematical models. The original disposal concept included a non-sorbing vault backfill, with a strong reliance on the wasteform as a barrier. This concept was altered to decrease reliance on the wasteform by replacing the original backfill with a sand/clinoptilolite mix, which is a better sorber of metal cations. This change lead to changes in the PA models which in turn altered the safety case for the facility. This, and other changes that impacted performance assessment modelling are the subject of this paper.

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

    Directory of Open Access Journals (Sweden)

    W. Shen

    2015-01-01

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

  16. The disposal of Canada's nuclear fuel waste: engineering for a disposal facility

    International Nuclear Information System (INIS)

    This report presents some general considerations for engineering a nuclear fuel waste disposal facility, alternative disposal-vault concepts and arrangements, and a conceptual design of a used-fuel disposal centre that was used to assess the technical feasibility, costs and potential effects of disposal. The general considerations and alternative disposal-vault arrangements are presented to show that options are available to allow the design to be adapted to actual site conditions. The conceptual design for a used-fuel disposal centre includes descriptions of the two major components of the disposal facility, the Used-Fuel Packaging Plant and the disposal vault; the ancillary facilities and services needed to carry out the operations are also identified. The development of the disposal facility, its operation, its decommissioning, and the reclamation of the site are discussed. The costs, labour requirements and schedules used to assess socioeconomic effects and that may be used to assess the cost burden of waste disposal to the consumer of nuclear energy are estimated. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

  17. Monitoring and information management system at the Underground Research Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Strobel, G.S.; Chernis, P.J.; Bushman, A.T.; Spinney, M.H.; Backer, R.J. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1996-07-01

    Atomic Energy of Canada Limited (AECL) has developed a customer oriented monitoring and information management system at the Underground Research Laboratory (URL) near Lac du Bonnet, Manitoba. The system is used to monitor instruments and manage, process, and distribute data. It consists of signal conditioners and remote loggers, central schedule and control systems, computer aided design and drafting work centres, and the communications linking them. The monitoring and communications elements are designed to meet the harsh demands of underground conditions while providing accurate monitoring of sensitive instruments to rigorous quality assured specifications. These instruments are used for testing of the concept for the deep geological disposal of nuclear fuel waste as part of the Canadian Nuclear Fuel Waste Management Program. Many of the tests are done in situ and at full-scale. The monitoring and information management system services engineering, research, and support staff working to design, develop, and demonstrate and present the concept. Experience gained during development of the monitoring and information management system at the URL, can be directly applied at the final disposal site. (author)

  18. From scientific evidence to radiation protection: a perspective of four decades

    International Nuclear Information System (INIS)

    I have had the good fortune to have been involved in a wide spectrum of radiation protection activities - instrument development, dosimetry and biokinetics, environmental radioactivity and biological effects (these four, the 'evidence' side of my title), and developments in practical radiological protection. In this short presentation, I shall highlight just some of these involvements. First will be the measurements of fallout and natural radioactivity that in 1959 started me in the business of radiological protection; second will be the R and D on tritium-related matters that occupied much of my hands-on research career through the 1960s and 1970s with AECL at Chalk River; and the final topic will be the studies involving the application of collective dose in radiological protection. The first two are examples of the R and D around the world that now supports the complex system of protection recommended by the ICRP. The third raises fundamental issues in the protection system, related to the assumption of linearity of response to dose, to individual variability and to the uncertainties in predictions of exposures and doses over long times. The current rapid advances in biological understanding of genetics and disease, while resolving some of these issues, may well lead to a more complex approach to protection, with a concomitant need for new directions in R and D. (author)

  19. Nuclear Legislation in OECD and NEA Countries. Regulatory and Institutional Framework for Nuclear Activities - Canada

    International Nuclear Information System (INIS)

    This country profile provide comprehensive information on the regulatory and Institutional Framework governing nuclear activities as well as a detailed review of a full range of nuclear law topics, including: mining regime; radioactive substances; nuclear installations; trade in nuclear materials and equipment; radiation protection; radioactive waste management; non-proliferation and physical protection; transport; and nuclear third party liability. The profile is complemented by reproductions of the primary legislation regulating nuclear activities in the country. Content: I. General regulatory regime: 1. Introduction (Licensing system; Offences, compliance and enforcement; Regulatory documents; Other relevant legislation); 2. Mining regime; 3. Nuclear substances and radiation devices; 4. Nuclear facilities; 5. Trade in nuclear materials and equipment (Exports, Other imports); 6. Radiation protection; 7. Radioactive waste management; 8. Non-proliferation and nuclear security; 9. Transport; 10. Nuclear third party liability; II. Institutional Framework: 1. Regulatory and supervisory authorities (Governor in council; Minister of natural resources; Other Ministerial authorities; Canadian Nuclear Safety Commission - CNSC); 2. Public and semi-public agencies (National Research Council - NRC; Natural Sciences and Engineering Research Council; Atomic Energy of Canada Ltd. - AECL)

  20. Temperature effect of DUPIC fuel in CANDU reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Nuclear fuel waste management - biosphere program highlights - 1978 to 1996

    International Nuclear Information System (INIS)

    The biosphere program in support of the development of the disposal concept for Canadian nuclear fuel waste since 1978 is scheduled for close-out. AECL's Environmental Science Branch (ESB) was mainly responsible for work in this program. In order to preserve as much information as possible, this report highlights many of the key achievements of the program, particularly those related to the development of the BIOTRAC biosphere model and its supporting research. This model was used for the assessment and review of the disposal concept in an environmental impact statement (EIS). The report also treats highlights related to alternative models, external scientific/technical reviews, EIS feedback, and the international BIOMOVS model validation program. Furthermore, it highlights basic aspects of future modelling and research needs in relation to siting a disposal facility. In this, feedback from the various reviews and the EIS is taken into account. Appendices of the report include listings of key ESB staff involved in the program, all the scientific/technical reports and papers produced under the program, contracts let to outside agencies, and issues raised by various participants or intervenors during the EIS review. Although the report is concerned with close-out of the biosphere program, it also provides valuable information for a continuing program concerned with siting a disposal facility. One of the conclusions of the report is that such a program is essential for successfully siting such a facility. (author)

  3. CATHENA Code Assessment for Pressure Tube and Calandria Tube Contact Phenomena

    International Nuclear Information System (INIS)

    Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA), has been validated against full-scale Contact Boiling Experiments conducted using specific channel power, pressure, and moderator subcooling as pre-test conditions. The pressure tube (PT) and calandria tube (CT) temperatures, the extent of dryout and failures of the pressure tube or the calandria tube (if any) are the outcome of these experiments. Recently, an IAEA International Collaborative Standard Problem (ICSP) to provide contact boiling experimental data to participants for assessing the subcooling requirements for a heated pressure tube, plastically deforming into contact with the calandria tube during a postulated large break LOCA condition has been performed. The CATHENA code assessment results against the experimental data distributed for the ICSP are provided in this paper. The CATHENA code is used to simulate the experiment on pressure tube ballooning conducted at the AECL. The overall code's predictions show good agreements with the experimental data. The contact timing by the pressure tube ballooning is predicted accurately, however, it is found that the code largely underpredict the peak temperature at the pressure tube and the calandria tube. This discrepancy seems to be induced from multi-dimensional flow effects in the water tank. For more accurate calculations, detailed modeling of the water tank is required

  4. Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

    International Nuclear Information System (INIS)

    The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU1 reactor fuel bundle. Based primarily on the CANFLEX2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

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

    International Nuclear Information System (INIS)

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

  6. Argo packing friction research update

    International Nuclear Information System (INIS)

    This paper focuses on the issue of valve packing friction and its affect on the operability of motor- and air-operated valves (MOVs and AOVs). At this time, most nuclear power plants are required to perform postmaintenance testing following a packing adjustment or replacement. In many cases, the friction generated by the packing does not impact the operability window of a valve. However, to date there has not been a concerted effort to substantiate this claim. To quantify the effects of packing friction, it has become necessary to develop a formula to predict the friction effects accurately. This formula provides a much more accurate method of predicting packing friction than previously used factors based strictly on stem diameter. Over the past 5 years, Argo Packing Company has been developing and testing improved graphite packing systems at research facilities, such as AECL Chalk River and Wyle Laboratories. Much of this testing has centered around reducing and predicting friction that is related to packing. In addition, diagnostic testing for Generic Letter 89-10 MOVs and AOVs has created a significant data base. In July 1992 Argo asked several utilities to provide running load data that could be used to quantify packing friction repeatability and predictability. This technical paper provides the basis to predict packing friction, which will improve calculations for thrust requirements for Generic Leter 89-10 and future AOV programs. In addition, having an accurate packing friction formula will improve packing performance when low running loads are identified that would indicate insufficient sealing force

  7. The effects of gamma radiation on soybean isoflavones contents

    International Nuclear Information System (INIS)

    Soybean (Glycine max) is the most common source of isoflavones in human feeding. It was suggested that there is a correlation among antioxidant activity of flavonoids and total phenolics content. Plants use isoflavones and their derivatives as part of the plant's defensive arsenal, to ward off disease-causing pathogenic fungi and other microbes. Highly processed foods made from legumes, such as tofu, retain most of their isoflavone content, with the exception of fermented miso, which has increased levels. Little is known about the influence of oxidative stress induced by radiation on the isoflavones contents. In the present paper, the effects of gamma irradiation on soybean isoflavones contents are presented. Samples from several Brazilian soybean cultivars were gamma irradiated with doses of 0, 1, 2, 5 e 10 kGy, dose rate about 3 kGy/h in a 60Co (Gammacell 220 - AECL). Isoflavones contents were determined after extraction with 70% ethanol containing 0.1% acetic acid by an HPLC method. The total isoflavone content remained almost unchanged with the increase of radiation dose up to 10 kGy. Although a general correlation among total isoflavone content and radiation dose was not found, some data suggest that for a few of the isoflavones from specific cultivars, the increase in the radiation dose induced a decrease in their content as for glucosyl glucosides and malonyl isoflavones, as well as an increase in their aglycone content. (author)

  8. Characterization of hydrophobic catalysts for hydrogen isotope exchange

    International Nuclear Information System (INIS)

    Domestic hydrophobic catalysts, KC-1 and KC-2, which were developed for the liquid phase catalytic exchange process separating hydrogen isotopes, were tested against Japanese catalyst, Kogel, which is being used in the Fugen's heavy water upgrader in Japan. KC-1 and KC-2 have different characteristics due to the differences of the solvent and solvent composition used. The test results of domestic hydrophobic catalysts characteristics such as pore distribution, specific surface area, platinum loading, and platinum dispersion from AECL agreed well with the results obtained by KEPRI/KAERI. The shape of KC-1 and KC-2 were 4x4 mm cylindrial pellet and that of Kogel catalyst was 4∼5.5mm sphere. The platinum loading of all catalysts were 0.8 wt%. The BET surface areas were 442, 247, 514m2 ·g-1 for KC-1, KC-2, and Kogel respectively, among which the BET surface area of KC-2 was the smallest. The platinum dispersion area was 2.47, 2.07, 1.90 m2g-1 and the platinum dispersion was 100, 100, 92% for KC-1, KC-2, and Kogel respectively, which showed domestic catalysts had higher values than Kogel catalyst. The average pore size was the largest in KC-2

  9. Professional aspects of nuclear safety

    International Nuclear Information System (INIS)

    Design and operation of nuclear facilities in Ontario are performed by professionals who have more at stake in the nuclear scene than the average resident of the province. Their technical expertise is constantly under scrutiny by their employers, the Atomic Energy Control Board, and the dissenting factions in the community. They and their families live close to nuclear facilities. It is highly unlikely that these professionals would assume a less than cautious approach to their work. The professional staff at both AECL-CANDU Operations and at Ontario Hydro have employee associations that date back many years. The presence of these associations has helped professional employees to divorce their labour-related concerns from their technical responsibilities to the advantage of the public. With the backing of their associations, the professional employees have encouraged the employers to sponsor career development programs to help them maintain state-of-the-art expertise. Employers have sponsored attendance and participation at technical seminars, many of them international. These benefits and privileges have contributed to improved standards in design, but most importantly the protection afforded by collective agreements to professional integrity has permitted engineers and other professionals to insist on the highest possible design standards

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  11. Sensorial analysis evaluation in cereal bars preserved by ionizing radiation processing

    Science.gov (United States)

    Villavicencio, A. L. C. H.; Araújo, M. M.; Fanaro, G. B.; Rela, P. R.; Mancini-Filho, J.

    2007-11-01

    Gamma-rays utilized as a food-processing treatment to eliminate insect contamination is well established in food industries. Recent troubles in Brazilian cereal bars commercialization require a special consumer's attention because some products were contaminated by insects. To solve the problem, food-irradiation treatment was utilized as a safe and effective solution. The final product was free of insect contamination. The aim of this study was to determine the best radiation dose processing utilized to disinfestations and detect some change on sensorial characteristic by sensorial analysis in cereal bars. In this study, three different kinds of cereal bars were purchased in São Paulo (Brazil) in supermarkets and irradiated with 1.0, 2.0 and 3.0 kGy at "Instituto de Pesquisas Energéticas e Nucleares" (IPEN-CNEN/SP). The samples were treated with ionizing radiation using a 60Co gamma-ray facility (Gammacell 220, A.E.C.L.). That radiation doses were used successfully as an anti-insect treatment in the cereal bars, since in some food industries doses up to 3.0 kGy are used to guarantee at least a dose of 1.0 kGy in internal cereal bars package. Sensorial analysis was necessary since cereal bars contain ingredients very sensitive to ionizing radiation process.

  12. X-ray diffraction residual stress measurement in the rolled-joint zone of Zr - 2.5 % Nb pressure tube

    International Nuclear Information System (INIS)

    The in-service experience of Zr - 2.5 % Nb pressure tubes in CANDU-type nuclear reactors has demonstrated very good performance over a long period of time. However, analyses done by AECL specialists on most failure cases, showed that a big percentage of defects are manufacturing defects, which appear mostly at the beginning of the rolled-joint zone. It has been observed that a correct rolling ensures an acceptable distribution of residual stress, but an incorrect one leads to an accumulation of big values of residual stress. This determines a preferential radial orientation of hydrides, which during operation in the reactor can produce DHC. To ensure a suitable performance of the Zr - 2.5 % Nb pressure tubes in the CANDU reactor, it is very important to have a correct rolling as mentioned in the procedure. This work presents a methodology for the measurement of the stressing state in the surfaces layers of the rolled-joint zone. The X-ray diffraction method can also be used for establishing the residual stress distribution across the tub wall, in order to ensure a good performance at Cernavoda nuclear plant. The results obtained for the investigated tube have led to the conclusion that the rolling process was correctly applied in this case, the values obtained for the residual stress being in good agreement with those accepted in literature. (Author) 2 Figs., 2 Tabs

  13. Recent advances in ultrasonic downcomer flow-measurement techniques for recirculating steam generators

    International Nuclear Information System (INIS)

    Non-intrusive ultrasonic measurements of downcomer flow velocity have been successfully used in the past to determine recirculation ratios and water inventory in CANDU steam generators. Knowledge of these process conditions allows operators to assess the effectiveness of maintenance programs, monitor the effects of tube fouling, and observe flow conditions following component modifications. It also provides designers with a means to verify or improve code predictions. Ultrasonic measurement systems have recently been installed on sixteen steam generators at the Bruce B Nuclear Generating Station, as part of an investigation into the possible effects of long-term boiler degradation. The most recent version of AECL's downcomer-flow technology was used, which features high-temperature transducers that are attached magnetically and then welded to the steam-generator outer shell. This method eliminates the complications of precision surface preparation, high-temperature couplants and awkward mechanical attachments. The paper will outline the method and summarize flow velocities measured during normal operation, over extended periods of time. It will also describe how the information might be used, e.g., to assess thermalhydraulic conditions, verify design calculations and support the case for reactor uprating. Further improvements that may allow the reliable measurement of flow in steam generators with steam carry-under are suggested, and preliminary results are presented from a dual-purpose single- and two-phase flow-measurement system. (author)

  14. Sensorial analysis evaluation in cereal bars preserved by ionizing radiation processing

    International Nuclear Information System (INIS)

    Gamma-rays utilized as a food-processing treatment to eliminate insect contamination is well established in food industries. Recent troubles in Brazilian cereal bars commercialization require a special consumer's attention because some products were contaminated by insects. To solve the problem, food-irradiation treatment was utilized as a safe and effective solution. The final product was free of insect contamination. The aim of this study was to determine the best radiation dose processing utilized to disinfestations and detect some change on sensorial characteristic by sensorial analysis in cereal bars. In this study, three different kinds of cereal bars were purchased in Sao Paulo (Brazil) in supermarkets and irradiated with 1.0, 2.0 and 3.0 kGy at 'Instituto de Pesquisas Energeticas e Nucleares' (IPEN-CNEN/SP). The samples were treated with ionizing radiation using a 60Co gamma-ray facility (Gammacell 220, A.E.C.L.). That radiation doses were used successfully as an anti-insect treatment in the cereal bars, since in some food industries doses up to 3.0 kGy are used to guarantee at least a dose of 1.0 kGy in internal cereal bars package. Sensorial analysis was necessary since cereal bars contain ingredients very sensitive to ionizing radiation process

  15. Radioactive waste management in Canada - Science, strategy, and outlook

    International Nuclear Information System (INIS)

    Nuclear electric power amounts to >15% of the electric power in Canada with 12,750-MW Canada deuterium uranium (CANDU) pressurized heavy-water reactors installed and 3,500 MW under construction. Three nuclear utilities, Ontario Hydro, Quebec Hydro, and New Brunswick Power, together with Atomic Energy of Canada Limited (AECL) and the nuclear industries share the burden of radioactive waste management in Canada. Radioactive wastes in Canada include uranium mine and mill tailings, low- and intermediate-level wastes from the nuclear fuel cycle and the Canadian nuclear industry, and used fuel from the once-through CANDU fuel cycle. In all cases, the objective is to manage these wastes in a way that creates minimal detriment to humans and the environment. In Canada, plans are for a gradual transition from interim facilities to long-term facilities, meeting the requirements of protection of humans and the environment in the long term. As in most nations, the solutions are evolving in a methodological manner, taking into account public sensitivities and concerns

  16. Status of the development of CANFLEX 0.9% SEU

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. (AECL) participates in international collaboration programs on slightly enriched uranium (SEU) with the Korean Atomic Energy Research Institute, British Nuclear Fuel plc. and most recently with Nucleoelectrica Argentina S. A. (NASA). In Argentina, NASA has successfully converted Atucha I from natural uranium to 0.85% SEU. Significant fuel cycle cost reductions were realized, and NASA wishes to explore similar concepts to reduce operating costs of its Embalse reactor. This collaboration covers the first phase of a 3-phase program. If the study confirms the feasibility and benefits of SEU, then the program could move onto a demonstration irradiation and, potentially, full-core implementation of SEU fuel at Embalse. This paper will provide an overview of the CANFLEX 0.9% fuel concept. Reactor physics assessment of the conversion of a natural uranium core to an enriched core will be presented. The feasibility of conducting a demonstration irradiation of SEU fuel in a reactor core of natural uranium fuel will be covered. Preliminary assessments of the safety and licensing implications will be summarized. (author)

  17. Development of Zirconium alloys (for pressure tubes)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1997-09-01

    The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

  18. Results of 14C analysis on low-level reference materials

    International Nuclear Information System (INIS)

    An invitation from the University of Glasgow to participate in an international intercomparison of 14C analysis and the increased need for quality assurance for measurements performed at the Low-Level Laboratory by AECL's Environmental Technologies Branch (ETB), Chalk River Laboratories, prompted the initiation of this work. Standard 14C samples from the IAEA and from the ETB were analyzed and compared to their reference values. All analyses agreed well with reported values and confirmed that the ETB can reliably measure 14C between 50 and 500 Bq/kg C. Samples from the University of Glasgow, Scotland, were also analyzed and reported. The results of counting various CO2 blank materials (dry ice from a CO2-siphon and limestone) showed that the source of CO2 does not have an effect on background count rate or counter efficiency. The ETB's lower limit of detection (LLD) has been calculated to be 10 -20 Bq/kg C and the determination limit (LQ) for samples during the ETB's routine operations has been documented to be 44 Bq/kg C. (author)

  19. The third generation CANDU control room

    International Nuclear Information System (INIS)

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

  20. Techniques developed to determine KIH of Zircaloy-4 cladding material

    International Nuclear Information System (INIS)

    Zircaloy-4, used as a fuel cladding material, is known to be susceptible to delayed hydride cracking (DHC). The study of the DHC mechanism and development of an approach to mitigate its occurrence, are of importance to the nuclear industry worldwide. Coordinated by the International Atomic Energy Agency (IAEA), an international research program was established in 2011 with the objective of experimentally determining the critical stress intensity factor (KIH) of DHC for various Zircaloy-4 cladding materials. Representing Canada, AECL Chalk River Laboratories (CRL) participates in this program. During 2011 to 2013, various techniques were developed at CRL with the objective of accurately determining the KIH of Pressurized Heavy Water Reactor (PHWR)-type Zircaloy-4 cladding and other zirconium-based cladding materials. These techniques include: 1) charging hydrogen into thin-wall test specimens with a gaseous approach, 2) determining hydrogen concentration in the specimens using differential scanning calorimetry, 3) fatigue pre-cracking of the specimens, and 4) establishing an empirical relationship between the stress-intensity factor (KI) and crack length of the specimens being studied. This paper describes the working principle of the techniques, and associated experimental results. (author)

  1. Fabrication of CANFLEX bundle kit for irradiation test in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon Sung; Kwon, Hyuk Il; Ji, Chul Goo; Chang, Ho Il; Sim, Ki Seob; Suk, Ho Chun

    1997-10-01

    CANFLEX bundle kit was prepared at KAERI for the fabrication of complete bundle at AECL. Completed bundle will be used for irradiation test in NRU. Provisions in the `Quality Assurance Manual for HWR Fuel Projects,` `Manufacturing Plan` and `Quality Verification, Inspection and Test Plan` were implemented as appropriately for the preparation of CANFLEX kit. A set of CANFLEX kit consist of 43 fuel sheath of two different sizes with spacers, bearing pads and buttons attached, 2 pieces of end plates and 86 pieces of end caps with two different sizes. All the documents utilized as references for the fabrication such as drawings, specifications, operating instructions, QC instructions and supplier`s certificates are specified in this report. Especially, suppliers` certificates and inspection reports for the purchased material as well as KAERI`s inspection report are integrated as attachments to this report. Attached to this report are supplier`s certificates and KAERI inspection reports for the procured materials and KAERI QC inspection reports for tubes, pads, spacers, buttons, end caps, end plates and fuel sheath. (author). 37 refs.

  2. Isotopic methods in hydrogeology and their application to the Underground Research Laboratory, Manitoba

    International Nuclear Information System (INIS)

    This review examines isotopic methods used to determine groundwater sources, residence times and processes of geochemical evolution that have been published in the international literature, with specific reference to AECL's experience in these methods and applications to groundwaters at the Underground Research Laboratory (URL), Manitoba. The program of groundwater sampling and analysis currently being planned for the URL area over the next several years will concentrate on specific isotopic measurements that may assist in understanding the groundwater flow system at the URL site. These results will add to the existing data for the URL area and indicate which isotopes are most useful when applied to the known groundwater flow system of the URL. This program of study is especially important because it not only uses standard geochemical and isotopic measurements (e.g., major ion, trace elements, 2H/18O, 14C, 34S) of groundwaters, but will determine values of more exotic and unusual ratios, such as 6Li/7Li, and B11/B10, whose potential for understanding groundwater geochemical evolution is largely unknown at present. In addition, the more established but equally complex methods of isotopic analysis, to determine 3He/4He, 36Cl/Cl and 129I/I, will be used to assess their potential for adding to the hydrogeochemical understanding of flow paths in crystalline rock. (author). 182 refs., 11 tabs., 27 figs

  3. CANDU 9 design

    International Nuclear Information System (INIS)

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

  4. Accuracy and Uncertainty Analysis of PSBT Benchmark Exercises Using a Subchannel Code MATRA

    Directory of Open Access Journals (Sweden)

    Dae-Hyun Hwang

    2012-01-01

    Full Text Available In the framework of the OECD/NRC PSBT benchmark, the subchannel grade void distribution data and DNB data were assessed by a subchannel code, MATRA. The prediction accuracy and uncertainty of the zone-averaged void fraction at the central region of the 5 × 5 test bundle were evaluated for the steady-state and transient benchmark data. Optimum values of the turbulent mixing parameter were evaluated for the subchannel exit temperature distribution benchmark. The influence of the mixing vanes on the subchannel flow distribution was investigated through a CFD analysis. In addition, a regionwise turbulent mixing model was examined to account for the nonhomogeneous mixing characteristics caused by the vane effect. The steady-state DNB benchmark data with uniform and nonuniform axial power shapes were evaluated by employing various DNB prediction models: EPRI bundle CHF correlation, AECL-IPPE 1995 CHF lookup table, and representative mechanistic DNB models such as a sublayer dryout model and a bubble crowding model. The DNBR prediction uncertainties for various DNB models were evaluated from a Monte-Carlo simulation for a selected steady-state condition.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  6. Multivariable control in nuclear power stations -survey of design methods

    International Nuclear Information System (INIS)

    The development of larger nuclear generating stations increases the importance of dynamic interaction between controllers, because each control action may affect several plant outputs. Multivariable control provides the techniques to design controllers which perform well under these conditions. This report is a foundation for further work on the application of multivariable control in AECL. It covers the requirements of control and the fundamental mathematics used, then reviews the most important linear methods, based on both state-space and frequency-response concepts. State-space methods are derived from analysis of the system differential equations, while frequency-response methods use the input-output transfer function. State-space methods covered include linear-quadratic optimal control, pole shifting, and the theory of state observers and estimators. Frequency-response methods include the inverse Nyquist array method, and classical non-interactive techniques. Transfer-function methods are particularly emphasized since they can incorporate ill-defined design criteria. The underlying concepts, and the application strengths and weaknesses of each design method are presented. A review of significant applications is also given. It is concluded that the inverse Nyquist array method, a frequency-response technique based on inverse transfer-function matrices, is preferred for the design of multivariable controllers for nuclear power plants. This method may be supplemented by information obtained from a modal analysis of the plant model. (auth)

  7. Panel presentation: innovations in nuclear industry restructuring

    International Nuclear Information System (INIS)

    Innovations in nuclear industry structuring is the theme of this panel presentation and I would like to take a few minutes to share with you one of the recent innovations in the Canadian nuclear industry. Namely, the creation of NPM Nuclear Managers Canada Inc. The company mandate and charter is to specialize in providing expert project management, construction management and commissioning management services for CANDU nuclear power projects world-wide. Nuclear Project Managers was incorporated in 1982 by AECL and four of Canada's largest engineering, construction, and management companies, who have been participating in the Canadian nuclear industry. NPM through its participating companies, which include its owners as well as Canatom, represents a resource base of more than 15 000 professional personnel, with skills and experience in all disciplines required to successfully manage large complex projects, such as CANDU power stations. This large technical resource assures that NPM can provide the qualified project team essential for successful project implementation. In addition to staff, the participant companies of NPM also provide proven operating policies and control systems

  8. Canada's disposal concept

    International Nuclear Information System (INIS)

    A concept for the safe and permanent disposal of nuclear fuel wastes from Canada's CANDU reactors has been developed by Atomic Energy of Canada Ltd. (AECL). The waste would be placed in an engineered disposal vault 500 to 1000 m below the surface in plutonic rock of the Canadian Shield. The multiple barriers to retain the waste and retard the release of radioactivity would be the waste form, the containers, buffer and backfill, and the rock overlying the vault. Numerous research programmes have been carried out to develop the technology for the concept. These include work on materials corrosion and failure mechanisms to assess the performance of the used fuel containers. Predictive modelling has shown that more than 97% of ASTM Grade 2 titanium containers will retain their integrity, even under pessimistic assumptions, for 1200-6000 years after emplacement, and even longer times may be achieved with other grades of titanium or copper. Other research has been aimed at vault sealing, at site characterization for an underground research laboratory and at the development of a methodology for assessing radiological and environmental effects after closure of the facility. A review of the safety and environmental impacts of the concept is now being carried out by an independent panel appointed by the government. (2 figures, 3 references) (UK)

  9. 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.

  10. Does nuclear energy have a role in the development of Canada's oil sands?

    International Nuclear Information System (INIS)

    The Canadian Energy Research Institute (CERI) completed a study for Atomic Energy of Canada Limited (AECL) that compares the economics of a modified ACR-700 Advanced CANDU Reactor with the economics of a natural gas-fired facility to supply steam to a hypothetical Steam Assisted Gravity Drainage (SAGD) project located in northeastern Alberta. The results were initially presented at the Petroleum Society's Canadian International Petroleum Conference 2003, Calgary, Alberta, Canada, June 10-12, 2003. The comparison was made by using discounted cash-flow methodology to estimate the levelized unit cost of steam that could be supplied to the SAGD project from either a nuclear or a gas-fired facility. The unit cost of steam was determined by treating the steam supply facility as a standalone business; it would ensure that all costs are recovered including capital costs, operating costs, fuel costs, and a return on investment. The study indicated that steam supply form an ACR-700 nuclear facility is economically competitive with stea supply from a gas-fired facility. An examination of key variables indicated that the cost of steam form the nuclear facility is very sensitive to the capital cost of the facility, while the cost of steam from the gas-fired facility is very sensitive to the price of natural gas and possible Kyoto Protocol compliance costs. (author)

  11. On the structure of Lattice code WIMSD-5B

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo

    2004-03-15

    The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel

  12. Operational improvements in the CANDU 9 control centre

    International Nuclear Information System (INIS)

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

  13. CANDU-OCR power station options and costs

    International Nuclear Information System (INIS)

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

  14. Identification of irradiated refrigerated pork with the DNA comet assay

    Science.gov (United States)

    Araújo, M. M.; Marin-Huachaca, N. S.; Mancini-Filho, J.; Delincée, H.; Villavicencio, A. L. C. H.

    2004-09-01

    Food irradiation can contribute to a safer and more plentiful food supply by inactivating pathogens, eradicating pests and by extending shelf-life. Particularly in the case of pork meat, this process could be a useful way to inactivate harmful parasites such as Trichinella and Taenia solium. Ionizing radiation causes damage to the DNA of the cells (e.g. strand breaks), which can be used to detect irradiated food. Microelectrophoresis of single cells (``Comet Assay'') is a simple and rapid test for DNA damage and can be used over a wide dose range and for a variety of products. Refrigerated pork meat was irradiated with a 60Co source, Gammacell 220 (A.E.C.L.) installed in IPEN (Sa~o Paulo, Brazil). The doses given were 0, 1.5, 3.0 and 4.5kGy for refrigerated samples. Immediately after irradiation the samples were returned to the refrigerator (6°C). Samples were kept in the refrigerator after irradiation. Pork meat was analyzed 1, 8 and 10 days after irradiation using the DNA ``Comet Assay''. This method showed to be an inexpensive and rapid technique for qualitative detection of irradiation treatment.

  15. Identification of irradiated refrigerated pork with the DNA comet assay

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, M.M. E-mail: villavic@net.ipen.br; Marin-Huachaca, N.S.; Mancini-Filho, J. E-mail: jmancini@usp.br; Delincee, H.; Villavicencio, A.L.C.H. E-mail: henry.delincee@bfe.uni-karlsruhe.de

    2004-10-01

    Food irradiation can contribute to a safer and more plentiful food supply by inactivating pathogens, eradicating pests and by extending shelf-life. Particularly in the case of pork meat, this process could be a useful way to inactivate harmful parasites such as Trichinella and Taenia solium. Ionizing radiation causes damage to the DNA of the cells (e.g. strand breaks), which can be used to detect irradiated food. Microelectrophoresis of single cells ('Comet Assay') is a simple and rapid test for DNA damage and can be used over a wide dose range and for a variety of products. Refrigerated pork meat was irradiated with a {sup 60}Co source, Gammacell 220 (A.E.C.L.) installed in IPEN (Sao Paulo, Brazil). The doses given were 0, 1.5, 3.0 and 4.5 kGy for refrigerated samples. Immediately after irradiation the samples were returned to the refrigerator (6 deg. C). Samples were kept in the refrigerator after irradiation. Pork meat was analyzed 1, 8 and 10 days after irradiation using the DNA 'Comet Assay'. This method showed to be an inexpensive and rapid technique for qualitative detection of irradiation treatment.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  18. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

  19. Public acceptance of small reactors

    International Nuclear Information System (INIS)

    The success of any nuclear program requires acceptance by the local public and all levels of government involved in the decision to initiate a reactor program. Public acceptance of a nuclear energy source is a major challenge in successful initiation of a small reactor program. In AECL's experience, public acceptance will not be obtained until the public is convinced that the specific nuclear program is needed, safe and economic and environmental benefit to the community. The title of public acceptance is misleading. The objective of the program is a fully informed public. The program proponent cannot force public acceptance, which is beyond his control. He can, however, ensure that the public is informed. Once information has begun to flow to the public by various means as will be explained later, the proponent is responsible to ensure that the information that is provided by him and by others is accurate. Most importantly, and perhaps most difficult to accomplish, the proponent must develop a consultative process that allows the proponent and the public to agree on actions that are acceptable to the proponent and the community

  20. The Comparative Study of Fuel Cost of The NPPs Proposed by Vendors

    International Nuclear Information System (INIS)

    The comparative study on fuel cost of the NPPs proposed by Vendors in the Feasibility Study of the first NPPs in Indonesia has been performed. The fuel cost is calculated based on the levelized cost -constant money method for single fuel batch at equilibrium core condition. The sensitivity study has also been performed in order to check the effects of the dominant economic parameters on fuel cost changing. For open cycle as well as closed cycle the study results generally show that: (a) the fuel cost of the Heavy Water Reactor proposed by AECL (PHWR and CANDU3) is 6 -43% lower compared with that of the Light Water Reactor proposed by MHl/WH, NPI, WH and GE; (b) among the Light Water Reactors, SBWR and ABWR-1000 proposed by GE have a fuel cost of 15 -300/0 higher than that of PWR proposed by MHl/WH, NPI and WH; (c) PWR proposed by NPI (Siemens and Framatome) has a fuel cost of 6 -9% lower than that of PWR proposed by MHl/WH and WH; (d) for the Light Water Reactor, the fuel cost for the closed cycle is 4 -6% higher compared with that of the open cycle; and (e) fuel cost of the Light Water Reactor is more sensitive to discount rate change compared with that of the Heavy Water Reactor

  1. Uncertainty analysis guide

    Energy Technology Data Exchange (ETDEWEB)

    Andres, T.H

    2002-05-01

    This guide applies to the estimation of uncertainty in quantities calculated by scientific, analysis and design computer programs that fall within the scope of AECL's software quality assurance (SQA) manual. The guide weaves together rational approaches from the SQA manual and three other diverse sources: (a) the CSAU (Code Scaling, Applicability, and Uncertainty) evaluation methodology; (b) the ISO Guide,for the Expression of Uncertainty in Measurement; and (c) the SVA (Systems Variability Analysis) method of risk analysis. This report describes the manner by which random and systematic uncertainties in calculated quantities can be estimated and expressed. Random uncertainty in model output can be attributed to uncertainties of inputs. The propagation of these uncertainties through a computer model can be represented in a variety of ways, including exact calculations, series approximations and Monte Carlo methods. Systematic uncertainties emerge from the development of the computer model itself, through simplifications and conservatisms, for example. These must be estimated and combined with random uncertainties to determine the combined uncertainty in a model output. This report also addresses the method by which uncertainties should be employed in code validation, in order to determine whether experiments and simulations agree, and whether or not a code satisfies the required tolerance for its application. (author)

  2. Proceedings of the OECD/NEA/CSNI workshop on the implementation of hydrogen mitigation techniques

    International Nuclear Information System (INIS)

    The Workshop on the Implementation of Hydrogen Mitigation Techniques was held in Winnipeg, Manitoba,Canada from 1996 May 13 to 15. It was organized in collaboration with the Whiteshell Laboratories of Atomic Energy of Canada Limited (AECL), Ontario Hydro and the CANDU Owner's Group (COG). Sixty-five experts from twelve OECD Member countries and the Russian Federation attended the meeting. Papers presented in the sessions included topics: accident management and analysis, relevant aspects of hydrogen production, distribution and mixing, engineering, technology, possible side-effects consequences and new designs. The objectives of the Workshop were the following: to establish the state of the art of hydrogen mitigation techniques, with emphasis on igniters and catalytic recombiners; to exchange information on Member countries' strategies in managing hydrogen mitigation, and to establish dialogue as to differences in approach; to determine whether there is now an adequate technical basis for such strategies or whether more work is needed; to exchange information on future plans for implementation of hydrogen mitigation techniques

  3. Underground Research Laboratory room 209 instrument array. Vol. 1,2

    International Nuclear Information System (INIS)

    An in situ excavation response test was conducted at the 240 Level of the Underground Research Laboratory (URL). The test was carried out in conjunction with the drill-and-blast excavation of a near-circular tunnel (Room 209), about 3.5 m in diameter. The tunnel was excavated through a tunnel axis. Three modelling groups made predictions of the response of the rock mass and hydraulic behaviour of the water-bearing fracture to excavation. The tunnel was excavated in two stages, a pilot tunnel followed by a slash, providing two complete sets of response measurements. Careful excavation was carried out to ensure the excavation shape after each blast round agreed closely with the planned shape incorporated in the numerical models. Instrumentation installed before the tunnel was extended monitored the complete strain tensor at eight locations around the tunnel, radial displacements and piezometric pressures at nine locations in the fracture. As well, tunnel convergence, water flows from the fracture, and hydraulic conductivity of the fracture at nine locations, were measured after each excavation step. The final tunnel profiles were accurately surveyed, and the geology was mapped in detail. The results are presented in this report for comparison with the modellers' predictions (reported in AECL--9566-2). Some preliminary conclusions and recommendations regarding the field testing are presented

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

    International Nuclear Information System (INIS)

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

  5. International

    International Nuclear Information System (INIS)

    This rubric reports on 10 short notes about international economical facts about nuclear power: Electricite de France (EdF) and its assistance and management contracts with Eastern Europe countries (Poland, Hungary, Bulgaria); Transnuclear Inc. company (a 100% Cogema daughter company) acquired the US Vectra Technologies company; the construction of the Khumo nuclear power plant in Northern Korea plays in favour of the reconciliation between Northern and Southern Korea; the delivery of two VVER 1000 Russian reactors to China; the enforcement of the cooperation agreement between Euratom and Argentina; Japan requested for the financing of a Russian fast breeder reactor; Russia has planned to sell a floating barge-type nuclear power plant to Indonesia; the control of the Swedish reactor vessels of Sydkraft AB company committed to Tractebel (Belgium); the renewal of the nuclear cooperation agreement between Swiss and USA; the call for bids from the Turkish TEAS electric power company for the building of the Akkuyu nuclear power plant answered by three candidates: Atomic Energy of Canada Limited (AECL), Westinghouse (US) and the French-German NPI company. (J.S.)

  6. Comparison of operating phase instrument readings with decommissioning observations in the clay bulkhead portion of the tunnel sealing experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, D.A.; Chernis, P.J.; Martino, J.B. [Atomic Energy of Canada Limited, AECL (Canada); Vignal, B. [Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA), 92 - Chatenay Malabry (France); Masumoto, K. [Kajima Technical Research Institute, Tokyo (Japan); Fujita, T. [Japan Nuclear Cycle Development Institute (JNC) (Japan)

    2005-07-01

    A major international experiment, demonstrating technologies for tunnel sealing at full-scale was conducted at Canada Underground Research Laboratory between 1998 and 2004. The participants in this Tunnel Sealing Experiment (TSX) were AECL, ANDRA, JNC and during the pre-thermal phase o f the experiment the USDoE (via the Waste Isolation Pilot Project (WIPP)). Two bulkheads were installed; one consisted of high-performance concrete and the other of highly compacted sand-bentonite material. The performance of these two bulkheads was monitored throughout the experiment in order to evaluate the influence of elevated hydraulic head (4 MPa) and chamber temperature (up to 85 C) on these materials. This paper compares the conditions measured by the monitoring instruments installed in and around the clay bulkhead portion of the TSX with the conditions observed during decommissioning. Instrumentation installed within and surrounding these bulkheads monitored the stress, strain, moisture, temperature and water transport through and adjacent to the clay bulkhead. This allowed for detailed interpretations of the evolution of the bulkhead in the course of the experiment operation. At the time of decommissioning of the TSX a sampling plan was implemented for taking many thousands of samples to allow generation of detailed end-of-test density, moisture content and chemical tracer profiles. (authors)

  7. Verification of safety critical software

    International Nuclear Information System (INIS)

    To assure quality of safety critical software, software should be developed in accordance with software development procedures and rigorous software verification and validation should be performed. Software verification is the formal act of reviewing, testing of checking, and documenting whether software components comply with the specified requirements for a particular stage of the development phase[1]. New software verification methodology was developed and was applied to the Shutdown System No. 1 and 2 (SDS1,2) for Wolsung 2,3 and 4 nuclear power plants by Korea Atomic Energy Research Institute(KAERI) and Atomic Energy of Canada Limited(AECL) in order to satisfy new regulation requirements of Atomic Energy Control Boars(AECB). Software verification methodology applied to SDS1 for Wolsung 2,3 and 4 project will be described in this paper. Some errors were found by this methodology during the software development for SDS1 and were corrected by software designer. Outputs from Wolsung 2,3 and 4 project have demonstrated that the use of this methodology results in a high quality, cost-effective product. 15 refs., 6 figs. (author)

  8. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

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

  9. Post-irradiation examination of MOX fuel with varying plutonium homogeneity

    International Nuclear Information System (INIS)

    The Atomic Energy of Canada Limited (AECL) Pu-containing mixed-oxide (MOX) fuel program includes fuel fabrication development, irradiation testing, post-irradiation examination (PIE), reactor physics and fuel-management studies. The BDL-446 experiment investigates one particular fabrication parameter, plutonium homogeneity, and its effect on fuel performance. Three different distributions of Pu in the UO2 matrix were tested: pure PuO2 particles within the matrix, regions of master mix particles (containing an intermediate Pu concentration in UO2) within the matrix, and a homogeneous, solid solution of (U+Pu)O2. The fuel pellets had a Pu content of 1.35 wt.% Pu in total heavy elements (HE), mixed with depleted uranium powder. The irradiation test began at outer element linear power ratings of ~52 kW/m, declining to a final linear power rating of ~21 kW/m at burnups of ~500 MWh/kgHE (21 MWd/kgHE). This paper discusses the PIE results of various fuel types. The PIE results showed that the performance of all three types of MOX fuel is typical of UO2 fuel irradiated under similar conditions of power and burnup. There is a difference in the fission gas release among the different fuel types, with the MOX fuel containing regions of pure Pu appearing to have higher gas release relative to the other two. (author)

  10. Progress in food irradiation: South Africa

    International Nuclear Information System (INIS)

    The most significant development to date is the commissioning of a pilot plant for the treatment of subtropical fruits at Tzaneen (N. Transvaal) in January 1977. The purpose of this facility is to generate commercial quantities of treated produce for process and product evaluation. The irradiator consists of a six-metre-deep pool filled with water. The product is lowered in two large, watertight containers which surround the plaque source. To provide for dose homogeneity, the containers are rotated at half the exposure time. With the present loading of cobalt-60, the treatment time for one batch of approximately 350 kg of fruit is approximately one hour. The pilot plant has been successfully used this year to treat approximately 6.5 t of mangoes in the commercial sea-shipment trial, as well as several small consignments of avocados and papayas sent to local markets for evaluation. The existing facilities are the same as described in the previous report. The commercial sterilization plant (AECL JS6500) has been up-graded to 400 kCi of cobalt-60. The research loop in this plant has been utilized to treat 10.5 t of potatoes which were supplied to local supermarkets for trial sale purposes. (orig.)

  11. Workers moving the industry forward

    International Nuclear Information System (INIS)

    The Power Workers' Union represents workers at Ontario Hydro's nuclear stations and AECL operators at Chalk River. Although labour relations are far from perfect, the union does its best to protect the industry. Avoiding confrontation as much as possible, this union is happy to be regarded as a partner in the business. The union is impressed by the consultants' report on Ontario Hydro's nuclear operations. Whatever the future may bring, the present is not really pleasant for nuclear workers generally, in that the work itself is very demanding technically, and must be performed with great diligence because the responsibility for safety is enormous. Considering the actual safety record, some caricatures or ''cheap shots'' from antinuclear politicians and special interest groups seem quite offensive. As a partner in public relations, the union has produced draft fact sheets on topics such as: transporting radioactive material; the burning of plutonium from dismantled weaponry; deep geological storage of nuclear waste; the sale of Candu reactors to China. The author closes with some advice on how to improve industrial relations, based on the union's experience

  12. Corrosion database for SCWR development

    International Nuclear Information System (INIS)

    The development of the Gen IV CANDU-Supercritical Water Reactor (SCWR) requires the identification and evaluation of candidate materials that can be used for in-core and out-of-core components. One initial goal of the Materials Database project under the Canadian National Gen IV program is to identify a short-list of candidate materials for longer-term testing. A large amount of data is now available on materials properties (e.g, corrosion, creep) under SCWR conditions (for temperature between 374 and 732°C, P = 35 MPa). To facilitate the collection and assessment of these data, a materials database has been developed by AECL and MTL. This paper describes the development of this database, outlining basic data requirements and design. Illustrations include the functional view of the database, query tables and user interface plots. Examples of corrosion rate assessments of some 3XX stainless steels and of Alloy 800H under various test conditions will be presented. The project is still in an early stage and development is underway, including data collection and Visual Basic programming. Even at this preliminary stage, the database is proving to be a valuable tool for the corrosion evaluation of alloys. (author)

  13. Micromechanical Modeling of Anisotropic Damage-Induced Permeability Variation in Crystalline Rocks

    Science.gov (United States)

    Chen, Yifeng; Hu, Shaohua; Zhou, Chuangbing; Jing, Lanru

    2014-09-01

    This paper presents a study on the initiation and progress of anisotropic damage and its impact on the permeability variation of crystalline rocks of low porosity. This work was based on an existing micromechanical model considering the frictional sliding and dilatancy behaviors of microcracks and the recovery of degraded stiffness when the microcracks are closed. By virtue of an analytical ellipsoidal inclusion solution, lower bound estimates were formulated through a rigorous homogenization procedure for the damage-induced effective permeability of the microcracks-matrix system, and their predictive limitations were discussed with superconducting penny-shaped microcracks, in which the greatest lower bounds were obtained for each homogenization scheme. On this basis, an empirical upper bound estimation model was suggested to account for the influences of anisotropic damage growth, connectivity, frictional sliding, dilatancy, and normal stiffness recovery of closed microcracks, as well as tensile stress-induced microcrack opening on the permeability variation, with a small number of material parameters. The developed model was calibrated and validated by a series of existing laboratory triaxial compression tests with permeability measurements on crystalline rocks, and applied for characterizing the excavation-induced damage zone and permeability variation in the surrounding granitic rock of the TSX tunnel at the Atomic Energy of Canada Limited's (AECL) Underground Research Laboratory (URL) in Canada, with an acceptable agreement between the predicted and measured data.

  14. Safe, permanent disposal of used CANDU fuel

    International Nuclear Information System (INIS)

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

  15. Radiation processing in Indonesia

    International Nuclear Information System (INIS)

    The laboratory experiments on the application of γ radiation in industrial process have already been started since a few years ago using Gamma Cell 220 from AECL Canada. The fields of study are those which the radiation application can have a maximum beneficial impact to the country such as radiation polymerization of natural latex, natural rubber, polypropylene fibers, cotton, and rayon fabrics. The following results are shown and discussed. 1. Dipping trials with irradiated concentrates and irradiated field latex gave products with high ultimate elongation, low modulus, and high permanent set. 2. The grafting of acrylonitrile onto natural rubber gave a product with better oil resistance and tensile strength. 3. Moisture regain, dye absorbtion, and the melting point of grafted polypropylene fibers were found to increase with the increasing degree of grafting. 4. Graft copolymerization of monomers into cellulose fabrics could improve the dimensional stability and crease recovery angle of the fabrics, while its tensile strength and its abrasion decreased with the increasing irradiation dose. (author)

  16. Technical report on implementation of reactor internal 3D modeling and visual database system

    International Nuclear Information System (INIS)

    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation's integrated computer aided engineering system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Power's PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new

  17. Qualification of Programmable Electronic System (PES) equipment based on international nuclear I and C standards

    International Nuclear Information System (INIS)

    Nuclear power plants (NPPs) are increasingly faced with the challenge of qualifying procured equipment, sub-components, and systems that contain digital programmed electronics for use in safety-related applications. Referred to as a 'programmable electronic system' (PES), such equipment typically contains both complex logic that is vulnerable to systematic design faults, and low voltage electronics hardware that is subject to random faults. Procured PES products or components are often only commercial grade, yet can offer reliable cost effective alternatives to custom-designed or nuclear qualified equipment, provided they can be shown to meet the quality assurance, functional safety, environmental, and reliability requirements of a particular application. The process of confirming this is referred to as application-specific product qualification (ASPQ) and can be challenging and costly. This paper provides an overview of an approach that has been developed at Atomic Energy Canada Limited (AECL) and successfully applied to PES equipment intended for use in domestic Candu R 6 nuclear power plants and special purpose reactors at Chalk River Laboratories. The approach has evolved over the past decade and has recently been adapted to be consistent with, and take advantage of new standards that are applicable to nuclear safety-related I and C systems. Also discussed are how recognized third-party safety-certifications of PES equipment to International Electrotechnical Commission (IEC) standards, and the assessment methods employed, may be used to reduce ASPQ effort. (authors)

  18. 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.

  19. The nuclear battery program: Progress and future possibilities

    International Nuclear Information System (INIS)

    The Nuclear Battery is an advanced small reactor being developed by Atomic Energy of Canada Limited (AECL) to produce electricity, and/or high-temperature steam heat, in locations remote form utility grids or natural gas pipelines. It features a novel passive primary heat transport system based on liquid-metal heat pipes, extraordinary passive safety based on the use of coated-particle fuel, and burnable neutron poisons in a solid graphite moderator. The reference design is capable of producing about 600 kW of electricity or about 2400 kW of steam heat in a base-load mode for 15 full-power years without refuelling. This paper reviews the technical progress, present activities and future goals of the Nuclear Battery R and D support program. Also, some of the alternate design approaches to increase the thermal power output from the Nuclear Battery and, hence, to further reduce its unit energy cost, are briefly outlined and compared. (author). 8 refs., 4 figs

  20. Wet sipping system at Wolsong-1

    International Nuclear Information System (INIS)

    After many years of operation, the on-power failed fuel detection and location systems along with alarm area gamma monitors at Wolsong-1 have successfully demonstrated that most, if not all, defective and suspect fuel bundles can be located before discharge to the fuel bay. Today, discharged bundles are now being transferred from the fuel bay to the AECL designed Modular Air-Cooled Storage (MACSTOR) canister facilities. Since these canisters are licensed for storing intact fuel bundles only, a procedure was needed at Wolsong-1 to separate any suspect or defective bundles that do not release fission products in detectable quantities. Therefore, KNF designed and built a wet sipper to enclose an irradiated bundle inside a sealed container at the bottom of the fuel bay. Various techniques were then used to enhance the release of water soluble fission products from defective fuel elements before circulating water samples from the immediate vicinity of an irradiated fuel bundle to an inspection station located at the top of the fuel bay. Any water samples with elevated levels of gamma activity were direct indications of a fuel cladding breach. The presence of defective fuel elements were then verified by visual inspection. The system performance test was performed in the Wolsong-1 nuclear power plant on March 2009.This paper describes the results of the wet sipping tests. (author)

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

    International Nuclear Information System (INIS)

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

  2. Gamma radiation influence on technological characteristics of wheat flour

    Science.gov (United States)

    Teixeira, Christian A. H. M.; Inamura, Patricia Y.; Uehara, Vanessa B.; Mastro, Nelida L. d.

    2012-08-01

    This study aimed at determining the influence of gamma radiation on technological characteristics of wheat (Triticum sativum) flour and physical properties of pan breads made with this flour. The bread formulation included wheat flour, water, milk, salt, sugar, yeast and butter. The α-amylase activity of wheat flour irradiated with 1, 3 and 9 kGy in a Gammacell 220 (AECL), one day, five days and one month after irradiation was evaluated. Deformation force, height and weight of breads prepared with the irradiated flour were also determined. The enzymatic activity increased—reduction of falling number time—as radiation dose increased, their values being 397 s (0 kGy), 388 s (1 kGy), 343 s (3 kGy) and 293 s (9 kGy) respectively, remaining almost constant over the period of one month. Pan breads prepared with irradiated wheat flour showed increased weight. Texture analysis showed that bread made of irradiated flour presented an increase in maximum deformation force. The results indicate that wheat flour ionizing radiation processing may confer increased enzymatic activity on bread making and depending on the irradiation dose, an increase in weight, height and deformation force parameters of pan breads made of it.

  3. Effects of irradiation on natural antioxidants of cinnamon ( Cinnamomum zeylanicum N.)

    Science.gov (United States)

    Kitazuru, E. R.; Moreira, A. V. B.; Mancini-Filho, J.; Delincée, H.; Villavicencio, A. L. C. H.

    2004-09-01

    Food irradiation to reduce the number of spoilage microorganisms and insects is an ionizing process that induces free radical formation in proteins, lipids, carbohydrates and other molecular structures in food. Antioxidants generally decrease the level of oxidation in such systems by transferring hydrogen atoms to the free radical structure. In the present paper, the effect of ionizing radiation on natural cinnamon antioxidants is studied. Cinnamon samples were purchased from retailers and irradiated with a 60Co source, Gammacell 220 (A.E.C.L.) installed at IPEN (São Paulo, Brazil) using 0, 5, 10, 15, 20, 25 kGy at room temperature. After irradiation 3 kinds of sequential extractions were performed. One was submitted to antioxidant extraction using ethyl ether, the second with ethanol and the last with water. The antioxidant activity was determined by β-carotene/linoleic acid co-oxidation. Irradiation in the dose range applied did not have any effect on the antioxidant potential of the cinnamon compounds. Further studies will be performed to study the possibility to use cinnamon extracts in preserving food from oxidative damage induced by ionizing radiation.

  4. The effects of gamma radiation on soybean isoflavones contents

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Marcos R.R. de; Mastro, Nelida L. del [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], e-mail: nlmastro@ipen.br, e-mail: mrramos@ipen.br; Mandarino, Jose M.G. [EMBRAPA Soybean, Londrina, PR (Brazil)], e-mail: jmarcos@cnpso.embrapa.br

    2009-07-01

    Soybean (Glycine max) is the most common source of isoflavones in human feeding. It was suggested that there is a correlation among antioxidant activity of flavonoids and total phenolics content. Plants use isoflavones and their derivatives as part of the plant's defensive arsenal, to ward off disease-causing pathogenic fungi and other microbes. Highly processed foods made from legumes, such as tofu, retain most of their isoflavone content, with the exception of fermented miso, which has increased levels. Little is known about the influence of oxidative stress induced by radiation on the isoflavones contents. In the present paper, the effects of gamma irradiation on soybean isoflavones contents are presented. Samples from several Brazilian soybean cultivars were gamma irradiated with doses of 0, 1, 2, 5 e 10 kGy, dose rate about 3 kGy/h in a {sup 60}Co (Gammacell 220 - AECL). Isoflavones contents were determined after extraction with 70% ethanol containing 0.1% acetic acid by an HPLC method. The total isoflavone content remained almost unchanged with the increase of radiation dose up to 10 kGy. Although a general correlation among total isoflavone content and radiation dose was not found, some data suggest that for a few of the isoflavones from specific cultivars, the increase in the radiation dose induced a decrease in their content as for glucosyl glucosides and malonyl isoflavones, as well as an increase in their aglycone content. (author)

  5. Sensorial analysis evaluation in cereal bars preserved by ionizing radiation processing

    Energy Technology Data Exchange (ETDEWEB)

    Villavicencio, A.L.C.H. [Instituto de Pesquisas Energeticas e Nucleares-IPEN-CNEN/SP, Centro de Tecnologia das Radiacoes, Lab. de Deteccao de Alimentos Irradiados, Travessa R. No. 400, Cidade Universitaria, CEP 05508-910, Sao Paulo (Brazil)], E-mail: villavic@ipen.br; Araujo, M.M.; Fanaro, G.B.; Rela, P.R. [Instituto de Pesquisas Energeticas e Nucleares-IPEN-CNEN/SP, Centro de Tecnologia das Radiacoes, Lab. de Deteccao de Alimentos Irradiados, Travessa R. No. 400, Cidade Universitaria, CEP 05508-910, Sao Paulo (Brazil); Mancini-Filho, J. [Faculdade de Ciencias Farmaceuticas-FCF/USP, Departamento de Alimentos e Nutricao Experimental, Lab. de Lipides, Sao Paulo (Brazil)], E-mail: jmancini@usp.br

    2007-11-15

    Gamma-rays utilized as a food-processing treatment to eliminate insect contamination is well established in food industries. Recent troubles in Brazilian cereal bars commercialization require a special consumer's attention because some products were contaminated by insects. To solve the problem, food-irradiation treatment was utilized as a safe and effective solution. The final product was free of insect contamination. The aim of this study was to determine the best radiation dose processing utilized to disinfestations and detect some change on sensorial characteristic by sensorial analysis in cereal bars. In this study, three different kinds of cereal bars were purchased in Sao Paulo (Brazil) in supermarkets and irradiated with 1.0, 2.0 and 3.0 kGy at 'Instituto de Pesquisas Energeticas e Nucleares' (IPEN-CNEN/SP). The samples were treated with ionizing radiation using a {sup 60}Co gamma-ray facility (Gammacell 220, A.E.C.L.). That radiation doses were used successfully as an anti-insect treatment in the cereal bars, since in some food industries doses up to 3.0 kGy are used to guarantee at least a dose of 1.0 kGy in internal cereal bars package. Sensorial analysis was necessary since cereal bars contain ingredients very sensitive to ionizing radiation process.

  6. An ESR study of the gamma radiolysis of aromatic polyesters containing isomeric naphthalene links

    Science.gov (United States)

    Hill, David J. T.; Choi, Bong-Ku; Ahn, Hung-Kun; Choi, E.-Joon

    2001-07-01

    Six polyesters were synthesised from 4,4'-oxy-bis(benzoyl chloride) and 1,4-, 1,5-, 1,6-, 2,3-, 2,6-, and 2,7-naphthalenediol isomers. The structures of the polyesters were characterised by means of IR, inherent viscosities in tetrachloroethane (TCE), solutions at 303 K and thermal analysis. The glass transition temperatures were in the range of 425-494 K by DSC thermal analysis. All of the polyesters were irradiated in an AECL Gammacell 220 unit at a dose rate of approximately 6.7 kGy/h to doses in the range of 0-15 kGy at 77 and 300 K. ESR spectroscopy was used to examine the radicals formed during radiolysis and to measure their yields. The G-values for radical formation in the polyesters were found to be in the range 0.18-1.41 at 77 K and 0.19-0.78 at 300 K. At 77 K, up to 15% of the radicals formed on radiolysis were found to be photo-bleachable anion radicals. Annealing experiments were carried out in order to identify the neutral radicals, which were assigned to naphthyl- or phenyl- and phenoxyl-type radicals.

  7. Effects of irradiation on natural antioxidants of cinnamon (Cinnamomum zeylanicum N.)

    Energy Technology Data Exchange (ETDEWEB)

    Kitazuru, E.R. E-mail: erkitazu@ipen.br; Moreira, A.V.B.; Mancini-Filho, J. E-mail: jmamcini@usp.br; Delincee, H.; Villavicencio, A.L.C.H. E-mail: villavic@ipen.br

    2004-10-01

    Food irradiation to reduce the number of spoilage microorganisms and insects is an ionizing process that induces free radical formation in proteins, lipids, carbohydrates and other molecular structures in food. Antioxidants generally decrease the level of oxidation in such systems by transferring hydrogen atoms to the free radical structure. In the present paper, the effect of ionizing radiation on natural cinnamon antioxidants is studied. Cinnamon samples were purchased from retailers and irradiated with a {sup 60}Co source, Gammacell 220 (A.E.C.L.) installed at IPEN (Sao Paulo, Brazil) using 0, 5, 10, 15, 20, 25 kGy at room temperature. After irradiation 3 kinds of sequential extractions were performed. One was submitted to antioxidant extraction using ethyl ether, the second with ethanol and the last with water. The antioxidant activity was determined by {beta}-carotene/linoleic acid co-oxidation. Irradiation in the dose range applied did not have any effect on the antioxidant potential of the cinnamon compounds. Further studies will be performed to study the possibility to use cinnamon extracts in preserving food from oxidative damage induced by ionizing radiation.

  8. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  9. Canadian fusion fuels technology project

    International Nuclear Information System (INIS)

    The Canadian Fusion Fuels Technology Project was launched in 1982 to coordinate Canada's provision of fusion fuels technology to international fusion power development programs. The project has a mandate to extend and adapt existing Canadian tritium technologies for use in international fusion power development programs. 1985-86 represents the fourth year of the first five-year term of the Canadian Fusion Fuels Technology Project (CFFTP). This reporting period coincides with an increasing trend in global fusion R and D to direct more effort towards the management of tritium. This has resulted in an increased linking of CFFTP activities and objectives with those of facilities abroad. In this way there has been a continuing achievement resulting from CFFTP efforts to have cooperative R and D and service activities with organizations abroad. All of this is aided by the cooperative international atmosphere within the fusion community. This report summarizes our past year and provides some highlights of the upcoming year 1986/87, which is the final year of the first five-year phase of the program. AECL (representing the Federal Government), the Ministry of Energy (representing Ontario) and Ontario Hydro, have given formal indication of their intent to continue with a second five-year program. Plans for the second phase will continue to emphasize tritium technology and remote handling

  10. The race for megavoltage. X-rays versus telegamma.

    Science.gov (United States)

    Robison, R F

    1995-01-01

    Roentgen's discovery was announced in January, 1896, and x-ray therapy trials followed in 1897. Becquerel rays and radioactive minerals were identified during 1896 through 1898. Radium was used for therapy by 1901, even though a pure standard was not achieved until 1910-1912. Quantities of radium finally became available after 1919, and for 20 years telegamma therapy machines underwent progressive development. Their megavoltage beam was much preferred over the standard 200-250 KV x-ray units of that time. Nuclear physicists during the Great Depression modified electron accelerators into giant 600-900 KV medical x-ray therapy machines and achieved one MV by 1937-1939. These were huge, complex, expensive, and unique to major academic and/or metropolitan centers. During World War II nuclear reactors superseded cyclotrons as efficient factories for few new radioisotopes, including "artificial radium". Few seemed interested in the latter for use in telegamma therapy until 1949-1951, when three competing teams from Canada and the USA designed telecobalt machines. From this competition, among then unknown innovators, emerged three future giants in radiation therapy: A.E.C.L., H. Johns, and G.H. Fletcher. The clinical application of telecobalt therapy was to revolutionize cancer care in community hospitals worldwide.

  11. On the structure of Lattice code WIMSD-5B

    International Nuclear Information System (INIS)

    The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel

  12. Replacement of Cobalt base alloys hardfacing by NOREM alloy; EDF experience and development, some metallurgical considerations. Valves application (CLAMA, RAMA)

    Energy Technology Data Exchange (ETDEWEB)

    Carnus, M. [EDF DPN UTO Direction Expertise Technique, Noisy le Grand (France); Confort, X. [VELAN SAS, Lyon (France)

    2011-07-01

    Cobalt base alloys, such as Stellite 6 and 21, are used extensively in applications where superior resistance to wear and corrosion are required. However the use of Cobalt alloys hardfacing materials, especially on valves, is a major contributor to the level of radioactive contamination of nuclear facilities. NOREM alloys, an iron base and cobalt free materials, have been developed through an Electric Power Research Institute (EPRI) long running program during the eighties as an alternative of Stellite. This alloy has relatively good weldability properties, it was developed initially for repairing Stellite hardfacing (deposit over existing hardfacing alloys). This alloy has good corrosion resistance properties associated with elevated hardness (HRC 36-42). Technological properties (such as galling resistance, wear resistance) have been evaluated through different testing programs led by EPRI, AECL(Atomic Energy of Canada Limited), Valves manufacturers, EDF and others during the nineties. More recently EDF (for replacement of globe valves) has carried out testing program focused on weld deposit chemistry and mechanical properties. NOREM is a candidate for replacement of stellite hardfacing on valves. However this alloy is not so versatile as stellite alloys regarding technological properties (such as wear resistance) at elevated temperature and under high contact pressure. As a consequence some limits have to be considered for application on valves operating at elevated temperature and under high contact pressure (> 20 Mpa). Examples of application on valves, from VELAN manufacturer, for EDF PWR equipment are given. The industrial feedback from installed equipment (CLAMA, RAMA) since 2006 on EDF PWR has been good

  13. Scenario analysis for the postclosure assessment of the Canadian concept for nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    AECL Research has developed and evaluated a concept for disposal of Canada's nuclear fuel waste involving deep underground disposal of the waste in intrusive igneous rock of the Canadian Shield. The postclosure assessment of this concept focusses on the effects on human health and the environment due to potential contaminant releases into the biosphere after the disposal vault is closed. Both radiotoxic and chemically toxic contaminants are considered. One of the steps in the postclosure assessment process is scenario analysis. Scenario analysis identifies factors that could affect the performance of the disposal system and groups these factors into scenarios that require detailed quantitative evaluation. This report documents a systematic procedure for scenario analysis that was developed for the postclosure assessment and then applied to the study of a hypothetical disposal system. The application leads to a comprehensive list of factors and a set of scenarios that require further quantitative study. The application also identifies a number of other factors and potential scenarios that would not contribute significantly to environmental and safety impacts for the hypothetical disposal system. (author). 46 refs., 3 tabs., 3 figs., 2 appendices

  14. Progress report. Physics and Health Sciences. Physics section. 1988 July 01-December 31

    International Nuclear Information System (INIS)

    Progress in overcoming the difficulties encountered in commissioning the superconducting cyclotron, particularly those associated with the π-mode and beam extraction, has been greatly accelerated, largely as a result of the establishment of a TASCC Beam Commissioning Group to give focus and coordination to the work. The '8π spectrometer', which is jointly owned and operated at TASCC by AECL and the Universities of Montreal and McMaster, was relocated during the shutdown and has been producing exciting information on a new type of nucleus-one that has approximately the shape of a football, and is called 'superdeformed'. The physicists working at Chalk River have found the key to making superdeformed nuclei and have already discovered two of their own: 149Gd and 153Dy. A very fine Recoil Doppler apparatus has been constructed and used inside the gamma ball, and a CsI(Tl) miniball is being constructed to do particle measurements with large solid angle, also inside the large ball. In cooperation with scientists from Reactor Development, a new way to identify and measure hydrogen diffusion in Zr(2.5% Nb), the alloy used in CANDU pressure tubes, has been identified and is being developed. It makes use of the superconductivity of zirconium-niobium alloys in the beta phase

  15. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  16. Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

    International Nuclear Information System (INIS)

    Three prototype fuel rods containing Al-64 wt% U3Si2 (3.15 gU/cm3) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U3Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U3Si2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U3Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

  17. Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

    International Nuclear Information System (INIS)

    This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D2O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U3Si fuel irradiated in the NRU reactor. (author)

  18. Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

    International Nuclear Information System (INIS)

    The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)

  19. Radiation sensitivity of different citric pectins

    International Nuclear Information System (INIS)

    Pectic substances are important soluble polysaccharides of plant origin of considerable interest for food industry as gelling agent and stabilizer in jams, fruit jellies, yogurt drinks and lactic acid beverages. Polysaccharides can be degraded by ionizing radiation due to the free radical induced scission of the glycosidic bonds. Viscosity methods had been used to determine the efficiency of hydroxyl radical induced chain breaks generation in macromolecules. In the present work samples of pectin with different degree of methoxylation were employed in order to study their radiation sensitivity by means of viscosity measurements. Samples of citric pectin 1% solutions were irradiated with gamma rays at different doses, ranging from 0 to 15 kGy, using a 60Co Gammacell 220 (AECL), dose rate about 2 kGy/h. After irradiation the viscosity was measured on the viscometer Brookfield model LV-DVIII at 50, 60 and 70 deg C within a period of 48h. Pectin viscosity with high degree of methoxylation decreased sharply with the radiation dose remaining almost constant from 10 kGy. Pectin with low degree of methoxylation presented initially higher values of viscosity and the radiation induced decrease was also pronounced. Viscosity measurements decreased with the increase of the temperature applied for both kind of samples. The effect of radiation induced chain breaks generation in pectin molecules was evident through the viscosity reduction of irradiated pectin solutions although the viscosity presented diverse values depending of the degree of methoxylation of carboxyl groups in the backbone of polysaccharide macromolecules. (author)

  20. Uncertainty analysis guide

    International Nuclear Information System (INIS)

    This guide applies to the estimation of uncertainty in quantities calculated by scientific, analysis and design computer programs that fall within the scope of AECL's software quality assurance (SQA) manual. The guide weaves together rational approaches from the SQA manual and three other diverse sources: (a) the CSAU (Code Scaling, Applicability, and Uncertainty) evaluation methodology; (b) the ISO Guide,for the Expression of Uncertainty in Measurement; and (c) the SVA (Systems Variability Analysis) method of risk analysis. This report describes the manner by which random and systematic uncertainties in calculated quantities can be estimated and expressed. Random uncertainty in model output can be attributed to uncertainties of inputs. The propagation of these uncertainties through a computer model can be represented in a variety of ways, including exact calculations, series approximations and Monte Carlo methods. Systematic uncertainties emerge from the development of the computer model itself, through simplifications and conservatisms, for example. These must be estimated and combined with random uncertainties to determine the combined uncertainty in a model output. This report also addresses the method by which uncertainties should be employed in code validation, in order to determine whether experiments and simulations agree, and whether or not a code satisfies the required tolerance for its application. (author)

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

    International Nuclear Information System (INIS)

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

  2. The Canadian approach to microbial studies in nuclear waste management and disposal

    International Nuclear Information System (INIS)

    Many countries considering radioactive waste disposal have, or are considering programs to study and quantify microbial effects in terms of their particular disposal concept. Although there is an abundance of qualitative information, there is a need for quantitative data. Quantitative research should cover topics such as the kinetics of microbial activity in geological media, microbial effects on radionuclide migration in host rock (including effects of biofilms), tolerance to extreme conditions of radiation, heat and desiccation, microbially-influenced corrosion of waste containers and microbial gas production. The research should be performed in relevant disposal environments with the ultimate objective to quantify those effects that need to be included in models for predictive and safety assessment purposes. The Canadian approach to dealing with microbial effects involves a combination of pertinent, quantitative measurements from carefully designed laboratory studies and from large scale engineering experiments in AECL's Underground Research Laboratory (URL). The validity of these quantitative data is measured against observations from natural environments and analogues. An example is the viability of microbes in clay-based scaling materials. Laboratory studies have shown that the clay content of these barriers strongly affects microbial activity and movement. This is supported by natural environment and analogue observations that show clay deposits to contain very old tree segments and dense clay lenses in sediments to contain much smaller, less diverse and less active microbial populations than more porous sediments. This approach has allowed for focused, quantitative research on microbial effects in Canada. (author)

  3. Mobile robots for hazardous environments

    Energy Technology Data Exchange (ETDEWEB)

    Bains, N.; Scott, D.A.; Tran, K.; Campbell, T. (Atomic Energy of Canada Ltd., Mississauga, Ontario (Canada))

    1992-01-01

    This paper describes the development of a mobile robot ARK-2 (Autonomous Robot for Known Environments) that utilizes a number of sensors for navigation in a known relatively structured indoor environment. At present, there are robots that can be preprogrammed and that move along a specified path, but they use dead-reckoning to evaluate position at any point along their paths, and this can lead to major error accumulation through wheel slippage and running over unforeseen objects on the floor. The ARK-2 robot will have the intelligence to determine its position utilizing natural landmarks at any point along its path; it is this feature that gives ARK-2 its uniqueness as well as its ability to operate in an industrial environment. The project started in September 1991 and will last 4 yr. There are five organizations involved in the project: Ontario Hydro, Atomic Energy of Canada Limited (AECL) CANDU, US Nuclear Regulatory Commission (NRC), University of Toronto, and York University. Funding is provided by the organizations involved as well as the federal and provincial governments and PRECARN Associates, which is a nonprofit precompetitive research consortium made up of 38 members.

  4. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-01

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

  5. Technical Advisory Committee on the nuclear fuel waste management program : thirteenth annual report

    International Nuclear Information System (INIS)

    Since the last reporting period by the Technical Advisory Committee (TAC) the emphasis of the work in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) has been on the writing of the Environmental Impact Statement (EIS) and the associated set of nine primary reference documents as well as supporting documents. These are in preparation for submission to the Environmental Assessment Review Panel who will lead the national evaluation of the disposal concept under the auspices of the Federal Environmental Assessment Review Office (FEARO). The disposal concept developed over the last fourteen years by Atomic Energy of Canada Limited (AECL) and anticipated to be presented by means of the EIS in 1994, is based on a multiple system of natural and man-made barriers wherein nuclear waste is first enclosed in corrosion-resistant containers, designed to last at least 500 years, and then placed in a vault excavated 500 - 1000 m deep in granitic rocks of the Canadian Shield. After container emplacement either in or on the floor of the vault, and with a surrounding buffer material of a bentonite clay/sand mixture, the vault will be backfilled and sealed with crushed rock, buffer and sand, as will be the shafts and exploratory boreholes. The case study being presented by AECL to demonstrate the safety of this concept and the technology to implement it, relies on computer simulations of a hypothetical disposal site with geological characteristics similar to those at the Underground Research Laboratory (URL) in the Whiteshell Research Area (WRA) located in Manitoba. The preliminary simulation results suggest that safe containment can be achieved provided that the waste is surrounded by a sparsely-fractured zone of rock wherein movement of contaminants carried by groundwater is modelled as a diffusive as opposed to a advective process. The principal focus of work during the past year within the environmental and safety assessment has been to complete the Post

  6. Appendix II: Training of personnel for the commissioning of Cernavoda NPP Unit 2 (Romania)

    International Nuclear Information System (INIS)

    Nuclear power generation is a reliable source of energy for Romania and an important contributor to the national electricity supply. Romania has one nuclear power plant, Cernavoda, which operates one CANDU 6 reactor, 707 MWe, designed by Atomic Energy of Canada Ltd (AECL). It provides about 10% of the total country electricity generation. The state-owned company responsible for the production and supply of energy from Cernavoda NPP, as well as for its development, is Societatea Nationala 'Nuclearelectrica'. The plant was designed to have five similar units. The studies performed for the plant, prior to the construction's start, established feasible technical solutions for all problems related to a 5- units plant, including the environmental impact, which was determined to be entirely acceptable. Construction of the first unit started in 1980, and units 2-5 in 1982. The pressing problems encountered during construction of Unit 1 (import restrictions, delays etc.) restrained the progress at Unit 2. Starting in 1990, the work on Cernavoda site was focused on Unit 1. Construction at the other units was suspended; during many years; only preservation works were carried-out. The re-start of work at Unit 2 began in 1995, under the management of the AECLANSALDO Consortium. Some work to further construction progress was carried-out (e.g. installation of the fuel channels) and a thorough assessment was performed of the condition of the equipment procured for Unit 2 and stored on site, or already installed. The absence of a clear contractual framework and the lack of resources hindered significant advances in the construction of Unit 2. Recently, estimates in Romania are that annual electricity production will become insufficient unless Unit 2 of Cernavoda NPP is commissioned. In this environment and considering the relatively low cost of the electricity produced at Unit 1 of Cernavoda NPP versus the energy cost in the conventional thermal stations, the Romanian Government

  7. Appendix III: Cernavoda Nuclear Power Plant: Phased implementation of similar units - Human resource management and training programmes (Romania) (Case study of human resource issues faced by NPP operating organizations, and how they were (or are being) addressed)

    International Nuclear Information System (INIS)

    Nuclear power generation is a reliable source of energy for Romania and an important contributor to the national electricity supply. Romania has one nuclear power plant, Cernavoda, which operates one CANDU 6 reactor, 707 MW(e), designed by Atomic Energy of Canada Ltd (AECL). Provides about 10 percent of the total country energy consumption. The state-owned company responsible for the production and supply of energy from Cernavoda NPP, as well as for its development, is Societatea Nationala 'Nuclearelectrica'. The plant was designed to have five similar units. The studies performed for the plant, prior to the construction's start, established feasible technical solutions for all problems related to a 5-unit site, including the environmental impact, which would be entirely acceptable. Construction of the first unit started in 1980 and of units 2-5 in 1982. The pressing problems encountered during construction of Unit 1 (import restrictions, delays etc.) restrained the progress at Unit 2. Starting with 1990, the work on Cernavoda site was focused on Unit 1 and the construction at the other units was suspended. Since then, only preservation activities have been carried-out. The re-start of work at Unit 2 took place in 1995, under the management of the AECLANSALDO Consortium. Some construction work was carried-out (e.g. installation of the fuel channels) and a thorough assessment was performed concerning the condition of the equipment procured for Unit 2 and stored on site, or already installed. The absence of a clear contractual framework and the lack of resources hindered significant advance in the construction of Unit 2. Since then, many economic estimates foresee that, starting with winter 2005 when Romanian electricity consumption is expected to increase with about 1000 MW(e), the annual electricity production will become insufficient and will be not covered unless Unit 2 of Cernavoda NPP is commissioned. In this environment and considering the relatively low cost

  8. Liquid radwaste processing with spiral wound reverse osmosis

    International Nuclear Information System (INIS)

    Two different reverse osmosis systems were investigated in this work. The first was a 50-element plant-scale system that is used to treat 2200 cubic metres of AECL (low to intermediate level) liquid radwastes annually. It uses thin film composite (TFC) membranes and operates at an applied pressure of 2760 kPa, with a fixed crossflow of about 40 L/min. The other system uses the same thin film composite membranes for waste processing, but is a 2-element pilot-scale system. It is operated at pressures ranging between 1500 kPa and 7000 kPa, at a fixed crossflow of 55 L/min. The average lifetime of the thin film composite membranes in the plant-scale processing application at AECL is about 3000 hours. After the service life has expired the rejection efficiency (for bulk conductivity) declines rapidly from 99.5% to about 95% as the membranes become impaired from chemical cleaning procedures that are required after each hundred cubic metres of waste are treated. The permeation flux for the plant-scale system decreases from about 2.2 L/min/element to below 0.5 L/min/element at the end of the membrane's useful service life. The plant-scale membrane elements, fouled by an assortment of chemicals including calcium phosphate, ferric oxide, and various organics, were successfully regenerated by exposing them to a three-step chemical cleaning procedure (in the pilot-scale system), using detergent, HCl, and an alkaline-based cleaning with EDTA respectively. The 3-step procedure was successful in elevating the flux from 0.5 L/min for the spent membrane, to 1.2 L/min after the three step cleaning procedure. The 1.2 L/min post-cleaning flux could be maintained at a crossflow velocity of 55 L/min/vessel. The decontamination factor (DF), which is the feed to permeate concentration ratio, was determined for cesium and strontium. For the plant-scale system (at the operating pressure of 2760 kPa), the DF of cesium decreased from about 100 when the membranes were new, to about 30 after

  9. Status of Korean nuclear industry and Romania-Korea cooperation in nuclear field

    International Nuclear Information System (INIS)

    , constructed by AECL turnkey contract, started commercial operations in 1983. Units 2,3 and 4 were constructed by a non turn-key contract scheme, which was constructed by KHNP with assistance from AECL for some areas. The second part of the paper deals with the Romania-Korea cooperation status. The cooperation between Romania-Korea in the nuclear power field got into stride in March 2001. A technical agreement was signed between Romanian Company SNN and KHNP in March 2002 for cooperation in the Cernavoda projects. An amount of 32 tones of Romanian heavy water was supplied to KHNP, The Technical Assistance Agreement between SNN and KHNP stipulates provisions for technical services for operation of Unit 1, construction and commissioning of Cernavoda Unit 2. This Technical Assistance Agreement will be the basis to enhance economy and safety of Cernavoda Units 1,2 and 3. In the frame of the cooperation in Cernavoda Unit 3 Project Romania can enjoy benefits from Korea's world-top class technologies and experience. Korea can support Romania utilizing the systematically established nuclear infrastructure. Korea, both government and nuclear power industry represented by KHNP, will fully support Romania so that new feasibility study may proceed in accordance with the required schedule. The paper has the following structure: Part 1- Korean nuclear industry status: 1. Current status of electric power in Korea; 2. Long term energy plan; 3. Status of nuclear power projects; 4. Operational performance; 5. Outlines of Wolsong CANDU units; Part 2 - Romania-Korea cooperation status: 1. History for cooperation; 2. Technical assistance for Cernavoda Units 1 and 2; 3. Joint development of Cernavoda Unit 3 Project; 4. Cooperation in Cernavoda Unit 3 Project

  10. The CANTEACH project: preserving CANDU technical knowledge

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. Parametric studies of radiolytic oxidation of iodide solutions with and without paint: comparison with code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Poletiko, C.; Hueber, C. [Inst. de Protection et de Surete Nucleaire, C.E. Cadarache, St. Paul-lez-Durance (France); Fabre, B. [CISI, C.E. Cadarache, St. Paul-lez-Durance (France)

    1996-12-01

    In case of severe nuclear accident, radioactive material may be released into the environment. Among the fission products involved, are the very volatile iodine isotopes. However, the chemical forms are not well known due to the presence of different species in the containment with which iodine may rapidly react to form aerosols, molecular iodine, hydroiodic acid and iodo-organics. Tentative explanations of different mechanisms were performed through benchscale tests. A series of tests has been performed at AEA Harwell (GB) to study parameters such as pH, dose rate, concentration, gas flow rate, temperature in relation to molecular iodine production, under dynamic conditions. Another set of tests has been performed in AECL Whiteshell (CA) to study the behaviour of painted coupons, standing in gas phase or liquid phase or both, with iodine compounds under radiation. The purpose of our paper is to synthesize the data and compare the results to the IODE code calculation. Some parameters of the code were studied to fit the experimental result the best. A law, concerning the reverse reaction of iodide radiolytic oxidation, has been proposed versus: pH, concentrations and gas flow-rate. This law does not apply for dose rate variations. For the study of painted coupons, it has been pointed out that molecular iodine tends to be adsorbed or chemically absorbed on the surface in gas phase, but the mechanism should be more sophisticated in the aqueous phase. The iodo-organics present in liquid phase tend to be partly or totally destroyed by oxidation under radiation (depending upon the dose delivered). These points are discussed. (author) 18 figs., 3 tabs., 15 refs.

  13. The reduction of I{sub 2} by H{sub 2}O{sub 2} in aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Ball, J.M.; Hnatiw, J.B. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.; Sims, H.E. [AEA Technology, Harwell Laboratory, Didcot (United Kingdom)

    1996-12-01

    The reduction of iodine by hydrogen peroxide is an important process which leads to a lower amount of molecular iodine in irradiated solutions of iodide as the pH is increased. There is quite a large amount of information on the reaction now but no consensus in the literature on the mechanisms for reaction and the generally accepted mechanism does not appear to be correct. A number of studies of the kinetics of the reaction in the pH range 2-7 have been carried out where the iodine reduction process exhibited a 1/[H{sup +}]{sup 2} dependence consistent with the proposed mechanism which were attributed primarily to the reaction of H{sub 2}O{sub 2} with IO{sup -}. Deviations were observed in the pH range 6-7 and were explained by incorporating the reaction of I{sub 2}OH{sup -} with H{sub 2}O{sub 2}. In some other experiments it was suggested that the failure to maintain a 1/[H{sup +}]{sup 2} dependence at high pH was due to the iodine hydrolysis being rate determining. Data from an experimental program performed at AECL described in this paper confirms that the 1/[H{sup +}]{sup 2} dependence does not hold at high pH. These studies were carried out as a function of acid, iodide, peroxide and buffer concentration for three buffers, barbital, citrate and phosphate. This paper discuss two mechanisms which involve the formation of an HOOI intermediate in the rate determining step and which adequately describe the experimental data. (author) 4 figs., 2 tabs., 23 refs.

  14. Low LET radiolysis escape yields for reducing radicals and H2 in pressurized high temperature water

    Science.gov (United States)

    Sterniczuk, Marcin; Yakabuskie, Pamela A.; Wren, J. Clara; Jacob, Jasmine A.; Bartels, David M.

    2016-04-01

    Low Linear Energy Transfer (LET) radiolysis escape yields (G values) are reported for the sum (G(radH)+G(e-)aq) and for G(H2) in subcritical water up to 350 °C. The scavenger system 1-10 mM acetate/0.001 M hydroxide/0.00048 M N2O was used with simultaneous mass spectroscopic detection of H2 and N2 product. Temperature-dependent measurements were carried out with 2.5 MeV electrons from a van de Graaff accelerator, while room temperature calibration measurements were done with a 60Co gamma source. The concentrations and dose range were carefully chosen so that initial spur chemistry is not perturbed and the N2 product yield corresponds to those reducing radicals that escape recombination in pure water. In comparison with a recent review recommendation of Elliot and Bartels (AECL report 153-127160-450-001, 2009), the measured reducing radical yield is seven percent smaller at room temperature but in fairly good agreement above 150 °C. The H2 escape yield is in good agreement throughout the temperature range with several previous studies that used much larger radical scavenging rates. Previous analysis of earlier high temperature measurements of Gesc(radOH) is shown to be flawed, although the actual G values may be nearly correct. The methodology used in the present report greatly reduces the range of possible error and puts the high temperature escape yields for low-LET radiation on a much firmer quantitative foundation than was previously available.

  15. Regional groundwater flow in the Atikokan Research Area : simulation of 18O and 3H distributions

    International Nuclear Information System (INIS)

    AECL is investigating a concept for disposing of nuclear fuel waste deep in plutonic rock of the Canadian Shield. As part of this investigation, we have performed a model simulation of regional groundwater flow in the Atikokan Research Area, a fractured plutonic rock environment of the Canadian Shield, and used the distribution of oxygen-18 (18O) and tritium (3H) in groundwater to test the model. At the first stage of model calibration, groundwater flow was simulated using a three-dimensional finite-element code, MOTIF, in conjunction with a conceptual framework model derived from field geological, geophysical and hydrogeological data. Hydraulic parameters (permeability and porosity) were systematically varied until simulated recharge rates to the water table compared favourably with estimated recharge rates based on stream flow analysis. At the second stage, vertical average linear groundwater velocities from the first stage of the calibration process were combined with conceptualized one-dimensional models of the system to generate depth concentration profiles of 18O and 3H. Recharge-, midline-and discharge area models of both the fracture zones and the rock mass were employed. The simulated profiles formed 'envelopes' around all field 18O and 3H data, indicating that the calibrated velocities used in the model are reasonable. The models demonstrate that the scatter of δ18O and 3H field data from the Atikokan Research Area is consistent with the groundwater flow model predictions and can be explained by the complexity arising from different hydraulic regimes (recharge, midline, discharge) and hydrogeologic environments (fracture zones, rock mass) of the regional flow system. 50 refs., 14 figs., 3 tabs

  16. CANDU steam generator life management

    International Nuclear Information System (INIS)

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

  17. Irradiation of mangoes as a quarantine treatment

    Energy Technology Data Exchange (ETDEWEB)

    Bustos R, M.E.; Enkerlin H, W.; Toledo A, J.; Reyes F, J.; Casimiro G, A

    1991-06-15

    This research project was conducted following guidelines of research protocols for post-harvest treatments developed by the United States Department of Agriculture CUSA. Laboratory bioassays included the irradiation of mangoes infested with third instar larvae of Anastrepha serpentina (Wied), A. ludens (Loew), A. obliqua (Macquart) and Ceratitis capitata (Wied) , at doses from 10 to 250 Gy. Irradiation doses were applied using a Co-60 AECL Model JS-7400 irradiator. The design was chosen to obtain a maximum to minimum ratio equal to, or less than, 1.025. C. capitata was the species most tolerant to irradiation. A dose of 60 Gy applied to third instar fruit fly larvae sterilized this species and prevented emergence of adults of the other three species. A dose of 250 Gy was required to prevent emergence of C. capitata. In fertility tests using emerged adults of A . Iudens, and A. obliqua a dose of 30 Gy gave 45 % and 27 % fertility, respectively. Adults of A. serpentina that emerged, died before reaching sexual maturity. The confirmatory tests, at probit-9 security level, were done at 100 Gy for the three species of Anastrepha and at 150 Gy for C. capitata. The quality of mangoes irradiated up to 1000 Gy was evaluated by chemical, physiological, and sensorial tests. The determination of vitamin C indicated that there was no loss of the nutritive value of the fruit. It also was observed that fruit metabolism was not accelerated since no significant increase in respiration or transpiration was registered and consumers accepted both treated and untreated fruit in the same way. (Author)

  18. Communicating Risk to a Concerned Public in Historic Low-Level Radioactive Waste (LLRW) Projects

    International Nuclear Information System (INIS)

    The Low-Level Radioactive Waste Management Office (LLRWMO) was established in 1982 to carry out federal government responsibility for historic low-level radioactive waste across Canada. Funded through Natural Resources Canada (NRCan) and administered by Atomic Energy of Canada Limited (AECL), the LLRWMO has conducted waste characterization, delineation and remediation projects in British Columbia, the Northwest Territories, Alberta and Ontario. Most (95%) of the historic low-level radioactive wastes for which the LLRWMO assumes responsibility are located in and around Port Hope, Ontario, the site of the refining operations of the former federal Crown Corporation, Eldorado Nuclear. Additional contamination is connected to the transportation of the ore along a route extending from Port Radium on Great Bear Lake (Northwest Territories) to Port Hope, a Lake Ontario community of 16,000 in south/central Ontario. This paper will focus on the successful strategies employed by the LLRWMO over the past 25 years to find solutions to the problems posed by the historic waste. Risks associated with each project must be defined prior to initiating any communication plan. Using this approach, the LLRWMO has addressed socio-economic, health and well being and environmental risks through strategies that embrace proactive communications and full public involvement. Recognizing the diversity of the Canadian communities affected by the historic waste, the LLRWMO has tailored its consultation programs both to the type of solutions proposed and to the character of the communities (i.e. urban, First Nation, rural etc.). In Port Hope, community solutions are being realized through a Legal Agreement negotiated by the affected communities and the federal government in 2000 and an intensive Environmental Assessment process under the Canadian Environmental Assessment Act, which is expected to reach completion early in 2007. (authors)

  19. Steam generator life management

    International Nuclear Information System (INIS)

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, 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. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

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

    International Nuclear Information System (INIS)

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

  1. The effect of gamma irradiation on the microbiological analysis on commercial functional Brazilian green banana flour

    Energy Technology Data Exchange (ETDEWEB)

    Taipina, Magda S.; Lamardo, Leda C.A.; Santos, Josefina S.; Silva Junior, Eneo A. da [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Balian, Simone C., E-mail: balian@usp.b [Universidade de Sao Paulo (USP), SP (Brazil). Fac. de Medicina Veterinaria e Zootecnia

    2011-07-01

    In Brazil, although it is qualified as a major world producers, however, the production losses are high. Nevertheless, these losses can be reduced by processing the fruit 'unsuitable' for consumption into products based on green banana (pulp, rind and flour). The green banana flour shows enhanced nutrition value, with higher contents of mineral, dietary fiber, resistant starch, and total phenolics, for use in Brazilian irradiated ready - to eat foods, such as bread, macaroni, among others. Food irradiation has been identified as safe technology to reduce risk of foodborne illness as part of high-quality food production, processing, handling and preparation. Food irradiation utilizes a source of ionizing energy that passes through food to destroy harmful bacteria and other organisms. Often referred to as 'cold pasteurization', food irradiation offers negligible loss of nutrients or sensory qualities in food as it does not substantially raise the temperature of the food during processing. The object of this work was to determine the effect of gamma irradiation on microbiological analyses of the: the number of mesophiles, total coliforms at 35 deg C, coliforms at 45 deg C, Staphylococcus aureus and Salmonella spp of the green banana flour, commercially found in the Brazilian market. The microbiological analyses were carried out in conformity with the methodologies described at the Faculty of Veterinary Medicine, according to the current legislation. Irradiation was performed in a {sup 60}Co Gammacell 220 (AECL) source, with dose of 3kGy at IPEN/CNEN-SP. In samples of Brazilian green banana flour, irradiated at 3 kGy, the growth of all microorganisms (mesophiles, total coliforms at 35 deg C, coliform at 45 deg C and Staphylococcus coagulase positive) were reduced. As a result, the application of the irradiation technique may be recommended to enhance the food safety. (author)

  2. Use of the 'DRAGON' program for the calculation of reactivity devices

    International Nuclear Information System (INIS)

    DRAGON is a computer program developed at the Ecole Polytechnique of the University of Montreal and adopted by AECL for the transport calculations associated to reactivity devices. This report presents aspects of the implementation in NASA of the DRAGON program. Some cases of interest were evaluated. Comparisons with results of known programs as WIMS D5, and with experiments were done. a) Embalse (CANDU 6) cell without burnup and leakage. Calculations of macroscopic cross sections with WIMS and DRAGON show very good agreement with smaller differences in the thermal constants. b) Embalse fresh cell with different leakage options. c) Embalse cell with leakage and burnup. A comparison of k-infinity and k-effective with WIMS and DRAGON as a function of burnup shows that the differences ((D-W)/D) for fresh fuel are -0.17% roughly constant up to about 2500 MWd/tU, and then decrease to -0.06 % for 8500 MWd/tU. Experiments made in 1977 in ZED-2 critical facility, reported in [3], were used as a benchmark for the cell and supercell DRAGON calculations. Calculated fluxes were compared with experimental values and the agreement is so good. d) ZED-2 cell calculation. The measured buckling was used as geometric buckling. This case can be considered an experimental verification. The calculated reactivity with DRAGON is about 2 mk, and can be considered satisfactory. WIMS k-effective value is about one mk higher. e) Supercell calculations for ZED-2 vertical and horizontal tube and rod adjuster using 2D and 3D models were done. Comparisons between measured and calculated fluxes in the vicinity of the adjuster rods. Incremental cross sections for these adjusters were calculated using different options. f) ZED-2 reactor calculations with PUMA reveal a good concordance with critical heights measured in experiments. The report describes also particular features of the code and recommendations regarding its use that may be useful for new users. (author)

  3. FEAST 3.1: finite-element modeling of sheath deformation such as longitudinal ridging and collapse into axial gap

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X.; Xu, Z.; Kim, Y-S.; Lai, L.; Cheng, G.; Xu, S. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2010-07-01

    During normal operation, the collapsible CANDU® fuel sheath deforms, especially, it may deform into longitudinal ridges or collapse instantaneously into the axial gaps between the end pellet and endcap or between two neighbouring pellets. These phenomena occur under certain conditions, such as the coolant pressure exceeding critical pressures for longitudinal ridging or axial collapse. Both longitudinal ridging and axial collapse phenomena result from plastic instability in the sheath under coolant pressure. Longitudinal ridging features one or multiple lobes or 'ridges' (outward from the sheath surface) formed along the sheath in the longitudinal direction. Axial collapse features a 'valley' around the sheath circumference. Both phenomena can lead to sheath overstrain, which in turn potentially leads to sheath failure. The LONGER code, which contains empirical correlations, has been used to predict the critical pressures for these two sheath deformation phenomena. To study fuel behaviour outside of the application ranges of the LONGER empirical correlations, a mechanistic model is needed. FEAST (Finite Element Analysis for Stresses) is an AECL computer code used to assess the structural integrity of the CANDU fuel element. The FEAST code has recently been developed (to Version 3.1) to model processes occurring during longitudinal ridge formation and instantaneous collapse into the axial gap. The new models include those for geometric non-linearity (large deformation, large material rotation), non-linear stress-strain curve for plastic deformation, Zr-4 sheath creep law, and variable Young’s Modulus etc. This paper describes the mechanistic model (FEAST 3.1) development for analyses of longitudinal ridging and instantaneous collapse into axial gap, and the comparison with the results from empirical correlations in LONGER. (author)

  4. Doses from aquatic pathways in CSA-N288.1: deterministic and stochastic predictions compared

    Energy Technology Data Exchange (ETDEWEB)

    Chouhan, S.L.; Davis, P

    2002-04-01

    The conservatism and uncertainty in the Canadian Standards Association (CSA) model for calculating derived release limits (DRLs) for aquatic emissions of radionuclides from nuclear facilities was investigated. The model was run deterministically using the recommended default values for its parameters, and its predictions were compared with the distributed doses obtained by running the model stochastically. Probability density functions (PDFs) for the model parameters for the stochastic runs were constructed using data reported in the literature and results from experimental work done by AECL. The default values recommended for the CSA model for some parameters were found to be lower than the central values of the PDFs in about half of the cases. Doses (ingestion, groundshine and immersion) calculated as the median of 400 stochastic runs were higher than the deterministic doses predicted using the CSA default values of the parameters for more than half (85 out of the 163) of the cases. Thus, the CSA model is not conservative for calculating DRLs for aquatic radionuclide emissions, as it was intended to be. The output of the stochastic runs was used to determine the uncertainty in the CSA model predictions. The uncertainty in the total dose was high, with the 95% confidence interval exceeding an order of magnitude for all radionuclides. A sensitivity study revealed that total ingestion doses to adults predicted by the CSA model are sensitive primarily to water intake rates, bioaccumulation factors for fish and marine biota, dietary intakes of fish and marine biota, the fraction of consumed food arising from contaminated sources, the irrigation rate, occupancy factors and the sediment solid/liquid distribution coefficient. To improve DRL models, further research into aquatic exposure pathways should concentrate on reducing the uncertainty in these parameters. The PDFs given here can he used by other modellers to test and improve their models and to ensure that DRLs

  5. CANDU refurbishment - managing the life cycle

    International Nuclear Information System (INIS)

    All utilities that operate a nuclear power plant have an integrated plan for managing the condition of the plant systems, structures and components. With a sound plant life management program, after about 25 years of operation, replacement of certain reactor core components can give an additional 25 to 30 years of operation. This demonstrates the long-term economic strength of CANDU technology and justifies a long-term commitment to nuclear power. Indeed, replacement of pressure tubes and feeders with the most recent technology will also lead to increased capacity factors - due to reduced requirements for feeder inspections and repair, and eliminating the need for fuel channel spacer relocation which have caused additional and longer maintenance outages. Continuing the operation of CANDU units parallels the successful life extensions of reactors in other countries and provides the benefits of ongoing reliable operation, at an existing plant location, with the continued support of the host community. The key factors for successful, optimum management of the life cycle are: ongoing, effective plant life management programs; careful development of refurbishment scope, taking into account system condition assessments and a systematic safety review; and, a well-planned and well-executed retubing and refurbishment outage, where safety and risk management is paramount to ensure a successful project The paper will describe: the benefits of extended plant life; the outlook for refurbishment; the life management and refurbishment program; preparations for retubing of the reactor core; and, enhanced performance post-retubing. Given the potential magnitude of the program over the next 10 years, AECL will maintain a lead role providing overall support for retubing and plant Life Cycle Management programs and the CANDU Owners Group will provide a framework for collaboration among its Members. (author)

  6. The role of function analysis in the ACR control centre design

    International Nuclear Information System (INIS)

    An essential aspect of control centre design is the need to characterize: plant functions and their inter-relationships to support the achievement of operational goals, and roles for humans and automation in sharing and exchanging the execution of functions across all operational phases. Function analysis is a design activity that has been internationally accepted as an approach to satisfy this need. It is recognized as a fundamental and necessary component in the systematic approach to control centre design and is carried out early in the design process. A function analysis can provide a clear basis for: the control centre design for the purposes of design team communication, and customer or regulatory review, the control centre display and control systems, the staffing and layout requirements of the control centre, assessing the completeness of control centre displays and controls prior and supplementary to mock-up walkthroughs or simulator evaluations, and the design of operating procedures and training programs. This paper will explore the role for function analysis in supporting the design of the control centre. The development of the ACR control room will be used as an illustrative context for the discussion. The paper will also discuss the merits of using function analysis in a goal-or function-based approach resulting in a more robust, operationally compatible, and cost-effective design over the life of the plant. Two former papers have previously outlined, the evolution in AECL's application approach and lessons learned in applying function analysis in support of control room design. This paper provides the most recent update to this progression in application refinement. (author)

  7. Gamma radiation effects on the viscosity of green banana flour

    Energy Technology Data Exchange (ETDEWEB)

    Uehara, Vanessa B.; Inamura, Patricia Y.; Mastro, Nelida L. Del [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], e-mail: vanessa.uehara@usp.br, e-mail: patyoko@yahoo.com, e-mail: nlmastro@ipen.br

    2009-07-01

    Banana (Musa sp) is a tropical fruits with great acceptability among consumers and produced in Brazil in a large scale. Bananas are not being as exploited as they could be in prepared food, and research could stimulate greater interest from industry. The viscosity characteristics and a product consistency can determine its acceptance by the consumer. Particularly the starch obtained from green banana had been studied from the nutritional point of view since the concept of Resistant Starch was introduced. Powder RS with high content of amylose was included in an approved food list with alleged functional properties in Brazilian legislation. Ionizing radiation can be used as a public health intervention measure for the control of food-borne diseases. Radiation is also a very convenient tool for polymer materials modification through degradation, grafting and crosslinking. In this work the influence of ionizing radiation on the rheological behavior of green banana pulp was investigated. Samples of green banana pulp flour were irradiated in a {sup 60}Co Gammacell 220 (AECL) with doses of 0 kGy,1 kGy, 3 kGy, 5 kGy and 10 kGy in glass recipients. After irradiation 3% and 5% aqueous dilution were prepared and viscosity measurements performed in a Brooksfield, model DVIII viscometer using spindle SC4-18 and SC4-31. There was a reduction of the initial viscosity of the samples as a consequence of radiation processing, being the reduction inversely proportional to the flour concentration. The polysaccharide content of the banana starch seems to be degraded by radiation in solid state as shown by the reduction of viscosity as a function of radiation dose. (author)

  8. Process optimization of DUPIC fuel pellet fabrication

    International Nuclear Information System (INIS)

    DUPIC pellets are remotely fabricated by using DUPIC powder prepared by the OREOX treatment of spent fuel pellets. DUPIC pellets were successfully fabricated using spent PWR fuel material with an average discharge burn-up of 27,300 MWd/tU. Sintered density, grain size and surface roughness of the DUPIC pellets were investigated on the basis of CANDU fuel criteria. In order to optimize the DUPIC pellet manufacturing processes, 3 series of experiments for the pre-qualification and 3 series for the qualification were performed. In these experiments, the sintered densities of the pellets ranged from 10.35 g/cm3(95.7 % of T.D.) to 10.43 g/cm3(96.4 % of T.D.) and the average grain size ranged from 14.6 to 14.9 μm. Based on these results, the optimum manufacturing processes of DUPIC pellets have been established. Then, under the control of the QA program developed with the assistance of AECL, 8 series of production runs have been performed to make DUPIC pellets in a batch size of 1 kg. The sintered densities of the fabricated pellets ranged from 10.26 g/cm3 to 10.43 g/cm3. The surface roughness of the ground pellets was less than Ra 0.8 μm by the dry grinding process. As the results of the production runs, DUPIC fuel pellets meeting the standard CANDU fuel specifications were successfully fabricated by the established processes. (author)

  9. The Canadian Safeguards Support Program - A future outlook

    International Nuclear Information System (INIS)

    Full text: The Canadian Safeguards Support Program (CSSP) is one of the first safeguards support programs with an overall objective to assist the IAEA by providing technical assistance and other resources and by developing equipment to improve the effectiveness of international safeguards. This paper provides a brief discussion of the evolution of the CSSP, from the beginning when the program was under joint management between the Atomic Energy Control Board (AECB) and Atomic Energy of Canada Limited (AECL), a Canadian crown corporation, until recent years when the AECB became responsible for all projects and financial management. Recently, new legislation came into force and the AECB became the Canadian Nuclear Safety Commission (CNSC). However, the mandate and management of the CSSP under the CNSC remain fundamentally unchanged. Major CSSP activities are devoted to the following areas: (a) Human resource assistance through the provision of cost-free experts (CFEs) to the IAEA; (b) Training of IAEA inspectors and facility operators, development of training resources and integrated approaches for training; (c) System studies, e.g. the development of integrated safeguards approach for CANDU reactors, geological repository, and physical model; (d) Equipment development, e.g. the VXI Integrated Fuel Monitor, Digital Cerenkov Viewing Device, seals, remote monitoring, encryption and authentication; (e) Information technology which includes satellite imagery, Geographical Information System (GIS), and position tracking of spent fuel containers. The CSSP has continued to evolve during the past 25 years. Although formerly larger the CSSP budget has settled to a stable level of just slightly above (Canadian) $2M. Leveraging of the CSSP budget through collaborations with several Member State Support Programs and Canadian government departments has provided mutual benefits for all parties involved and useful results that have been put into practical use by the IAEA. (author)

  10. The reduction of I2 by H2O2 in aqueous solution

    International Nuclear Information System (INIS)

    The reduction of iodine by hydrogen peroxide is an important process which leads to a lower amount of molecular iodine in irradiated solutions of iodide as the pH is increased. There is quite a large amount of information on the reaction now but no consensus in the literature on the mechanisms for reaction and the generally accepted mechanism does not appear to be correct. A number of studies of the kinetics of the reaction in the pH range 2-7 have been carried out where the iodine reduction process exhibited a 1/[H+]2 dependence consistent with the proposed mechanism which were attributed primarily to the reaction of H2O2 with IO-. Deviations were observed in the pH range 6-7 and were explained by incorporating the reaction of I2OH- with H2O2. In some other experiments it was suggested that the failure to maintain a 1/[H+]2 dependence at high pH was due to the iodine hydrolysis being rate determining. Data from an experimental program performed at AECL described in this paper confirms that the 1/[H+]2 dependence does not hold at high pH. These studies were carried out as a function of acid, iodide, peroxide and buffer concentration for three buffers, barbital, citrate and phosphate. This paper discuss two mechanisms which involve the formation of an HOOI intermediate in the rate determining step and which adequately describe the experimental data. (author) 4 figs., 2 tabs., 23 refs

  11. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

  15. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  16. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  17. Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1997-02-01

    Inspection of steam generator tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness-for-service of the steam generators. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. A major challenge continues to be the detection and sizing of circumferential cracks. Utilities around the world have experienced this type of tube failure. Conventional in-service inspection, performed with eddy current bobbin probes, is ineffectual in detecting circumferential cracks in tubing. It has been demonstrated in CANDU steam generators, with deformation, magnetite and copper deposits that multi-channel probes with transmit-receive eddy current coils are superior to those using surface impedance coils. Transmit-receive probes have strong directional properties, permitting probe optimization according to crack orientation. They are less sensitive to lift-off noise and magnetite deposits and possess good discrimination to internal defects. A single pass C3 array transmit-receive probe developed by AECL can detect and size circumferential stress corrosion cracks as shallow as 40% through-wall. Since its first trial in 1992, it has been used routinely for steam generator in-service inspection of four CANDU plants, preventing unscheduled shutdowns due to leaking steam generator tubes. More recently, a need has surfaced for simultaneous detection of both circumferential and axial cracks. The C5 probe was designed to address this concern. It combines transmit-receive array probe technology for equal sensitivity to axial and circumferential cracks with a bobbin probe for historical reference. This paper will discuss the operating principles of transmit-receive probes, along with inspection results.

  18. The buffer/container experiment design and construction report

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, N.A.; Wan, A.W.L.; Roach, P.J

    1998-03-01

    The Buffer/Container Experiment was a full-scale in situ experiment, installed at a depth of 240 m in granitic rock at AECL's Underground Research Laboratory (URL). The experiment was designed to examine the performance of a compacted sand-bentonite buffer material under the influences of elevated temperature and in situ moisture conditions. Buffer material was compacted in situ into a 5-m-deep, 1.24-m-diameter borehole drilled into the floor of an excavation. A 2.3-m long heater, representative of a nuclear fuel waste container, was placed within the buffer, and instrumentation was installed to monitor changes in buffer moisture conditions, temperature and stress. The experiment was sealed at the top of the borehole and restrained against vertical displacement. Instrumentation in the rock monitored pore pressures, temperatures and rock displacement. The heater was operated at a constant power of 1200 W, which provided a heater skin temperature of approximately 85 degrees C. Experiment construction and installation required two years, followed by two and a half years of heater operation and two years of monitoring the rock conditions during cooling. The construction phase of the experiment included the design, construction and testing of a segmental heater and controller, geological and hydrogeological characterization of the rock, excavation of the experiment room, drilling of the emplacement borehole using high pressure water, mixing and in situ compaction of buffer material, installation of instrumentation in the rock, buffer and on the heater, and the construction of concrete curb and steel vertical restraint system at the top of emplacement borehole. Upon completion of the experiment, decommissioning sampling equipment was designed and constructed and sampling methods were developed which allowed approximately 2000 samples of buffer material to be taken over a 12-day period. Quality assurance procedures were developed for all aspects of experiment

  19. Building condition assessment program - risk evaluation and liability management

    International Nuclear Information System (INIS)

    Chalk River Laboratories (CRL) is a large nuclear research and development/industrial site operated by Atomic Energy of Canada Limited (AECL). The CRL site consists of a 70 hectare developed (industrial) site located within a larger undeveloped area (Supervised Area - 37 km2, or 3700 hectares). Construction of the CRL site started in 1944. The development and operating history includes the construction and operation of 7 research reactors and numerous associated supporting nuclear laboratories, including fuel fabrication facilities, research laboratories, test facilities, and waste processing facilities. Numerous other support facilities were also constructed, such as administrative and office buildings, manufacturing facilities, and buildings for essential services such as fire and security services. Altogether, the CRL site includes roughly 120 buildings (Figure 1), and the site continues to operate in the fields of nuclear research and development and medical isotope production. Within this operating environment, a number of buildings, facilities, structures, and reactors (hereinafter referred to as buildings), have become redundant and have been shut down for various reasons. Redundant buildings are currently shut down within the operating organization and turned over to the decommissioning organization for decommissioning, but in the early years, in the absence of a decommissioning program, redundant buildings were most often simply placed into storage for an undefined period. As a result, there are a significant number of buildings at CRL that have been declared redundant (roughly 20, or 1400 m2 ), particularly those constructed in the early years of site development. Further, with many buildings at CRL approaching the ends of their design life, a significant number of other buildings will become redundant during the next decade (an additional 20). (author)

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

    International Nuclear Information System (INIS)

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